• Nuclear Engineering and Technology
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• 제50권1호
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• pp.182-189
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• 2018

A neutron generation system was developed to induce fissile fission in a lead slowing down spectrometer (LSDS) system. The source neutron is one of the key factors for LSDS system work. The LSDS was developed to quantify the isotopic contents of fissile materials in spent nuclear fuel and recycled fuel. The source neutron is produced at a multilayered target by the (e,${\gamma}$)(${\gamma}$,n) reaction and slowed down at the lead medium. Activation analysis of the target materials is necessary to estimate the lifetime, durability, and safety of the target system. The CINDER90 code was used for the activation analysis, and it can involve three-dimensional geometry, position dependent neutron flux, and multigroup cross-section libraries. Several sensitivity calculations for a metal target with different geometries, materials, and coolants were done to achieve a high neutron generation rate and a low activation characteristic. Based on the results of the activation analysis, tantalum was chosen as a target material due to its better activation characteristics, and helium gas was suggested as a coolant. In addition, activation in a lead medium was performed. After a distance of 55 cm from the lead surface to the neutron incidence, the neutron intensity dramatically decreased; this result indicates very low activation.

• 한국방사선학회논문지
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• 제4권4호
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• pp.29-32
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• 2010

본 연구는 교토대학 원자로실험소의 46-MeV전자선형가속기에서 발생된 전자선을 Ta표적에 충돌시켜 발생된 중성자를 이용하였다. 발생된 중성자는 납감속측정장치(LSDS:Lead Slowing-down Spectrometer))속에서 감속되었고 중성자비행시간법(TOF:Time-of-Flight)을 이용하여서 중성자에너지를 선별하여 $^{197}Au$의 중성자포획단면적을 중성자에너지 0.1 eV에서 10 keV범위에 걸쳐 측정하였다. 발생된 중성자속을 측정하기 위하여 $BF_3$검출기속의 $^{10}B(n,{\gamma})$반응을 이용하였고 이것을 이용하여 중성자 반응 단면적을 상대적으로 얻었다. TOF방법으로 얻어진 결과는 1 eV에서의 결과(24.5 b)에 규격화되었다. 기존의 실험결과들과 평가결과들인 JENDL/D-99 Dosimetry File과 비교하였다.

• 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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• 제49권3호
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• pp.549-555
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• 2017

A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2~3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.921-928
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• 2013

A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

• 한국콘텐츠학회:학술대회논문집
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• 한국콘텐츠학회 2004년도 추계 종합학술대회 논문집
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• pp.495-500
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• 2004

The neutron capture cross section of $^{99}Tc$ has been measured relative to the $^{10}B$(n,g) standard cross section by the neutron time-of-flight(TOF) method in the energy range of 0.007 eV to 47keV using a 46-MeV electron linear accelerator(linac) at the Research Reactor Institute, Kyoto University(KURRI). In order to experimentally prove the result obtained, the supplementary cross section measurement has been made from 0.3 eV to 1 keV using the Kyoto University Lead slowing-down Spectrometer(KULS) coupling to the linac. The relative measurement by the TOF method has been normalized to the reference value(20.01 b) at 0.0253 eV and the KULS measurement to that by the TOF method. The existing experimental data and the evaluated capture cross sections in ENDF/B-VI, JENDL-3.2, and JEF-2.2 have been compared with the current measurements by the linac TOF and the KULS experiments. The energy dependency of the KULS data is close to that of the TOF data which are energy-broadened by the resolution function of the KULS.