• The Korean Journal of Zoology
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• v.11 no.3
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• pp.85-91
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• 1968

Artificial and natural radioactivity in airborne, rain-out and fallout dusts in Seoul district in the period of 1963-1967 were studied by measuring gross-activity and by analyzing nuclides by means of $\\gamma$-spectrometry. Short-lived radium and thorium decay products give rise to most of the airborne activity unless the fission product concentration is extremely high and it is likely to be said activity remaining after a few days is attributable to fission products. Of seven Chinese nuclear explosions performed at Lop Nor, Sinkiang Province, two exhibited the activity of extremely high concentration of fission product and reached Seoul district around 30 hours after the explosion. The activity was followed by a sudden decrease in less than a week, in contrast to the long-lasted activity of low concentration originating from the huge tests performed by the United States and the USSR in 1956-1962. The radioactive environmental contamination in Seoul district, due to the Chiness nuclear test explosions, largely depends on the height above the earth at which the nuclear explosion is performed and the type of nuclear device as well as the weather system at the time and immediately after the explosion, especially the jet stream in middle latitude in the upper troposphere.

• Journal of Radiation Protection and Research
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• v.14 no.2
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• pp.30-36
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• 1989

A FISSLIB, 51 group nuclear data set for 22 product nuclides, which are present in spent fuel and significantly affect the criticality of spent fuel, was generated from ENDF/B-IV using XLACS-II. The FISSLIB is ready to be used together with a data set generated from DLC-43/CSRL using AMPX system. The reliability of FISSLIB was verified by comparison with the data reported in BNL-325. Using FISSLIB, the criticality of KORI-1 spent fuel rod arranged infinitely was analyzed, and it was found that $K_{eff}$ of the spent fuel including fission products was lower by 9-14% than that calculated without fission products.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.169-179
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• 1984

An efficient procedure of evaluating the fuel cladding failures occurring in the normal operations of typical PWR's has been investigated through the analysis of fission product(FP) activities in the reactor coolant using an analytical model, FIPREL code. Performed by this code is an extensive study on the sensivities of FP activities to such physical parameters as enrichment, turnup, and operation temperature of failed fuel rod as well as the effective failure size quantified in terms of the magnitude of gap release coefficient. The results of study are generally in agreement with those by PROFIP method. In the presence of tramp uranium the portion of activities released from failed rod is separated by an iterative calculation based on the activity ratios of fission nuclides chemically more stable than iodines. Obtained are the linear power density and the number of failed rods, the effective failure size, and the mass of tramp uranium. The operation experiences of 4 cycles of Kori Unit 1 are analyzed and the results show that the model is highly reliable for the survey and evaluation of fuel rod conditions during reactor operations.

• Corrosion Science and Technology
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• v.18 no.5
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• pp.175-181
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• 2019

A failed spent fuel rod with 53,000 MWd/tU from a nuclear power plant was characterized, and the fission products and oxygen layer in the pellet-clad mechanical interaction region were observed using an EPMA (Electron Probe Micro-Analyzer). A sound fuel rod burned under similar conditions was used to compare and analyze, the results of the failed fuel rod. In the failed fuel rod, the oxide layer represented $10{\mu}m$ of the boundary of the cladding, and $35{\mu}m$ of the region outside the cladding. By comparison, in the sound fuel rod, the oxide layer was $8{\mu}m$, observed in the cladding boundary region. The cladding inner surface corrosion and the resulting fuel-cladding bonding were investigated using an EPMA. Zirconium existed in the bonding layer of the (U, Zr)O compound beyond the pellet cladding interaction gap of $20{\mu}m$, and composition of UZr2O3 was observed in the failed fuel rod. This paper presents the results of the EPMA examination of a spent fuel specimen, and a technique to analyze fission products in the pellet-clad mechanical interaction region.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3073-3084
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• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.484-488
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• 1997

Fission product release (FPR) assessment for 100% reactor outlet header (ROH) break in CANDU reactor loaded with CANFLEX-NU fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The fuel failure thresholds for the CANFLEX and standard bundle elements are very similar. All the sheaths at the corresponding fuel failure thresholds for the CANFLEX and standard bundles fail due to the significant cracks in the surface oxide, except those for the CANFLEX inner element at burnups of 220 to 240 MW.h/kg(U), which fail due to the excessive diametral strain. The fuel failure analysis predicts that the number of failed fuel elements for the CANFLEX bundle case is none, while that for the standard bundle case is 1827. The total (gap plus bound) I-131 releases for the CANFLEX and standard bundles are none and 5889 TBq, respectively The significant reduction of the number of failed fuel elements and FPR for the CABFKEX fuel bundle is attributed to the lower linear power of the CANFLEX fuel bundle compared with the standard fuel bundle.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.870-880
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• 2016

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

• Bulletin of the Korean Chemical Society
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• v.1 no.1
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• pp.35-39
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• 1980

A vacuum ultraviolet photolysis of ethyl bromide was studied in the pressure range of 0.5-19.9 torr and at 123.6 nm krypton resonance line. The pressure effect on the reaction was studied by increasing the reactant pressure and by adding an inert gas, e.g., He. In the observation the monatomic gas is found to be no effect in the reaction. A scavenger effect of the reaction was also performed by adding NO gas as a radical scavenger and was found to be quite efficient to scavenge a radical product $C_2H_6$. The observation of the major reaction product $C_2H_6$ was interpreted in terms of a molecular elimination. Nontheless the decreasing phenomenon of ${\phi}_{C_2H_4}/{\phi}_{C_2H_6}$ with pressure rise was attributed to the existence of the two electronically excited states. One state proceeds to the molecular elimination and the other to carbon-bromine bond fission. The excitation and the decomposition mechanisms between two excited states and the reaction products were interpreted in terms of the first excitation which proceeds the molecular elimination, and the second excitation which resulted from the first excited state by collisional cross over decomposes by carbon-bromine bond fission.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2702-2713
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• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.