Thermal Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

THERMAL HYDRAULIC EXPERIMENT TO TEST THE STABLE OPERATION OF A PIUS TYPE REACTOR
Ign. Djoko Irianto ¹⁾, T.Kanji ²⁾, Y.Kukita ^3)
1) Center for Nuclear Technology Assessment - BATAN
²⁾ Department of Nuclear Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya464-01 JAPAN
³⁾ Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki 319-11 JAPAN
ABSTRACT
THERMAL HYDRAULIC EXPERIMENT TO TEST THE STABLE OPERATION OF A PIUS TYPE REACTOR. An advanced type of reactor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal hydraulic test loop which simulated the PIUS principle. The experiments such as: start up and power ramping tests for normal operation simulation and a loss of feedwater test for an accident condition simulation, carried out in JAERI.

Keywords : Thermal-Hydraulic Experiment, PIUS type Reactor, Passive Safety, Pump Speed Control System, Density Lock, Start-Up Simulation Test, Power Ramping Test, Loss of Feedwater Test
Published : Proceeding "Pertemuan dan Presentasi Ilmiah Penelitian Dasar Ilmu Pengetahuan Dan Teknologi Nuklir", Yogyakarta, 25-27 Oktober 1995, ISSN 0216-3128

Read More

THERMAL-HYDRAULIC ANALYSIS OF THE NUCLEAR BATTERY COOLANT SYSTEM

THERMAL-HYDRAULIC ANALYSIS OF THE NUCLEAR BATTERY COOLANT SYSTEM
Ign. Djoko Irianto, Budi S., Sarwo D.Danupoyo, Ahmad Syaukat, Budi Santoso
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
THERMAL-HYDRAULIC ANALYSIS OF THE NUCLEAR BATTERY COOLANT SYSTEM. Thermal-hydraulic analysis has been done for the coolant system of the nuclear battery, which utilizes heat pipes as the primary coolant system for heat removal from the reactor core to the secondary coolant system. This paper analysis the thermal hydraulic aspect of the nuclear battery coolant system by constructing the design model for heat pipe, as the primary coolant system, and analyzing the energy balance on the secondary coolant system. The model refers to the nuclear battery operated at a power of 2400 kW(t) and nominal core graphite temperature of 550 oC. The wrapped screen wick type heat pipe 3 m length and 5 cm diameter with potassium as working fluid has a maximum axial heat flow of 102957 W at operating temperature 482 oC. Using toluene as working fluid at maximum temperature of 370 oC the secondary coolant system equipped with a regenerator has a thermal efficiency of 26 %. The nuclear battery with capacity of 2400 kW(t) requires 24 heat pipes.

Keywords : Thermal-Hydraulic Analysis, Nuclear Battery, Primary coolant System, Heat Pipe, Thermal Efficiency
Published : Proceeding "Seminar Seperempat Abad Reaktor Nuklir Mengabdi Ilmu Pengetahuan dan Teknologi", Bandung, 16-17 Oktober 1989, ISSN No. : 1410-1769

Read More