• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.277-286
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• 1989

Destructive methods are used for the turnup determination of an irradiated PWR fuel. One of the methods includes U, Pu, Nd-148 and Nd-(145+146) determination by an isotope dilution mass spectrometry using triple spikes (U-233, Pu-242 and Nd-150). The method involves two sequential ion exchange resin separation procedures. Pu is eluted from the first anion exchange resin column (Dowex AG 1$\times$8) with 12 M HCl-0.1 M HI mixed solution, followed by U elution with 0.1 M HCl. Nd is isolated from other fission products on the second anion exchange resin column (Dowex AG 1$\times$4) with a nitric acid-methanol eluent. Each fraction is analysed by thermal ionization mass spectrometry. The difference between Nd-148 and Nd-(145+146) method is found with an average 2.07%. The results are compared with those by the heavy element method using U and Pu isotopes and by the destructive y-spectrometric measurement of Cs-137. The dependences of isotope composition of U and Pu on burn-up, and correlation between those isotopes are illustrated graphically.

• Analytical Science and Technology
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• v.14 no.4
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• pp.291-299
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• 2001

A method of separation and purification of Sm for quantitation of Sm isotopes from various fission products in PWR spent nuclear fuels has been studied. Simulated solution containing inactive metal ions(Cs, Ba, Gd, Eu, Sm and Nd) in place of radioactive fission products was prepared. Sm was separated with 0.5 M $HNO_3$/80% MeOH after washing with 1 M $HNO_3$/90% MeOH on AG $1{\times}8$, anion exchange resin. Sm was purified on cation exchange resin, AG $50W{\times}8$, pretreated with 0.2 M alpha-hydroxisobutyric acid(pH 4.5-4.6) to remove Ba causing isobaric effect Sm from PWR spent fuel. As a result of mass spectrometric measurement, eluted Sm portion did not include isobars form other elements such as Gd, Eu, Pm, Nd and BaO. The contents of Sm and its isotopes($^{147}Sm$, $^{148}Sm$, $^{149}Sm$, $^{150}Sm$, $^{151}Sm$, $^{152}Sm$ and $^{154}Sm$) in spent fuel were determined by isotope dilution mass spectrometric method spiking $^{154}Sm$.

• Analytical Science and Technology
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• v.16 no.2
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• pp.117-124
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• 2003

Oxygen to metal ratio has been measured by wet and dry chemical analysis to study the properties of sintered $UO_2$ pellets and $U_3O_8$ in the lithium reduction process of spent pressurized water reactor fuels. Uranium dioxide pellets simulated for the spent PWR fuels with burnup values of 20,000~60,000 MWd/MtU were prepared by mixing $UO_2$ powder and oxides of fission product elements, pelleting the powder mixture and sintering it at $1,700^{\circ}C$ under a hydrogen atmosphere. For wet chemical analysis, the simulated spent fuels were dissolved with mixed acid (10 M HCl : 8 M $HNO_3$, 2.5 : 1, v/v) using acid digestion bomb technique. The total amount of uranium and fission products added in the simulated spent fuels were measured using inductively coupled plasma atomic emission spectrometry. Weight change of the simulated fuel during its oxydation was measured by thermogravimetry and then the O/M ratio result was compared to that obtained by wet chemical analysis. Influence of $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$, quaternary alloy, on the determination of O/M ratio was investigated.

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.9-14
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• 2009

High density polyethylene (HDPE) is degraded due to a radiation-induced oxidation when it is used as a neutron moderator in a neutron counter for a nuclear material accounting of spent fuels. The HDPE exposed to the gamma-ray emitted from the fission products in a spent nuclear fuel results in a radiation-induced degradation which changes its original molecular structure to others. So a neutron moderating power variation of HDPE, irradiated by a gamma radiation, was investigated in this work. Five HDPE moderator structures were exposed to the gamma radiation emitted from a $^{60}Co$ source to a level of $10^5-10^9$ rad to compare their post-irradiation properties. As a result of the neutron measurement test with 5 irradiated HDPE structures and a neutron measuring system, it was confirmed that the neutron moderating power for the $10^5$ rad irradiated HDPE moderator revealed the largest decrease when the un-irradiated pure one was used as a reference. It implies that a neutron moderating power variation of HDPE is not directly proportional to the integrated gamma dose rate. To clarify the cause of these changes, some techniques such as a FTIR, an element analysis and a densitometry were employed. As a result of these analyses, it was confirmed that the molecular structure of the gamma irradiated HDPEs had partially changed to others, and the contents of hydrogen and oxygen had varied during the process of a radiation-induced degradation. The mechanism of these changes cannot be explained in detail at present, and thus need further study.

• Nuclear Engineering and Technology
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• v.2 no.2
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• pp.67-72
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• 1970

The measurements of the fast neutron flux and its spectrum have been carried out by the threshold detectors in the central thimble of TRIGA Mark-II reactor operating at 250 KW. The following reactions have been employed for these measurements, viz : Ni$^{58}$ (n, p) Co$^{58}$ ； $Mg^{24}$ (n, p) Na$^{24}$ ； $Al^{27}$ (n, $\alpha$) Na$^{24}$ . From the activation data the fast neutron spectrum were calculated by CDC-3600 computer making use of two semi-empirical methods. It has been verified that the validity of assumption of a fission spectrum in the central thimble exists only above 1 to 2 Mev energy level. With this spectrum, a fast neutron flux in the range of 1 $\times$ 10$^{12}$ n/$\textrm{cm}^2$-sec above the energy of 2.6 Mev was observed in the central thimble of TRIGA MARK-II reactor.

