• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.160.2-160
• /
• 1999

• Proceedings of the Korean Nuclear Society Conference
• /
• 2003.10a
• /
• pp.81.2-81.2
• /
• 2003

• The KSFM Journal of Fluid Machinery
• /
• v.19 no.2
• /
• pp.49-54
• /
• 2016

A conceptual study of an air-cooled heat exchanger is conducted to achieve the long-term passive cooling of an integral reactor. A newly designed air-cooled heat exchanger is introduced in the present study and preliminary thermal sizing is demonstrated. This study mainly focuses on feasibility of an innovative air-cooled heat exchanger to extend the cooling period of the passive residual heat removal system(PRHRS) only in passive manners. A vertical shell-and-tube air-cooled heat exchanger is installed at the top of the emergency cooldown tank(ECT) to collect evaporated steam into condensate, which enables water inventory of the ECT to be kept. Finally, thermal sizing of an air-cooled heat exchanger is presented. The length and the number of tubes required, and also the height of a stack are calculated to remove the designated heat duty. The present study will contribute to an enhancement of the passive safety system of an integral reactor.

• Nuclear Engineering and Technology
• /
• v.55 no.12
• /
• pp.4335-4349
• /
• 2023

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 ℃/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.

• Journal of Energy Engineering
• /
• v.25 no.4
• /
• pp.124-132
• /
• 2016

In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.

• Nuclear Engineering and Technology
• /
• v.41 no.8
• /
• pp.1015-1024
• /
• 2009

The presence of a non-condensable gas can considerably reduce the level of condensation heat transfer. The non-condensable gas effect is a primary concern in some passive systems used in advanced design concepts, such as the Passive Residual Heat Removal System (PRHRS) of the System-integrated Modular Advanced ReacTor (SMART) and the Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR). This study examined the capability of the Multi-dimensional Analysis of Reactor Safety (MARS) code to predict condensation heat transfer in a vertical tube containing a non-condensable gas. Five experiments were simulated to evaluate the MARS code. The results of the simulations showed that the MARS code overestimated the condensation heat transfer coefficient compared to the experimental data. In particular, in small-diameter cases, the MARS predictions showed significant differences from the measured data, and the condensation heat transfer coefficient behavior along the tube did not match the experimental data. A new method for calculating condensation heat transfer coefficient was incorporated in MARS that considers the interfacial shear stress as well as flow condition determination criterion. The predictions were improved by using the new condensation model.

• Nuclear Engineering and Technology
• /
• v.45 no.3
• /
• pp.347-360
• /
• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
• /
• v.45 no.4
• /
• pp.439-458
• /
• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.