• Nuclear Engineering and Technology
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• 제2권2호
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• pp.85-95
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• 1970

두개의 threshold detector로서 인자로의 폭발사고시에 방출되는 속 중성자의 속도분포를 측정하고 그로부터 속 중성자의 인체흡수선량을 계산하였다. 이때 속 중성자의 속도분포는 하나의 스펙트럼 매개변수에 의하여 결정된다는 가정으로부터 얻어지는데 이 매개변수는 threshold detector의 반응율을 측정하므로서 구해진다. 속 중성자의 인체흡수선량은 속 중성자의 속도분포 변화에 따라 큰 변동이 없었으나 threshold detector의 평균반응단면적은 크게 변하였다. 따라서 속 중성자의 속도분포에 관계없이 threshold detector의 평균반응단면적을 고정된 값으로 취하여 속 중성자선량을 계산한다면 큰 오차를 일으키게 될 것이라는 것을 보여주었다. 한편 핵분열에서 방출되는 속 중성자의 속도분포에 대한 세 해석적 표현인 즉 Watt, Cranberg및 Maxwellian 공식들로부터 속 중성자 선량을 계산하여 서로 비교하였다. Watt 및 Cranberg 공식들로 부터 얻어진 속 중성자선량은 Maxwellian 공식으로부터 얻어진 그것보다 약간 높은 값을 보여 주었으며 Watt 공식에 의한 선량계산치는 Cranberg 공식에 의한 그것과 비슷한 값을 보여주었다.

• 대한전기학회논문지:전기물성ㆍ응용부문C
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• 제52권12호
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• pp.545-552
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• 2003

When a Si PIN diode is exposed to fast neutrons, it results in displacement damage to the Si lattice structure of the diode. Defects induced from structural dislocation become effective recombination centers for carriers which pass through the base of a PIN diode. Hence, increasing the resistivity of the diode decreases the current for the applied forward voltage. This paper involves the development of a neutron sensor based on the phenomena of the displacement effect damaged by neutron exposure. The neutron effect on the semiconductor was analyzed. Several PIN diode arrays with various thickness and cross-section area of the intrinsic layer(I layer) were fabricated. Under irradiation tests with a neutron beam, the manufactured diodes have a good linearity to neutron dose and show that the increase of thickness of I layer and the decrease of cross-section of PIN diodes improve the sensitivity. Newly developed PIN diodes with thicker I layer and various cross section, were retested and then showed the best neutron sensitivity at the condition that the I layer thickness was similar to a side length. On the basis of two test results, final discrete PIN diodes with a rectangular shape were manufactured and the characteristics as neutron detectors were analyzed through the neutron beam test using on-line electronic dosimetry system. Developed PIN diode shows a good linearity as dosimetry in the range of 0 to 1,000cGy(Tissue) and its neutron sensitivity is 13mV/cGy at constant current of 5mA, that is three times higher than that of commercially available neutron detectors. And the device shows little dependency on the orientation of the neutron beam and a considerable stability in annealing test for a long period.

• 콘크리트학회논문집
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• 제28권2호
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• pp.197-204
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• 2016

본 논문에서는 타 재료에 비해 수소 원소 함유량이 높아 고속 중성자 차폐에 유리한 합성 고분자 화합물과 중성자 포획단면적이 큰 붕소화합물을 각각 혼입한 모르타르를 대상으로 중성자 차폐성능을 분석하였다. 합성 고분자 화합물의 종류, 형상, 크기, 함량 및 붕소화합물의 함량에 따라 총 16개 모르타르 배합을 설계하였으며, 각 배합의 슬럼프 플로우, 28일 압축 및 인장강도를 측정하고, 고속 중성자 및 열중성자에 대한 차폐실험을 수행하였다. 합성 고분자 화합물의 선량 투과율은 모르타르 대비 최대 38.5%까지 감소하였으며, 공기 중 선량의 26.3%까지 차폐하였다. 붕소화합물을 혼입한 모르타르의 열중성자 차폐율은 최대 90.3%로 대부분의 열중성자를 차폐하였다. 비록 화합물 혼입에 의해 모르타르의 기본 특성은 저하되었으나, 표면 개질, 특수 혼화제 첨가 등 지속적인 연구를 통하여 성능 저하를 최소화할 수 있을 것으로 판단되며, 중성자 투과성능 역시 다양한 타입의 시험체 조합을 통한 레이어 시스템 도입 등으로 다양한 투과성능에 따른 맞춤형 설계를 제공할 수 있을 것이다.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2002년도 Proceedings
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• pp.161-163
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• 2002

