• 분석과학
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• 제14권4호
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• pp.1075-1081
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• 2001

기기중성자방사화 분석법은 핵 분석기술의 대표적인 방법으로 비파괴-동시 다원소분석의 장점과 함께 절대측정값에 의해 정량 할 수 있다는 특징을 갖고 있다. 최근에는 정확도와 편의성을 만족할 수 있는 $k_0$-정량법을 사용한 기기중성자방사화 분석법이 세계적으로 일반화되고 있다. 본 연구에서는 $k_0$-법의 적용을 위하여 이 방법의 전체적인 개념의 소개와 함께 $k_0$-파라미터를 측정하고자 하였다. 이를 위하여 $k_0$-법의 개념이해와 정량에 필요한 인자들의 정의 및 파라미터인 $Q_0$(${\alpha}$)와 f 값을 결정하기 위한 수식과 실험적 측정방법 등을 요약하였고 중성자조사공에 따라 특성 값을 갖는 ${\alpha}$와 f 값을 하나로 연구용원자로의 방사화분석용 조사공(NAA#1)에서 측정하여 $k_0$-법의 도입을 위한 기반을 마련하였다.

• Nuclear Engineering and Technology
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• 제24권2호
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• pp.163-167
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• 1992

고순도알루미늄중 불순물의 Parameter로 이용될수 있는 구리의 비파괴 방사화분석법의 고찰 및 23종의 극미량불순성분원소의 함량을 분석하였다. 즉 구리의 분석은 원자로의 속중성자에 의한 27Al(n,$\alpha$)24Na반응으로 생성되는 24Na의 방사능을 감소시키기 위하여 Thermal Column을 이용하였고 다른 조사공을 이용한 경우보다 약 100 배 정 도 방해 요인을 감소시킬 수 있었다. 24Na 에 의한 영향은 2-3 %범위 이하이었다. 이 방법에 의해 표준알루미늄(6 nine class)시료로부터 구리를 정량하였고 아울러 기타 불순원소들을 일상 방사화분석법에 의해 정량하였다. 구리의 함량은 0.54$\pm$0.08 ppm이었다. 이러한 결과는 문헌값과 비교할때 타당성이 있었고 일상분석에 이용할 수 있는 좋은 방법으로 여겨진다.

• 비파괴검사학회지
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• 제33권2호
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• pp.212-218
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• 2013

A transfer method of neutron radiography was developed to measure the size of the end plug and a gap of an intact K102L-2, the irradiated fuel of a ballooned K174L-3, a ballooned and ruptured K98L-3. A typical irradiation time of 25 min. was determined to obtain a film density of between 2 and 3 of SR X-ray film with neutrons of $1.5{\times}10^{11}n{\cdot}cm^{-2}$. To validate and calibrate the results, a RISO fuel standard sample, Cd plate and ASTM-BPI/SI were used. An activated latent image formed in the $100{\mu}m$ Dy foil was subsequently transferred in a dark room for more than 8 hours to the SR film which is a maximum of three half-lives. Due to the L/D ratio an unsharpness of $9.82-14{\mu}m$ and a magnification of 1.0003 were given. After digitizing an image of SR film, the ballooning gap of the plug was discernible by an H/V filter of image processing. The gap size of the ballooned element, K174L-3, is equal to or greater than 1.2 mm. The development of a transfer method played a pivotal role in developing high burn-up of Wolsung and PWR nuclear fuel type.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2007년도 동계학술발표대회 논문집
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• pp.310-315
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• 2007

The FTL has been developed to be able to irradiate test fuels at the irradiation hole(IR1 hole) by considering its utility and user's irradiation requirements. FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. For safety assurance IPS is designed to have dual stainless steel pressure vessel and OPS is composed of main cooling water system, emergency cooling water system, LMP(letdown, make-up, purification) system, etc. FTL Conceptual design was set up in 2001, basic design had completed including a design requirement, basic piping & instrument diagram (P&ID), and the detail design in 2004. In 2005, the development team carried out purchase and manufacture hardware and make a contract for construction work. FTL construction work began on August, 2006 and ended on March, 2007. After FTL development which is expected to be finished by 2008, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

