• Nuclear Engineering and Technology
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• v.32 no.3
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• pp.261-273
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• 2000

During an extensive review made as part of the Integrated Diagnosis System project of the Maiden Reactor Project, MOAS (Maryland Operator Advisory System) was identified as one of the most thorough systems developed thus far. MOAS is an integrated on-line diagnosis system that encompasses diverse functional aspects that are required for an effective process disturbance management: (1) intelligent process monitoring and alarming, (2) on-line sensor data validation and sensor failure diagnosis, (3) on-line hardware (besides sensors) failure diagnosis, and (4) real-time corrective measure synthesis. The MOAS methodology was used at the Maiden Man-Machine Laboratory HAMMLAB of the OECD Maiden Reactor Project. The performance of MOAS, developed in G2 real-time expert system shell for the high-pressure preheaters of the NORS process in the HAMMLAB, was tested against a variety of transient scenarios, including failures of the control valves and sensors, and tube leakage of the preheaters. These tests showed that MOAS successfully carried out its intended functions, i.e., quickly recognizing an occurring disturbance, correctly diagnosing its cause, and presenting advice on its control to the operator. The lessons learned and insights gained during the implementation and performance tests also are discussed.

• 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.

• Journal of Soil and Groundwater Environment
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• v.17 no.5
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• pp.49-55
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• 2012

A 1.28 L-batch reactor and continuous-flow stirred tank reactor (CFSTR) fed with formate and trichloroethene (TCE) were operated for 120 days and 56 days, respectively, to study the effect of formate as electron donor on anaerobic reductive dechlorination (ARD) of TCE to cis-1,2-dichloroethylene (c-DCE), vinyl chloride (VC), and ethylene (ETH). In batch reactor, injected 60 ${\mu}mol$ TCE was completely degraded in the presence of 20% hydrogen gas ($H_2$) in less than 8 days by anaerobic dechlorination mixed-culture (300 mg-soluble protein), Evanite Culture with ability to completely degrade tetrachloroethene (PCE) and -TCE to ETH under anaerobic conditions. Once the formate was used as electron donor instead of hydrogen gas in batch or chemostat system, the TCE-dechlorination rate decreased and acetate production rate increased. It indicates that the concentration of hydrogen produced in both systems is possibly more close to threshold for homoacetogenesis process. Soluble protein concentration of Evanite culture during the batch test increased from 300 mg to 688 mg for 120 days. Through the protein monitoring, we confirmed an increase of microbial population during the reactor operation. In CFSTR test, TCE was fed continuously at 9.9 ppm (75.38 ${\mu}mol/L$) and the influent formate feed concentration increased stepwise from 1.3 mmol/L to 14.3 mmol/L. Injected TCE was accumulated at 18 days of HRT, but TCE was completely degraded at 36 days of HRT without accumulation of the injected-TCE during the left of experiment period, getting $H_2$ from fermentative hydrogen production of injected formate. Although c-DCE was also accumulated for 23 days after beginning of CFSTR operation, it reached steady-state in the presence of excessive formate. We also evaluated microbial dynamic of the culture at different chemical state in the reactor by DGGE (denaturing gradient gel electrophoresis).

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.575-576
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• 2004

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.04a
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• pp.293-294
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• 2014

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.99.1-99.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.287.1-287
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.120-120
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• 2002

• Journal of the Korean Society of Safety
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• v.30 no.5
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• pp.123-130
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• 2015

Wolsong Unit 2,3&4 has been performing a containment integrity test during power operation. This test could impact to the safe operation during test. If an accident occurs during pressure dropping phase, reactor trip can be delayed because of the increased pressure difference which causes a time delay to reach the trip set-point. On the contrary, if an accident occurs during pressure increasing phase, reactor trip could be accelerated because the pressure difference to the trip set-point decrease. Point Lepreau nuclear power plant, which installed GCLMS (Gross Containment Leakage Monitoring System) in 1990, has discontinued the test since 1992 due to these adverse effects. Therefore, we evaluated the risk to obviate the GCLMS test based on PWR's ILRT (Integrated Leak Rate Test) extension methodologies. The results demonstrate that risk increase rate is not high in case of performing only ILRT test at every 5 years instead of doing GCLMS test at every 1.5 years. In addition, the result shows that GCLMS test can be removed on a risk-informed perspective since risk increasement is in acceptable area of regulatory acceptance criteria.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.