• Journal of Radiation Protection and Research
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• v.11 no.2
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• pp.139-145
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• 1986

An improved mothod of assessing fuel status by analyzsis of the fission product in the reactor coolant system is proposed. The release mechanism of specific fission products is established for determination of the coefficients in the equations which relate the radioactivities with the amount of defected fuel. Knock-out and migration models are employed in the formulation of the release mechanism. The influence of the tramp uranium is quantified. Sample calculations were made for KNU 1 reactor system using the I-131 and I-133 concentrations in the primary coolant. The estimated number of defected fuel pins in the third and sixth cycles appeared to be $9.34{\pm}1.13\;and\;0.294{\pm}0.092$, respectively.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.17-36
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• 2010

In this study, we compare the mass release rates of radionuclides(1) from waste forms arising from the KIEP-21 pyroprocessing system with (2) those from the directly-disposed pressurized-water reactor spent fuel, to investigate the potential radiological and environmental impacts. In both cases, most actinides and their daughters have been observed to remain in the vicinity of waste packages as precipitates because of their low solubility. The effects of the waste-form alteration rate on the release of radionuclides from the engineered-barrier boundary have been found to be significant, especially for congruently released radionuclides. the total mass release rate of radionuclides from direct disposal concept is similar to those from the pyroprocessing disposal concept. While the mass release rates for most radionuclides would decrease to negligible levels due to radioactive decay while in the engineered barriers and the surrounding host rock in both cases even without assuming any dilution or dispersal mechanisms during their transport, significant mass release rates for three fission-product radionuclides, $^{129}I$, $^{79}Se$, and $^{36}Cl$, are observed at the 1,000-m location in the host rock. For these three radionuclides, we need to account for dilution/dispersal in the geosphere and the biosphere to confirm finally that the repository would achieve sufficient level of radiological safety. This can be done only after we have known where the repository site would by sited. the footprint of repository for the KIEP-21 system is about one tenth of those for the direct disposal.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.386-392
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• 2021

MACCS (MELCOR Accident Consequence Code System), WinMACCS, and MelMACCS now facilitate a multi-unit consequence analysis. MACCS evaluates the consequences of an atmospheric release of radioactive gases and aerosols into the atmosphere and is most commonly used to perform probabilistic safety assessments (PSAs) and related consequence analyses for nuclear power plants (NPPs). WinMACCS is a user-friendly preprocessor for MACCS. MelMACCS extracts source-term information from a MELCOR plot file. The current development can combine an arbitrary number of source terms, representing simultaneous releases from a multi-unit facility, into a single consequence analysis. The development supports different release signatures, fission product inventories, and accident initiation times for each unit. The treatment is completely general except that the model is currently limited to collocated units. A major practical consideration for performing a multi-unit PSA is that a comprehensive treatment for more than two units may involve an intractable number of combinations of source terms. This paper proposes and evaluates an approach for reducing the number of calculations to be tractable, even for sites with eight or ten units. The approximation error introduced by the approach is acceptable and is considerably less than other errors and uncertainties inherent in a Level 3 PSA.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.551-562
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• 1996

This paper introduces the five-step methodology for identifying information needs and assessing instrument availability during the course of severe accidents in nuclear power plants. The methodology is applied to the Yonggwang (YGN) 3&4 to shed light on accident management. It constructs three safety objective trees to prevent the reactor vessel failure, to prevent the containment failure, and to mitigate the fission product release from the containment. The study assesses information needs and instrument availability under severe conditions for preventing the reactor vessel failure of YGN 3&4, and recommends additional instrument that m8y prove to be of vital importance in managing the accident.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1629-1635
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• 2017

Reliable prediction of nuclear fuel rod behavior of nuclear power reactors constitutes a basic demand for steady-state calculations, design purposes, and fuel performance assessment. Perfect design of fuel rods as the first barrier against fission product release is very important. Simulation of fuel rod performance with a code or software is one of the fuel rod design steps. In this study, a software program called MARCODE is developed in MATLAB environment that can analyze the temperature distribution, gap conductance value, and fuel and clad displacement in both solid and annular fuel rods. With a comparison of the maximum fuel temperature, fuel average temperature, fuel surface temperature, and gap conductance in solid and annular fuel, the effects of a central hole on the fuel temperature distribution are investigated.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.536-545
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• 2024

