• Proceedings of the Korean Vacuum Society Conference
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• 2016.02a
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• pp.425.1-425.1
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• 2016

고온의 플라즈마를 긴 펄스 및 장시간 연속운전 유지기술 개발 및 연구를 위해서는 플라즈마는 더욱 가열되어야 하고, 고온 고밀도의 플즈마 상태를 유지시켜야 한다. 이러한 고성능 플라즈마 개발은 향후 핵융합 에너지의 상용화를 위한 절대필수적 기반기술이다. 현재 KSTAR 토카막에서는 플라즈마를 가열하기 위한 장치들 중 하나로서, 출력 6 MW 급의 중성입자빔을 입사하는 NBI (Neutral Beam Injection) 가열장치가 설치 운영 중에 있다. 이 NBI 가열장치는 진공환경에서 고온, 고압, 고전압 방전 및 수냉 등이 작동 및 운전되고 있기 때문에, 구성 부품 들의 미세한 구조적 결함에도 장치의 치명적 failed로 이어질 수 있다. 이번 연구에서는 NBI 가열장치의 특성상 극한 운전 환경에 있는 진공용기 부품 중 하나 인 빔인출을 위한 가속 그리드 (accelerating grid)의 구조적 손상및 결함 여부를 고속중성자 이미지 기법을 적용하여 내부를 투시 진단하였다. 가속 그리드는 copper로 제작되었고, 빔인출을 위한 원형의 구멍과 냉각관을 가진 평면판 형태로 되었다. 본 연구에서 내부투시 및 진단할 수 있는 고속중성자 이미징 기법의 적용으로 진공용기 부품 및 장치의 구조적 결함 및 손상 여부를 판단 가능하다는 연구 결과를 얻었다.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.6
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• pp.693-700
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• 2004

The probabilistic fracture mechanics(PFM) is a useful analytical tool to assess the integrity of reactor pressure vessel(RPV) at the event of pressurized thermal shock(PTS). In PFM, the probabilities of flaw initiation and propagation are estimated by comparing the applied stress intensity factor with the fracture toughness calculated by the simulation of various stochastic variables. It is known that the results of PFM analyses are dependent on the choice of the stochastic parameters and assumptions. Of the various variables and assumptions, we investigated the effects of the RT$_{NDT}$ shift equations, fracture toughness curves, and flaw distributions on the PFM results for the three PTS transients. The results showed that the combined effects of the RT$_{NDT}$ shift equations and fracture toughness curves are complicated and dependent on the characteristics of the transients, the chemistry of the materials, the fast neutron fluence, and so on.

• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.355-362
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• 2012

The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.165-169
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• 2018

An assessment for determining N-16 activity concentrations during the operation condition of Bangladesh Atomic Energy Commission TRIGA Research Reactor was performed employing several governing equations. The radionuclide N-16 is a high energy (6.13 MeV) gamma emitter which is predominately created by the fast neutron interaction with O-16 present in the reactor core water. During reactor operation at different power level, the concentration of N-16 at the reactor bay region may increase causing radiation risk to the reactor operating personnel or the general public. Concerning the safety of the research reactor, the present study deals with the estimation of N-16 activity concentrations in the regions of reactor core, reactor tank, and reactor bay at different reactor power levels under natural convection cooling mode. The estimated N-16 activity concentration values with 500 kW reactor power at the reactor core region was $7.40{\times}10^5Bq/cm^3$ and at the bay region was $3.39{\times}10^5Bq/cm^3$. At 3 MW reactor power with active forced convection cooling mode, the N-16 activity concentration in the decay tank exit water was also determined, and the value was $4.14{\times}10^{-1}Bq/cm^3$.

