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
HEAT TRANSFER THROUGH TWO PHASE FLOW ON THE POROUS MATERIAL
Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
The experiment using porous heat pipe model with internal heat generation has been done to study the heat transfer characteristic on the porous material. Heat pipe model comprises a quartz tube with 300 mm length, 20 mm inner diameter and 3.5 mm thickness. Porous media were simulated using small steel balls. Working fluid which is used are water, propanol, and octane. Internal heat generation was simulated using high frequency induction. The experiment results showed that if boils have not occurred, heat transfer occurred by conduction process. Using the higher power, two phase flows will occur so that the heat transfer coefficient will increase.
Keywords : heat pipe, porous material, heat transfer, two phase flow
Published : Proceedings "XVIth National Symposium On Physics And ASEANIP Regional Seminar On The Physics Of Metals and Alloys", Bandung, December 12-14, 1996, ISBN 979-8580-14-1
RELIABILITY STUDY OF THE ABWR SAFETY SYSTEM TO THE ABNORMAL TRANSIENT
Sarwo D.Danupoyo and Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
RELIABILITY STUDY OF THE ABWR SAFETY SYSTEM TO THE ABNORMAL TRANSIENT. Abnormal transient is an event that always happens to all equipments made by human being including NPP (Nuclear Power Plant). Disruptions of the equipments, miss operation or loss of site power are the cause of the abnormal transient. In order to reduce the effects of abnormal transient and to avoid the accident, the validity of NPP safety system design must be confirmed. In this paper, the reliability study of the ABWR-type-NPP safety system that was recently constructed in Kashiwazaki-Kariwa Japan is discussed. The study was carried out by learning the results of the ABWR safety system tests by computer simulation in Japan to overcome the abnormal transient conditions. The results show that the design of the ABWR safety system is reliable enough to overcome the abnormal transient
Keywords : Safety System, Abnormal Transient, Computer Simulation, Accident Risk, ABWR
Published : Proceeding "Seminar ke-III Teknologi dan Keselamatan PLTN Serta Fasilitas Nuklir", Serpong, 5-6 September 1995, ISSN No. : 0854-2910
ALTERNATIF PENYEDIAAN ENERGI LISTRIK ABAD 21 DI INDONESIA
Arnold Y.Sutrisnanto and Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRAK
Akhir abad 20 ditandai dengan peningkatan industrialisasi di semua negara, baik di negara maju maupun di negara yang sedang berkembang. Laju kegiatan industri akan selalu diikuti oleh peningkatan konsumsi energi listrik. Padahal pemakaian beberapa sumber energi dunia yang utama pada saat ini akan jauh berkurang pada awal abad 21. Untuk itu perlu dicarikan jalan pemecahannya yang antara lain dengan melakukan penghematan energi dan substitusi energi pengganti. Dari sekian banyak energi pengganti baik yang terbarukan maupun yang tidak terbarukan, hanyalah pemakaian bersama energi batubara dan energi nuklir yang nampaknya dapat dipertanggungjawabkan dari segi densitas energi, faktor ekonomi, kontinuitas persediaan bahan bakar, kontinuitas pembangkitan dan problem ekologinya. Dengan demikian pemakaian bersama energi batubara dan energi nuklir dapat diharapkan mampu mengatasi lonjakan kebutuhan energi abad 21 khususnya dalam era perdagangan bebas mendatang yang akan dimulai pada awal abad 21.
Keywords : National Industry, domestic participation
Published : Proceedings "Hasil-Hasil Lokakarya Energi 1995", Jakarta, 25-27 Juli 1995
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.Kukita2)
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
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.Kukita2)
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
THE ROLE OF NATIONAL INDUSTRY TO COMMEMORATE BUILDING NPP IN INDONESIA
N.P.Ginting, M.S.Kasim, M.T.Razak, A.Syaukat, R.Setiadi, Ign. Djoko Irianto, Puradwi, J.Situmorang
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
THE ROLE OF NATIONAL INDUSTRY TO COMMEMORATE BUILDING NPP IN INDONESIA. The need for energy in Indonesia keeps increasing steadily and will reach a value of 27,000 MWe by the year 2015. Conventional sources of energy will not be able to cope with this demand. Nuclear energy is considered favourably as an alternate energy source. BATAN in cooperation with sofratome has made surveys on possible participation of local industry in building PLTN (Nuclear Power Plant) in Indonesia. In the present paper the results of this survey together with evaluation of possible participation from the domestic industry will be reported.
