Studi Perhitungan Reaktor HTR Pebble-Bed Dengan Berbagai Opsi Desain Matriks Bahan Bakar

STUDI PERHITUNGAN REAKTOR HTR PEBBLE-BED DENGAN
BERBAGAI OPSI DESAIN MATRIKS BAHAN BAKAR

Zuhair dan Suwoto
Pusat Teknologi Reaktor dan Keselamatan Nuklir � BATAN
ABSTRACT
STUDY ON PEBBLE-BED HTR REACTOR CALCULATION WITH SEVERAL OPTIONS OF FUEL MATRIX DESIGNS. Pebble-bed HTR core is able to accommodate various types of fuel without significant core modification. This paper presents study of calculation of pebble-bed HTR core with three options of fuel matrix designs: UO2 (8.2% U235 enrichment), PuO2 (53.85% Pu239 enrichment) and ThO2/UO2 (7.47% U233 enrichment). Core calculation includes cell calculation using infinite array model of pebble-bed fuel with reflective boundary and full core calculation uses cylindrical model (2-D R-Z) with 300 cm in diameter and 943 cm in height. All computations are carried out using Monte Carlo transport code MCNP5 at temperature of 293.6 K and 1000 K. In general, MCNP5 calculations indicate consistency with kinf and keff values of UO2 core which always almost higher than those of PuO2 and ThO2/UO2 cores. Compared to the other Monte Carlo simulation show that MCNP5 produces the value of kinf which is closer to that obtained by MCNP-4B than that obtained by MONK9 with the computation bias less than 1.3%. The MCNP5's keff calculation reflect a close tendency to that achieved by MCNP-4B, KENO-V.a, and MONK9, however, its computation bias is relatively high compared to the TRIPOLI4, especially for reactor core with PuO2. It can be concluded that MCNP5 estimations excist in the range of all Monte Carlo calculation codes and are expected to be the most precision if the experimental data found later.

Keywords: HTR pebble-bed, fuel, MCNP5, ENDF/B-VI
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009

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Review Of Helium Impurities Effect On Corrosion Process Of HTGR Reactor Coolant

REVIEW OF HELIUM IMPURITIES EFFECT ON CORROSION PROCESS OF HTGR REACTOR COOLANT

Sriyono
Pusat Teknologi Reaktor dan Keselamatan Nuklir - BATAN,
Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310

ABSTRACT
REVIEW OF HELIUM IMPURITIES EFFECT ON CORROSION PROCESS OF HTGR REACTOR COOLANT. HTGR is an advanced nuclear reactor which has helium gas as a coolant and operates safely at high temperature. Corrosion is one of serious problem must be solved in HTGR caused by its impurities. The impurities of helium gas are H2O, CO, CO2, N2, H2 and CH4 which has various concentrations in HTGR coolant system. Corrosion in HTGR system caused by oxidation and carburization-decarburization process. The results of oxidation are oxide scale and carburization-decarburization promotes the carbide compound. Both of these adhered in the material surface and degraded it. The many experiments have been done to understand the effect of impurities to material in HTGR. The purpose of review is to know the effect of helium impurities to material surrounding and determine the suitable material in HTGR design. Among the materials, 2 1/4Cr-1Mo and modified 9Cr1Mo ferrite steels are considered for application in reactor pressure vessels. Fe-Cr-Ni alloys such as Alloy 800H and austenitic stainless steels are considered for recuperates and reactor internals. Alloys such as 617, Hastelloy X, and Hastelloy XR are considered for components that will be exposed to helium coolant at temperatures up to 900�C. Alloys such as 713LC and Mo-TZM are considered for the turbine blade. Alloys such as A286, 706, and 718, are examined for turbine disk application.

Keywords: impurities, helium, corrosion, coolant, HTGR
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.

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Comparisan Of Nuclear Hydrogen Production Between Sulfur-Iodine Cycle Of Thermochemical And Natural Gas Steam Reforming Process

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.