• Journal of computational fluids engineering
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• v.16 no.3
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• pp.95-103
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• 2011

In the block type VHTR core, there are inevitable gaps among core blocks for the installation and refueling of the fuel blocks. These gaps are called bypass gap and the bypass flow is defined as a coolant flows through the bypass gap. Distribution of core bypass flow varies according to the reactor operation since the graphite core blocks are deformed by the fast neutron irradiation and thermal expansion. Furthermore, the cross-flow through an interfacial gap between the stacked blocks causes flow mixing between the coolant holes and bypass gap, so that complicated flow distribution occurs in the core. Since the bypass flow affects core thermal margin and reactor efficiency, accurate prediction and evaluation of the core bypass flow are very important. In this regard, experimental and computational studies were carried out to evaluate the core bypass flow distribution. A multi-block experimental apparatus was constructed to measure flow and pressure distribution. Multi-block effect such as cross flow phenomenon was investigated in the experiment. The experimental data were used to validate a CFD model foranalysis of bypass flow characteristics in detail.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.189-199
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• 2016

Fast neutrons with a broad energy spectrum, with which it is possible to evaluate nuclear data for various research fields such as medical applications and the development of fusion reactors, can be generated by irradiating proton beams on target materials such as beryllium. To generate short-pulse proton beam, we adopted a deflector and slit system. In a simple deflector with slit system, most of the proton beam is blocked by the slit, especially when the beam pulse width is short. Therefore, the available beam current is very low, which results in low neutron flux. In this study, we proposed beam modulation using a buncher cavity to increase the available beam current. The ideal field pattern for the buncher cavity is sawtooth. To make the field pattern similar to a sawtooth waveform, a multiharmonic buncher was adopted. The design process for the multiharmonic buncher includes a beam dynamics calculation and three-dimensional electromagnetic simulation. In addition to the system design for pulsed proton generation, a test bench with a microwave ion source is under preparation to test the performance of the system. The design study results concerning the pulsed proton beam generation and the test bench preparation with some preliminary test results are presented in this paper.

• KOREAN JOURNAL OF CROP SCIENCE
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• v.10
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• pp.73-77
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• 1971

The experiment was carried out to investigate post-irradiation storage effect which was related to temperature(i.e. at 2$^{\circ}C$, 17$^{\circ}C$ and 4$0^{\circ}C$) on wheat seeds; Weibull's Svenno, treated with gamma-ray and fast neutron. Results obtained showed that the seedling height in both radiation sources was decreased with prolongation of storage period, especially when the seeds were treated with high dosage and stored at high temperature(4$0^{\circ}C$). The results of this trial, however, showed that storage effect was influenced by irradiation dose, temperature and storage time.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4209-4214
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• 2022

Precise modelling of the interaction of ions with materials is important for many applications including material characterization, ion implantation in devices, thermonuclear fusion, hadron therapy, secondary particle production (e.g. neutron), etc. In this study, a new approach using the Geant4 toolkit in combination with the Bayesian regularization (BR) learning algorithm of the feed-forward neural network (FFNN) is developed to estimate the range of ions in materials accurately and quickly. The different incident ions at different energies are interacted with the target materials. The Geant4 is utilized to model the interactions and to calculate the range of the ions. Afterward, the appropriate architecture of the FFNN-BR with the relevant input features is utilized to learn the modelled ranges and to estimate the new ranges for the new cases. The notable achievements of the proposed approach are: 1- The range of ions in different materials is given as quickly as possible and the time required for estimating the ranges can be neglected (i.e. less than 0.01 s by a typical personal computer). 2- The proposed approach can generalize its ability for estimating the new untrained cases. 3- There is no need for a pre-made lookup table for the estimation of the range values.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.195-202
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• 2018

As a rule, geological disposal is considered a safe method for final disposal of high-level radioactive waste. However, some long-lived fission products like $^{99}Tc$ and $^{129}I$ contained in spent nuclear fuel are highly mobile as less sorbing anionic species in the subsurface environment and can mainly cause exposure dose to the ecosystem by emission of beta rays in the hundreds of keV range. Therefore, if these two nuclides can be separated and converted with high efficiency into radioactively unharmful nuclides, this would have a positive effect on disposal safety. One candidate method is to transmute these two nuclides in nuclear reactors into short-lived nuclides or into stable nuclides. For this purpose, it is necessary to evaluate which reactor type is more efficient in burning these two nuclides. In this study, the simulation results of nuclear transmutation of $^{99}Tc$ and $^{129}I$ in light water reactor (PWR), heavy water reactor (CANDU) and fast neutron reactor (SFR, MET-1000) are compared and discussed.