• 방사성폐기물학회지
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• 제2권2호
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• pp.97-104
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• 2004

NIPS 시스템은 중성자 핵반응 결과 방출되는 즉발 감마선을 정량적으로 측정하는 장치이며 고체 및 액체 폐기물 중 존재하는 다양한 원소를 비파괴적으로 분석할 수 있는 장점이 있다. 본 연구에서는 NIPS 시스템에 이용된 고순도반도체 검출기의 계측효율을 $^{l33}$Ba 및 $^{152}$Eu 방사성 동위원소 선원과 $^{35}$ Cl(n, ${\gamma}$)$^{36}$ Cl 핵반응 시 발생되는 즉발감마선을 이용하여 80 keV에서 8 MeV까지 넓은 영역에 대하여 구하였다. $^{35}$ Cl(n, ${\gamma}$)$^{36}$ Cl 핵반응을 이용한 고에너지 감마선의 계측효율은 즉발감마선의 방사능 값을 정확히 알 수 없기 때문에 저 에너지 영역에서 정확히 알고 있는 검출기 효율곡선에 규격화시켜 전 에너지 영역에서의 효율보정곡선을 구하였다. 또한 KCl 표준용액에 $^{252}$ Cf 중성자 선원을 조사시켜 표준용액으로부터 방출되는 즉발 감마선을 고순도반도체 검출기로 측정하고 광대역 계측효율 곡선을 이용하여 수용액 시료에서의 평균 열중성자 속을 예측하였다. NIPS 측정시스템은 주변 재료 물질의 핵반응으로 방출되는 감마선 background를 줄이기 위해 두 개의 고순도반도체 검출기를 이용한 동시계수 장치가 고안되었으며, 동시계수 모드에서의 계측효율도 함께 고려되었으며, 표준선원을 이용하여 전 계수 또는 동시계수모드에서의 background에 대한 측정감도를 비교하였다.다.

• Restorative Dentistry and Endodontics
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• 제13권1호
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• pp.185-190
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• 1988

The study was designed to establish quantitative method for assessing the marginal leakage of dental restorations. 18 Class V cavities with $45^{\circ}$ bevel joint were prepared and replicas of these teeth were made with polyethylene wax. and classified with three groups. First group was filled with Scotch bond and silux. Second group was filled with glass ionomer cement:scotchbond/silux. Third group was filled with Dentin-Adhesit/Heliosit. After finishing, all specimens were subjected manually to 100 thermal cycles at $0^{\circ}C$ and $100^{\circ}C$ Samarium nitrate solution, irradiated with flux of $6{\times}12^{12}$ neutrons/$cm^2$/sec for 11 hours, woled for 200 hours, counted with the HpGe detector and the tracer uptake was determined by comparison with a standard of samarium ($10{\mu}g$). The following results were obtained. 1) The group filled with glass ionomer cement base showed least marginal leakage. 2) The group filled with Dentin-Adhesit/Heliosit showed less marginal leakage than the group filled with scotchbond/silux.

• Restorative Dentistry and Endodontics
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• 제13권1호
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• pp.69-78
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• 1988

The study was designed to establish a more nearly quantitative method for assessing the marginal leakage of dental restorations. 27 Class V cavities with $45^{\circ}$ bevel joint were prepared and classified into 2 groups. One group was filled with Scotchbond and silux. The other group was filled with glass ionomer cement, Scotchbond and silux. After finishing, all specimens were subjected manually to 100 thermal cycles at $0^{\circ}C$ and $100^{\circ}C$ water-bath. They were soaked in a samarium nitrate solution for 3 hours, irradiated with flux of $6{\times}10^{12}$ neutrons/$cm^2$/sec for 11 hours, cooled for 200 hours, counted with the HPGE detector and the tracer uptake was determined by comparison with a standard of samarium (10 ${\mu}g$). The following results were obtained. 1. Both of the two groups showed a considerable amounts of marginal leakage. 2. The group filled without glass ionomer cement base showed more marginal leakage than the group filled with glass ionomer cement base. 3. Neutron Activation Analysis produced a good quantitative method to measure the marginal leakage and samarium was appropriate to measure the marginal leakage of resin restorations using neutron activation analysis.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2670-2689
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• 2022

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS - also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark.

• 방사성폐기물학회지
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• 제18권2_spc호
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• pp.327-336
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• 2020

HANARO (High-flux Advanced Neutron Application Reactor)는 우라늄의 핵분열 연쇄반응에서 생성된 중성자를 이용하여 다양한 연구개발을 수행하는 열출력 30 MW 규모의 연구용 원자로이다. 탈기탱크는 HANARO의 부속시설에 설치되어 있다. 탈기탱크는 내부환경요인으로 인해 기체오염물질을 발생시킨다. 탈기탱크는 기체오염물질을 허용 가능한 수준 이하로 유지하기위해 필요하며 기체시료채취판넬의 분석기에 의해 모니터링 된다. 응축수가 발생하여 기체시료채취판넬의 분석기 내부로 유입된다면, 분석기의 측정 챔버 내부에 부식이 발생하여 고장을 야기한다. 응축수의 생성 원인은 탈기탱크에 존재하는 기체가 분석기로 유입되는 과정에서 탈기탱크와 분석기사이 온도 차이다. 응축수 생성을 억제하고 계통 내부에 생성된 응축수를 효율적으로 제거하기 위해 탈기탱크와 기체시료채취판넬 사이에 히팅시스템이 설치되었다. 이 연구에서 우리는 히팅시스템의 효율성을 알고자 한다. 또한 Wall Condensation Model을 이용하여 유체 입구온도, 외부온도 및 히팅 케이블 설정온도 변화에 따른 파이프 온도와 평균응축량의 변화를 모델링하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.

