• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.05a
• /
• pp.287-288
• /
• 2017

• The Korean Journal of Zoology
• /
• v.11 no.3
• /
• pp.85-91
• /
• 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
• /
• v.14 no.2
• /
• pp.30-36
• /
• 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
• /
• v.50 no.6
• /
• pp.820-828
• /
• 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%.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1907-1916
• /
• 2020

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

• Nuclear Engineering and Technology
• /
• v.49 no.2
• /
• pp.434-441
• /
• 2017

Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the "black-box" code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.

• Journal of Radiation Protection and Research
• /
• v.42 no.2
• /
• pp.106-113
• /
• 2017

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

• Nuclear Engineering and Technology
• /
• v.48 no.4
• /
• pp.965-974
• /
• 2016

The article describes the results of experiments conducted on pigs to determine the effect of plutonium, which is the most radiotoxic and highly active element in the range of mixed fuel $(U_{0.8}Pu_{0.2})O_2$ fission products, on living organisms. The results will allow empirical prediction of the emergency plutonium radiation dose for various organs and tissues of humans in case of an accident in a reactor running on mixed fuel $(U_{0.8}Pu_{0.2})O_2$.

• Nuclear Engineering and Technology
• /
• v.24 no.3
• /
• pp.228-235
• /
• 1992

Some soluble species may not be solubility-limited or congruent-released with the matrix species. For example, during the operation of the nuclear reactor, the fission products can be accumulated in the fuel-cladding gap, voids, and grain boundaries of the fuel rods. In the waste package for spent-fuel placed in a geologic repository, the high solubility species of these fission products accumulated in the“gap”, e.g. cesium or iodine are expected to dissolve rapidly when ground water penetrates fuel rods. The time and space dependent mass transport for high solubility nuclides in the gap is analyzed, and its numerical illustrations are demonstrated. The approximate solution that is valid for all times is developed, and validated by comparison with an asymptotic solution and the solution obtained by the numerical inversion of Laplace transform covering the entire time span.

• Journal of the Korean institute of surface engineering
• /
• v.56 no.1
• /
• pp.84-93
• /
• 2023

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.