• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.237-244
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• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Journal of the Korean Institute of Intelligent Systems
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• v.3 no.2
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• pp.18-35
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• 1993

The water level control system of the steam generator in a pressurized water reactor and its control problems are analysed. In this work the stable control strategy during the low power operation and transient states is studied. To solve the problem, a fuzzy logic control method is applied as a basic algorithm of the controller. The control algorithm is based on the operator's knowledges and the experiences of manual operation for water level control at the compact nuclear simulator set up in Korea Atomic Energy Research Institute. From a viewpoint of the system realization, the control variables and rules are established considering simpler tuning and the input-output relation. The control strategy includes the dynamic tuning method and employs a substitutional information using the bypass valve opening instead of incorrectly measured signal at the low flow rate as the fuzzy variable of the flow rate during the pressure control mode of the steam generator. It also involves the switching algorithm between the control valves to suppress the perturbation of water level. The simulation results show that both of the fine control action at the small level error and the quick response at the large level error can be obtained and that the performance of the controller is improved.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.186-195
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• 1982

The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f $t^2$ steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of $F_{{\Delta}H}$=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.

• KIISE Transactions on Computing Practices
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• v.22 no.3
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• pp.145-150
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• 2016

Finding the exact location of faults in a program requires enormous time and effort. Several fault localization methods based on control flows of a program have been studied for decades. Unfortunately, these methods are not applicable to programs based on data-flow languages. A recently proposed mutation-based fault localization method is applicable to data-flow languages, as well as control-flow languages. However, there are no studies on the effectiveness of the mutation-based fault localization method for data-flow based programs. In this paper, we provided an experimental case study to evaluate the effectiveness of mutation-based fault localization on programs implemented in Function Block Diagram (FBD), a widely used data-flow based language in safety-critical systems implementation. We analyzed several real faults in the implementation of FBD programs of a nuclear reactor protection system, and evaluated the mutation-based fault localization effectiveness for each fault.

• Analytical Science and Technology
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• v.16 no.5
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• pp.418-425
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• 2003

Inorganic elemental content of bottom ash generated from municipal solid waste incinerator was determined by neutron activation analysis. Bottom ash samples were collected monthly from incinerator located in city D, strained out by the 5 mm sieve, dried by oven and pulverized by agate mortar. The samples were irradiated at NAA #1 irradiation hole in HANARO research reactor of Korea Atomic Energy Research Institute and the irradiated samples were measured by HPGe-gamma-ray spectrometer. From the activity of measured nuclides, 33 elements including As, Cr, Cu, Fe, Mn, Sb and Zn were determined applying activity creation formula and nuclear data. The quality control was conducted by simultaneous analysis with NIST standard reference materials.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.83-86
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• 2010

Purpose: The Molybdenum which is the raw material of $^{99}Mo$-$^{99m}Tc$ generator (generator) is produced from the nuclear reactor. However, output has dwindled as the two nuclear reactors supplying the bulk of radioactive material-one in Chalk River, Ontario and the other in Petten, the Netherlands-have been closed for repairs or maintenance. This resulted in the enhancement of its price. Therefore we have tried to seek the new method which could run generator to increase activity of $^{99m}Tc$ in this study. Materials and Methods: The $^{99m}Tc$ activity obtained from 5 times elution for 5 days from Monday to Friday using two generators was compared with 10 times elution. Appearance test, pH test, LAL test, sterility test, chemical impurity(Al) test, radio chemical purity test, ratio of $^{99}Mo$/$^{99m}Tc$ activity test have been done to check the stability of $^{99m}Tc$ eluting from generator respectively. Results: The $^{99m}Tc$ activity obtained from 5 times elution for 5 days was 168.2 GBq (4545 mCi) and 10 times was 230.5 GBq (6230 mCi). All quality control tests were within normal limit. Conclusion: We got to know that 2 times elution a day obtained more $^{99m}Tc$ activity than one time elution in this study.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1115-1130
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• 2012

In nuclear power plants, the reactor pressure vessel (RPV) upper head control rod drive mechanism (CRDM) penetration nozzles are fabricated using J-groove weld geometry. Recently, the incidences of cracking in Alloy 600 CRDM nozzles and their associated welds have increased significantly. The cracking mechanism has been attributed to primary water stress corrosion cracking (PWSCC), and it has been shown to be driven by welding residual stresses and operational stresses in the weld region. The weld-induced residual stress is the main factor contributing to crack growth. Therefore, an exact estimation of the residual stress is important for ensuring reliable operation. This study presents the residual stress computation performed for an RPV CRDM penetration nozzle in Korea. Based on two and three dimensional finite element analyses, the effect of welding variables on the residual stress variation is estimated for sensitivity analysis.

• Proceedings of the Korea Society for Simulation Conference
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• 2000.11a
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• pp.61-66
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• 2000

The KEPRI has been upgrading the KNPEC(Korea Nuclear Power Education Center) #2 simulator, a replica of Yonggwang Unit 1 & 2, due to the outdated systems. The scenarios, such as the continuous load change, are selected to verify and validate the simulator, and the data required to V&V are generated with the best-estimated codes, RETRAN and MARS. The reactor coolant system and steam generator system are cut up into volumes and junctions for the accurate model of the scenarios, and other components and control systems are modeled. For the model the operation and design data of the plants is used and in some cases the data of Kori Unit 3 & 4 is used to fill up the lack of required data. The results of some selected analyses with the models are compared with the operating data of the plants to verify the models, and the analyses of the scenarios are carried out to generated the data for the V&V of the simulator

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.107-111
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• 2011

The release of a site and building from regulatory control is the final stage of the decommissioning process. The MARSSIM (Multi-Agency Radiation Survey and Site Investigation Manual) provides overall framework for conducting data collection for a final status survey to demonstrate compliance with site closure requirements. The KAERI carried out establishing a final status survey by using the guidance provided in the MARSSIM for of a site and building of the Korea Research Reactor. The release criteria for a site and building were set up based on these results of the site specific release levels which were calculated by using RESRAD and RESRAD-Build codes. The survey design for a site and building was classified by using the survey dataset and potential contamination. The number of samples in each survey unit was calculated by through a statistical test using the collected data from a scoping and characterization survey. The results of the final status survey were satisfied the release criteria based on an evaluation of the measured data.