• Nuclear Engineering and Technology
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• 제40권1호
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• pp.9-20
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• 2008

Since plutonium accounts for 40-50% of the power produced by uranium fuels, spent fuel contains only residual plutonium. Management of this plutonium is one of the aspects influencing the choice of a fuel cycle back-end option: reprocessing, direct disposal or wait-and-see. Different grades and qualities of plutonium exist depending from their specific generation conditions; all are valuable fissile material. Safeguard authorities watch the inventories of civil plutonium, but access to those data is restricted. Independent evaluations have led to an estimated current inventory of 220t plutonium in total (spent fuel, separated civil plutonium and military plutonium). If used as MOX fuel, it would be sufficient to feed all the PWRs and BWRs worldwide during 7 years or to deploy a FBR park corresponding to 150% of today' s installed nuclear capacity worldwide, which could then be exploited for centuries with the current stockpile of depleted and spent uranium. The energy potential of plutonium deteriorates with storage time of spent fuel and of separated plutonium, due to the decay of $^{241}Pu$, the best fissile isotope, into americium, a neutron absorber. The loss of fissile value of plutonium is more pronounced for usage in LWRs than in FBR. However, keeping the current plutonium inventory for an expected future deployment of FBRs is counterproductive. Recycling plutonium reduce the required volume for final disposal in an underground repository and the cost of final disposal. However, the benefits of utilizing an energy resource and of reducing final disposal liabilities are not the only aspects that determine the choice of a back-end policy.

• 방사성폐기물학회지
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• 제19권4호
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• pp.447-457
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• 2021

Graphite Isotope Ratio Method (GIRM) can be used to estimate plutonium production in a graphite-moderated reactor. This study presents verification results for the GIRM combined with a 3-D polynomial regression function to estimate cumulative plutonium production in a graphite-moderated reactor. Using the 3-D Monte-Carlo method, verification was done by comparing the cumulative plutonium production with the GIRM. The GIRM can estimate plutonium production for specific sampling points using a function that is based on an isotope ratio of impurity elements. In this study, the ¹⁰B/¹¹B isotope ratio was chosen and calculated for sampling points. Then, 3-D polynomial regression was used to derive a function that represents a whole core cumulative plutonium production map. To verify the accuracy of the GIRM with polynomial regression, the reference value of plutonium production was calculated using a Monte-Carlo code, MCS, up to 4250 days of depletion. Moreover, the amount of plutonium produced in certain axial layers and fuel pins at 1250, 2250, and 3250 days of depletion was obtained and used for additional verification. As a result, the difference in the total cumulative plutonium production based on the MCS and GIRM results was found below 3.1% with regard to the root mean square (RMS) error.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.727-732
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• 2016

A simple and rapid spectrophotometric method has been developed to enable the determination of plutonium concentration in an irradiated fuel solution in the presence of all fission products. An excess of ceric ammonium nitrate solution was employed to oxidize all the valence states of plutonium to +6 oxidation state. Interference due to the presence of fission products such as ruthenium and zirconium, and corrosion products such as iron in the envisaged concentration range, as in the irradiated fuel solution, was studied in the determination of plutonium concentration by the direct spectrophotometric method. The stability of plutonium in +6 oxidation state was monitored under experimental conditions as a function of time. Results obtained are reproducible, and this method is applicable to radioactive samples resulting before the solvent extraction process during the reprocessing of fast reactor spent fuel. An analysis of the concentration of plutonium shows a relative standard deviation of <1.2% in standard as well as in simulated conditions. This reflects the fast reactor fuel composition with respect to uranium, plutonium, fission products such as ruthenium and zirconium, and corrosion products such as iron.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.510-517
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• 2019

