• Applied Chemistry for Engineering
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• v.17 no.6
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• pp.604-608
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• 2006

Polymer composites for measuring the radioactive contamination are prepared by coating ZnS(Ag) powders as a scintillator on polysulfone base layer. The composites consist of the active layer for a scintillation reaction with radioactive wastes and the transparent support layer for transmittance of light photons emitted by scintillation in the active layer. The binding of the active layer, including ZnS(Ag), on the support layer is proceeded via coating with polysulfone as a binder, without any extra adhesive. The coating was obtained by either casting via a Doctor Blade as applicator or screen printing. The prepared composites feature a monolithic structure, resulting in the complete adhesion between two layers. The composite prepared by the casting technique using an applicator holds a good detection efficiency in measuring the alpha radionuclide, but its structure becomes fragile because of warping in morphology. On the contrary, the composite prepared by the screen printing shows a good detection capacity as well as a good stability in a mechanical shape.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.3
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• pp.193-198
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• 2013

In general, boron recovery of 40-90% could be achieved by Reverse Osmosis (RO) membranes in neutral pH condition. As an emerging technology, Forward Osmosis (FO) membrane has attracted growing interest in wastewater treatment and desalination. The objective of this study is to evaluate the possibility of the boron removal in radioactive liquid waste by FO. In this study, the performance of FO was investigated to remove boron in the simulated liquid waste as the factors such as pH, osmotic pressure, ionic strength of solution, etc. The pH of feed solution is a major operating parameter which strongly influences to the permeation of boron and more than 80% of boron content can be separated when conducted at pH values less than 7. The water flux is not influenced but the boron flux and permeation rate tends to decrease in the low salt concentration of 1,000 mg/L. The boron flux increases linearly, but the permeation ratio of reducing boron is nearly constant even with changes in the draw solution concentration.

• Tunnel and Underground Space
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• v.12 no.1
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• pp.52-69
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• 2002

The WIPP (Waste Isolation Pilot Plant), located 42km east of Carlsbad, New Mexico (NM), in bedded salt 655m below the surface, is a mined repository constructed by the US DOE for the permanent disposal of transuranic (TRU) wastes generated by activities related to defence of the US since 1970. Its historical disposal operation began in March 1999 following receipt of a final permit from the State of NM after a positive certification decision for the WIPP was issued by the EPA in 1998, as the first licensed facility in the US for the deep geologic disposal of radioactive wastes. The CCA (Compliance Certification Application) for the WIPP that the DOE submitted to the EPA in 1966 was supported by an extensive performance assessment (PA) carried out by Sandia National Laboratories (SNL), with so-called 1996 PA. Even though such PA methodologies could be greatly different from the way we consider for HLW disposal in Korea largely due to quite different geologic formations in which repository are likely to be located, a review on lots of works done through the WIPP PA studies could be the most important lessons that we can learn from in view of current situation in Korea where an initial phase of conceptual studies on HLW disposal has been just started. The objective of this art report is an overview of the methodology used in the recent WIPP PA to support the US DOE WIPP CCA and some relevant results completed by SNL.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.21-34
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• 2015

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.767-774
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• 1995

Behavior of radionuclides such $^{60}$ Co, $^{54}$ Mn and $^{137}$ Cs in the incineration Process was Studied by trial burns of simulated wastes with radio-isotope tracers. Behavior of nonvolatiles, $^{60}$ Co and $^{54}$ Mn, was similar to that of particulate matters in the process. Decontamination factors(DFs) for $^{60}$ Co and $^{54}$ Mn were 4.7$\times$10$^{5}$ and 6.2$\times$10$^{5}$ , respectively. Behavior of semivolatile radio-isotope, $^{137}$ Cs, was temperature dependent. DFs for $^{l37}$Cs at In different incineration temperature of 85$0^{\circ}C$ and $700^{\circ}C$ were 2.8$\times$10$^3$ and 2.6$\times$10$^4$, respectively. Trial bums of dry active waste(DAW) transported from nuclear power station(NPS) Kori 3,4 were also performed. DF for gross $\beta$/${\gamma}$ radioactivity in DAW was 1.1$\times$10$^{5}$ . This was a little higher than the estimated value, which was calculated from the tracer test results and nuclides distribution in the DAW. Average emission concentration was 0.019 Bq/N $m^3$, which could meet the maximal permissible concentration(MPC) in stack emission.n.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.191-198
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• 2014