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Study On Model Of VHTR And GFR Cores For Reactor Multiplication Calculation

STUDY ON MODEL OF VHTR AND GFR CORES FOR REACTOR MULTIPLICATION CALCULATION

Maman Mulyaman dan Zuhair
Pusat Teknologi Reaktor dan Keselamatan Nuklir (PTRKN), BATAN
Kawasan Puspiptek, Serpong, Tangerang 15310

ABSTRACT
STUDY ON MODEL OF VHTR AND GFR CORES FOR REACTOR MULTIPLICATION CALCULATION. VHTR and GFR are two candidates of Generation IV reactors which have received a lot of attention specifically to meet world energy needs in the future. These two reactors have different neutron spectrums, but use the same coolant material, namely helium. The outlet core temperatures of VHTR and GFR which are 1000oC and 850oC respectively enable them to produce hydrogen. In this paper, the cores of VHTR and GFR are modeled homogeneously, in which the heterogeneous cells are parsed into isotopic density and new materials consisting of weighted nuclides are formed. The model of cell shape is hexagon with 30 cm distance of flat to flat and 40 cm height. Because of incomplete specific temperature data for materials of nuclear fuel, cladding, and moderator/reflector, the calculation was done in room temperature and outlet core temperature. The calculation results using the Monte Carlo transport code MCNP5 and continuous energy nuclear data library ENDF/B-VI show that the GFR and VHTR have negative temperature effects with coefficient reactivity temperature of -5,1259E-5 ?k/k/oC and -5,5177E-5 ?k/k/oC, respectively. The value of keff VHTR, which is greater than 4.54% compared with those of GFR concludes that the presence of the role of graphite composition which dominates U-235 in VHTR and the effect of neutrons resonance absorption in U-238 which is significant in GFR.

Keywords: VHTR, GFR, reactor multiplication
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.

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The Feasibility Of Heat Transfer System Aspect On Very High Temperature Reactor (VHTR)

THE FEASIBILITY OF HEAT TRANSFER SYSTEM ASPECT ON VERY HIGH TEMPERATURE REACTOR (VHTR)

Sudarmono
Pusat Teknologi Reaktor dan Keselamatan Nuklir - BATAN
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310

ABSTRACT
THE FEASIBILITY of HEAT TRANSFER SYSTEM ASPECT on VERY HIGH TEMPERATURE REACTOR (VHTR). Very high temperature reactor is a generation IV reactor has been enhancing to support the innovation nuclear energy system. VHTR is a concept reactor for challenging technology goals for Generation IV nuclear energy systems and heat utility for hydrogen production and thermo-chemical applications. The VHTR is a next step in the evolutionary development of high-temperature gas cooled reactors. VHTR system are purposed to enhance of reactor safety and reliability, economics electricity production and new products, nuclear waste reduction and proliferation resistance and physical protection. Reactor operations on very high temperature give an effect for generate electricity with high efficiency, over 50%. Reactor technical specification that operated on very high temperature needs all components have to be developed for temperatures well above the present state of 1000oC. Safety aspect of reactor system should be separate against petrochemical system. As a preliminary conclusion, it�s needed to enhance the heat transfer material however to continue follow the VHTR development, by concerning to the others aspects, VHTR can be choose as an alternative to fulfill electricity and hydrogen production in Indonesia.

Key words: VHTR, concept reactor, hydrogen production, high efficiency
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.

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Study On Pebble-Bed HTR Reactor Calculation With Several Options Of Fuel Matrix Designs

STUDY ON PEBBLE-BED HTR REACTOR CALCULATION WITH SEVERAL OPTIONS OF FUEL MATRIX DESIGNS

Zuhair dan Suwoto
Pusat Teknologi Reaktor dan Keselamatan Nuklir � BATAN

ABSTRACT
STUDY ON PEBBLE-BED HTR REACTOR CALCULATION WITH SEVERAL OPTIONS OF FUEL MATRIX DESIGNS. Pebble-bed HTR core is able to acomodate various types of fuel without significant core modification. This paper presents study of calculation of pebble-bed HTR core with three options of fuel matrix designs: UO2 (8.2% U235 enrichment), PuO2 (53.85% Pu239 enrichment) and ThO2/UO2 (7.47% U233 enrichment). Core calculation includes cell calculation using infinite array model of pebble-bed fuel with reflective boundary and full core calculation uses cylindrical model (2-D R-Z) with 300 cm in diameter and 943 cm in height. All computations are carried out using Monte Carlo transport code MCNP5 at temperature of 293.6 K and 1000 K. In general, MCNP5 calculations indicate consistency with k_inf and k_eff values of UO2 core which always almost higher than those of PuO2 and ThO2/UO2 cores. Compared to the other Monte Carlo simulation show that MCNP5 produces the value of k_inf which is closer to that obtained by MCNP-4B than that obtained by MONK9 with the computation bias less than 1.3%. The MCNP5's k_eff calculation reflect a close tendency to that achieved by MCNP-4B, KENO-V.a, and MONK9, however, its computation bias is relatively high compared to the TRIPOLI4, especially for reactor core with PuO2. It can be concluded that MCNP5 estimations exist in the range of all Monte Carlo calculation codes and are expected to be the most precision if the experimental data found later.

