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

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SMALL SCALE THERMAL-HYDRAULIC EXPERIMENT FOR STABLE OPERATION OF A PIUS-TYPE REACTOR

SMALL SCALE THERMAL-HYDRAULIC EXPERIMENT FOR STABLE OPERATION OF A PIUS-TYPE REACTOR
K.Tasaka¹⁾, M.Tamaki¹⁾, S.Imai¹⁾, Ign. Djoko Irianto¹⁾, Y.Tsuji¹⁾, Y.Kukita^2)
1) Department of Nuclear Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya464-01 JAPAN
²⁾ Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki 319-11 JAPAN
ABSTRACT
Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiment. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor.

Keywords : Pump speed control system, density lock, PIUS-type reactor, start-up test, stable operation
Published : Proceedings of "Seventh International Conference on Emerging Nuclear Energy Systems", Makuhari, Chiba, Japan, 20-24 September 1993

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THERMAL-HYDRAULIC EXPERIMENT FOR SAFE AND STABLE OPERATION OF A PIUS-TYPE REACTOR

THERMAL-HYDRAULIC EXPERIMENT FOR SAFE AND STABLE OPERATION OF A PIUS-TYPE REACTOR
K.Tasaka¹⁾, S.Imai¹⁾, H.Masaoka¹⁾, Ign. Djoko Irianto¹⁾, H.Kohketsu¹⁾, M.Tamaki¹⁾, Y.Anoda²⁾, H.Murata²⁾ and Y.Kukita^2)
1) Department of Nuclear Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya464-01 JAPAN
²⁾ Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki 319-11 JAPAN
ABSTRACT
A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to 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 experiment such as start-up and power ramping tests for normal operation simulation and a loss of feedwater test for an accident condition simulation, using a small scale atmospheric pressure test loop which simulated the PIUS principle.

Keywords : Pump speed control system, density lock, PIUS-type reactor, start-up test, power ramping test, normal operation simulation, accident condition simulation
Published : Proceedings of "International Conference on Design and Safety of Advanced Nuclear Power Plant", Tokyo, Japan, October 25-29, 1992

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LOSS OF FLOW TRANSIENT SIMULATION IN PIUS REACTOR USING RELAP5/MOD2

LOSS OF FLOW TRANSIENT SIMULATION IN PIUS REACTOR USING RELAP5/MOD2
Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
LOSS OF FLOW TRANSIENT SIMULATION IN PIUS REACTOR USING RELAP5/MOD2. Accident transient due to the loss of flow in the Process Inherent Ultimate Safety (PIUS) reactor have been simulated. This simulation constitute a part of the safety analysis in the advanced nuclear reactor design which inherently safe such as PIUS reactor having the thermal power of 2000 MWth (PIUS 2000). The attention of analysis has been focused on the accidental transient of loss of flow due to the loss of power supply to the primary pumps that can result from a loss of off site power. Simulation has been done in Nagoya University of Japan using computer code RELAP5/MOD2. The effect of heat structures on the coolant flow rate through the core during the phase of natural circulation through density lock and borated water pool has been analyzed. The simulation results showed that the cyclic behavior of natural circulation through the pool, density lock and riser (core) will be occurred due to the trip of the primary pump.

Keywords : PIUS type Reactor, Primary Pump Speed, Pressure Balance, Transient Simulation Test, Density Lock, RELAP5/MOD2.
Published : Proceeding "Pertemuan dan Presentasi Ilmiah Penelitian Dasar Ilmu Pengetahuan dan Teknologi Nuklir", Yogyakarta, 23-25 April 1996, ISSN No. : 0216-3128

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FEEDBACK CONTROL SYSTEM OPTIMIZATION OF THE PRIMARY PUMP FOR PIUS REACTOR

FEEDBACK CONTROL SYSTEM OPTIMIZATION OF THE PRIMARY PUMP FOR PIUS REACTOR
Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
FEEDBACK CONTROL SYSTEM OPTIMIZATION OF THE PRIMARY PUMP FOR PIUS REACTOR. Thermal hydraulic experiments using a small scale thermal-hydraulic test loop have been performed for the Process Inherent Ultimate Safety (PIUS) reactor simulation to develop the pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments that is carried out in JAERI Japan. Each of the feedback control systems had been verified in the simulation experiment such as a start up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective and optimal to maintain the stratified interface between hot primary water and cool poison water for stable operation of the PIUS reactor.

