• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.38-47
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• 1986

A statistical procedure, which uses response surface method and Monte Carlo simulation technique, is proposed for analyzing the thermal margin of light water reactor core. The statistical thermal margin analysis method performs the best.estimate thermal margin evaluation by the probabilistic treatment of uncertainties of input parameters. This methodology is applied to KNU-1 core thermal margin analysis under the steady state nominal operating condition. Also discussed are the comparisons with conventional deterministic method and Improved Thermal Design Procedure of Westinghouse. It is deduced from this study that the response surface method is useful for performing the statistical thermal margin analysis and that thermal margin improvement is assured through this procedure.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.379-384
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• 1996

Thermal-hydraulic performance has been analyzed for an advanced reactor core loaded with hexagonal fuel assemblies. Currently available CHF prediction models and data base for triangular lattice bundles have been thoroughly reviewed, and as a result the KfK-3 CHF correlation with limit CHFR of 1.235 has been determined to be most appropriate. The pressure drop model in COBRA-IV-I code has been modified for the analysis of triangular lattice rod bundles. In view of maximizing the thermal margin, the geometry of a hexagonal fuel assembly, such as rod diameter and rod pitch, has been optimized with a fixed fuel assembly cross sectional area The optimum value of the moderator-to-fuel volume ratio is estimated to lie between 0.65 to 1 with 9.5 mm rod diameter. The thermal margin of these hexagonal fuel assemblies in the AP600 core has been evaluated and compared with that of square lattice fuel assemblies such as VANTAGE-5H and KOFA. The analysis result shows that the performances of hexagonal fuel assemblies are more favorable than the square fuel assemblies in the aspect of steady-state overpower margin.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.155-160
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• 1984

Analyzed is the thermal margin of the Korea Nuclear Unit 1 (KNU-1) reactor core consisting of either 14 x 14 standard fuel assemblies (SFA) or optimized fuel assemblies (OFA). Employed for the analysis are two different thermal design methods; traditional and statistical thermal design method. Compared to the traditional design thermal method, the statistical thermal design method improves the core thermal margin utilizing best-estimate values for the core operating parameters combining their uncertainties in a statistical manner. Calculations are performed using a steady state and transient thermal-hydraulic analysis computer program, COBRA-IV-i. Calculated results show that the statistical thermal design method significantly improves the thermal margin and satisfies the core thermal design base of the KNU-1 SFA and OFA core. However, the thermal design base can not be met, if the traditional thermal design method is employed for the OFA role analysis.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.518-531
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• 1995

A study has been Peformed to investigate the thermal margin increase by replacing the single-channel analysis model with a multichannel analysis model. h new critical heat flux(CHF) correlation, which is applicable to a 17$\times$17 Korean Fuel Assembly(KOFA)-loaded core, was developed on the basis of the local conditions predicted by the subchannel analysis code, TORC. The hot sub-channel analysis was carried out by using one-stage analysis methodology with a prescribed nodal layout of the core. The result of the analysis shooed that more than 5％ of the thermal margin can be recovered by introducing the TORC/KRB-1 system(multichannel analysis model) instead of the PUMA/ERB-2 system(single-channel anal)sis model). The thermal margin increase was attributed not only to the effect of the local thermal hydraulic conditions in the hot subchannel predicted by the code, but also to the effect of the characteristics of the CHF correlation.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.58 no.12
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• pp.2311-2315
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• 2009

This paper presents a method to apply energy margin for assesment of total transfer capability (TTC). In order to calculate energy margin, two values of the transient energy function have to be computed. The first value is transient energy that is the sum of kinetic and potential energy at the end of fault. The second is critical energy that is potential energy at controlling UEP(Unstable Equilibrium Point). It is seen that TTC level is determined by not only bus voltage magnitudes and line thermal limits but also transient stability. TTC assessment is compared by the repeated power flow(RPF) method and optimization method.

• International Journal of Aerospace System Engineering
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• v.9 no.2
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• pp.1-9
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• 2022

The NEA release mechanism is used to provide restraint and release functions with low shock for critical deployment operations on solar arrays after launch. The GK3 solar array consists of 2 wings and 6 hold down points per panel. The NEA SSD9103S1 is a part of the GK3 solar array hold-down and release mechanism. Each NEA unit is equipped with two Z-diodes which provide power to a NEA unit connected in series after actuation of the fuse wire. This paper presents the hot firing test results of a quadrature NEA SSD9103S1 configuration. One output powers a maximum of 4 NEA SSD9103S1 units simultaneously. The necessary actuation pulse duration has been determined to meet margin requirement for thermal energy of minimum 4. Actuation thermal energy difference is about 6.6% between each half of two fired serial NEAs. Thermal energy margin at worst case is minimum 5.9 in case of an actuation pulse duration of 500 ms. Two series Zener impedance depend on current applied has been characterized by an additional actuation after all fuse wires are open circuit. Total number of actuation commands to the GK3 NEA unit reduce drastically from 24 in case of single NEA configuration down to 8 in case of parallel and quadrature NEA configurations. It can be accommodated by the existing HP2U Pyro design without any impact.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.45-57
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• 1990

A steady-state margin comparison study was performed between analog and digital protection systems. The systems compared are the thermal overpower and overtemperature delta T system of Westinghouse, and Core Protection Calculator System of Combustion Engineering, Inc. No dynamic offset was considered to eliminate the margin differences by different safety analysis methodologies. The result shows that the digital protection system has about 30% more rated power margin than the analog system in protecting against the fuel rod centerline melting. The digital protection system is shown to have almost same margin with the analog protection system in preventing the DNB at EOC (End of Cycle) even if the digital protection system has about 10% more margin at BOC(Beginning of Cycle).

• Progress in Superconductivity
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• v.8 no.2
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• pp.193-201
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• 2007

Since the margins for the minimum quench energy and for the operating current in the superconducting magnet determine the operating regime of the magnet, a thermal stability analysis for the KSTAR superconducting magnet system is performed using 1-D Gandalf code. The result shows that the minimum quench energy is about 500 mJ/cc and the operating current margin is about 70 %. These values are larger than those of the KSTAR design criteria and the KSTAR superconducting magnet system can be operated stably under various experimental environments.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.374-379
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• 1997

The thermal margin of CANDU-6 reactor is estimated by the CCP, which is dependent on fuel channel hydraulics and the CHF of fuel bundle. This paper intents to describe the characteristics of CCP behavior for the CANDU-6 channel in which CANFLEX-NU fuel bundles are assumed to be loaded. Also, it includes the thermal margin evaluation of the CANDU-6 channel loaded with a mixed CANFLEX-NU and 37-element fuel bundles as a simulation of the partial loading of CANFLEX-NU fuel bundle in the CANDU-6 reactor. For the mixed fuel channels, the effects of axial flux distribution(AFD) on CCP were investigated by using the AFD tilted in the downstream. The CCP of CANFLEX-NU fuel bundle was found to be improved by the CHF enhancement, despite of the slight flow decrease, in case of both full and partial loading, compared with those of a standard 37-element fuel bundle.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.684-701
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• 2016

The Gaussian process model (GPM) is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU) and Level 1 probabilistic safety assessment (PSA) success criteria definitions while dealing with a large number of uncertainties.