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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PRIMARY SYSTEM DEPRESSURIZATION ACCIDENT ANALYSIS IN THE HTTR

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

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STUDY OF SEVERE ACCIDENT MANAGEMENT STRATEGY FOR PWR TYPE REACTOR

STUDY OF SEVERE ACCIDENT MANAGEMENT STRATEGY FOR PWR TYPE REACTOR
Ign. Djoko Irianto 1) and Edison Sihombing 2)
1) Center for Nuclear Technology Assessment - BATAN
2) Center for Multipurpose Reactor
Email: igndjoko@batan.go.id
ABSTRACT
STUDY OF SEVERE ACCIDENT MANAGEMENT STRATEGY FOR PWR TYPE REACTOR. Accident management is establishment of all the actions which aim to anticipate accidents and mitigate the consequences due to the accidents in nuclear power plants such as PWR type reactors. There are two categories of accidents, that are; the accidents in the DBA (design basis accident) category and the accidents in the beyond DBA category, namely severe accident. Accident management comprises the prevention measure to maintain the core cool ability and containment integrity for design basis accident events using safety-related and inside the design limits. There are two phases of severe accident management. One is the countermeasure to prevent severe damage to the reactor core that is called "phase-1 accident management". The other is the countermeasure to mitigate the consequences of severe accident that is called "phase-2 accident management", this countermeasure particularly to keep the containment vessel integrity in order to mitigate the material radioactive release to the environment. This paper describes a concept of some countermeasure according to the severe accident management such as to maintain core cooling, containment integrity and some support system to the safety function.

Keywords : Severe Accident Management, Design Basis Accident, PWR.
Published : Proceeding "Seminar Teknologi dan Keselamatan PLTN serta Fasilitas Nuklir - IV", Serpong, 10-11 Desember 1996, ISSN No. : 0854-2910

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NEUTRON FLUX OPTIMIZATION USING HAMILTON PONTRYAGIN EQUATION

NEUTRON FLUX OPTIMIZATION USING HAMILTON PONTRYAGIN EQUATION
Hasan, Sarwo D.Danupoyo, B.Santoso, Ign. Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
Email: igndjoko@batan.go.id
ABSTRACT
NEUTRON FLUX OPTIMIZATION USING HAMILTON PONTRYAGIN EQUATION. An analysis using Hamilton Pontryagin equation to optimize neutron flux in nuclear reactor have been done. Neutron flux could be optimized using control system. Control system condition can be represented by index performance (J) that can be adjusted to make control system optimal. This paper describes analysis of neutron flux multiplication in a certain time using Hamilton Pontryagin equation and compare the result with the conventional equation. An analysis result shows that to make neutron flux becomes 100 times in 0.6 sec, index performance which be calculated using conventional equation is J=0.0078, by using Hamilton Pontryagin is J=0.00588.

Keywords : Neutron Flux, Hamilton Pontryagin, Optimize Neutron Flux.
Published : Proceeding of "Seminar Sains dan Teknologi Nuklir PPTN - BATAN", Bandung, 12-13 Maret 1996. ISSN.: 1410-1769

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Assessment Of Generation IV Reactors Secondary System For Hydrogen Production

ASSESSMENT OF GENERATION IV REACTORS SECONDARY SYSTEM FOR HYDROGEN PRODUCTION
Ign. Djoko Irianto
Center for Reactor Technology and Nuclear Safety � BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan
ABSTRACT
ASSESSMENT OF GENERATION IV REACTORS SECONDARY SYSTEM FOR HYDROGEN PRODUCTION. Nuclear power plans (NPP) is one of electrical generation plans which are environmental friendly. It means that NPP does not release exhaust gases contaminating environment. Continuous and increasing demand of clean energy quantitatively triggers the development of the next generation of nuclear energy systems (NES), which have to fulfill several criteria, such as economic, sustainable, safe and reliable, proliferation resistant and have physical protection concept. Nowadays, Generation IV Nuclear Energy Systems consist of Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), Sodium-cooled Fast Reactor (SFR), Super-Critical-cooled Water Reactor (SCWR), and Very High Temperature Reactor (VHTR). The application of the next generation of nuclear reactor designs or Generation IV reactors has been diversified including electricity and non-electricity. One of non electricity applications is the use of nuclear power plant for hydrogen production. There are three methods underlying the hydrogen production processes : water electrolysis, steam reforming of methane, and sulfur-iodine cycle. Parameters of efficiency, outlet temperature of secondary system, and other characteristics of secondary system make the combination of VHTR and steam reforming of methane process or VHTR and sulfur-iodine cycle process become good alternative for hydrogen production installation.
Keywords : Gen.IV Reactors, Secondary System, Hydrogen Production, VHTR, .
Published : 2007, Technical Report, Proceeding "PRESENTASI ILMIAH TEKNOLOGI KESELAMATAN NUKLIR XIII", ISSN No. : 1410-0533

