• Journal of Radiation Protection and Research
• /
• 제26권3호
• /
• pp.167-175
• /
• 2001

Semiconductor diode detectors coated with neutron reactive material are presently under investigation for various uses, such as remote sensing of thermal neutrons, fast neutron counting, and thermal neutron radiography. Theory indicates that single-coated devices can yield thermal neutron efficiencies from 4% to 11 %, which is supported by experimental evidence. Radiation endurance measurements indicate that the devices function well up to a limiting thermal neutron fluence of $10^{13}/cm^2$, beyond which noticeable degradation occurs. Thermal neutron contrast images of step wedges and simple phantoms, taken with dual in-line pixel devices, show promise for thermal neutron imaging detectors.

• Nuclear Engineering and Technology
• /
• 제55권7호
• /
• pp.2366-2372
• /
• 2023

Are borate and silicate glasses suitable for working as shieling materials against fast neutrons? To correctly answer the above question, some silicate, and borate-based glasses were fabricated and irradiated with various doses of fast neutrons varied between 1.73 and 12.10 MGy. The color and hardness of the fabricated glasses were affected by the fast neutron fluence where the transparent glasses turned colored as well as the hardness of the fabricated glasses was decreased. The gamma-ray spectrometric analysis shows a high activity concentration produced in the barium borate glasses due to the formation of radioisotopes Ba-131 and Ba-133 reaches to 5.92E+05 Bq and 4.25E+03 Bq, respectively for sample Cd-5 Batch 3. Additionally, the gamma-ray spectrometric analysis for the sodium silicate glasses shows low activity concentrations emitted from isotopes formed due to the activation of Y₂O₃-associated impurities. These activities are low compared to that emitted by barium borate-based glasses.

• Nuclear Engineering and Technology
• /
• 제48권2호
• /
• pp.351-359
• /
• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

• Radiation Oncology Journal
• /
• 제19권1호
• /
• pp.66-73
• /
• 2001

목적 : 붕소-중성자 포획치료법(Boron Neutron Capture Therapy, BNCT)을 위해 원자력병원 싸이클로트론에서 발생되는 최대에너지 34.4 MeV의 속중성자(Fast neutron)를 70 cm 파라핀으로 감속시킨 후 선량 특성을 조사하였다. 그 결과를 토대로 열외중성자(Epithermal neutron) 선량 측정법에 대한 프로토콜을 확립하여 원자로에서 방출되는 열외 중성자 선량 특성 평가의 기초를 삼고, 가속기를 이용한 BNCT 연구에 대한 타당성 여부를 조사하고자 한다. 대상 및 방법 : 공기 중 선량 및 물질 내 선량 분포 측정을 위해 Unidos 10005 (PTW, Germany) 전기계와 조직 등가 물질인 A-150 플라스틱으로 제작된 IC-17 (Far West, USA) 및 IC-18, ElC-1 이온함을 사용하였고, 감마선의 측정을 위해서는 마그네슘으로 제작된 IC-l7M 이온함을 이용하였으며 조직등가 기체와 아르곤 기체를 분당 5cc 씩 주입하며 측정하였다. 중성자, 광자, 전자가 혼합된 장의 모의 수송 해석을 위해 이용되는 Monte Carlo N-Particle (MCNP) transport code를 사용하여 2차원적 선량 분포 및 에너지 분포를 계산하였으며 이 결과를 측정값과 비교하였다. 결과 : BNCT에서의 유효 치료 깊이인 물 팬텀 4 cm에서의 선량은 치료기 1 MU 당 $6.47\times10^{-3}\;cGy$로 미세하였으며, 이때 감마 오염도(contamination)는 $65.2{\pm}0.9\%$로 중성자보다는 감마선에 의한 선량 기여분이 우세하였다. 깊이에 따른 선량 분포 특성에서는 중성자 선량은 선형적으로 감쇠 되었고, 감마선량은 지수적으로 보다 급격히 감쇠되는 경향을 보였으며 전체 선량의 $D_{20}/D_{10}$은 0.718 이었다. MCNP에 의한 에너지 분포 전산 계산의 결과 2.87 MeV 이하에서 중성자 피크가 나타났으며, 저에너지 영역에서는 감마선이 연속적으로 분포되는 양상을 보였다. 결론 : 벽 물질이 서로 다른 두 개의 이온함을 사용한 직접 선량 측정과 MCNP 전산 시뮬레이션을 이용한 공간 선량분포 계산으로 미세 속중성자 빔에 대한 선량 특성을 파악할 수 있었으며, 원자로 열외중성자 주(Epithermal neutron column)에 대한 선량 평가 자료로 확보하였다. 아울러 가속기에 대한 연구가 진행되어 고전압, 고전류를 발생시키는 전원 공급장치와 표적핵(Target) 물질이 개발되고 비스무스나 납 등에 의해 감마 오염도를 줄일 경우, 싸이크로트론에 의한 보론-중성자 포획치료도 가능해질 것으로 판단된다.