Keywords: HTR pebble-bed, fuel, MCNP5, ENDF/B-VI
Published : Prosiding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009

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Intermediate Heat Exchanger (IHX) Effectiveness Calculation Of The Cogeneration System Of High Temperature Gas-Cooled Reactor (HTGR)

INTERMEDIATE HEAT EXCHANGER (IHX) EFFECTIVENESS CALCULATION OF THE COGENERATION SYSTEM OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR)

Ign. Djoko Irianto
Center For Reactor Technology and Nuclear Safety - BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan

ABSTRACT
INTERMEDIATE HEAT EXCHANGER (IHX) EFFECTIVENESS CALCULATION OF THE COGENERATION SYSTEM OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR). Very High Temperature Reactor (VHTR) is a high temperature gas-cooled reactor (HTGR) which be a one of Generation IV reactors which is conceptually designed using cogeneration configuration for electric generation and for hydrogen production. VHTR employs a helium-coolant with operating pressure 5,0 MPa and 950oC outlet temperature. The main energy conversion component in VHTR cogeneration is intermediate heat exchanger (IHX). Thermal energy passes the IHX from the reactor system to the cogeneration system for electric generation and for hydrogen production or another application. The success of VHTR design is affected by many factors, one of which is the performance of IHX. To support the conceptual design, many factors which affect the IHX performance particularly high temperature IHX have to be examined, calculated and analyzed. Many factors which affect the IHX performance is namely effectiveness, total heat transfer, etc. In this research, the effectiveness of the conceptually designed of IHX which refer to IHX in GTHTR300C and the total heat transfer of IHX for the cogeneration systems have been calculated using variant of inlet temperature. Total IHX heat transfer and the effectiveness are calculated using e-NTU (Number of Transfer Unit) method. With assumption of a helium-coolant used in the both side of IHX, the optimal effectiveness IHX is 0.95. Conceptually, it can be concluded that this IHX is effective to be used in the HTGR cogeneration based on VHTR.

Keywords: High temperature gas-cooled reactor (HTGR), cogeneration, IHX, effectiveness, efektivitas, metode number of transfer units (NTU method)
Published : Prosiding Seminar Nasional Ke-16 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surabaya, 28 Juli 2010

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Effect Of IHX Configuration Concerning The Effectiveness And Efficiency Of VHTR Reactor Cogeneration

EFFECT OF IHX CONFIGURATION CONCERNING THE EFFECTIVENESS AND EFFICIENCY OF VHTR REACTOR COGENERATION

Ign. Djoko Irianto
Center For Reactor Technology and Nuclear Safety - BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan

ABSTRACT
EFFECT OF IHX CONFIGURATION CONCERNING THE EFFECTIVENESS AND EFFICIENCY OF VHTR REACTOR COGENERATION. Very High Temperature Reactor (VHTR) is one of Generation IV reactors which is conceptually designed using cogeneration configuration for electric generation and for hydrogen production. VHTR employs a helium-coolant with operating pressure 7 MPa and 1000oC outlet temperature. The main energy conversion component in VHTR cogeneration is intermediate heat exchanger (IHX). Thermal energy passes the IHX from the reactor system to the cogeneration system for electric generation and for hydrogen production. The value of efficiency and effectiveness of the system make certain to the success of VHTR design. This paper describes the results of the analysis of IHX configuration effect concerning the effectiveness and efficiency of VHTR cogeneration system. There are three configuration systems analyzed: direct cycle with secondary heat exchanger (SHX), direct cycle without SHX, and indirect cycle. The results of the analysis show that the highest efficiency is given by the direct cycle with SHX and the IHX is set in parallel with PHX. Despite its low efficiency, the third configuration obtains the highest effectiveness. This configuration in which electricity is generated indirectly has possibility to be used for a compact system design.

Keywords : Cogeneration, VHTR, IHX
Published : Journal "Sigma Epsilon", Volume 13 Nomor 3, Agustus 2009

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Energy Conversion System Modeling Based On HTGR Cogeneration For Electric Generation And Hydrogen Production