• Journal of Magnetics
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• 제21권2호
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• pp.183-186
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• 2016

In this study, we report on an investigation of proton radiation resistance of 410 martensitic stainless steels under 3 MeV proton with the doses ranging from $1.0{\times}10^{15}$ to $1.0{\times}10^{17}p/cm^2$ at the temperature 623 K. Vibrating sample magnetometer (VSM) and X-ray diffractometer (XRD) were used to study the variation of magnetic properties and structural damages by virtue of proton irradiation, respectively. VSM and XRD analysis revealed that the 410 martensitic stainless steels showed proton radiation resistance up to $10^{17}p/cm^2$. Proton energy degradation and flux attenuations in 410 stainless steels as a function of penetration depth were calculated by using Stopping and Range of Ions in Matter (SRIM) code. It suggested that the 410 stainless steels have the radiation resistance up to $5.2{\times}10^{-3}$ dpa which corresponds to neutron irradiation of $3.5{\times}10^{18}n/cm^2$. These results could be used to predict the maintenance period of SUS410 stainless steels in fission power plants.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2723-2733
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• 2023

Transuranic nuclides (such as ²³⁸Pu, ²⁵²Cf, ²⁴⁹Bk, etc.) have a wide range of application in industry, medicine, agriculture, and other fields. However, due to the complex conversion chain and remarkable fission losses in the process of transuranic nuclides production, the generation amounts are extremely low. High flux reactor with high neutron flux and flexible irradiation channels, is regarded as the promising candidate for producing transuranic nuclides. It is of great significance to increase the conversion ratio of transuranic nuclides, resulting in higher efficiency and better economy. In this paper, we perform an optimization design evaluation of producing transuranic nuclides in high flux reactor, which includes optimization design of irradiation target and influence study of reactor core loading. It is demonstrated that the production rate increases with appropriately determined target material and target structure. The target loading scheme in the irradiation channel also has a significant influence on the production of transuranic nuclides.

• 대한방사선기술학회지:방사선기술과학
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• 제26권3호
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• pp.51-57
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• 2003

자체 개발한 수소화금속을 이용하여 고속 중성자 방사선을 효율적으로 차폐할 수 있다면 방사선 안전신기술 개발과 확립에 큰 기여를 할 것으로 생각되어 본 연구를 시행하였다. 여러 수소화 안정 금속들을 대상으로 핵적 특성, 단위 부피당 수소원자함유 수 등의 예비평가를 통하여 수소화금속($ZrH_2,\;TiH_2$) 등과 낮은 중성자 흡수 단면적과 높은 에너지 감쇄능력을 고려하여 중수소화 금속($ZrD_2,\;TiD_2$) 등을 추가하여 개발하였다. MCNP 코드를 이용하여 각각의 흡수율과 에너지 감소율을 평가하였다. 전산 모사 계산과 실험과의 비교평가를 위해 실험과 동일한 조건의 모사를 수행하였는데, 즉 중성자 선원은 Cf-252(10 mCi)을 사용하였으며 각 수소화금속의 0, 1, 3, 5 cm 두께를 통과한 중성자속의 강도와 에너지별 분포변화를 계산하였다. 코드 계산을 통해 평가된 $TiH_2/TiD_2,\;ZrH_2,/ZrD_2$ 등의 수소화금속에 대한 중성자 감소율은 각 수소화금속 두께의 증가에 따라 중성자 감소율이 지수적으로 증가함을 보였다. 또한 이 때 중수소 함유 금속, $ZrD_2$과 $TiD_2$는 중성자 흡수에 있어 $ZrH_2$와 $TiH_2$의 각각 보다 적게 나타냈다. 본 연구를 통하여 개발된 수소화금속의 중성자 방사선 차폐에 관한 결과는 과학 기술적으로 많은 인용과 아울러 학술적 연구뿐만 아니라 실제 실용화를 위한 연구의 기초자료로 충분한 활용이 있을 것으로 기대한다.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.469-476
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• 2020

