• Title/Summary/Keyword: Neutron capture

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An Assessment on the Contribution of $^3$He to the Tritium Generation in the CANDU PHWR (가압중수로에서 헬륨-3이 삼중수소의 생성에 미치는 영향평가)

  • Kwak, Sung-Woo;Chung, Bum-Jin
    • Journal of Radiation Protection and Research
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    • v.22 no.2
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    • pp.119-125
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    • 1997
  • PHWR achieves high neutron economy by adopting heavy water as its moderator and coolant. On the other hand it permits much tritium generation, compared to LWR, due to the neutron capture reaction of deuterium in heavy water. Meanwhile in the reactor core, $^3He formed as the result of-decay of tritium, captures a thermal neutron and transforms to tritium again. The existing calculation models on tritium generation in PHWR neglect the contribution of $^3He$ in both moderator and coolant due to its relatively low solubility. However the neutron capture cross-section of $^3He$ is almost $1.6{\times}10^7$ times as large as that of deuterium. That means that the dissolved amount of 0.03 ppm of $^3He$ in heavy water is enough to generate the same amount of tritium as that generated by the deuterium of total heavy water in the system. This study dealt with the contribution of $^3He$ to tritium generation. As a sample case, the contribution of $^3He$ to the tritium generation in Wolsong #1 was evaluated and compared to the measured values. According to the result of this study, it is concluded that $^3He$ in coolant contributes very much to the tritium generation but that in moderator shows negligible effects due to the low solubility and $^4He$ cover gas. At the beginning of the plant operation, the contribution of $^3He$ is slightly greater than the measured value but agrees well with the measured as the operating time increases.

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Reference based simulation study of detector comparison for BNCT-SPECT imaging

  • Kim, Moo-Sub;Shin, Han-Back;Choi, Min-Geon;Monzen, Hajime;Shim, Jae Goo;Suh, Tae Suk;Yoon, Do-Kun
    • Nuclear Engineering and Technology
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    • v.52 no.1
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    • pp.155-163
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    • 2020
  • To investigate the optimal detector material for prompt gamma imaging during boron neutron capture therapy, in this study, we evaluated the characteristic regarding radiation reaction of available detector materials using a Monte Carlo simulation. Sixteen detector materials used for radiation detection were investigated to assess their advantages and drawbacks. The estimations used previous experimental data to build the simulation codes. The energy resolution and detection efficiency of each material was investigated, and prompt gamma images during BNCT simulation were acquired using only the detectors that showed good performance in our preliminary data. From the simulation, we could evaluate the majority of detector materials in BNCT and also could acquire a prompt gamma image using the six high ranked-detector materials and lutetium yttrium oxyorthosilicate. We provide a strategy to select an optimal detector material for the prompt gamma imaging during BNCT with three conclusions.

A Study on the Technology of Measuring and Analyzing Neutrons and Gamma-Rays Using a CZT Semiconductor Detector (CZT 반도체 검출기를 활용한 중성자 및 감마선 측정과 분석 기술에 관한 연구)

  • Jin, Dong-Sik;Hong, Yong-Ho;Kim, Hui-Gyeong;Kwak, Sang-Soo;Lee, Jae-Geun
    • Journal of radiological science and technology
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    • v.45 no.1
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    • pp.57-67
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    • 2022
  • CZT detectors, which are compound semiconductors that have been widely used recently for gamma-ray detection purposes, are difficult to detect neutrons because direct interaction with them does not occur unlike gamma-rays. In this paper, a method of detecting and determining energy levels (fast neutrons and thermal neutrons) of neutrons, in addition of identifying energy and nuclide of gamma-rays, and evaluating gamma dose rates using a CZT semiconductor detector is described. Neutrons may be detected by a secondary photoelectric effect or compton scattering process with a characteristic gamma-ray of 558.6 keV generated by a capture reaction (113Cd + 1n → 114Cd + 𝛾) with cadmium (Cd) in the CZT detector. However, in the case of fast neutrons, the probability of capture reaction with cadmium (Cd) is very low, so it must be moderated to thermal neutrons using a moderator and the material and thickness of moderator should be determined in consideration of the portability and detection efficiency of the equipment. Conversely, in the case of thermal neutrons, the detection efficiency decreases due to shielding effect of moderator itself, so additional CZT detector that do not contain moderator must be configured. The CZT detector that does not contain moderator can be used to evaluate energy, nuclide, and gamma dose-rate for gamma-rays. The technology proposed in this paper provides a method for detecting both neutrons and gamma-rays using a CZT detector.