The BNCT(Boron Neutron Capture Therapy) facility has been developed in Hanaro(High-flux Advanced Neutron Application Reactor), a research reactor of Korea Atomic Energy Research Institute. A typical tangenial beam port is utilized with this BNCT facility. Thermal neutrons can be penetrated within the limits of the possible maximum instead of being filtered fast neutrons and gamma rays as much as possible using the silicon and bismuth single crystals. In addition to, the liquid nitrogen (LN$_2$) is used to cool down the silicon and bismuth single crystals for the increase of the penetrated thermal neutron flux. Neutron beams for BNCT are shielded using the water shutter. The water shutter was designed and manufactured not to interfere with any other subsystem of Hanaro when the BNCT facility is operated. Also, it is replaced with conventional beam port plug in order to cut off helium gas leakage in the beam port. A circular collimator, composed of $\^$6/Li$_2$CO$_3$ and polyethylene compounds, is installed at the irradiation position. The measured neutron flux with 24 MW reactor power using the Au-198 activation analysis method is 8.3${\times}$10$\^$8/ n/cm$^2$ s at the collimator, exit point of neutron beams. Flatness of neutron beams is proven to ${\pm}$ 6.8% at 97 mm collimator. According to the result of acceptance tests of the water shutter, the filling time of water is about 190 seconds and drainage time of it is about 270 seconds. The radiation leakages in the irradiation room are analyzed to near the background level for neutron and 12 mSv/hr in the maximum for gamma by using BF$_3$ proportional counter and GM counter respectively. Therefore, it is verified that the neutron beams from BNCT facility in Hanaro will be enough to utilize for the purpose of clinical and pre-clinical experiment.

• 분석과학
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• 제6권5호
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• pp.501-508
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• 1993

$^{241}Am-Be$(${\alpha}$, n) 중성자 선원의 중성자/감마선(n/${\gamma}$) 복합 방사선장에서 $^6Li$(n, ${\alpha}$)T 핵반응을 이용하고 두 섬광체, BC 501($C_8H_{10}$)과 Cerium의 섬광 감쇠 시간차와 동일 섬광체 내에서의 n/${\gamma}$에 대한 서로 다른 섬광감쇠 시간차를 병용하여 PSD(Pulse Shape Disciminator) 및 CFD(Constant Fraction Discriminator) 방법으로 n와 ${\gamma}$성분을 분리 측정하였으며 $^6Li$ 속중성자 분광계의 figure of merit는 1.36으로 평가되었다.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.313-342
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• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

• 한국자기학회지
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• 제15권1호
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• pp.12-16
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• 2005

본 연구에서는 비정질 시료 $Fe_{87}Zr_{7}B_{6}$ 에 중성자르 조사시킨 후 자기화의 온도의존성, X-ray 회절과 복소 투자율을 측정하여 중성자 조산 전, 후의 자기적 특성 변화를 보았으며, 다양한 중성자 조사량에 따른 복소 투자율과 자기이력곡선을 측정하여 측정하여 중성자 조사량에 따른 지구벽 운동과 자기화 회전을 보았다. 실험결과 중성자 조사에 의한 지구벽 운동은 감소하였으며, 자기화 회전은 증가 한 것을 볼 수 잇었다. 또한, 자기이완 주파수는 지구벽 운동의 경우 증가하였으며, 자기화 회전에 의한 이완 주파수는 감소하였다. 자기이력곡선의 결과 중성자 조사 후 포화 자기화 값이 감소하는 것을 볼 수 있었는데 이러한 실험 결과로부터 중성자 조사 후 생긴 결함에 의해 지구벽 운동은 억제되었으며, 자기화 회전부분은 증가 한 것을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.837-846
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• 2014

A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ~E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.

• Nuclear Engineering and Technology
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• 제34권1호
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• pp.42-59
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• 2002

In order to transmute long-lived radioactive nuclides such as transuranics(TRU), Tc-99, and I- l29 in LWR spent fuel, a preliminary conceptual design study has been performed for the accelerator driven subcritical reactor system, called HYPER(Hybrid Power Extraction Reactor) The core has a hybrid neutron energy spectrum: fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU is loaded into the HYPER core as a TRU-Zr metal form because a metal type fuel has very good compatibility with the pyre- chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb-Bi is chosen as a primary coolant because Pb-Bi can be a good spallation target and produce a very hard neutron energy spectrum. As a result, the HYPER system does not have any independent spallation target system. 9Cr-2WVTa is used as a window material because an advanced ferritic/martensitic steel is known to have a good performance under a highly corrosive and radiation environment. The support ratios of the HYPER system are about 4∼5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4 ∼5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is faced with.

• Corrosion Science and Technology
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• 제20권5호
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• pp.308-314
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• 2021

Neutron dose level at bottom head of a reactor pressure vessel (RPV) was calculated using reactor vessel neutron transport for a Korean nuclear power plant A. At 34 EFPY with a 40-year (2042) design life after plating repair, irradiation fast neutron effect was 6.6x10¹⁵ n/cm². As helium(He) gas can be generated by Ni only at 1/10⁶ level of 5 × 10²¹ n/cm², He generation possibility in the Ni plating layer is very little during 40 years of operation (2042, 34 EFPY). Thermal neutrons can significantly affect the generation of He from Ni metal. At 10 years after a repair, He can be generated at a level of about 0.06 appm, a level that can add general welding repair without any consideration. After 40 years of repair, 9.8 appm of He may be generated. Although this is a rather high value, it is within the range of 0.1 to 10 appm when welding repair can be applied. Clad repair by Ni electroplating technology is expected to greatly improve the operation efficiency by improving the safety and shortening the maintenance period of the nuclear power plant.