• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1871-1877
• /
• 2023

The main goal of the coordinated project development of therapeutic radiopharmaceuticals of Y-90 and Re-188 is to exploit advancements in radionuclide production technology. Here, direct and indirect production methods with medium reactor and cyclotron are compared to evaluate derived neutron flux and production yield. First, nano-sized ¹⁸⁶W and ⁸⁹Y specimens are suspended in water in a quartz vial by FLUKA simulation. Then, the solution is irradiated for 4 days under 9E+14 n/cm2/s neutron flux of reactor. Also, a neutron activator including three layers-lead moderator, graphite reflector, and polyethylene absorbent- is simulated and tungsten target is irradiated by 60 MeV protons of cyclotron to generate induced neutrons for ¹⁸⁸W and ⁹⁰Sr production via neutron capture. As the neutron energy reduced, the flux gradually increased towards epithermal range to satisfy (n/2n,γ) reactions. The obtained specific activities at saturation were higher than the reported experimental values because the accumulated epithermal flux and nano-sized specimens influence the outcomes. The beta emitters, which are widely utilized in brachytherapy, appeal an alternative route to locally achieve a rational yield. Therefore, the proposed method via neutron activator may ascertain these broad requirements.

• 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.

• 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.

• Proceedings of the KIEE Conference
• /
• 2005.07c
• /
• pp.2184-2186
• /
• 2005

Neutron production is very important to apply fusion energy through SCBF(Spherically Convergent Beam Fusion) device and its rate is Proportional to the square of the ion current$({\propto}I^2)$. Also the ion current has a close relation with the potential well structure in grid cathode. In this paper, the ion current is calculated for the increasement of neutron production rate in a variety of grid cathode geometry. The atomic and molecular collision are taken into account by Monte Carlo Method and Potential is calculated by Finite Element Method. Main processes of the discharge is the ionization of $D_2$ by fast $D_2^+$ ion. As the number of a cathode ring is small and gap distance decreases, the ion current increases and neutron production rate will increase.

• Nuclear Engineering and Technology
• /
• v.15 no.2
• /
• pp.160-168
• /
• 1983

About 25 years ago, halfway along the recorded history of the neutron as a separate entity, Korea entered the nuclear age and initiated its own neutron research and development programs. Since that time Korean scientists have taken all possible advantages of the special opportunities offered by the neutron. Scientists the world over, in the Far East, hear East, and the West, have adapted these opportunities to their special needs. These needs are manifested in all phases of modern life, including power generation by nuclear means, food preservation, production of new types of food-bearing plants, commercial uses of activation analysis, irradiations, and isotope production, nuclear medicine, industrial quality control through nuclear measurements, and direct use of neutrons in research in many areas including solid state physics, chemistry, physics, biology, and medicine. Research with neutrons has been successfully conducted using nuclear research reactors of all sizes ranging from the very small (∼10 kilowatts) to the very large(50-100 Megawatts). This speaker has teen associated with nuclear research since 1945 and directly with neutron research since 1957. From this continuous research and development activity, he will report on some of the prospects in the second 50 years of the neutron.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
• /
• v.20 no.5
• /
• pp.467-470
• /
• 2007

Application field of neutron beam is very broad including industry, medicine and science. But the research and development and use of neutron beam is restricted within in narrow limits in this country, because neutron beam facility is insufficient - a big research facility of nuclear reactor(HANARO) and some small industrial facilities which use radioisotope neutron source are available. This paper compare and investigate the results of experiment and numerical analysis of the discharge in the spherically convergent beam fusion device which were expected as a portable neutron source. The spherically convergent beam fusion device will offer stability in neutron production, possibility of movement for convenience, low construction cost and higher neutron flux than radioisotope neutron source. The star mode discharge which efficiently generate neutron, were observed at both results.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2334-2337
• /
• 2022

Deuterium-tritium reaction is the most promising one in term of the highest nuclear fusion cross-section for the reactor. So it is one of urgent issues to develop materials and components that are simultaneously resistant to high heat flux and high energy neutron flux in realization of the fusion energy. 2.45 MeV neutron production was reported in D-D reaction in KSTAR and regarded as beam-target is the dominant process. The feasibility study of KSTAR to wide area neutron source facility is done in term of D-D and D-T reactions from the empirical scaling law from the mixed fast and thermal stored energy and its projection to cases of heating power upgrade and DT reaction is done.

• Nuclear Engineering and Technology
• /
• v.52 no.7
• /
• pp.1517-1523
• /
• 2020

The widely used medical isotope technetium-99 m (^99mTc) is a daughter of Molybdenum-99 (⁹⁹Mo), which is mainly produced using dedicated research reactors from the nuclear fission of uranium-235 (²³⁵U). ^99mTc has been used for several decades, which covers about 80% of the all the nuclear diagnostics procedures. Recently, the instability of the supply has become an important topic throughout the international radioisotope communities. The aging of major ⁹⁹Mo production reactors has also caused frequent shutdowns. It has triggered movements to establish new research reactors for ⁹⁹Mo production, as well as the development of various ⁹⁹Mo production technologies. In this context, a new research reactor project was launched in 2012 in Korea. At the same time, the development of fission-based ⁹⁹Mo production process was initiated by Korea Atomic Energy Research Institute (KAERI) in 2012 in order to be implemented by the new research reactor. The KAERI process is based on the caustic dissolution of plate-type LEU (low enriched uranium) dispersion targets, followed by the separation and purification using a series of columns. The development of proper waste treatment technologies for the gaseous, liquid, and solid radioactive wastes also took place. The first stage of this process development was completed in 2018. In this paper, the results of the hot test production of fission ⁹⁹Mo using HANARO, KAERI's 30 MW research reactor, was described.

• 한국초전도학회:학술대회논문집
• /
• 2008.08a
• /
• pp.87-87
• /
• 2008

• Nuclear Engineering and Technology
• /
• v.54 no.2
• /
• pp.637-643
• /
• 2022

Uranium and plutonium are required to be accounted in spent fuel head-end and major recovery area in pyro-process for safeguards purpose. The possibility of neutron resonance technique, as a nondestructive analysis, was simulated on isotopic fissile analysis for large scale process. Neutron resonance technique has advantage to distinguish uranium from plutonium directly in mixture. Simulation was performed on U235 and Pu239 assay in spent fuel and for scoping examination of assembly type. The resonance energies were determined for U235 and Pu239. The linearity in the neutron transmission was examined for the selected resonance energies. In addition, the limit for detection was examined by changing sample density, thickness and content for actual application. Several factors were proposed for neutron production and the moderated neutron source was simulated for effective and efficient transmission measurement. From the simulation results, neutron resonance technique is promising to analyze U235 and Pu239 for spent fuel assembly. An accurate fissile assay will contribute to an increased safeguards for the pyro-processing system and international credibility on the reuse of fissile materials in the fuel cycle.