Keywords : National Industry, domestic participation
Published : Proceedings "Seminar Pendayagunaan Reaktor Nuklir Untuk Kesejahteraan Masyarakat", Bandung, 26-27 September 1990, ISSN No. : 1410-1769
STUDY OF VERIFICATION TECHNIQUES FOR NUCLEAR MATERIAL SAFEGUARDS AND SECURITY
Ign. Djoko Irianto
Center for Reactor Technology and Nuclear Safety � BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan
Email: igndjoko@yahoo.com
ABSTRACT
STUDY OF VERIFICATION TECHNIQUES FOR NUCLEAR MATERIAL SAFEGUARDS AND SECURITY. The presence of nuclear materials at any nuclear facilities must be known as safeguards purpose through the knowledge of position, form or type and the amount. The clarification of the position, the type and the amount must be reported to International Atomic Energy Agency (IAEA) as the international regulatory body. Then IAEA will verify that report. The verification must be done to know that there is no difference of the amount, and to give assurance to the international community that the nuclear material used only to non military purpose. To carry out the verification, several verification techniques such non-destructive analysis such as gamma spectrometry, neutron counting, surveillance technique, unattended and remote monitoring and environmental sampling are explained in this paper to give the impression how those techniques are implemented.
Keywords: Safeguards Technology, Verification Techniques, Nuclear Material, Non Destructive Analysis
Published : Journal of Nuclear Equipments, Volume 1, Nomor 1, Mei 2007, ISSN:1978-3515
NUCLEAR MATERIAL SAFEGUARDS AND SECURITY SYSTEM ANALYSIS BASED ON MEASUREMENT
Ign. Djoko Irianto
Center For Safeguards Technology � Batan
Kawasan Puspiptek Gedung 90 Lantai 3, Serpong � 15310
Email: igndjoko@yahoo.com
ABSTRACT
NUCLEAR MATERIAL SAFEGUARDS AND SECURITY SYSTEM ANALYSIS BASED ON MEASUREMENT. Nuclear material safeguards and security are the important aspect in the nuclear facility management due to the nuclear material could be terrorism object. The two aspect of nuclear material security are nuclear material safeguards system and physical protection system. The most important in safeguards system is how to report the existence of nuclear material and the quantity of nuclear material. To perform the safeguards system the data of nuclear material are needed. The data of quality and quantity of nuclear material could be found by destructive analysis (DA) technique and non destructive analysis (NDA) technique. The DA technique are used to analysis the nuclear material that forming in powder, the NDA technique are used to analysis the nuclear material in spent fuel. In BATAN, the technique of measurement of nuclear material weight is more dominant than the other technique to be used in nuclear material safeguards and security systems.
Keywords: Safeguards Technology, Physical Protection, Nuclear Material, Destructive Analysis, Non Destructive Analysis, Security Systems.
Published : A Scientific Journal "PRIMA : Aplikasi dan Rekayasa Dalam Bidang Iptek", Volume 4, Nomor 8, Nopember 2007, ISSN:1411-0296.
ABWR SAFETY SYSTEM ASSESSMENT FOR LOCA ANTICIPATION
Ign. Djoko Irianto and Sarwo D.Danupoyo
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
ABWR SAFETY SYSTEM ASSESSMENT FOR LOCA ANTICIPATION. Loss of coolant accident (LOCA) is the accident which is assumed as the event that initiated from pipe rupture of primary cooling system. One of the risk of LOCA is the increase of fuel cladding temperature that will destroy the integrity of fuel cladding and the release of fission product from the fuel. Emergency core cooling system (ECCS) on BWR has been designed to anticipate the accident such as LOCA. ECCS modification on ABWR is emphasized to improve the ECCS performance. ECCS of ABWR is designed on 3 independent division that operated simultaneously and be able to repair on line automatically until 72 hours without any action of operator. Internal pumps utilization as the recirculation pump make possible to reduce large break LOCA probability due to rupture of primary coolant piping. In this case, the accident risk which could not disparage is the breach that occurred in the High pressure Core Flooder (HPCF) system. In case of such accident occurred the reactor core can be maintain still submerged.
