• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2690-2701
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• 2022

In this paper, we select important fission products for the estimation of the source term during a severe accident of a PWR. The selection is based on the numerical results obtained from depletion calculations for the typical PWR fuel via the in-house code named DEGETION (Depletion, Generation, and Transmutation of Isotopes on Nuclear Application), release fractions of the fission products derived from NUREG-1465, and effective dose conversion coefficients from ICRP 119. Then, for the selected fission products, we obtain the adjoint solutions of the Bateman equations for radioactive decay in order to determine the importance of precursors producing the aforementioned fission products via radioactive decay, which would provide insights into the assumption used in MACCS 2 for a level 3 PSA analysis in which up to six precursors are considered in the calculations of radioactive decays for the fission product after release from the reactor.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2977-2983
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• 2023

During normal operation, some parts of the fission product in the defective fuel rods can release into the primary loops in PWR and the escape rate coefficients are widely used to assess quantitatively the release behaviors of fission products in the industry. The escape rate coefficients have been standardized and have been validated by some drilling experiments before the 1970s. In the paper, the model to determine the escape rate coefficients of fission products has been established and the typical escape rate coefficients of noble gas and iodine have been deduced based on the measured radiochemical data in one operating PWR. The result shows that the apparent escape rate coefficients vary with the release-to-birth and decay constants for different fission products of the same element. In addition, it is found that the escape rate coefficients from the defective rod with large defects are much higher than the standard escape rate coefficients, i.e., averagely 4.4 times and 1.8 times for noble gas and iodine respectively. The enhanced release of fission products from the severe secondary hydriding of several defective fuel rods in one cycle may lead to the potential risk of the temporary shutdown of the operating reactors.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.350-353
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• 2016

Background: Various characteristics such as size distribution, chemical component and radio-activity have been analyzed for radioactive aerosols released from Fukushima Daiichi Nuclear Power Plant. Measured results for radioactive aerosols suggest that the potential transport medium for radioactive cesium was non-sea-salt sulfate. This result indicates that cesium isotopes would preferentially attach with sulfate compounds. In the present work the attachment behavior of fission products to aqueous solution aerosols of sodium salts has been studied using a generation system of solution aerosols and spontaneous fission source of $^{248}Cm$. Materials and Methods: Attachment ratios of fission products to the solution aerosols were compared among the aerosols generated by different solutions of sodium salt. Results and Discussion: A significant difference according as a solute of solution aerosols was found in the attachment behavior. Conclusion: The present results suggest the existence of chemical effects in the attachment behavior of fission products to solution aerosols.

• Journal of the Korean Chemical Society
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• v.9 no.3
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• pp.113-115
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• 1965

In previous paper, the determination procedure of fission products mainly consists of many steps which are usually derived from classical analysis schemes. In the present paper a simple and more rapid method of the determination of fission products by means of a smaller amount of reagent than that correspond to the stoichiometric ratio of element to be determined is investigated. The yield of Ba-140 in fission products is determined by this method.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.825-836
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• 2014

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

• Nuclear Engineering and Technology
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• v.11 no.4
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• pp.295-301
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• 1979

For thermal-neutron-induced fission of $U^{235}$, nuclear charge distribution in the light part of the primary products has been calculated by using several postulates of charge distribution in the fission fragments. By comparing these values with the experimental results, it is revealed that those models are not appropriate for predicting the nuclear charge distribution in the fission fragments. The variation in the most probable charge, $Z_{P}$, of the isobaric distribution for the fission fragments and the charge for a mass given by unchanged charge density, $Z_{UCD}$, is turned out to be small as a function of mass. The parameter, $Z_{P}$ $-Z_{UCD}$, varies from 0.45 to 0.5 in charge units. The nuclear charge dispersion, $\sigma$, shows about 0.5 charge units for the fission fragments. Neutron odd-even effect in fission products could not be revealed clearly without considering the odd-even effect of prompt neutron emission.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.70-75
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• 1997

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14$\times$14 PWR fuel assembly and the results are given in this paper.

• Nuclear Engineering and Technology
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• v.13 no.4
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• pp.254-263
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• 1981

Paper electrophoretic separation of fission products has been carried out by using the specially designed migration apparatus. In general, the isolation of rubidium, strontium, zirconium, ruthenium, cesium, cerium, molybdenum, and some short-lived fission products is more efficient under 0.1M HCl electrolyte as compared with 0.1M NaOH electrolyte. In addition to Np-239, 1-131∼135 were, in particular, observed with different iodine chemical species obtained by the paper-electrophoretic separation of short, neutron-irradiated uranyl nitrate solution.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.