Radioactive iodine is a representative fission product to be quantified for the safety assessment of nuclear facilities. In integral severe accident analysis codes, the iodine behavior is usually described by a multi-physical model of iodine chemistry in aqueous phase under radiation field and mass transfer through gas-liquid interface. The focus of studies on iodine source term evaluations using the combination approach is usually put on the chemical aspect, but each contribution to the iodine amount released to the environment has not been decomposed so far. In this study, we attempted the decomposition by revising the two-film theory of molecular-iodine mass transfer. The model involves an effective overall mass transfer coefficient to consider the iodine chemistry. The decomposition was performed by regarding the coefficient as a product of two functions of pH and the overall mass transfer coefficient for molecular iodine. The procedure was applied to the EPICUR experiment and suppression chamber in BWR.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.53-64
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• 2007

The dynamic release behavior of Cs from high burn-up spent PWR fuel was experimentally performed under the conditions of a thermal treatment process such as voloxidation and sintering conditions. In voloxidation process, influence of the oxidation and reduction atmosphere on the Cs release characteristic using fragment type of spent fuel heated up to $1,500^{\circ}C$ was compared. In sintering process, temperature history effect on Cs release behavior was evaluated using green pellet under 4% $H_2/Ar$ environment. Temperature range for complete Cs release from spent fuel fragment under voloxidation condition was about $800^{\circ}C{\sim}1,200^{\circ}C$, but that of green pellet under the reduction atmosphere was $1,100^{\circ}C{\sim}1,400^{\circ}C$. Key parameters on Cs release behavior from spent fuel was powder formation as well as the diffusion rate of Cs compound to grain boundary and fuel surface.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.692-699
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• 2017

The determination of the shape and rising velocity of gas bubbles in a liquid pool is of great importance in analyzing the radioactive aerosol emissions from nuclear power plant accidents in terms of the fission product release rate and the pool scrubbing efficiency of radioactive aerosols. This article suggests a simple parameterization for the gas bubble rising velocity as a function of the volume-equivalent bubble diameter; this parameterization does not require prior knowledge of bubble shape. This is more convenient than previously suggested parameterizations because it is given as a single explicit formula. It is also shown that a bubble shape diagram, which is very similar to the Grace's diagram, can be easily generated using the parameterization suggested in this article. Furthermore, the boundaries among the three bubble shape regimes in the $E_o-R_e$ plane and the condition for the bypass of the spheroidal regime can be delineated directly from the parameterization formula. Therefore, the parameterization suggested in this article appears to be useful not only in easily determining the bubble rising velocity (e.g., in postulated severe accident analysis codes) but also in understanding the trend of bubble shape change due to bubble growth.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4049-4061
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• 2022

TRISO-coated particle fuel is widely used in high temperature gas cooled reactors and other advanced reactors. The performance of coated fuel particle is one of the fundamental bases of reactor safety. The failure probability of coated fuel particle should be evaluated and determined through suitable fuel performance models and methods during normal and accident condition. In order to better facilitate the design of coated particle fuel, a new TRISO fuel performance code named FRAT (Fission product Release Analysis Tool) was developed. FRAT is designed to calculate internal gas pressure, mechanical stress and failure probability of a coated fuel particle. In this paper, FRAT was introduced and benchmarked against IAEA CRP-6 benchmark cases for coated particle failure analysis. FRAT's results agree well with benchmark values, showing the correctness and satisfactory applicability. This work helps to provide a foundation for the credible application of FRAT.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.301-307
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• 2005

The morphology of the high burnup $UO_2$ spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI's PIEF. The test specimen was used with about 200 mg of the spent $UO_2$ fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at $1400^{\circ}C$ for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to $500^{\circ}C$ and held for 3hrs in the mixed gas (60 ml He and 100 ml STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the $UO_2$ matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to $10\;{\mu}m$.