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.2
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• pp.69-76
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• 2006

TRISO(tri-isotropic)-coated fuel particle technology is utilized owing to its higher stability at a high temperature and Its efficient retention capability for fission products In the HTGR(high temperature gas-reeled reactor). The typical spherical TRISO fuel panicle with a diameter of about 1mm is composed of a nuclear fuel kernel and outer coating layers. The outer coating layers consist of a buffer PyC(pyrolytic carbon) layer, Inner PyC(1-PyC) layer, SiC layer, and outer PyC(O-PyC) layer Most of the Inspection Items for the TRTSO-coated fuel particle depend on destructive methods. The coating thickness of the TRISO fuel particle can be nondestructively measured by the X-ray radiography without generating radioactive wastel. In this study, the coaling thickness for the simulated TRISO-coated fuel particle with $ZrO_2$ kernel Instead of $%UO_2$ kernel was measured by using micro-focus X-ray radiography with micro-focus X-ray generator and flat panel detector The radiographic image was also enhanced by image processing technique to acquire clear boundary lines between coating layers. The coaling thickness wat effectively measured by applying the micro-focus X-ray radiography The inspection process for the TRISO-coated fuel particles will be improved by the developed micro-focus X-ray radiography and digital image processing technology.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.327-336
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• 2020

The high-flux advanced neutron application reactor (HANARO) is a research reactor with thermal power of 30 MW applied in various research and development using neutrons generated from uranium fission chain reaction. A degasifier tank is installed in the ancillary facility of HANARO. This facility generates gas pollutants produced owing to internal environmental factors. The degasifier tank is designed to maintain the gas contaminants below acceptable levels and is monitored using an analyzer in the gas sampling panel. If condensate water is generated and flows into the analyzer of the gas sampling panel, corrosion occurs inside the analyzer's measurement chamber, which causes failure. Condensate water is generated because of the temperature difference between the degasifier tank and analyzer when the gas flows into the analyzer. A heating system is installed between the degasifier tank and gas sampling panel to suppress condensate water generation and effectively remove the condensate water inside the system. In this study, we investigated the efficiency of the heating system. In addition, the variations in the pipe temperature and the amount of average condensate water were modeled using a wall condensation model based on the changes in the fluid inlet temperature, outside air temperature, and heating cable-setting temperature.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.79-84
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• 2007

Inventories of soluble elements in the gap and grain boundaries of domestic used PWR fuel pellets were measured to estimate the quantities of radionuclides that are liable to be rapidly released into the groundwater of a disposal site. The gap inventory of cesium for the pellets in the used fuel with a burn-up range of 45 to 66 GWD/MTU showed 0.85 to 1.7% of its total inventory, which was close to 1/6 to 1/3 of the fission gas release fraction (FGRF). However, the amounts of cesium released from the gaps of the pellets below 40 GWD/MTU of a burn-up and less than 1% FGRF were so erratic that the gap inventory could not be defined by ie FGRF. Strontium inventories in the gap and grain boundaries of the pellets in the same rod were not significantly varied, and the iodine inventory in the gap of the used PWR fuels was estimated to be less than or the same as the FGRF.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.301-308
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• 2018

Anthropogenic radioactive noble gases formed by nuclear fission are significant indicators used to monitor the nuclear activity of neighboring countries. In particular, radioactive xenon, owing to its abundant generation and short half-life, can be used to detect nuclear testing, and radioactive krypton has been used as a tracer to monitor the reprocessing of nuclear fuels. Released radioactive noble gases are in the atmosphere at infinitesimal amounts due to their dilution in the air and their short half-life decay. Therefore, to obtain reliable and significant data when performing measurement of noble gases in the atmosphere, the minimum detectable activity (MDA) for noble gases should be defined as low as possible. In this study, the MDA values for radioactive xenon and krypton were theoretically obtained based on the BfS-IAR system by collecting both noble gases simultaneously. In addition, various MDA methods, confidence level and analysis conditions were suggested to reduce and optimize MDA with an assessment of the factors affecting MDA. The current investigation indicated that maximizing the pretreatment efficiency and performance maintenance of the counter were the most important aspects for Xe. In the case of Kr, since sample activities are much higher than those of Xe, it is possible to change the target MDA or to simplification of the analysis system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.191-197
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• 2017

The Korea Atomic Energy Research Institute (KAERI) has developed a safeguards technology for pyroprocessing based on the Safeguards-By-Design (SBD) concept. KAERI took part in a Member-State Support Program (MSSP) to establish a pyroprocessing safeguards approach. A Reference Engineering-scale Pyroprocessing Facility (REPF) concept was designed on which KAERI developed its safeguards system. Recently the REPF is being upgraded to the REPF+, a scaled-up facility. For assessment of the nuclear-material accountancy (NMA) system, KAERI has developed a simulation program named Pyroprocessing Material Flow and MUF Uncertainty Simulation (PYMUS). The PYMUS is currently being upgraded to include a Near-Real-Time Accountancy (NRTA) statistical analysis function. The Advanced Spent Fuel Conditioning Process Safeguards Neutron Counter (ASNC) has been updated as Non-Destructive Assay (NDA) equipment for input-material accountancy, and a Hybrid Induced-fission-based Pu-Accounting Instrument (HIPAI) has been developed for the NMA of uranium/transuranic (U/TRU) ingots. Currently, performance testing of Compton-suppressed Gamma-ray measurement, Laser-Induced Breakdown Spectroscopy (LIBS), and homogenization sampling are underway. These efforts will provide an essential basis for the realization of an advanced nuclear-fuel cycle in the ROK.