Using results of groundwater flow system modeling for a hypothetical deep geological repository site, a distribution of groundwater flow rates at the disposal depth was analyzed and a method of applying this distribution to a safety assessment for a disposal of radioactive wastes was suggested. The distribution of groundwater flow rates was produced by hydraulic heads simulated from regional and local scale groundwater flow models for the hypothetical disposal site. The flow rates at the locations where deposition holes would be located were estimated. These rates were normalized by the maximum of the flow rates in order to probabilistically illustrate a possibility of canister failures at the deposition holes. From the normalized distribution, probabilistic expectations for mass discharges of radionuclides released from the canisters assumed to be failed were calculated and compared with those deterministically estimated under the assumption that the canisters at the same deposition holes were definitely failed. The suggested method can be contributed to constructing a methodology for safety assessment of a geological repository by reflecting natural conditions of a disposal site in more detail.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.113-122
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• 2015

⁵⁵Fe and ⁶³Ni are key factors in deciding the proper handling of the decommissioning of radioactive waste from nuclear facilities. For determining beta emitting radionuclides, the dismantled waste samples should be completely decomposed and separated from the sample matrix. This study reports the comparison results of the recovering efficiencies of Iron and Nickel with wet digestion methods that use various acids and alkali-fusion methods. Various matrices of NIST SRMs (1646a, 1944, 8704, 2709a, and 1633c), the recovering efficiencies of using alkali-fusion methods ranged from 95.3 to 98.3% for Iron, and from 86.6 to 88.1% for Nickel within about 2% of relative standard deviation. On the other hand, those using one of the three wet digestion methods ranged from 77.9 to 105.3% for Iron and from 40.1 to 78.5% for Nickel with over 10% of relative standard deviation. Therefore, one may draw the conclusion that the analytical results derived from Iron and Nickel using alkali-fusion methods are fairly reliable due to the recovering efficiencies observed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.305-315
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• 2020

For geological disposal of radioactive wastes, a method was proposed to evaluate the radionuclide transport in the biosphere by calculating the elapsed time of nuclide migration. The radionuclides were supposed to be introduced from a natural barrier and reached a large surface water body following a groundwater flow in a shallow subsurface. The biosphere was defined as a shallow subsurface environment that included aquifers on a host rock. Using the proposed method, a calculation algorithm was established, and a computer code that implemented the algorithm was developed. The developed code was verified by comparing the simulation results of the simple cases with the results of the analytical solution and a public program, which has been widely used to evaluate the radiation dose using the radionuclide transport near the surface. A case study was constructed using the previous research for radionuclide transport from the hypothetical geological disposal repository. In the case study, the code calculated the mass discharge rate of radionuclide to a stream in the biosphere. Because the previous research only demonstrated the transport of radionuclides from the hypothetical repository to the host rock, the developed code in the present study could help identify the total transport of radionuclide along the complete pathway.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.283-295
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• 2007

The technical feasibility of a pyroprocessing of PWR spent fuels to recover nuclear fuel materials, uranium and transuranic elements group(TRU), was examined in this study. Also its applicability as a new fuel cycle technology in terms of non-proliferation was investigated. First, various unit processes were combined to a pyroprocess. Then the flow aspects of such materials of issue as uranium, transuraniums, rare earth, noble metals and heat generating elements were examined on the flowsheet, which was obtained by the assumptions on the basis of various experimental results in this work or separation data collected from literatures. Consequently, the calculated results of the material balance for the whole process showed that uranium and TRU could be recovered as products by 98.0 % and 97.0 %, respectively, from a PWR spent fuel while removing the other elemental groups into radioactive wastes. On the one hand, the TRU product was found to emit a considerable amount of ${\gamma}$-ray as well as neutrons favorably contributing to the strategy of proliferation resistance.

• Tunnel and Underground Space
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• v.9 no.4
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• pp.306-314
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• 1999

In underground spaces including nuclear waste repository, prediction of air quantity, temperature/humidity and pollutant concentration is utmost important for space construction and management during the normal state as well as for determining the measures in emergency cases such as underground fires. This study aims at developing a model for underground space environment which has capabilities to take into account the effects of autocompression for the natural ventilation head calculation, to find the optimal location and size of fans and regulators, to predict the temperature and humidity by calculating the convective heat transfer coefficient and the sensible and latent heat transfer rates, and to estimate the pollutant levels throughout the network. The temperature/humidity prediction model was applied to a military storage underground space and the relative differences of dry and wet temperatures were 1.5 ~ 2.9% and 0.6 ~ 6.1%, respectively. The convection-based pollutant transport model was applied to two different vehicle tunnels. Coefficients of turbulent diffusion due to the atmospheric turbulence were found to be 9.78 and 17.35$m^2$/s, but measurements of smoke and CO concentrations in a tunnel with high traffic density and under operation of ventilation equipment showed relative differences of 5.88 and 6.62% compared with estimates from the convection-based model. These findings indicate convection is the governing mechanism for pollutant diffusion in most of the tunnel-type spaces.