• Journal of Energy Engineering
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• v.13 no.1
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• pp.51-59
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• 2004

The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited compulsational capability at that time, they usually used very simplified physical models for a real-time simulation of NSSS thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called "negative training", especially for complicated two-phase flows in the reactor coolant system. In resolve the problem, KEPRI developed a realistic NSSS T/H program ARTS which was based on the RETRAN-3D code for the improvement of the Nuclear Power Plant full-scope simulator. The ARTS (based on the RETRAN-3D code) guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there is some possibility of failing to calculate in the case of large break loss of coolant accident (LBLOCA) and low-pressure low-flow transient. In this case, the backup calculation system cover automatically the ARTS. The backup calculation system was expected to provide substantially more accurate predictions in the analysis of the system transients involving LBLOCA. The results were reasonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with FSAR and the AMSI/ANS-3.5-1998 simulator software performance criteria.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.821-827
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• 2001

As one of the advanced design features of the Korea next generation reactor, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLl) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, a new scaling method, using time and velocity reduced linear scaling law, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in PWR downcomer.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.421-432
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• 2002

As one of the advanced design features of the APR1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, test results of direct ECC bypass performed in the steam-water test facility tailed MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) are presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’with the l/4.93 length scale . From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region are obtained.

• 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.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.481-488
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• 2014

An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks' formula. The uncertainty range and distribution of each input parameter associated with the LOCA phenomena were determined based on previous PIRT results and documentation during the BEMUSE project. Calulations were conducted on 3,500 cases within a 2-week CPU time on a 14-PC cluster system. The Monte-Carlo exercise shows that the 95% upper limit PCT value can be obtained well, with a 95% confidence level using the Wilks' formula, although we have to endure a 5% risk of PCT under-prediction. The results also show that the statistical fluctuation of the limit value using Wilks' first-order is as large as the uncertainty value itself. It is therefore desirable to increase the order of the Wilks' formula to be higher than the second-order to estimate the reliable safety margin of the design features. It is also shown that, with its ever increasing computational capability, the Monte-Carlo method is accessible for a nuclear power plant safety analysis within a realistic time frame.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.979-988
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• 2017

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.265-276
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• 1994

Several improvements to the RELAP5/MOD3 reflood model hate been made. These improvement were made to correct deficiencies in the reflood model identified by the assessment of the RELAP5/MOD3 code against FLECHT-SEASET experiments. The improvements consist of modification of reflood wall heat transfer package and adjusting the droplet size in dispersed flow regime. The time smoothing of wall vaporization and level tracking of transition flow are also added to eliminate the pressure spikes and level oscillation during reflood process. Assessment of the improved model against FLECHT-SEASET experimental data and application of LBLOCA analysis for plant shows that the deficiencies have been corrected.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.742-754
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• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.