• Journal of Radiation Protection and Research
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• 제45권3호
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• pp.130-141
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• 2020

Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd₂O₃, B₄C, polyethylene (PE), and borated (5 wt% ^natB) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B₄C and Gd₂O₃ were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1- or 0.4-cm-thick B₄C or Gd₂O₃ behind 10 cm-thick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cm-thick Gd₂O₃+10 cm-thick PE or 0.1 cm-thick B₄C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B₄C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.

• 천문학회지
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• 제54권1호
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• pp.1-8
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• 2021

We report X-ray timing and spectral properties of the pulsar PSR J0205+6449 measured using NuSTAR and Chandra observatories. We measure the pulsar's rotation frequency ν = 15.20102357(9) s-1 and its derivative $\dot{\nu}=-4.5(1){\times}10^{-11}\;s^{-2}$ during the observation period, and model the 2-30 keV on-pulse spectrum of the pulsar with a power law having a photon index Γpsr = 1.07 ± 0.16 and a 2-30 keV flux F2-30 keV = 7.3±0.6 × 10-13 erg cm-2 s-1. The Chandra 0.5-10 keV data are analyzed for an investigation of the pulsar's thermal emission properties. We use thermal and non-thermal emission models to fit the Chandra spectra and infer the surface temperature T∞ and luminosity Lth of the neutron star to be T∞ = 0.5 - 0.8 MK and Lth = 1 - 5 × 1032 erg s-1. This agrees with previous results which indicated that PSR J0205+6449 has a low surface temperature and luminosity for its age of 800-5600 yrs.

• Nuclear Engineering and Technology
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• 제27권6호
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• pp.811-824
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• 1995

CANDU 핵연료봉의 휨 열적 휨 멘트와 수력학적 견인력 및 기계적 하중에 기인하는 휨 모멘트에 의하여 일어난다. 여기서, 연료봉 휨은 연료봉 축방향 중심선으로부터의 측면 처짐으로 정의한다. 본 논문에서는 연료봉 축방향 중심선에 대한 비대칭 온도불포에 의해 핵연료 피복관 자체와 피복관과 소결체의 상호작용 부위에서 발생하는 열적 휨만을 취급한다. 이를 위해 1).소결체와 피복관사이의 기계적 상호작용을 무시한 조건에서의 핵연료 피복관의 휨과 2) 소결체와 피복관의 온도 변화에 기인하여 발생하는 소결체와 피복관 사이의 기계적 상호작용을 고려한 조건에서의 연료봉 휨을 혼합 고려하고, 각각에서 피복관의 비대칭 온도분포가 (i) 냉각재의 불완전한 혼합에 따른 비균질 냉각재 온도, (ii) 핵연료 피복관과 냉각재 사이의 비균질한 열전달 계수, (iii) 핵연료내 반경 방향으로의 중성자속 감쇄에 의한 비대칭 열 발생 등의 복합적효과에 의해 발생되는 것으로 고려하여 피복관의 대칭온도 분포까지 포함 할 수 있는 열적 휨의 일반적 해석 공식을 제시하였다. 본 휨 공식에 사용되는 모든 변수에 대한 민감도 분석을 통해, 핵연료봉 길이, 피복관 내경, 냉각재 평균 온도 및 변화 인자, 소결체 -피복관 기계적 상호 작용 인자, 중성자속 감쇄 인자, 핵연료 열팽창 계수, 피복관-냉각재 열전도 계수 등의 변화가 피복관 두께, 피복관-냉각재 열전달 계수, 피복관 열팽창 계수, 핵연료-피복관 열전달 계수 등의 변화보다 핵연료봉의 열적 휨에 상대적으로 더욱 영향을 미치는 것으로 밝혀졌다.

• Nuclear Engineering and Technology
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• 제56권10호
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• pp.4335-4354
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• 2024

Some core designs integrate high-enriched fuel and moderator materials to enhance neutron utilization. This combination results in a broad spectrum within the system, posing challenges in resonance calculation. This paper introduces a general framework to realize resonance self-shielding treatment in broad-spectrum fuel lattice problems. The framework consists of three components. First, a new energy group structure is devised to support resonance calculation in the entire energy range and capture spectral transition and thermalization effects during eigenvalue calculation. Second, the subgroup method based on narrow approximation is selected as a universal method to perform resonance calculation. Finally, transport equations for each fissionable region are solved for neutron flux to collapse the fission spectrum. The proposed method is verified against fast, intermediate, and thermal spectrum pin cell problems and an assembly problem featuring a fast-thermal coupled spectrum. Numerical results affirm the accuracy of the proposed method in handling these scenarios, with eigenvalue errors below 154 pcm for pin cell problems and 106 pcm for the assembly problem. The verification results revealed that the proposed method enables accurate resonance self-shielding treatment for broad-spectrum problems.