• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.73-86
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• 2015

The use of complimentary indicators, other than radiation dose and risk, to assess the safety of radioactive waste disposal has been discussed in a number of publications for providing the reasonable assurance of disposal safety and convincing the public audience. In this study, the radionuclide flux was selected as performance indicator to appraise the performance of engineered barriers and natural barrier in the Wolsong low- and intermediate-level waste disposal facility. Radionuclide flux showing the retention capability by each compartment of the disposal system is independent of assumptions in biosphere model and exposure pathways. The scenario considered as the normal scenario of disposal facility has been divided into intact or degraded silo concrete conditions. In the intact silo concrete, the radionuclide flux has been assessed with respect to the radionuclide retardation performance of each engineered barrier. In the degraded silo concrete, the radionuclide flux has been explored based on the performance degradation of engineered barriers and the relative significance of natural barrier quantitatively. The results can be used to optimally design the near-surface disposal facility being planned as the second project phase. In the future, additional complimentary indicators will be employed for strengthening the safety case for improving the public acceptance of low- and intermediate-level waste disposal facility.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.228-228
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• 2004

Nuclear Environment Technology Institute (KHNP-NETEC) developed the conceptual design of the low and intermediate-level radioactive waste (LILW) repository. Among many engineering challenges, it is of particular importance to find out an optimum arrangement of near-surface disposal vaults in the repository area to minimize the radionuclide flux and concentration at the interface between the geo-sphere and bio-sphere. (omitted)

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1871-1877
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• 2023

The main goal of the coordinated project development of therapeutic radiopharmaceuticals of Y-90 and Re-188 is to exploit advancements in radionuclide production technology. Here, direct and indirect production methods with medium reactor and cyclotron are compared to evaluate derived neutron flux and production yield. First, nano-sized ¹⁸⁶W and ⁸⁹Y specimens are suspended in water in a quartz vial by FLUKA simulation. Then, the solution is irradiated for 4 days under 9E+14 n/cm2/s neutron flux of reactor. Also, a neutron activator including three layers-lead moderator, graphite reflector, and polyethylene absorbent- is simulated and tungsten target is irradiated by 60 MeV protons of cyclotron to generate induced neutrons for ¹⁸⁸W and ⁹⁰Sr production via neutron capture. As the neutron energy reduced, the flux gradually increased towards epithermal range to satisfy (n/2n,γ) reactions. The obtained specific activities at saturation were higher than the reported experimental values because the accumulated epithermal flux and nano-sized specimens influence the outcomes. The beta emitters, which are widely utilized in brachytherapy, appeal an alternative route to locally achieve a rational yield. Therefore, the proposed method via neutron activator may ascertain these broad requirements.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.274-279
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• 1996

Diffusion coefficient is a critical parameter for predicting radiological source term(migration rate and flux of radionuclide) through given near field conditions in spent fuel or high level waste repository. The effect of exchangeable cation-$Na^+$ and $Ca^{2+} - on the diffusion of $I^- \;and^3H$ (as HTO) in compacted bentonite was examined using a through-diffusion method. Bentonite material used here was compacted to a density of 1.3 Mg/m$^3$, and Na-bentonite was saturated with a solution of 100 mol NaCl/m$^3$ and Ca-bentonite with 50 $mol\;CaCl_2$/m$^3$. The results show that effective diffusion coefficients are generally higher by a factor of two to five in Ca-than Na-clay. This is attributed to the larger particle size of Ca-compared to Na-bentonite; hence, Ca-bentonite has a greater proportion of relatively large pores, which make a greater contribution to mass transport than small pores. Although the nature of the exchangeable cation affects mass diffusion in compacted bentonite, the effect is small and not likely to influence performance assessment modeling of compacted bentonite-based barriers.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1324-1329
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• 2018