Elemental Analysis by Neutron Induced Nuclear Reaction - Prompt Gamma Neutron Activation Analysis for Chemical Measurement - (중성자 핵반응을 이용한 원소 검출기술 - 즉발감마선 중성자 방사화분석법을 이용한 검출기술 -)

  • Song, Byung Chul;Park, Yong Joon;Jee, Kwang Yong
    • Analytical Science and Technology
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    • v.16 no.5
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    • pp.1041-1051
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    • 2003
  • Neutron induced prompt gamma activation analysis (PGAA) offers a nondestructive, sensitive and relatively rapid method for the determination of trace and major elements and is proven to be convenient for online analysis of minerals, metals, coal, cement, petrochemical, coating, paper as well as many other materials and products. The technique has found many uses in medicine, industry, research, security and the detection of contraband items. This report reviews the present status and future trends of the PGAA techniques. Requirements for the system are neutron source, high resolution HPGe detectors with a high-voltage power supply, an amplifier, analog-to-digital converter, and a multichannel analyzer for the detection and measurement of prompt ${\gamma}$-ray emit form the neutron capture elements. Introducing a ${\gamma}$-${\gamma}$ coincidence system also improves the quality of the ${\gamma}$-ray spectrum by suppressing the background created from the Compton scattering of high energy prompt ${\gamma}$-rays. A PGAA system using a $^{252}Cf$ neutron source is currently under construction for the on-line measurement of several elements in aqueous samples at KAERI. The system can be applied for the detection of chemical weapons and explosives as well as various narcotics.

New Boron Compound, Silicon Boride Ceramics for Capturing Thermal Neutrons (Possibility of the material application for nuclear power generation)

  • Matsushita, Jun-ichi
    • Proceedings of the Materials Research Society of Korea Conference
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    • 2011.05a
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    • pp.15-15
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    • 2011
  • As you know, boron compounds, borax ($Na_2B_4O_5(OH)_4{\cdot}8H_2O$) etc. were known thousands of years ago. As for natural boron, it has two naturally occurring and stable isotopes, boron 11 ($^{11}B$) and boron 10 ($^{10}B$). The neutron absorption $^{10}B$ is included about 19~20% with 80~81% $^{11}B$. Boron is similar to carbon in its capability to form stable covalently bonded molecular networks. The mass difference results in a wide range of ${\beta}$ values between the $^{11}B$ and $^{10}B$. The $^{10}B$ isotope, stable with 5 neutrons is excellent at capturing thermal neutrons. For example, it is possible to decrease a thermal neutron required for the nuclear reaction of uranium 235 ($^{235}U$). If $^{10}B$ absorbs a neutron ($^1n$), it will change to $^7Li+^1{\alpha}$ (${\alpha}$ ray, like $^4He$) with prompt ${\gamma}$ ray from $^{11}B$ $^{11}B$ (equation 1). $$^{10}B+^1n\;{\rightarrow}\;^{11}B\;{\rightarrow}\; prompt \;{\gamma}\;ray (478 keV), \;^7Li+4{\alpha}\;(4He)\;\;\;\;{\cdots}\; (1)$$ If about 1% boron is added to stainless steel, it is known that a neutron shielding effect will be 3 times the boron free steel. Enriched boron or $^{10}B$ is used in both radiation shielding and in boron neutron capture therapy. Then, $^{10}B$ is used for reactivity control and in emergency shutdown systems in nuclear reactors. Furthermore, boron carbide, $B_4C$, is used as the charge of a nuclear fission reaction control rod material and neutron cover material for nuclear reactors. The $B_4C$ powder of natural B composition is used as a charge of a control material of a boiling water reactor (BWR) which occupies commercial power reactors in nuclear power generation. The $B_4C$ sintered body which adjusted $^{10}B$ concentration is used as a charge of a control material of the fast breeder reactor (FBR) currently developed aiming at establishment of a nuclear fuel cycle. In this study for new boron compound, silicon boride ceramics for capturing thermal neutrons, preparation and characterization of both silicon tetraboride ($SiB_4$) and silicon hexaboride ($SiB_6$) and ceramics produced by sintering were investigated in order to determine the suitability of this material for nuclear power generation. The relative density increased with increasing sintering temperature. With a sintering temperature of 1,923 K, a sintered body having a relative density of more than 99% was obtained. The Vickers hardness increased with increasing sintering temperature. The best result was a Vickers hardness of 28 GPa for the $SiB_6$ sintered at 1,923K for 1 h. The high temperature Vickers hardness of the $SiB_6$ sintered body changed from 28 to 12 GPa in the temperature range of room temperature to 1,273 K. The thermal conductivity of the SiB6 sintered body changed from 9.1 to 2.4 W/mK in the range of room temperature to 1,273 K.

