• Structural Engineering and Mechanics
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• v.19 no.1
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• pp.97-106
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• 2005

The standard practice is to seismically qualify the safety related equipment and structural components used in the nuclear power plants. Among several qualification approaches the qualification by the analysis using finite element (FE) method is the most common approach used in practice. However the predictions by the FE model for a structure is known to show significant deviations from the dynamic behaviour of 'as installed' structure in many cases. Considering such limitation, few researchers have advocated re-qualification of such structures after installation at site to enhance the confidence in qualification vis-$\grave{a}$-vis plant safety. For such an exercise the validation of FE model with experimental modal data is important. A validated FE model can be obtained by the Model Updating methods in conjugation with the in-situ experimental modal data. Such a model can then be used for qualification. Seismic analysis using the updated FE model and its advantage has been presented through an example of an in-core component - a perforated horizontal tube of a nuclear reactor.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Journal of Radiation Protection and Research
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• v.14 no.2
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• pp.37-44
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• 1989

An improved water chemistry technique was studied to reduce radioactivity build-up in reactor coolant system. The technique is convering the current coordinated lithium-boron chemistry regime to the elevated lithium chemistry regime in order to maintain high pH. Correlations between reactor coolant pH and radioactivity build-up were analized by using pH data from domestic PWRs. Consequently, it was founded that high pH chemistry was moer effective for radioactivity build-up reduction than current chemistry regime. This fact had revealed that much portion of reactor coolant corrosion products were nickel ferrite rather than magnetite, and that pH value ranging 7.0-7.4 was appropriate for high-pH chemistry operation.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.559-566
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• 2015

The thermalehydraulic characteristics for the CANadian Deuterium Uranium Flexible (CANFLEX)-burnable poison (BP) fuel channel, which is loaded with a BP at the center ring based on the CANFLEX-RU (recycled uranium) fuel channel, are evaluated and compared with that of standard 37-element and CANFLEX-NU (natural uranium) fuel channels. The distributions of fuel temperature and critical channel power for the CANFLEX-BP fuel channel are calculated using the NUclear Heat Transport CIRcuit Thermohydraulics Analysis Code (NUCIRC) code for various creep rate and burnup. CANFLEX-BP fuel channel has been revealed to have a lower fuel temperature compared with that of a standard 37-element fuel channel, especially for high power channels. The critical channel power of CANFLEX-BP fuel channel has increased by about 10%, relative to that of a standard 37-element fuel channel for 380 channels in a core, and has higher value relative to that of the CANFLEX-NU fuel channel except the channels in the outer core. This study has shown that the use of a BP is feasible to enhance the thermal performance by the axial heat flux distribution, as well as the improvement of the reactor physical safety characteristics, and thus the reactor safety can be improved by the use of BP in a CANDU reactor.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.929-944
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• 2009

The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Light Water Reactor (LWR) lower plenum during hypothetical severe accident progression. The ECM uses heat transfer characteristic velocities to describe turbulent natural convection of a melt pool. The simple approach of the ECM method allows implementing different models of the characteristic velocity in a mushy zone for non-eutectic mixtures. The Phase-change ECM (PECM) was examined using three models of the characteristic velocities in a mushy zone and its performance was compared. The PECM was validated using a dual-tier approach, namely validations against existing experimental data (the SIMECO experiment) and validations against results obtained from Computational Fluid Dynamics (CFD) simulations. The results predicted by the PECM implementing the linear dependency of mushy-zone characteristic velocity on fluid fraction are well agreed with the experimental correlation and CFD simulation results. The PECM was applied to simulation of melt pool formation heat transfer in a Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lower plenum. The study suggests that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2812-2820
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• 2020

House load operation (HLO) occurs when the generator supplies power to the house load without triggering reactor trips during grid disturbances. In Korea, the HLO capability of optimized power reactor 1000 (OPR1000) plants has prevented several reactor trips. Operational experiences demonstrate the difference in the reactor trip incidence due to grid disturbances between OPR1000 plants and Westinghouse plants in Korea, attributable to the availability of the HLO capability. However, probabilistic safety assessments (PSAs) for OPR1000 plants have not considered their specific design features in the initiating event analyses. In an at-power PSA, the HLO capability can affect the initiating event frequencies of general transients (GTRN) and loss of offsite power (LOOP), resulting from transients within the grid system. The initiating event frequencies of GTRN and LOOP for an OPR1000 plant are reduced by 17.7% and 78.7%, respectively, compared to the Korean industry-average initiating event frequencies, and its core damage frequency from internal events is reduced by 15.2%. The explicit consideration of the HLO capability in initiating event analyses makes significant changes in the risk contributions of the initiating events. Consequently, for more realistic at-power PSAs in Korea, we recommend incorporating plant-specific HLO-related design features when estimating initiating event frequencies.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1306-1313
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• 2018

The randomness and incipient nature of certain faults in reactor systems warrant a robust and dynamic detection mechanism. Existing models and methods for fault diagnosis using different mathematical/statistical inferences lack incipient and novel faults detection capability. To this end, we propose a fault diagnosis method that utilizes the flexibility of data-driven Support Vector Machine (SVM) for component-level fault diagnosis. The technique integrates separately-built, separately-trained, specialized SVM modules capable of component-level fault diagnosis into a coherent intelligent system, with each SVM module monitoring sub-units of the reactor coolant system. To evaluate the model, marginal faults selected from the failure mode and effect analysis (FMEA) are simulated in the steam generator and pressure boundary of the Chinese CNP300 PWR (Qinshan I NPP) reactor coolant system, using a best-estimate thermal-hydraulic code, RELAP5/SCDAP Mod4.0. Multiclass SVM model is trained with component level parameters that represent the steady state and selected faults in the components. For optimization purposes, we considered and compared the performances of different multiclass models in MATLAB, using different coding matrices, as well as different kernel functions on the representative data derived from the simulation of Qinshan I NPP. An optimum predictive model - the Error Correcting Output Code (ECOC) with TenaryComplete coding matrix - was obtained from experiments, and utilized to diagnose the incipient faults. Some of the important diagnostic results and heuristic model evaluation methods are presented in this paper.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2445-2453
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• 2021

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutualcomparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21^(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4014-4021
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• 2021

If safety injection systems (SISs) do not work in the event of a loss-of-coolant accident (LOCA), the accident can progress to a severe accident in which the reactor core is exposed and the reactor vessel fails. Therefore, it is considered that a technology that provides recoverable maximum time for SIS actuation is necessary to prevent this progression. In this study, the corresponding time was defined as the golden time. To achieve the objective of accurately predicting the golden time, the prediction was performed using the deep fuzzy neural network (DFNN) with rule-dropout. The DFNN with rule-dropout has an architecture in which many of the fuzzy neural networks (FNNs) are connected and is a method in which the fuzzy rule numbers, which are directly related to the number of nodes in the FNN that affect inference performance, are properly adjusted by a genetic algorithm. The golden time prediction performance of the DFNN model with rule-dropout was better than that of the support vector regression model. By using the prediction result through the proposed DFNN with rule-dropout, it is expected to prevent the aggravation of the accidents by providing the maximum remaining time for SIS recovery, which failed in the LOCA situation.