Increasing proliferation resistance of plutonium by way of increased $^{238}Pu$ content is of interest to the nuclear nonproliferation and international safeguards community. Considering the high alpha decay heat of $^{238}Pu$, increasing the isotopic fraction leads to a noticeably higher amount of heat generation within the plutonium. High heat generation is especially unattractive in the scenario of weaponization. Upon weaponization of the plutonium, the plutonium may generate enough heat to elevate the temperature in the high explosives to above its self-explosion temperature, rendering the weapon useless. In addition, elevated temperatures will cause thermal expansion in the components of a nuclear explosive device that may produce thermal stresses high enough to produce failure in the materials, reducing the effectiveness of the weapon. Understanding the technical limit of $^{238}Pu$ required to reduce the possibility of weaponization is key to reducing the current limit on safeguarded plutonium (greater than 80 at. % $^{238}Pu$). The plutonium vector evaluated in this study was found by simulating public information on Lightbridge's fuel design for pressurized water reactors. This study explores the temperature profile and maximum stress within a simple (first generation design) hypothetical nuclear explosive device of four unique scenarios over time. Analyzing the transient development of both the temperature profile and maximum stress not only establishes a technical limit on the $^{238}Pu$ content, but also establishes a time limit for which each scenario would be useable.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.820-828
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• 2018

The growing nuclear threat has amplified the need for developing diverse and accurate nuclear forensics analysis techniques to strengthen nuclear security measures. The work presented here is part of a research effort focused on developing a methodology for reactor-type discrimination of weapons-grade plutonium. To verify the developed methodology, natural $UO_2$ fuel samples were irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) and produced approximately $20{\mu}g$ of weapons-grade plutonium test material. Radiation transport simulations of common thermal reactor types that can produce weapons-grade plutonium were performed, and the results are presented here. These simulations were needed to verify whether the plutonium produced in the natural $UO_2$ fuel samples during the experimental irradiation at MURR was a suitable representative to plutonium produced in common thermal reactor types. Also presented are comparisons of fission product and plutonium concentrations obtained from computational simulations of the experimental irradiation at MURR to the nondestructive and destructive measurements of the irradiated natural $UO_2$ fuel samples. Gamma spectroscopy measurements of radioactive fission products were mostly within 10%, mass spectroscopy measurements of the total plutonium mass were within 4%, and mass spectroscopy measurements of stable fission products were mostly within 5%.

• 대한화학회지
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• 제44권6호
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• pp.581-586
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• 2000

가압경수로형 사용후핵연료에 포함된 플루토늄을 분리하여 전기량 적정방벙으로 정량하였다. 플루토늄은 음이온 교환수지(AG MP-1)를 이용하여 분리하였으며 플루토늄의 정량은 조절전위전기량 적정방법을 이용하였다. 본 연구에서는 플루토늄 분리와 관련된 산화상태 조절, 분리 회수율 및 정량과 관련된 몇 가지 전기량 적정조건들에 대해 검토하였다. 실험결과 플루토늄은 산화상태를 조절하여 분리하는 것이 정량결과에 대한 정확도를 높였으며, 플루토늄의 전기량 적정결과에 미치는 몇 가지 금속이온들에 의한 영향은 나타나지 않았다. 플루토늄 표준용액을 이용하여 측정한 전기량 정량결과에 대한 정확도는 회수율로 나타냈을 경우 0.230∼3.02 mg 범위에서 99.36%이었다. 가압경수로형 사용후핵연료 용해용액들에 포함된 0.250∼0.450 mg의 플루토늄을 적정하여 얻은 전기량 정량결과에 대한 평균 정밀도(상대표준편차, RSD)는 0.38%이었다.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.384-393
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• 2019