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Determination of Boron Isotopic Ratio by Using an Alpha Track Technique

  • Park, Yong-Joon;Pyo, Hyung-Yeal;Song, Kyu-Seok;Song, Byoung-Chul;Jee, Kwang-Yong;Kim, Won-Ho
    • Bulletin of the Korean Chemical Society
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    • v.27 no.10
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    • pp.1609-1612
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    • 2006
  • The determination of the boron isotopic ratio in solutions was achieved by means of a solid state track detector by using an alpha track. The neutron flux was optimized by using a Cd-foil to find the optimum conditions for counting the number of alpha tracks on the selected solid detector caused by the (n, $\alpha$) nuclear reaction of boron. The home-made multi-dot detector plate was utilized in this study to increase the reproducibility of the measurement by uniformly drying the boron solution within the marked circle area on the detector plate. The experimental results of this study verified that the $^{11}B/^{10}B $ isotopic ratio can be measured by observing the number of alpha tracks for different concentrated standard solutions with various isotopic compositions. This technique was applied to the determination of $^{10}B$ enrichment factor in a biological sample for a boron neutron capture therapy.

Recoil Effects of Neutron-irradiated Metal Salts

  • Lee, Byung-Hun;Lee, Jong-Du
    • Nuclear Engineering and Technology
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    • v.12 no.2
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    • pp.99-105
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    • 1980
  • The distribution of $^{55}$ Mn and $^{38}$ Cl recoil species following radiative neutron capture in permanganates, chlorates and perchlorates has been investigated by using ion-exchange chromatography method. The whole of the $^{55}$ Mn radioactivity in permanganates appeared in two valence states, the $^{38}$ Cl radioactivity in chlorates in two valence states and also the $^{38}$ Cl radioactivity in perchlorates in three valence states. Recoil energy was calculated. The internal conversion of $^{38}$ Cl isomer transition affects the retention value. The greater the radii of the cation, the higher is the probability of the recoil atom breaking through the secondary cage. In ammonium salt, the ammonium ion behaves as a reducing agent. Crystal structures with their greater free space have shown low retention.

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ANALYSIS OF DOPPLER-BROADENED PEAK IN THERMAL NEUTRON INDUCED 10B(n,α γ)7Li REACTION USING HYPERGAM

  • Choi, Hee-Dong;Jung, Nam-Suk;Park, Byung-Gun
    • Nuclear Engineering and Technology
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    • v.41 no.1
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    • pp.113-124
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    • 2009
  • The line shape functions for the Doppler-broadened gamma ray spectrum are considered in the $^{10}B(n,{\alpha}{\gamma})^7Li$ reaction occurring in a surrounding medium where the excited $^7Li$ nucleus is slowed down and stopped before decay. The phenomenological form of the stopping power was used for the broadening effect. Convolution with the detailed response of a germanium detector is taken into consideration for the simplest case of solely electronic stopping. A numerical study for the analysis of $^{10}B$ by thermal neutron capture is conducted by performing a parametric search and fitting the measured spectrum in a least-squares approach. In comparison with the previous numerical approach using the same analysis, the computational speed is increased and reliable information concerning the stopping power of the medium is obtained while estimating the uncertainty. Implementation of the routine analysis of $^{10}B$ is facilitated on a recent version of the gamma ray spectrum analysis package HyperGam.

A Study on the Radioactive Products of Components in Proton Accelerator on Short Term Usage Using Computed Simulation (몬테칼로 시뮬레이션을 활용한 양성자가속기 단기사용 시 구성품의 방사화 평가)

  • Bae, Sang-Il;Kim, Jung-Hoon
    • Journal of radiological science and technology
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    • v.43 no.5
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    • pp.389-395
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    • 2020
  • The evaluation of radioactivated components of heavy-ion accelerator facilities affects the safety of radiation management and the exposure dose for workers. and this is an important issue when predicting the disposal cost of waste during maintenance and dismantling of accelerator facilities. In this study, the FLUKA code was used to simulate the proton treatment device nozzle and classify the radio-nuclides and total radioactivity generated by each component over a short period of time. The source term was evaluated using NIST reference beam data, and the neutron flux generated for each component was calculated using the evaluated beam data. Radioactive isotopes caused by generated neutrons were compared and evaluated using nuclide information from the International Radiation Protection Association and the Korea Radioisotope association. Most of the nuclides produced form of beta rays and electron capture, and short-lived nuclides dominated. However, In the case of 54Mn, which is a radioactive product of iron, the effect of gamma rays should be considered. In the case of tritium generated from a material with a low atomic number, it is considered that handling care should be taken due to its long half-life.