Keywords : Feedback Control System, PIUS type Reactor, Primary Pump Speed, Pressure Balance, Start-Up Simulation Test, Density Lock.
Published : Proceeding "Seminar Sains dan Teknologi Nuklir, Menyongsong Reaktor Triga Mark II Bandung 2 MegaWatt Sebagai Sarana Peningkatan Mutu Litbang Iptek", Bandung, 12-13 Maret 1996, ISSN No. : 1410-1769

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BORON DILUTION IN THE PRIMARY COOLANT INFLUENCE TO THE FUEL CHARACTERISTIC OF PIUS REACTOR

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

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FEEDBACK CONTROL SYSTEM ANALYSIS FOR PIUS TYPE REACTOR USING RELAP5/MOD2

FEEDBACK CONTROL SYSTEM ANALYSIS FOR PIUS TYPE REACTOR USING RELAP5/MOD2
Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
FEEDBACK CONTROL SYSTEM ANALYSIS FOR PIUS TYPE REACTOR USING RELAP5/MOD2. A reactor concept such as Process Inherent Ultimate Safety (PIUS) is a modification of Pressurized Water Reactor (PWR) in which the primary coolant system is submerged in a pool of poison water. Operating principle of this reactor is to maintain the pressure balance along the density lock which be a connection between the primary coolant system and the pool, so that no inflow through density lock. The pressure balance could be maintaned by controlling the primary pump speed using feedback control system. This feedback control system was simulated by experimental using thermal hydraulic test loop and by numerical simulation using RELAP5/MOD2. Numerical simulation is done by setting the thermal hydraulic test loop nodalization based on RELAP5/MOD2 norm input. The system nodalization is composed of 119 volumes, 127 junctions, and 106 heat structures. Analysis is carried out based on the result of experimental and numerical simulation. The two simulations gave some particularly parameters such as the primary pump speed, the primary flow rate, the poison water flow rate, and the primary coolant temperatures. Conclusion from the two simulation results, that is, feedback control system to the primary pump speed could be introduced to the operation of the PIUS reactor.

Keywords : Feedback Control System, PIUS type Reactor, Primary Pump Speed, RELAP5/MOD2, Pressure Balance, Numerical Simulation, Density Lock.
Published : Proceeding "Seminar Sains dan Teknologi Nuklir, Peningkatan Keselamatan Reaktor Triga Mark II Bandung", Bandung, 19-20 Maret 1997, ISSN No. : 1410-1769

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LOSS OF SECONDARY COOLANT ACCIDENT ANALYSIS FOR PIUS TYPE REACTOR USING RELAP5/MOD2

LOSS OF SECONDARY COOLANT ACCIDENT ANALYSIS FOR PIUS TYPE REACTOR USING RELAP5/MOD2

Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id

ABSTRACT
LOSS OF SECONDARY COOLANT ACCIDENT ANALYSIS FOR PIUS TYPE REACTOR USING RELAP5/MOD2. Process Inherent Ultimate Safety (PIUS) reactor concept is a reactor concept that intrinsically based on passive safety. This reactor refer to Pressurized Water Reactor (PWR) wherein the primary system is submerged in a pool of poison water. The operating principle is to maintain the pressure balance, so that no inflow from pool to the primary system. On loss of secondary coolant accident, primary coolant temperature increases, it is followed by the increase of primary pump speed. When the upper limit is reached, the pump is tripped. Due to the pressure balance disturbance, poison water flows from pool to the primary system, then reactor shut down. This accident condition was simulated by experimental and numerical simulation using RELAP5/MOD2. Numerical simulation was done to the experimental apparatus nodalization that was set on the norm of RELAP5/MOD2. This nodalization consist of 119 volumes, 127 junctions, and 106 heat structures. Analysis was carried out using both experimental and numerical simulation results. It can be concluded that PIUS type reactor is able to anticipate loss of coolant accident because its capability of self shut down.

Keywords : LOCA, PIUS type Reactor, RELAP5/MOD2, PWR.
Published : Proceeding of "Komputasi Dalam Sains dan Teknologi Nuklir VI", Jakarta, 16-17 Januari 1997.

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