• Journal of Korean Society for Quality Management
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• v.32 no.4
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• pp.1-14
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• 2004

Nuclear project company establishes and maintains its QA system in the conformance with applicable Codes & Standards and contract provisions. Certifying agency(e.g., Korea Electric Association) or project owner appraises the effectiveness of the company QA system by surveying or auditing it. However, although the company fulfills the requirements of this survey/audit, the capability level of its nuclear QA system is not appraised numerically and developed in that KEPIC QAP-1, used by the company as a nuclear QA guidebook, does not describe an approach for managing the capability level of the QA system. This study evaluates the characteristics of QAP-1 from the viewpoint of SE-CMM(Capability Maturity Model for Systems Engineering) in order to suggest a possible solution.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.417-422
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• 1998

he Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3&4(UCN 3&4) has been improved from the Yonggwang Nuclear Power Plant Units 3&4(YGN 3&4) based on the evaluation results for depressurization capability tests performed at YGN 3&4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown Phenomena in order to optimize the orifice size of UCN 3&4 RCGVS. Baesd on these analyses results, the RCGVS orifice size for UCN 3&4 has been reduced to 9/32 inch from the l1/32 inch for YGN 3&4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3&4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.968-976
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• 2019

This paper presents a number of verification case studies for a recently developed sensitivity/uncertainty code package. The code package, ROMUSE (Reduced Order Modeling based Uncertainty/Sensitivity Estimator) is an effort to provide an analysis tool to be used in conjunction with reactor core simulators, in particular the Virtual Environment for Reactor Applications (VERA) core simulator. ROMUSE has been written in C++ and is currently capable of performing various types of parameter perturbations and associated sensitivity analysis, uncertainty quantification, surrogate model construction and subspace analysis. The current version 2.0 has the capability to interface with the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) code, which gives ROMUSE access to the various algorithms implemented within DAKOTA, most importantly model calibration. The verification study is performed via two basic problems and two reactor physics models. The first problem is used to verify the ROMUSE single physics gradient-based range finding algorithm capability using an abstract quadratic model. The second problem is the Brusselator problem, which is a coupled problem representative of multi-physics problems. This problem is used to test the capability of constructing surrogates via ROMUSE-DAKOTA. Finally, light water reactor pin cell and sodium-cooled fast reactor fuel assembly problems are simulated via SCALE 6.1 to test ROMUSE capability for uncertainty quantification and sensitivity analysis purposes.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1907-1916
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• 2020

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

• 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.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.78-79
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• 2019

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.681-682
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.643-644
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.362-367
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• 1997

To assess the coolability of particle bed, which is formed in reactor cavity, it is important to assess the prediction capabilities of Dryout Heat flux correlations. The existing DHF correlations (Sowa et al., Dhir-Catton (a), Dhir-Catton (b), Hardee-Nilson, Ostesen, Shires-Stevens, Lipinski, Jones et al., Dhir-Barleon, Theofanous-Saito, Henry-Fauske) for particle beds are assessed using developed DHF database. Eleven DHF correlations are chosen for assessment based on literature survey. Among them, five are based on flooding correlation, which are used for chemical engineering and others are based on conservation equations. The parameters in DHF correlations are directly substituted into correlations. Totally 202 data are classified into 6 groups based on bed thickness and particle diameter. In each group, prediction capabilities of correlations are assessed and shown by standard deviation and root mean square (RMS) error. Prediction capability of each correlation depends on the data group and none of correlations shows best prediction capability on entire groups. According to present study, even if those correlations show poor prediction capability, Lipinski correlation is best correlation considering entire groups.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.410-410
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• 2001