• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.148-157
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• 1997

A new capsule that has a unique structure in which the test environments including neutron flux and fluence, and irradiation temperature can be controlled precisely during irradiation, was conceptually designed. The capsule structure and instrumentation were successfully designed according to the JMTR's standard procedures of capsule design. Based on the target irradiation, the details of the irradiation such as neutron fluence and irradiation temperature ore calculated and the related capsule safety was evaluated. In addition, the effects of design parameters including the changes in inner-capsule configuration, heater capacity, and Helium gas pressure on the specimen temperature were analyzed with a computer program. Through these thermal and strength evaluations, this capsule was proved to be safe during the irradiation in the JMTR.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.315-319
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• 2001

It is shown, that both cosmic radiation (external source) and natural radioactive gases (inner source) are sources of neutrons near the Earth crust. Correlation between the Earth crust dynamics and variations of thermal and slow neutron flux near the Earth surface is studied. It is shown, that variations of neutron flux near the Earth crust can be used for short-term predicting of natural hazards.

• Journal of the Korean Chemical Society
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• v.7 no.2
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• pp.165-169
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• 1963

12 samples from an assay process chain were submitted to qualitative and quantitative neutron activation analysis for the determination of gold. Gold was detected and quantitatively determined in three samples after a chemical separation consisting of solvent extraction and precipitation steps. Recoveries ranged between 81.0 and 93.6% and results of duplicated determinations were reproducible. Quantitative data were obtained from gamma-spectrometric photopeak-area counting. Interference from fast neutron reactions was negligible.

• Journal of the Korean Chemical Society
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• v.5 no.1
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• pp.26-28
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• 1961

Microgram quantities of thallirum in meteorites, rocks, minerals, alloys and biological samples have been analyzed by rapid neutron activation analysis. A l0-minute radiochemical separation procedure coupled with a gas-flow proportional detector for 4.2-minute half life measurement and a gamma or beta scintillation detector placed in close proximity to the sample permitted detection of the 4.2-min Tl206. Samples were irradiated for 10-minutes at a thermal neutron flux of approximately $0.95{\times}10^{11}$ neutron-$cm^{-2}-sec^{-1}.$ The low limits of detection was about $10^{-7}$ gm of thallium.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.299-303
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• 2016

Prompt gamma activation analysis using a thermal neutron-guided beam at Japan Atomic Energy Agency JRR-3M was applied for the precise determination of Si in silicon nitride ceramic reference materials [Japan Ceramic Reference Material (JCRM) R 003]. In this study, the standard addition method coupled with internal standard was used for the nondestructive determination of Si in the sample. Cadmium was used as internal standard to obtain the linear calibration curves and to compensate for the neutron beam variability. The analytical result of determining Si in JCRM R 003 silicon nitride fine powder ceramic reference materials using prompt gamma activation analysis was in good agreement with that obtained by classical gravimetric analysis. The relative expanded measurement uncertainty (k = 2) associated with the determined value was 2.4%.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• Journal of Radiopharmaceuticals and Molecular Probes
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• v.8 no.1
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• pp.17-23
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• 2022

Boron neutron capture therapy is a precision treatment technology that selectively destroys only tumor cells by irradiating thermal neutrons after accumulating boron drugs in tumor cells. Brain tumor is difficult to diagnose and treat due to the low permeability and targeting of drugs caused by the blood-brain-barrier. Crossing the BBB is essential for drug delivery to the brain. In this study, we designed and synthesized a novel compound incorporating benzothiazole to develop a boron drug with high BBB permeability and selectivity for brain tumor cells. In addition, their potential as a BNCT drugs was evaluated.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3668-3685
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• 2024

The operation of a nuclear power plant relies on precalculated nuclear design predictions based on core calculations of various reactor states. The fuel temperature is a crucial factor in determining the reactor fuel behavior, but assessing the temperature variation in a fuel pellet taking into account neutron transport is challenging. Detailed simulation of the temperature behavior within the fuel pellet can be obtained by coupling of Monte Carlo neutron transport codes with thermal-hydraulics solvers. However, this approach is not practical for standard nuclear design calculations, and computationally cheaper and faster methods must be used. In nuclear core simulators, a concept of a single "effective temperature" that yields the same neutron response as in the case of the actual temperature shape is mainly applied. This paper evaluates various fuel temperature models used in nuclear core simulation calculations, ultimately recommending a new effective temperature model that considers the burnup correction.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2717-2724
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• 2020

Silicon, especially silicon in the form of SiO₂, is a major component of rocks. Final spent fuel storages, which are being designed, are located in suitable rock formations in the Earth's crust. Reduction of the uncertainty of silicon neutron scattering and capture is needed; improved silicon evaluations have been recently produced by the ORNL/IAEA collaboration within the INDEN project. This paper deals with the nuclear data validation of that evaluation performed at the LR-0 reactor by means of critical experiments and measurement of reaction rates. Large amounts of silicon were used both as pure crystalline silicon and SiO₂ sand. The critical moderator level was measured for various core configurations. Reaction rates were determined in the largest core configuration. Simulations of the experimental setup were performed using the MCNP6.2 code. The obtained results show the improvement in silicon cross-sections in the INDEN evaluations compared to existing evaluations in major libraries. The new Thermal Scattering Law for SiO₂ published in ENDF/B-VIII.0 additionally reduces the discrepancy between calculation and experiments. However, an unphysical peak is visible in the neutron spectrum in SiO₂ obtained by calculation with the new Thermal Scattering Law.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.411-416
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• 2006

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the comer 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.