• Nuclear Engineering and Technology
• /
• 제41권7호
• /
• pp.979-988
• /
• 2009

Tensioned metastable fluids provide a powerful means for low-cost, efficient detection of a wide range of nuclear particles with spectroscopic capabilities. Past work in this field has relied on one-component liquids. Pure liquids may provide very good detection capability in some aspects, such as low thresholds or large radiation interaction cross sections, but it is rare to find a liquid that is a perfect candidate on both counts. It was hypothesized that liquid mixtures could offer optimal benefits and present more options for advancement. However, not much is known about radiation-induced thermal-hydraulics involving destabilization of mixtures of tensioned metastable fluids. This paper presents results of experiments that assess key thermophysical properties of liquid mixtures governing fast neutron radiation-induced cavitation in liquid mixtures. Experiments were conducted by placing liquid mixtures of various proportions in tension metastable states using Purdue's centrifugally-tensioned metastable fluid detector (CTMFD) apparatus. Liquids chosen for this study covered a good representation of both thermal and fast neutron interaction cross sections, a range of cavitation onset thresholds and a range of thermophysical properties. Experiments were devised to measure the effective liquid mixture viscosity and surface tension. Neutron-induced tension metastability thresholds were found to vary non-linearly with mixture concentration; these thresholds varied linearly with surface tension and inversely with mixture vapor pressure (on a semi-log scale), and no visible trend with mixture viscosity nor with latent heat of vaporization.

• 대한방사선기술학회지:방사선기술과학
• /
• 제31권4호
• /
• pp.415-417
• /
• 2008

현재 방사선치료는 선형가속기에 의하여 대다수 이루어지고 있으며 사용되는 방사선인 광자도 의학의 발전에 의해 고에너지화 고선량화 되고 있다. 본 연구에서는 방사선치료 조사면에서 중성자 측정이 가능한 CR-39를 이용한 중성자 검출법을 이용하였다. 선형가속기에서 발생되는 X선(광자)치료 시 발생 되는 중성자의 선량을 CR-39를 이용한 중성자 검출법을 이용하여 측정하고, 임상적 응용으로 고에너지 광자를 이용하여 암 치료에 사용할 때 중성자의 발생이 환자치료 선량과 연관되는 어떤 문제를 발생시키는지를 연구한 결과는 다음과 같다. 속중성자의 경우 광자 1Gy 조사 시 평균 0.35mSv, 2Gy 조사 시 평균 0.65mSv, 5Gy 조사 시 평균 1.82mSv, 열중성자의 경우 광자 1Gy 조사 시 평균 0.26mSv, 2Gy 조사 시 평균 0.56mSv, 5Gy 조사 시평균 1.23mSv의 중성자 발생하였다. Wedge Filter를 사용하여 중성자의 발생을 측정한 결과 Wedge Filter를 사용했을 때 중성자의 발생이 증가하였다. 고선량을 요구하는 SRS Cone을 사용했을 때에는 기존의 실험결과 보다 많은 중성자가 검출되었다. 속중성자의 경우 광자 5Gy 조사 시 평균 2.85mSv, 열중성자의 경우 광자 5Gy 조사 시 평균 1.37mSv의 중성자가 발생하였다. 일반 치료 시 광자 5Gy 조사했을 때 보다 속중성자의 경우 약 1.6배, 열중성자의 경우 약 1.12배 정도의 중성자가 더 발생하는 것으로 나타났다.

• Nuclear Engineering and Technology
• /
• 제52권5호
• /
• pp.1002-1007
• /
• 2020

The purpose of the current study was to evaluate a spectrum formulation set employed to modify the neutron spectrum of D-D fusion neutrons in a IS plasma focus device using GEANT4, MCNPX2.6, and FLUKA codes. The set consists of a moderator, reflector, collimator and filters of fast neutron and gamma radiation, which placed on the path of 2.45 MeV neutron energy. The treated neutrons eliminate cancerous tissue with minimal damage to other healthy tissue in a method called neutron therapy. The system optimized for a total neutron yield of 10⁹ (n/s). The numerical results indicate that the GEANT4 code for the cubic geometry in the Beam Shaping Assembly 3 (BSA3) is the best choice for the energy of epithermal neutrons.

• Nuclear Engineering and Technology
• /
• 제50권1호
• /
• pp.80-87
• /
• 2018

Consideration is given to the possibility to use changes in buoyancy as a negative reactivity feedback mechanism during temperature transients in heavy liquid metal fast reactors. It is shown that by the proper use of heavy pellets in the fuel elements, fuel rods could be endowed with a passive self-ejection mechanism and then with a negative feedback. A first estimate of the feasibility of the mechanism is calculated by using a simplified geometry and model. If in addition, a neutron poison pellet is introduced at the bottom of the fuel, then when the fuel element is displaced upward by buoyancy force, the reactivity will be reduced not only by disassembly of the core but also by introducing the neutron poison from the bottom. The use of induced buoyancy opens up the possibility of introducing greater amounts of actinides into the core, as well as providing a palliative solution to the problem of positive coolant temperature reactivity coefficients that could be featured by the heavy liquid metal fast reactors.

• Nuclear Engineering and Technology
• /
• 제53권2호
• /
• pp.423-429
• /
• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

• Nuclear Engineering and Technology
• /
• 제4권1호
• /
• pp.47-58
• /
• 1972