An experimental validation of a nuclear forensics methodology for the source reactor-type discrimination of separated weapons-useable plutonium is presented. The methodology uses measured values of intra-element isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results. The experimental validation of the methodology was achieved by performing two low-burnup experimental irradiations, resulting in distinct fuel samples containing sub-milligram quantities of weapons-useable plutonium. The irradiated samples were subjected to gamma and mass spectrometry to measure several intra-element isotope ratios. For each reactor in the library, a maximum likelihood calculation was utilized to compare the measured and simulated intra-element isotope ratio values, producing a likelihood value which is proportional to the probability of observing the measured ratio values, given a particular reactor in the library. The measured intra-element isotope ratio values of both irradiated samples and its comparison with the simulation predictions using maximum likelihood analyses are presented. The analyses validate the nuclear forensics methodology developed.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2004-2010
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• 2022

Quantification of sensitive material is of vital importance when it comes to the movement of nuclear fuel throughout its life cycle. Within the electrorefiner vessel of electrochemical separation facilities, the task of quantifying plutonium by neutron analysis is especially challenging due to it being in a constant mixture with curium. It is for this reason that current neutron multiplicity methods would prove ineffective as a safeguards measure. An alternative means of plutonium verification is investigated that utilizes the (𝛼,n) signature that comes as a result of the eutectic salt within the electrorefiner. This is done by utilizing the multiplicity variable a and breaking it down into its constituent components: spontaneous fission neutrons and (𝛼,n) yield. From there, the (𝛼,n) signature is related to the plutonium content of the fuel.

• Mass Spectrometry Letters
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• 제2권4호
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• pp.80-83
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• 2011

Isotope ratio analysis of nuclear materials in individual particles is of great importance for nuclear safeguards. Although secondary ion mass spectrometry (SIMS) and thermal ionization mass spectrometry (TIMS) are utilized for the analysis of individual uranium particles, few studies were conducted for the analysis of individual uranium-plutonium mixed oxide particles. In this study, we applied SIMS and inductively coupled plasma mass spectrometry (ICP-MS) to the isotope ratio analysis of individual U-Pu mixed oxide particles. In the analysis of individual U-Pu particles prepared from mixed solution of uranium and plutonium standard reference materials, accurate $^{235}U/^{238}U$, $^{240}Pu/^{239}Pu$ and $^{242}Pu/^{239}Pu$ isotope ratios were obtained with both methods. However, accurate analysis of $^{241}Pu/^{239}Pu$ isotope ratio was impossible, due to the interference of the $^{241}Am$ peak to the $^{241}Pu$ peak. In addition, it was indicated that the interference of the $^{238}UH$ peak to the $^{239}Pu$ peak has a possibility to prevent accurate analysis of plutonium isotope ratios. These problems would be avoided by a combination of ICP-MS and chemical separation of uranium, plutonium and americium in individual U-Pu particles.

• 분석과학
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• 제14권1호
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• pp.28-33
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• 2001

자연수에 존재할 가능성이 있는 $Pu^{3+}$, $Pu^{4+}$, $PuO_2{^+}$, $PuO_2{^{2+}}$ 상태의 플루토늄을 이온크로마토그래피로 분리하는 방법을 연구하였다. 각각 $SiO^-$, SiO-$SO_3{^-}$작용기를 가진 이온교환체를 충전한 두 컬럼을 이용하는 방법과, Dionex사 AG11컬럼에 oxalate/$HNO_3$ 용리액을 사용하여 IC로 분리하는 두 가지 방법을 시도하였다. 예비실험으로서 플루토늄과 화학적 성질이 유사하며 산화수가 안정한 $Eu^{3+}$, $Th^{4+}$, $NpO_2{^+}$, $UO_2{^{2+}}$을 사용하여 분리조건을 얻었다. 이 분리조건을 플루토늄에 적용시켰을 때, $SiO^-$, SiO-$SO_3{^-}$ 이온교환체를 이용한 방법은 산화수별 플루토늄을 성공적으로 분리할 수 있었다. 그러나 IC를 사용한 방법은 oxalate 용액에서 $Pu^{3+}$가 쉽게 $Pu^{4+}$로 산화되며, $PuO_2{^{2+}}$도 $PuO_2{^+}$로 환원되어 분리가 곤란하였다.