This study presents a practical application of complementary safety indicators, which can be applied in a safety assessment of a radioactive waste repository by excluding a biosphere simulation and comparing the artificial radiation originating from the repository with the background natural radiation. Complementary safety indicators (radiotoxicity flux from geosphere and radiotoxicity concentration in seawater) were applied in the safety assessment of a rock-cavern type low and intermediate level radioactive waste (LILW) repository in the Republic of Korea. The natural radionuclide ($^{40}K$, $^{226,228}Ra$, $^{232}Th$, and $^{234,235,238}U$) concentrations in the groundwater and seawater at the Gyeongju LILW repository site were measured. Based on the analyzed concentrations of natural radionuclides, the levels of natural radiation were determined to be $8.6{\times}10^{-5}$ - $8.0{\times}10^{-4}Sv/m^2/yr$ and $6.95{\times}10^{-5}Sv/m^3$ for radiotoxicity flux from the geosphere and radiotoxicity concentration in seawater, respectively. From simulation results obtained using a Goldsim-based safety assessment model, it was determined that the radiotoxicity of radionuclides released from the repository is lower than that of the natural radionuclides inherently present in the natural waters. The applicability of the complementary safety indicators to the safety case was discussed with regard to reduction of the uncertainty associated with biosphere simulations, and communication with the public.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.153-164
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• 2014

Currently, there are many programs which are now being developed or already developed to predict radionuclide and corrosion product at the stage of designing NPP. However, since there are not many developments in evaluating quantity of activation corrosion products occurring when disassembling a nuclear power plant there exist some difficulties in calculating accurately. In order to evaluate activation products inventory for the research of effect of neutron activation in the reactor vessel, component of nuclear reactor and adjacent structures, it should be evaluated by using operation history of nuclear reactor, material composition of structure and average neutron flux in every field representing fixed structure of nuclear reactor. In this study, CORA, PACTOLE, CRUDSIM, CREAT and ACE codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor which is used at the stage of designing. As a future study, the accuracy in calculating the quantity of product corrosion can be increase by finding out the possibility of use and improvement for evaluation of the decontamination.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.121-126
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• 2019

The radionuclide inventory prediction of a nuclear power plant can help establish decommissioning plan by providing information of radiation environment. Accumulated radionuclides in reactors and related facilities after reactor shutdown can be divided into neutron activated materials and contaminated materials. Among the neutron activated radionuclides, $^{36}Cl$ and $^{41}Ca$ are important from the viewpoint of disposal because of its long half-life and physiochemical characteristics. In this research, we calculated the radionuclides of $^{36}Cl$ and $^{41}Ca$ in bioshielding concrete by estimating the neutron flux and cross section using the MCNPX. And we evaluated the inventories of $^{36}Cl$ and $^{41}Ca$ using the activation calculation code ORIGEN2.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Journal of Korean Society for Atmospheric Environment
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• v.11 no.4
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• pp.351-359
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• 1995

In order to investigate geochemical behaviors of artificial radionuclide($^{137}$ Cs), the fallout deposition of arificial radioisotope($^{137}$ Cs) was measured from May to October in 1994 at the Korea Ocean Research & Development Institute(KORDI), Ansan, Kyunggido, Korea. And to study radioisotopic behavior and cumulative action in soil, soil samples were collected from Kwang-Leung Forest, Kyunggidom and artificial radioisotope ($^{137}$ Cs) and natural radioisotope($^{210}$ Pb) were identified. The amount of $^{137}$ Cs in atmosphere collected by wet deposition process in May was found to be 4.95 to 11.96mBq m$^{-2}$ whereas the amounts of $^{137}$ Cs by dry deposition process in May and October were found to be 4.0mBq g$^{-1}$ and 3.0mBq g$^{-1}$ , respectively. The amount of $^{137}$ Cs accumulated in soil was measured to be 311mBq cm$^{-2}$ , which contained 83% of the total inputs from atmospheric fallout (374 mBq cm$^{-2}$ ) since 1960s. In addition, the accumulation rate and the annual flux of $^{210}$ Pb into soils were 0.32cm yr$^{-1}$ and 34 mBq cm$^{-2}$ yr$^{-1}$ , respectively. Conclusively, it was found that arificial radioisotopes were mainly from the stratosphere and soil resupension of continental China through the troposphere.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4715-4721
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• 2022

A comprehensive approach for modeling the pulse height spectra of gamma-ray detectors from passing radioactive cloud in a case of accident at NPP has been developed. It involves modeling the transport of radionuclides in the atmosphere using Lagrangian stochastic model, WRF meteorological processor with an ARW core and GFS data to obtain spatial distribution of radionuclides in the air at a given moment of time. Applying representation of the cloud as superposition of elementary sources of gamma radiation the pulse height spectra are calculated based on data on flux density from point isotropic sources and detector response function. The proposed approach allows us to obtain time-dependent spectra for any complex radionuclide composition of the release. The results of modeling the pulse height spectra of the scintillator detector NaI(Tl) Ø63×63 mm for a hypothetical severe accident at a NPP are presented.