COMPARISON OF NUCLEAR HYDROGEN PRODUCTION BETWEEN SULFUR-IODINE CYCLE OF THERMOCHEMICAL AND NATURAL GAS STEAM REFORMING PROCESS
Djati H. Salimy, Ida N. Finahari
Pusat Pengembangan Energi Nuklir (PPEN) BATAN
Jl. Abdul Rohim Kuningan Barat, Mampang Prapatan
ABSTRACT
COMPARISON OF NUCLEAR HYDROGEN PRODUCTION BETWEEN SULFUR-IODINE CYCLE OF THERMOCHEMICAL AND NATURAL GAS STEAM REFORMING PROCESS. Paper describes comparison of nuclear hydrogen production for two technology processes: thermochemical of sulfur-iodine cycle and steam reforming of natural gas. The goal of the study is to understanding production characteristic of each processes. The comparison is analyzed from the point of advantages and disadvantages, thermal efficiency, and technology statues. Steam reforming of natural gas is the proven technology, while thermochemical process is still in the stage of research and development. Thermal efficiency of steam reforming (70-76%) is about three time of electrolysis (47-52%). Preliminary estimation of production cost also showed that steam reforming is cheaper. However, from the point of raw material, thermochemical is more advantage since the unlimited and renewable raw material of water, promising the process of hydrogen production without CO2 emission. While, steam reforming depend on non renewable raw material of natural gas. For nuclear application, test production of nuclear steam reforming has been going on since the mid of 2010 and will soon be operated by 2015. Couple thermochemical process with nuclear, will conducted in the end of 2010, hope be operated by 2025. For commercial operation both of the processes still wait the commercialization of HTGR.
Keywords: steam reforming, thermochemical, thermal efficiency, HTGR
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
STUDY ON HELIUM TURBINE FOR SECONDARY COOLANT SYSTEM OF MOLTEN SALT REACTOR
Sri Sudadiyo
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
STUDY ON HELIUM TURBINE FOR SECONDARY COOLANT SYSTEM OF MOLTEN SALT REACTOR. From the viewpoint of energy system and environment, concept for molten salt reactor (MSR) is one of advanced nuclear reactors which have good potential for electricity generation device. Within MSR, molten salt fuel flows through graphite core channels, to produce thermal neutron. The obtained heat of nuclear fuel was transferred to secondary coolant system through the heat exchanger using closed cycle of helium turbine. The resulted hot helium gas was expanded to the turbine for getting power. This study purposed to determine the performance of helium turbine as main components of secondary coolant cycle in MSR. The applied parameter was pressure ratio, specific heat ratio, and temperature. By placing both of helium turbine and compressor at single shaft, it was obtained approximate 49 % from turbine power output for driving compressor and the residual power to turn on electricity generator. The yielded turbine adiabatic efficiency is 85 % and it is able to improve thermal efficiency for secondary coolant system of MSR.
Keywords : Turbine, efficiency, helium
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
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