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ANALYSIS OF PHYSICAL PROTECTION SYSTEMS IN BATAN�S NUCLEAR FACILITIES

ANALYSIS OF PHYSICAL PROTECTION SYSTEMS IN BATAN�S NUCLEAR FACILITIES
Ign. Djoko Irianto, Sumijanto
Center for Reactor Technology and Nuclear Safety � BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan.
ABSTRACT
ANALYSIS OF PHYSICAL PROTECTION SYSTEMS IN BATAN�S NUCLEAR FACILITIES. Physical protection system for nuclear facilities has two objectives, i.e.: the first to minimize or eliminate the possibility of theft or unauthorized removal of nuclear materials and the occurrence of sabotage, the second is to deter incoming threat and localize and retrieve the missing nuclear materials effectively and immediately. In order to achieve these two objectives, it requires effective planning of physical protection system for nuclear facilities, which covers detection system, delaying system and response system. According to the Guidance for physical protection implementation, INFCIRC 225 Rev.4, the design of physical protection system for nuclear facilities should also consider the categorization of nuclear materials as the subject of security. This paper discusses the result of analysis of physical protection system implementation in BATAN refer to INFCIRC 225 Rev.4. The result of analysis shows that the implementation of physical protection for nuclear facilities in BATAN had satisfied with the international criteria established by IAEA.

Keywords : Physical Protection, Nuclear Material, BATAN�s Nuclear Facility, INFCIRC.
Published : 2006, Technical Report, Proceeding "PRESENTASI ILMIAH TEKNOLOGI KESELAMATAN NUKLIR XII", ISSN No. : 1410-0533

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ROLE OF PHYSICAL PROTECTION AND SAFEGUARDS TECHNOLOGY USED TO NUCLEAR MATERIAL SECURITY

ROLE OF PHYSICAL PROTECTION AND SAFEGUARDS TECHNOLOGY USED TO NUCLEAR MATERIAL SECURITY
Ign. Djoko Irianto
Center For Safeguards Technology � Batan
Kawasan Puspiptek Gedung 90 Lantai 3, Serpong � 15310
Email: igndjoko@yahoo.com
ABSTRACT
ROLE OF PHYSICAL PROTECTION AND SAFEGUARDS TECHNOLOGY USED TO NUCLEAR MATERIAL SECURITY. The presence of nuclear materials at any nuclear facility must be in secure and must be known as safeguards purpose such as its position, form or type and amount. The clarification of the amount be reported to the national regulatory body and International Atomic Energy Agency (IAEA) as the international regulatory body. The national regulatory body and IAEA will then verify that report. The verification must be done to know there is no difference of the amount, and to give the assurance to the international community that any diversion of safeguarded nuclear material from civil use to a prescribed military purpose would be detected. To carry out verification, several verification techniques such as non-destructive analysis, surveillance, unattended and remote monitoring and environmental sampling are explained to convey the impression how those techniques are implemented. According to the security requirement, the physical protection system including all components of physical protection system have to be effectively designed.

Keywords: Safeguards technology, physical protection, nuclear material
Published : A Scientific Journal for Safeguards Technology. ISSN:1907-0535. Vol.1 No.1. October 2005.

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SAFEGUARDS TECHNOLOGY DEVELOPMENT TO PREVENT NUCLEAR MATERIAL ILLICIT TRAFFICKING

SAFEGUARDS TECHNOLOGY DEVELOPMENT TO PREVENT NUCLEAR MATERIAL ILLICIT TRAFFICKING
Ign. Djoko Irianto
Center For Safeguards Technology � Batan
Kawasan Puspiptek Gedung 90 Lantai 3, Serpong � 15310
Email: igndjoko@yahoo.com
Abstract
SAFEGUARDS TECHNOLOGY DEVELOPMENT TO PREVENT NUCLEAR MATERIAL ILLICIT TRAFFICKING. Nuclear material have been a material which have a strategic value. The presence of nuclear material at any nuclear facility have to be in secure and surveillance. If there are a changing in form or type of nuclear material, a changing in amount or weight and also its position have to be recorded. It has intended to prevent the possibility of the deviation in usage or in storage, and to prevent the possibility of nuclear material illicit trafficking. This paper describe the safeguards technology development aiming in the effort to prevent the possibility of nuclear material illicit trafficking. Technology development was focused on the improvement of existing nuclear material measurement techniques, detection methods and measurements of micro-particles, and the science of nuclear forensic implementation especially the determination of isotope age with concern to prevent nuclear material illicit trafficking. Measurement techniques and method improvement covered isotope correlation implementation to estimate the 242Pu, verification technique of 237Np, and detection method and micro-particles measurement.