ENERGY CONVERSION SYSTEM MODELING BASED ON HTGR COGENERATION FOR ELECTRIC GENERATION AND HYDROGEN PRODUCTION

Ign. Djoko Irianto
Center For Reactor Technology and Nuclear Safety - BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan
Email: igndjoko@batan.go.id
ABSTRACT
ENERGY CONVERSION SYSTEM MODELING BASED ON HTGR COGENERATION FOR ELECTRIC GENERATION AND HYDROGEN PRODUCTION. Very High Temperature Reactor (VHTR) is a high temperature gas-cooled reactor (HTGR) which be a one of Generation IV reactors which is conceptually designed employs a helium-coolant with operating pressure 7,0 MPa and 1000 oC outlet temperature. Conceptually, VHTR is designed using cogeneration configuration for electric generation and for hydrogen production. The thermal power of the reactor could be determined according to the requirement which will be build in Bangka Belitung Province is 600 MWth. In this research, energy conversion system modeling based on HTGR cogeneration has been done in direct and indirect cycle configuration. There are two configuration in the direct cycle, which divide of the IHX and turbine in parallel or serial. With assumption of a helium-coolant used in the both side of IHX, the optimal effectiveness IHX is 0.95. Based on the effectivenes of heat exchanger, the heat transfer rate has been calculated. The system efficiency calculated for the three configuration. In generally, the efficiency of direct cycle is higher than the other. The efficiency is about 50%. Despite its low efficiency, the indirect cycle configuration obtains the highest effectiveness.

Keywords: HTGR, energy conversion system, cogeneration, effectiveness, efficiency
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR III 2010", Cilegon, Banten, 24 Juni 2010.

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Assessment Of The Performance Of Intermediate Heat Exchanger On VHTR Reactor Cogeneration System

ASSESSMENT OF THE PERFORMANCE OF INTERMEDIATE HEAT EXCHANGER ON VHTR REACTOR COGENERATION SYSTEM

Ign. Djoko Irianto
Center for Reactor Technology and Nuclear Safety � BATAN
Kompleks Puspiptek Serpong, Tangerang Selatan
ABSTRACT
ASSESSMENT OF THE PERFORMANCE OF INTERMEDIATE HEAT EXCHANGER ON VHTR REACTOR COGENERATION SYSTEM. Very High Temperature Reactor (VHTR) is one of Generation IV reactors which is conceptually designed using cogeneration configuration for electric generation and for hydrogen production. VHTR employs a helium-coolant with operating pressure 7 MPa and 1000oC outlet temperature. The main energy conversion component in VHTR cogeneration is intermediate heat exchanger (IHX). Thermal energy passes the IHX from the reactor system to the cogeneration system for electric generation and for hydrogen production. The success of VHTR design is affected by many factors, one of which is the performance of IHX. This paper describes the results of the assessment of IHX performance in VHTR cogeneration system. The performance of IHX is influenced by some parameters such as: effectiveness of IHX, efficiency of IHX, and the configuration of IHX in the cogeneration system. There are three configuration systems assessed: direct cycle with secondary heat exchanger (SHX), direct cycle without SHX, and indirect cycle. The assessment are performed by comparing the parameter of IHX performance in the three configuration systems. The results of the assessment show that the highest efficiency is given by the direct cycle with SHX and the IHX is set in parallel with PHX. Despite its low efficiency, the third configuration obtains the highest effectiveness. This configuration in which electricity is generated indirectly has possibility to be used for a compact system design.

Key words : VHTR cogeneration system, intermediate heat exchanger (IHX), IHX system configuration, IHX efficiency, IHX effectiveness
Published : Prosiding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009

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License For HTR Type Reactor

LICENSE FOR HTR TYPE REACTOR

N.Nababan ^1], Ign.Djoko Irianto ^2], Yus Rusdian A ^1]
1] Center for Multipurpose Reactor - BATAN
^2] Center for Nuclear Technology Assessment - BATAN
ABSTRACT
LICENSE FOR HTR TYPE REACTOR. The reason for liking the idea of introducing the modular High Temperature Reactor (HTR) technology can be quite different reasons like e.g.: "economic reasons based on electricity production prices and advantages of co-generation", "strategic reasons based on independency of fossil fuels", "long term industrial strategy", "environmental reasons based on confidence in the safety of the HTR and on the strong reduction of CO₂ emissions", or a combination of these aspect. The combination of a HTR with a gas turbine offers all features as a modern energy technology: safe, clean, and efficient. Licensability of a new reactor concept is more than just fulfilling a set of existing nuclear regulations and guidelines. An essential prerequisite for a satisfactory completion of a lincensing procedure for a new technology is : favorable climate in terms of a positive altitude of the general public, government and utilities. It may be expected that if, in a specific country, the idea of constructing and operating modular HTRs is attractive to those people who play a decisive role in the decision making, the modular HTR will be licensable as well.

Keywords : Modular, HTR, Co-generation, License,
Published : Proceeding "Seminar Ke-4 Reaktor Temperatur Tinggi dan Teknologi Nuklir", Jakarta, 15-16 February 1999, ISSN 0854-8803

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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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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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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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