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
RADIATION EXPOSURE ESTIMATION AROUND THE REACTOR DUE TO LOSS OF COOLANT ACCIDENT (LOCA) ON ABWR
Ign. Djoko Irianto and Sarwo D.Danupoyo
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
RADIATION EXPOSURE ESTIMATION AROUND THE REACTOR DUE TO LOSS OF COOLANT ACCIDENT (LOCA) ON ABWR. Loss of coolant accident in a nuclear reactor is an accident which is assumed be initialized by a leakage at the primary coolant piping or by a leakage at the reactor vessel. In this accident, the coolant will be released and can be followed by some radioactive fission product release to the primary containment, secondary containment and to the environment. Beside to the environment which be a risk to the public, fission product released also occurred to the control room that will be a risk to the operator. Releasing process, pathway, activity and the exposure dose to the environment and also to the control room are assessed and estimated in this paper. The activity which be released in the primary containment at the first two hours are 2.21E07 Ci for iodine and 3.43E08 Ci for noble gas. The increasing of the environment activity and in the control room are 3.1E04 Ci and 2.0E03 Ci for iodine and 7.0E06 Ci and 1.9E07 Ci for noble gas, respectively.
Keywords : Radiation Exposure, Loss of Coolant Accident, ABWR, Primary Coolant
Published : Proceeding "Presentasi Ilmiah Keselamatan Radiasi dan Lingkungan", Jakarta, 20-21 Agustus 1996, ISSN No. : 0854-4085
BORON DILUTION IN THE PRIMARY COOLANT INFLUENCE TO THE FUEL CHARACTERISTIC OF PIUS REACTOR
Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
BORON DILUTION IN THE PRIMARY COOLANT INFLUENCE TO THE FUEL CHARACTERISTIC OF PIUS REACTOR. The influence of boron dilution in the primary coolant to the fuel characteristic of Pressurized Water Reactor (PWR) have been studied as a preliminary study of fuel characteristic of Process Inherent Ultimate Safety (PIUS) reactor. The focus of study are fuel bundle with several enrichment of U-235 from 2.4% until 3.4% using boron dilution concentration in the primary coolant are 150 ppm and 200 ppm. The calculation result of the fuel bundle which is set as a matrix of 17x17 element and the dimension is 30.294 cm x 30.294 cm using WIMS-D4 personal computer (PC) version, shows that the value of infinite multiplication factor of the fuel bundle which 2.4% enrichment in the primary coolant with 200 ppm boron concentration is 1.01672. In general, it is concluded that the fuel bundle which enrichment of 2.4% and the concentration of boron dilution of 200 ppm, the PIUS reactor could be achieve the critical or could be able to start up.
Keywords : Boron Dilution, Primary Coolant, Fuel Characteristic, WIMS-D4, Start Up Reactor.
Published : Journal "Pengkajian Sains dan Teknologi Nuklir", Vol.3, No.1, 1998, Pusat Pengkajian Teknologi Nuklir, BATAN, ISSN No. : 0852-8047
PRIMARY SYSTEM DEPRESSURIZATION ACCIDENT ANALYSIS IN THE HTTR
Ign. Djoko Irianto, Sarwo D.Danupoyo, Sahala M.Lumbanraja
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
PRIMARY SYSTEM DEPRESSURIZATION ACCIDENT ANALYSIS IN THE HTTR. The behavior of the HTTR (High Temperature Testing Reactor) during depressurization accident caused by a primary pipe rupture was analyzed in a safety analysis of HTTR. The analysis was performed using TAC-NC computer code which is a thermal hydraulics code used to calculate thermal transient, including natural circulation. This paper describes analytical model, analytical condition and analytical results during the depressurization accident. The analytical results proved that thermal transient behavior during depressurization accident is slower than that of the PWR (Pressurized Water Reactor) for the similar accident. It also proved that the maximum fuel temperature does not exceed the normal operation temperature 1495 C, and the maximum pressure vessel temperature would remain below its limit of 550 C.
Keywords : Primary System, Accident Analysis, HTTR, Thermal Transient.
Published : Proceeding "Seminar dan Lokakarya ke-3 Teknologi dan Aplikasi Reaktor Temperatur Tinggi", Jakarta, 3-4 Juni 1996, ISSN No. : 0854-8803