Keywords : Safeguards technology, illicit trafficking, nuclear material.
Published : A Scientific Journal for Safeguards Technology. ISSN:1907-0535. Vol.1 No.2. Desember 2005

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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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DEVELOPMENT OF NUCLEAR DATA MODULE FOR VERIFY THE CRITICALITY OF VHTR CORE CALCULATION

DEVELOPMENT OF NUCLEAR DATA MODULE FOR VERIFY
THE CRITICALITY OF VHTR CORE CALCULATION

Zuhair, Syafarudin, Ign. Djoko Irianto, Hery Adrial, Maman Mulyaman
Center for Reactor Technology and Nuclear Safety � BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan

ABSTRACT
DEVELOPMENT OF NUCLEAR DATA MODULE FOR VERIFY THE CRITICALITY OF VHTR CORE CALCULATION. A data nuclear module interfacing between the MCNP (Monte Carlo N-Particle) transport code input and the cross section spectra resulted from experiments or simulations, was developed to support the core criticality calculation of VHTR (very high temperature reactor). The module was equipped by a sub-module converting experimental spectrum into cross section spectrum, and a sub-module of simulation that consists of a package of calculation codes. In the experimental sub-module, all kinematical steps starting from the state of incident particle colliding with atom target until scattering particle is emitted at a certain angle. The simulation sub-module was constructed by embedding the QMD+SDM Quantum Molecular Dynamics + Statistical Decay Model) into the main module. The module was also equipped by a graphical interface for the parameters of TRISO dimensions and the densities of its elements. By this simulation, energy spectra of elastic and inelastic reaction ratio for three major atoms constructing TRISO particle, i.e. C-12, O-16, and Si-32, were obtained. The results show that the lowest elastic ratio is found at the outgoing energy between 3 to 4 MeV, which is identical for those three atoms. This output is expected to become an essential feedback for MCNP calculation theory. However, more intensive studies for other atoms related to this issue are required.

Keywords: MCNP, QMD, SDM, pebble bed, TRISO, VHTR, elastic neutron reaction ratio.
Published: Technical report of Block Grant 2009

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Preliminary Study On Hydrogen Production Process By Applying VHTR Reactor Cogeneration Concept

PRELIMINARY STUDY ON HYDROGEN PRODUCTION PROCESS BY APPLYING VHTR REACTOR COGENERATION CONCEPT
Ign. Djoko Irianto
Center For Reactor Technology and Nuclear Safety - BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan
Email: igndjoko@batan.go.id
ABSTRACT
PRELIMINARY STUDY ON HYDROGEN PRODUCTION PROCESS BY APPLYING VHTR REACTOR COGENERATION CONCEPT. Increasing demand on clean energy quantitatively triggers the development of the renewable energy systems, which have to fulfill several criteria, such as abundant, safe and sustainable. As an energy carrier, hydrogen have been more researched even used as a settled energy generation as well as movable energy generation likely used in transportation. However, the proper and economically hydrogen production process still more necessary to be assessed and researched. There are three methods underlying the hydrogen production processes: water electrolysis, steam reforming of methane, and sulfur-iodine cycle. The energy resource necessities to the hydrogen production process supplied by fossil fuel combustion, resulting in the total efficiency are decreased. This paper explains the assessment of feasibility of several hydrogen production processes using nuclear energy. As a nuclear energy resource, the very high temperature reactor (VHTR) is used. VHTR is one of the six generation IV reactor concepts. The assessment are stressed in the comparison of the advantage and disadvantage for each hydrogen production process coupled with the reactor system of the VHTR as the concept of cogeneration, and the comparison of the efficiency of each process. The result shows that VHTR reactor cogeneration concept combined with sulfur-iodine cycle process become good alternative for hydrogen production installation.

Keyword : Hydrogen Production, cogeneration, very high temperature reactor, efficiency.
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.

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