• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.52-57
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• 1996

The coolant void reactivity is positive for CANDU reactor loaded with DUPIC fuel which has more fissile content compared to natural uranium. A parametric study was done to reduce the void reactivity of the fuel bundle and loss in discharge burnup was estimated. It is observed that the burnable absorbers like gadolinium, boron, europium are not able to keep the reduction in void reactivity uniform throughout fuel burnup. Dysprosium and erbium can keep the void reactivity reduction uniform throughout. fuel burnup but toss in discharge burnup for erbium case is more compared to that of dysprosium case.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.189-200
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• 2016

The Republic of Tajikistan has ten former uranium mining sites. The total volume of all tailings is approximately 55 million tonnes, and the covered area is more than 200 hectares. The safe management of legacy uranium mining and tailing sites has become an issue of concern. Depending on the performance requirements and site-specific conditions (location in an arid, semiarid or humid region), a cover system for uranium tailings sites could be constructed using several material layers using both natural and man-made materials. The purpose of this study is to find a feasible cost-effective cover system design for the Degmay uranium tailings site which could provide a long period (100 years) of protection. The HELP computer code was used in the evaluation of potential Degmay cover system designs. As a result of this study, a cover system with 70 cm thick percolation layer, 30 cm thick drainage layer, geomembrane liner and 60 cm thick barrier soil layer is recommended because it minimizes cover thickness and would be the most cost-effective design.

• Analytical Science and Technology
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• v.16 no.6
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• pp.450-457
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• 2003

The determination of uranium and its isotopes in spent nuclear fuels by isotope dilution mass spectrometry (IDMS) has been studied. The spent fuel samples were dissolved in 8 M $HNO_3$ or its mixture with 14 M $HNO_3-0.05M$ HF. The dissolved solutions were filterred on membrane filter with $1.2{\mu}m$ pore size. The uraniums in the spiked and unspiked sample solutions were quantitatively adsorbed by anion exchange resin, AG 1X8 and eluted with 0.1 M HCl. The contents of uranium and its isotopes ($^{234}U$, $^{235}U$, $^{236}U$ 및 $^{238}U$) in the spent fuel samples were determined by isotope dilution mass spectrometric method using $^{233}U$ as spike. The spike reference solution was standarized by reverse isotope dilution mass spectrometry (R-IDMS) using natural and depleted uranium. The results from IDMS were in average relative difference of 0.34% when compared with those by the potentiometric titration method.

• Journal of the Korean Society of Radiology
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• v.15 no.1
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• pp.37-43
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• 2021

Republic of Korea has many nuclear facilities in the country, and Democratic People's Republic of Korea(North Korea) locates in the surrounding country. Therefore, it is necessary to construct the target facility's nuclear forensic data in a preemptive response to the changing international situation. For this reason, this study suggests "signature" materials used to understand the origins and sources of nuclear and other radioactive materials, taking into account domestic nuclear facilities and the nuclear fuel cycle. In domestic, pressurized light water reactors and pressurized heavy water reactors are in operation, and enriched and natural uranium are used as fuels. In the front-end fuel cycle, the signature materials can be nature uranium and UF6 in the uranium enrichment process. The domestic back-end fuel cycle adopts a non-circulating cycle excluding the reprocessing process, and the primary signature material is spent nuclear fuel. According to IAEA recommendation, the importance of these materials as the signature and characteristic contents are suggested in this study. To prove the integrity of nuclear material and build a national nuclear forensics library, it is necessary to grasp the signature material and acquire the characteristic data considering the domestic nuclear facilities and the nuclear fuel cycle.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.1054-1059
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• 1995

국제원자력기구에서는 현재 적용되고 있는 보장조치(Safeguards) 방법을 보다 효과적이고 효율적으로 적용하기 위하여 1993년부터 'Program 93+2'라는 사업을 수행하고 있다. 이중 하나의 과제로 수행되고 있는 구역 보장조치는 기존의 보장조치 개념이 하나의 시설을 대상(Facility-Oriented Safeguards)으로 개발된 것과는 달리 동일한 범주의 핵물질을 취급하는 여러 개의 시설을 하나의 가상적인 구역(Fuel Cycle-Oriented Safeguards)으로 설정하여 보장조치를 적용하는 개념으로, 보다 강화된 사찰 활동에 의하여 보장조치 신뢰도를 향상시키면서 사찰 횟수 및 사찰량은 절감되고 있다. 우리나라는 한국원자력연구소의 중수로핵연료 가공시설과 월성 1호기를 천연우라늄 구역(Natural Uranium Zone)으로, 한국원전연료(주)의 경수로핵연료 가공시설과 국내의 모든 경수로를 저농축우라늄 구역(Low Enriched Uranium Zone)으로 설정하여 성공적으로 구역 보장조치를 적용하고 있다. 그러나 이러한 구역 보장조치의 적용에는 원자력산업 체제의 단순화와 같은 제약조건이 따른다. 앞으로 우리나라에서는 현재 적용되고 있는 구역 보장조치 방법이 보다 효율적으로 운영되고 시설 운영에 대한 방해를 최소화시키는 방안을 고려하여야 하며 이에 는 가공시설에서의 생산 및 수송 일정을 발전소 운영 및 사찰 일정과 적절히 조화시키는 방법, 가공시설에서 검증된 핵연료에 대하여 적절한 감시 및 봉인 장비를 적용하는 방법, 현재의 구역 이외의 시설 또는 핵물질에 새로운 구역을 설정, 적용하는 방안 등을 고려할 수 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.257-262
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• 1996

The characteristics of adjuster rods have been studied for the application to DUPIC core in two aspects: the half an hour xenon override capability and power flattening. The transient analysis has shown that the adjusters used for CANDU 6 have the reactivity worths more than required to override xenon load for DUPIC core. Parametric study has shown that removing 7 adjuster rods in the middle row and adjusting the strength of the rest of adjuster rods can provide the performances no worse than those of natural uranium core.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.651-656
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• 1997

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been peformed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of the total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.101-116
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• 2017

The interaction of U(VI) (hexavalent uranium) species with natural organic matter (NOM) in KURT (KAERI Underground Research Tunnel) groundwater is investigated using a laser spectroscopic technique. The luminescence spectra of the NOM are observed in the ultraviolet and blue wavelength regions by irradiating a laser beam at 266 nm in groundwater. The luminescence spectra of U(VI) species in groundwater containing uranium concentrations of $0.034-0.788mg{\cdot}L^{-1}$ are measured in the green-colored wavelength region. The luminescence characteristics (peak wavelengths and lifetime) of U(VI) in the groundwater agree well with those of $Ca_2UO_2(CO_3)_3(aq)$ in a standard solution prepared in a laboratory. The luminescence intensities of U(VI) in the groundwater are weaker than those of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution at the same uranium concentrations. The luminescence intensities of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution mixed with the groundwater are also weaker than those of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution at the same uranium concentrations. These results can be ascribed to calcium-U(VI)-carbonate species interacting with NOM and forming non-radiative U(VI) complexes in groundwater.

• Nuclear Engineering and Technology
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• v.9 no.3
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• pp.125-137
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• 1977

The temperature coefficient has been investigated on the Wolsung nuclear power reactor, in which fuel is natural uranium dioxide and moderator heavy water. The numerical computations are carried out in terms of changes of the effective neutron multiplication factor with respect to fuel, moderator, and coolant temperatures. Those results are compared with the computed values of temperature coefficient based on the LATREP computer code.

• 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.