Research reactors in-core experimental facilities are designed to provide the highest steady state flux for user's irradiation requirements. However, fuel conversion from highly enriched uranium (HEU) to low enriched uranium (LEU) driven by the ongoing effort to diminish proliferation risk, will impact reactor physics parameters. Preserving the reactor capability to produce the needed flux to perform its intended research functions, determines the conversion feasibility. This study investigates the neutron flux in the central experimental facility of two material test reactors (MTR), the IAEA generic10 MW benchmark reactor and the 22 MW s Egyptian Test and Research Reactor (ETRR-2). A 3D full core model with three uranium enrichment of 93%, 45%, and 20% was constructed utilizing the OpenMC particle transport Monte Carlo code. Neutronics calculations were performed for fresh fuel, the beginning of life cycle (BOL) and end of life cycle (EOL) for each of the three enrichments for both the IAEA 10 MW generic reactor and core 1/98 of the ETRR-2 reactor. Criticality calculations of the effective multiplication factor (K_eff) were executed for each of the twelve cases; results show a reasonable agreement with published benchmark values for both reactors. The thermal, epithermal and fast neutron fluxes were tallied across the core, utilizing the mesh tally capability of the code and are presented here. The axial flux in the central experimental facility was tallied at 1 cm intervals, for each of the cases; results for IAEA 10 MW show a maximum reduction of 14.32% in the thermal flux of LEU to that of the HEU, at EOL. The reduction of the thermal flux for fresh fuel was between 5.81% and 9.62%, with an average drop of 8.1%. At the BOL the thermal flux showed a larger reduction range of 6.92%-13.58% with an average drop of 10.73%. Furthermore, the fission reaction rate was calculated, results showed an increase in the peak fission rate of the LEU case compared to the HEU case. Results for the ETRR-2 reactor show an average increase of 62.31% in the thermal flux of LEU to that of the HEU due to the effect of spectrum hardening. The fission rate density increased with enrichment, resulting in 34% maximum increase in the HEU case compared to the LEU case at the assemblies surrounding the flux trap.

• 분석과학
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• 제16권5호
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• pp.391-398
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• 2003

하나로의 즉발감마선 방사화분석 장치를 이용하여 생물시료중의 붕소의 정량을 위한 기초연구를 수행하였다. 측정조건에 대한 특성조사를 위해 시료에 대한 중성자 조사 위치에서 중성자속 및 균질도를 측정하였다. 시료위치에서 열중성자 빔의 크기가 $2{\times}2cm^2$ 되도록 집속하였으며, 측정된 선속은 $1.0{\sim}6.5{\times}10^7n{\cdot}cm^{-2}{\cdot}s^{-1}$ 범위를 나타냈으며, 중심부로부터 반경 4.5 mm 이내 및 9 mm 이내에서 각각 $5.77{\pm}0.71{\times}10^7n{\cdot}cm^{-2}{\cdot}s^{-1}$, $4.68{\pm}1.64{\times}10^7n{\cdot}cm^{-2}{\cdot}s^{-1}$이었다. 따라서 양질의 균일한 조사를 위해서 시료의 크기를 10 mm 이내로 조정하였다. 검출 시스템은 컴프턴 산란에 의한 백그라운드 요인을 억제하고 분석감도를 높이기 위해 설계되었으며, 감마선 계측 시스템의 에너지 교정과 컴프턴 억제율을 조사하기 위해 NaCl 표준체를 이용하여 단일 및 컴프턴 모드로 백그라운드를 측정하였다. 또한 정확한 붕소의 측정을 위해 시료의 매질효과로서 발생하는 분광학적 Na의 472 keV 피이크에 대한 간섭효과를 결정하였으며, 세 가지 인증표준물질 (NIST SRM 1570a, 1547, 1573a)을 이용한 붕소농도 측정시험을 두 가지 모드로 실시한 후 결과를 비교하였다.