Keywords : Safety System, Loss of Coolant Accident, ABWR, ECCS, HPCF
Published : Proceeding "Seminar ke-III Teknologi dan Keselamatan PLTN Serta Fasilitas Nuklir", Serpong, 5-6 September 1995, ISSN No. : 0854-2910
THE EFFECT OF CONTROL ROD DISPLACEMENT AGAINST THE SAFETY OF REACTOR CORE OF NUCLEAR BATTERY
Ign. Djoko Irianto, Budi Santoso, Sahala Lumban Raja, Ahmad Syaukat
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
THE EFFECT OF CONTROL ROD DISPLACEMENT AGAINST THE SAFETY OF REACTOR CORE OF NUCLEAR BATTERY. The Nuclear Battery reactor is small, solid-state passively cooled reactor power supply being developed to produce electricity and/or steam heat in remote locations. A fundamental design principles of the Nuclear Battery is a very high level of safety with no short-term intervention required by either human operators or engineered safety devices. One of the limiting hypothetical accident scenarios for a nuclear reactor is the rapid ejection of a single control rod from and initial state of reactor criticality. This paper presents a safety aspect of reactor core of the Nuclear Battery, and shows that the Nuclear Battery would survive a rapid reactivity insertion of a least 20 mk without compromising fuel integrity.
Keywords : Nuclear Battery, Safety Aspect, Reactor Core, Reactivity
Published : Proceeding "Pertemuan dan Presentasi Ilmiah Penelitian Dasar Ilmu Pengetahuan dan Teknologi Nuklir", Yogyakarta, 21-22 Maret 1990, ISSN No.: 0216-3128
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
THE APPLICATION OF DIGITAL TECHNIQUE IN THE RESEARCH REACTOR CONTROL SYSTEM
Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
THE APPLICATION OF DIGITAL TECHNIQUE IN THE RESEARCH REACTOR CONTROL SYSTEM. Research reactor is a multifunction nuclear reactor that provide neutron sources and gamma radiation for use in some activities. This wide range of activities necessitates frequent adjustments of a research reactor power. To increase the flexibility of operation with high degree of safety, the research reactor control system should be automated. The automation can only be done when the control system is change from analog to digital. This paper describes the model-based control systems for used in research reactor that was designed by period-generated control method. This control system is designed by combining feedback and feed forward control techniques that can be control non-linear system. The rate of change of reactivity is used as actuator signal. Performance test has been conducted in MITR-II owned by Massachusetts Institute of Technology and ACRR that is operated by Sandia National Laboratories. The result shows that the model-based digital control system is superior to control the research reactor operation.
Keywords : Digital Technique, Reactor Control, Digital Control System.
Published : Technical Report, 1997.
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
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
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
DEFENCE-IN-DEPTH IN ABWR TYPE NPP FROM ENVIRONMENT SAFETY ASPECT POINT OF VIEW
Sarwo D.Danupoyo and Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
DEFENCE-IN-DEPTH IN ABWR TYPE NPP FROM ENVIRONMENT SAFETY ASPECT POINT OF VIEW. Study about application of "defence-in-depth" concept in ABWR type NPP has been done. The study was carried out in order to understand the performance of protection and safety structures in ABWR type NPP from environment safety aspect point of view. The study stressed on simplification of coolant system design, fuel-cladding interaction, core-melt probability, and the possibility of fission product release distributed in environment. The strength of protection and safety structures were observed at loss of coolant accident and severe accident conditions. The result shows that design improvement has increased the ABWR defence-in-depth performance better than conventional BWR type. And so, the radiation release possibility to environment can be reduced.
Keywords : Defence-in-Depth, Safety Structure, Loss of Coolant Accident, Severe Accident, ABWR.
Published : Journal "Pengkajian Sains dan Teknologi Nuklir", Vol.3, No.2, 1998, Pusat Pengkajian Teknologi Nuklir, BATAN, ISSN No. : 0852-8047
