• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.23 no.7
• /
• pp.831-840
• /
• 1999

In this work, an aerothermodynamic calculation model for cooled axial flow turbine blades with trailing edge ejection is suggested and a mean line performance analysis of a turbine stage with nozzle cooling is carried out. A unique model regarding the interaction between coolant and main gas is proposed, while existing correlations are adopted to predict viscous loss and blade outflow angle. The interactions considered are the heat transfer from main gas to coolant and the temperature and pressure losses by the mixing of two streams due to the trailing edge coolant ejection. For a stator blade without ejection, trailing edge loss calculated by the trailing edge analysis is compared with that calculated by loss correlation. The effect of heat transfer effectiveness of coolant passage on the mixing loss is analyzed. For a model turbine stage with nozzle cooling, parametric analyses are carried out to investigate the effect of main design variables(coolant mass flow ratio, temperature and ejection area) on the stage performance.

• Nuclear Engineering and Technology
• /
• v.55 no.7
• /
• pp.2438-2446
• /
• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.

• Nuclear Engineering and Technology
• /
• v.26 no.2
• /
• pp.179-189
• /
• 1994

A performance of reactor coolant pump in two-phase flow is examined using the pump geometric conditions and the performance of the pump in single-phase flow. Wall friction loss of the reactor coolant pump in single-phase flow is prdicted using the Truckenbrodt boundary layer theory, and the head loss in two-phase flow is predicted with calculated well friction loss and separation loss coefficients. The analysis results are compared with the Combustion Engineering pump test data. The effect of two-phase multiplier on the peak clad temperature in Loss-of-Coolant Accident is also examined using the RELAP5 and the results indicate the importance of its accuracy.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1532-1539
• /
• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• Nuclear Engineering and Technology
• /
• v.53 no.2
• /
• pp.474-483
• /
• 2021

To establish the exclusive role of hydrogen on burst behaviour of Zircaloy-4 during loss-of-coolant accident transients, an extensive single-rod burst tests were conducted on both unirradiated as-received and hydrogenated Zircaloy-4 cladding tubes at different heating rates and internal overpressures. The visual observations of cladding tubes during bursting as well as post-burst are presented in detail to understand the effect of hydrogen concentration, heating rate, and internal pressure. Impact of hydrogen on burst parameters-burst stress, burst strain, burst temperature-during loss-of-coolant accident transients are compared and discussed. Rupture at multiple locations for hydrogenated cladding at lower internal pressure and higher heating rate is reported for the very first time. A novel burst criterion accounting hydrogen concentration in nuclear fuel cladding is proposed.

• Nuclear Engineering and Technology
• /
• v.49 no.5
• /
• pp.968-978
• /
• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.812-817
• /
• 1998

Experimental studies on heat transfer and solidification of the molten metal pool with overlying coolant with boiling were performed The simulant molten pool material is tin (Sn) with the melting temperature of 232$^{\circ}C$. Demineralized water and R113 are used as the working coolant. This work examines the crust formation and the heat transfer characteristics of the molten metal pool immersed in the boiling coolant. The Nusselt number and the Rayleigh number in the molten metal Pool region of this study are compared between the water coolant case and the R113 coolant case. The experimental results or the water coolant are higher than those for R113. Also, the empirical relationship of the Nusselt number and the Rayleigh number is compared with the literature correlations measure from mercury. The present experimental results are higher than the literature correlations. It is believed that this discrepancy is caused by the effect of heat loss to the environment on the natural convection heat transfer in the molten pool.

• Journal of Energy Engineering
• /
• v.26 no.3
• /
• pp.71-77
• /
• 2017

In the event of loss of coolant accident (LOCA) and station black out (SBO) in the primary system of a nuclear reactor, the coolant water should be injected to reactor coolant system (RCS) without any intervention of operators or active components. To satisfy the requirements, hybrid safety injection tank (Hybrid SIT) was suggested by Korea Atomic Energy Research Institute (KAERI). The pressure equalizing time of Hybrid SIT is an important parameter to determine the timing of coolant injection. To predict the pressure equalizing time of the Hybrid SIT, a separate effect test facility was constructed and sensitivity tests were conducted in various conditions. The main parameter determining the pressure equalizing time was obtained from separate effect test (SET) results. The wall of condensation on the inner wall of SIT and direct contact condensation on the water surface affected to the pressure equalizing time very much. In this study, the effect of each condensation phenomena on pressure equalizing time was quantitatively analyzed from results of SET and a prediction method of pressure equalizing time was proposed.

• Nuclear Engineering and Technology
• /
• v.39 no.4
• /
• pp.257-272
• /
• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

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