• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3563-3579
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• 2021

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.431-439
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• 2018

Less mature nuclear reactor technologies are characterized by a greater uncertainty due to insufficient detailed design information, operational data, cost information, etc., but the expected performance characteristics of less mature options are usually more attractive in comparison with more mature ones. The greater uncertainty is, the higher economic risks associated with the project realization will be. Within a comparative evaluation of less and more mature nuclear reactor technologies, it is necessary to apply economic risk measures to balance judgments regarding the economic performance of less and more mature options. Assessments of any risk metrics involve calculating different characteristics of probability distributions of associated economic performance indicators and applying the Monte-Carlo method. This paper considers the applicability of statistical risk measures for different economic performance indicators within a trial case study on a comparative evaluation of less and more mature unspecified LWRs. The presented case study demonstrates the main trends associated with the incorporation of economic risk metrics into a comparative evaluation of less and more mature nuclear reactor technologies.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2395-2406
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• 2023

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.520-525
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• 2024

The optimization design problem of nuclear reactor radiation shield is a typical multi-objective optimization problem with almost 10 sub-objectives and the sub-objectives are always demanded to be under tolerable limits. In this paper, a design method combining multi-objective optimization algorithms with paralleling discrete ordinate transportation code is developed and applied to shield design of the Savannah nuclear reactor. Three approaches are studied for light-weighted and compact design of radiation shield. Comparing with directly optimization with 10 objectives and the single-objective optimization, the approach by setting sub-objectives representing weight and volume as optimization objectives while treating other sub-objectives as constraints has the best performance, which is more suitable to reactor shield design.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.354-358
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• 2016

Background: To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition Materials and Methods: MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. Results and Discussion: It was found that the leak rate is $0.1%{\cdot}day^{-1}$ of air, $0.004%{\cdot}day^{-1}$ of noble gas and $3.7{\times}10^{-5}%{\cdot}day^{-1}$ of aerosol during typhoon passing. The air leak rate of $0.1%{\cdot}day^{-1}$ can be converted into $1.36m^3{\cdot}hr^{-1}$, but the design leak rate in HANARO safety analysis report was considered as $600m^3{\cdot}hr^{-1}$ under the condition of $20m{\cdot}sec^{-1}$ wind speed outside of the building by typhoon. Conclusion: Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1311-1312
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• 2005

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.288-288
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• 2000

Reactor vessel is one of the most important equipment with regard to the safety of nuclear power plant. Thus nuclear regulation requires its periodical examination by certified inspection experts. Conventional reactor inspection machines are obsolete, hard to handle, and very expensive. To solve these problems we developed robotic reactor vessel inspection system which are small, easy to use for inspection, cost effective, and convenient in operation. This paper describes the main features of Main Control System which is one part of robotic inspection equipment we developed.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2004.10a
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• pp.1327-1330
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• 2004

An underwater robotic system has been developed and applied to visual inspection of reactor vessel internals. The Korea Electric Power Robot for Visual Test (KeproVt) consists of an underwater robot, a vision processor-based measuring unit, a master control station and a servo control station. The robot guided by the control station with the measuring unit can be controlled to have any motion at any position in the reactor vessel with $\pm$1 cm positioning and $\pm$2 degrees heading accuracies with enough precision to inspect reactor internals. A simple and fast installation process is emphasized in the developed system. The developed robotic system was successfully deployed at the Younggwang Nuclear Unit 1 for the visual inspection of reactor internals.

• Proceedings of the KIEE Conference
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• 2005.05a
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• pp.185-187
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• 2005

The reactor protection system is the most important function for the safe operation of nuclear powerplants (NPPs) in that such system protects a nuclear reactor tore whose damage can cause an enormous disaster to the nuclear facility and the public. A digital reactor protection system (DRPS) is being developed in KAERI for use in the newly-constructed NPPs and also for replacing the existing analog-type reactor Protection systems. In this paper, an software verification and validation (V&V) activities for DRPS, which are independent of the DRPS development processes, are described according to the software development life cycle. The main activities of DRPS V&V processes are the software planning documentations, the verification of software requirements specification (SRS) and software design specification (SDS), the verification of codes, the tests of the integrated software and system. Moreover, the software safety analysis and the software configuration management are involved in the DRPS V&V processes. All of the V&V activities are described, in detail, in this paper.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.728-744
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• 2024

The presence of hydrogen absorbed by zirconium-based cladding materials during reactor operation can trigger degradation mechanisms and endanger the rod integrity. Ensuring the durability of the rods in extended time-frames like dry storage requires anticipating hydrogen behavior using numerical modeling. In this context, the present paper describes a hydrogen post-processing tool for Falcon - HYPE, a PSI's in-house tool able to calculate hydrogen uptake, transport, thermochemistry, reorientation of hydrides and hydrogen-related failure criteria. The tool extracts all necessary data from a Falcon output file; therefore, it can be considered loosely coupled to Falcon. HYPE has been successfully validated against experimental data and applied to reactor operation and interim storage scenarios to present its capabilities.