• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1571-1584
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• 2023

In this study, a numerical analysis was performed as part of an international joint research project to reproduce a real cable tray fire that occurred in the heater bay area of the turbine building of a nuclear power plant. A sensitivity analysis was performed on various input parameters to derive results consistent with the sprinkler activation time obtained from the fire event analysis. For all sensitive parameters, the normalized sprinkler activation time correlated well with the power function of the normalized sprinkler height. A correlation equation was developed to identify the sprinkler activation time at any location when determining the slope or fire growth rate under the conditions assuming a linear or t-squared heat release rate (HRR) time curve. Various cable fire growth assumptions were used to determine which assumption was better to provide the prediction coincident with the information given from the fire event analysis in terms of the sprinkler activation time and total energy generated from cables damaged by fire. In the comprehensive analysis of all the sensitive parameters, the standard deviation of the input parameters increased as the sprinkler height decreased. Within the range of the sensitivity parameter values given in this study, when considering all sprinkler heights, the standard deviation of the cable model change was the largest and that of the overhang position change was the smallest.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1225-1233
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• 2024

Level 3 probabilistic safety assessment (PSA) is performed to calculate radionuclide concentrations and exposure dose resulting from nuclear power plant accidents. To calculate the external exposure dose from the released radioactive materials, the radionuclide concentrations are multiplied by two factors of dose coefficient and a finite cloud dose correction factor (FCDCF), and the obtained values are summed. This indicates that a standard set of FCDCFs is required for external exposure dose calculations. To calculate a standard set of FCDCFs, the effective distance from the release point to the receptor along the wind direction should be predetermined. The TID-24190 document published in 1968 provides equations to calculate FCDCFs and the resultant standard set of FCDCFs. However, it does not provide any explanation on the effective distance required to calculate the standard set of FCDCFs. In 2021, Sandia National Laboratories (SNLs) proposed a method to predetermine finite effective distances depending on the atmospheric stability classes A to F, which results in six standard sets of FCDCFs. Meanwhile, independently of the SNLs, the authors of this paper discovered that an infinite effective distance assumption is a very reasonable approach to calculate one standard set of FCDCFs, and they implemented it into the multi-unit radiological consequence calculator (MURCC) code, which is a post-processor of the level 3 PSA codes. This paper calculates and compares short- and long-range FCDCFs calculated using the TID-24190, SNLs method, and MURCC method, and explains the strength of the MURCC method over the SNLs method. Although six standard sets of FCDCFs are required by the SNLs method, one standard sets of FCDCFs are sufficient by the MURCC method. Additionally, the use of the MURCC method and its resultant FCDCFs for level 3 PSA was strongly recommended.

• Journal of the Korean Society of Safety
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• v.26 no.6
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• pp.85-89
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• 2011

Component reliability data were analyzed by using PRinS(Plant Reliability data information System) based on the latest operating experiences of eight KSNPs(Korea Standard Nuclear Power plants), and these new data were applied to the KSNP PSA models. In addition, the existing PSA models were revised for reflecting as-built and as-operated plant conditions. As a result of newly performing PSA in this paper, CDF and LERF were estimated 26.1% and 18.2% lower than the existing values, respectively. It was identified that the risk measures decreased not because of revising the models but because of applying the new component reliability data. The result and the method of this paper could be used when generating plant specific data and performing the living PSA in the future.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.117.1-117
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.381.1-381
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.341.1-341
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.797-798
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.213-213
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• 2002

• Journal of the Korean Institute of Intelligent Systems
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• v.5 no.3
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• pp.52-63
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• 1995

A neuro-fuzzy algorithm is presented for nuclear reactor power control in a pressurized water reactor. Automatic reacotr power control is complicated by the use of control rods because of highly nonlinear dynamics in the axial power shape. Thus, manual shaped controls are usually employed even for the limited capability during the power maneuvers. In an attempt to achieve automatic shape control, a neuro-fuzzy approach is considered because fuzzy algorithms are good at various aspects of operator's knowledge representation while neural networks are efficinet structures capable of learning from experience and adaptation to a changing nuclear core state. In the proposed neuro-fuzzy control scheme, the rule base is formulated based ona multi-input multi-output system and the dynamic back-propagation is used for learning. The neuro-fuzzy powere control algorithm has been tested using simulation fesponses of a Korean standard pressurized water reactor. The results illustrate that the proposed control algorithm would be a parctical strategy for automatic nuclear reactor power control.

• Journal of Energy Engineering
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• v.21 no.3
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• pp.292-300
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• 2012

KHNP CRI has been developing APR+ nuclear power plant since 2007, which is GEN III+ model with 4,361 MWth capacity. To develop safer and more economical nuclear power plant than APR1400, we studied domestic and foreign nuclear power plants under construction. We also reviewed nuclear power plants which are appropriate for domestic construction in Korea and also for export. Economic assessments were made twice during the second phase of standard detailed design of the plant. The result of the second phase of economic analysis for APR+ standard detailed design showed that APR+ N-th plant was 24.6% more economical than coal-fired 1,000MW power plant, and was evaluated to be competitive enough in global market for construction of the nuclear power plant.