STUDY OF NUCLEAR FUEL CYCLE INFRASTRUCTURE IN INDONESIA
Sahala M. Lumbanraja
Pusat Pengembangan Energi Nuklir (PPEN) BATAN
Jl. Kuningan Barat, Mampang Prapatan Jakarta 12710
ABSTRACT
STUDY OF NUCLEAR FUEL CYCLE INFRASTRUCTURE IN INDONESIA. The government of Indoensia planned to utilize nuclear power plant for future electric source. This implied in Government Regulation No. 5 Year 2006 regarding the National Energy Policy and Act No. 17 Year 2007 regarding the National Planning year 2005 � 2025. International Atomic Energy Agency (IAEA) Nuclear Energy Series No. NG-T-3.2 stated that there were 19 infrastructure topics that have to be evaluated and prepared, one of it was the insfrastructure study on nuclear fuel cycle. Spent fuel has a high economic value but still contain a high risk, therefore it required a proper management. Generally, nuclear fuel cycle consists of once through fuel cycle an closed fuel cycle. Each have their own advantages and disadvantages. This infrastructure study was need in order to support the stake holders in deciding appropriate and profitable nuclear fuel cycle strategy for Indonesia in the long term in nuclear power plant would be operated.
Key words: infrastructure, nuclear fuel, NPP
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
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.
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.
ANALYSIS OF CROSS SECTION SPECTRUM OF DIRECT REACTION FOR THE NEUTRONIC CALCULATION OF NEW GENERATION REACTOR
Syafarudin
Pusat Teknologi Reaktor dan Keselamatan Nuklir - BATAN
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
ANALYSIS OF CROSS SECTION SPECTRUM OF DIRECT REACTION FOR THE NEUTRONIC CALCULATION OF NEW GENERATION REACTOR. The demand on new spectrum of neutronic cross section comes from the Nuclear Energy System (NES) of new reaction generation since it could not be fulfilled anymore by the existing nuclear data. Neutron, uncharged particle, gives no magnetic interaction when it is passed onto a magnetic field, as the charged one does. Consequently, it is difficult to measure the neutronic spectrum with a good enough energy resolution. Taking advantage from the similarity of neutron and proton (spin and mass), it is possible to study the neutronic reaction using the analogue protonic one. In the current research, the general characteristics of cross section of nucleon is studied intensively as the most pronounced contribution in higher energy region. The cross section of nucleon is approached by the DWBA (Distorted Wave Born Approximation) theorem, and executed by the DWUCK4 code using global parameter of OMP (Optical Model Potential). The comparation between the calculation results and a data of reference, shows good agreement in the characteristics of all dominant single-states of nucleon.
Keywords: DWBA, OMP, single-state, cross section, direct reaction, neutronics
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
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.
THE ASSESSMENT OF THERMODYNAMIC MODEL FOR HYDROGEN PRODUCTION BY IS THERMOCHEMICAL CYCLE
Itjeu Karliana
Pusat Teknologi Reaktor dan Keselamatan Nuklir (PTKRN) - BATAN
Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
THE ASSESSMENT OF THERMODYNAMIC MODEL FOR HYDROGEN PRODUCTION BY IS THERMOCHEMICAL CYCLE. Thermodynamic model for hydrogen production by I-S thermochemical cycle has been studied on the Bunsen reaction. The alternative energy resource of hydrogen which water splitting is promising to produce hydrogen because it has efficient energy, environment acceptable, and competitive cost operation compared to fossil energy or renewable energy resources. For commercial scale of hydrogen production through the I-S thermochemical cycle as the aim of others. In this cycle, iodine and sulfur dioxide mixture with water forming iodide acid and sulfuric acid. Both phases to form two separated section, H2SO4: [H2SO4 + H2O]l at the upper layer and HIx : [2HI]g + [(x-1)I2]l + [H2O]l at the bottom. In the separation process known several factor has been problems, for instances: HI extraction from HIx mixture because azeotropic mixing within HIx section, solidification of iodine, and heterogenous H2O-HI-I2 ternery mixtures. In this paper are described thermodynamic model on the hydrogen production by I-S thermochemical cycle using ZRP/EoS/Gex and PR/MHV2/UNSolv combined with activity coefficient and Engel�s salvation model. The goal of assessment is to evaluate equilibrium system in HIx region of HIx : [2HI]g + [(x-1)I2]l + [H2O]l due too many dissolved fractions. From this assessment that thermodynamic model can explain the equilibrium of liquid-liquid phase and vapor-liquid phase in HIx mixture solution.
Keywords: Thermodynamic model, hydrogen production, thermochemical.
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
STUDY AND ASSESSMENT OF GENERATION IV REACTOR NUCLEAR DATA WITH FAST NEUTRON SPECTRA
Suwoto
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
STUDY AND ASSESSMENT OF GENERATION IV REACTOR NUCLEAR DATA WITH FAST NEUTRON SPECTRA. Generation IV International Forum (GIF) has evaluated and assessed NES of Gen-IV and selected six potential types of reactors to be deployed in the next decade. Those include GFR, LFR, SFR, MSR, SCWR and VHTR. The assessment focused on the nuclear data characteristic parameter and nuclear data uncertainties of Gen-IV reactor with fast neutron spectrum. Until 2008, the accuracy target of nuclear data cross-sections used it in fast reactor spectrum calculation are relatively significant especially for s-capture, s-fission, and s-inelastic. Several differences of nuclear data cross-sections on minor actinide isotopes between expected and targeted parameters are observed such as sfission of Cm-244 isotope up to 10 times larger and s-capture of 92-U-238 isotope around 1.5-2 times higher than targeted parameters. Uncertainty and accuracy of minor actinide cross-sections for fast spectrum Gen-IV reactors provide relatively significant discrepancies (1.3 to 10 times higher) in term of accuracy between expected and targeted parameters. There are some differences of several evaluated nuclear data files. Some discrepancies on integral parameter of fast spectrum Gen-IV reactors between expected and targeted such k-eff, void reactivity and Doppler effects, peak power and burn-up are clearly observed. Accurate and precise cross-sections data of radiation captured and threshold reaction cross sections such as (n,2n), (n,3n), (n,p), (n,a) are necessary for fast reactors.
Keywords: cross-sections, fast neutron spectrum, GFR, LFR, SFR, uncertainty and target accuracy
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
STUDY OF SILICA MEMBRANE PERFORMANCE FOR SEPARATION HYDROGEN GAS FROM THE MIXTURE OF H2-H2O-HI TO SUPPORT EFFICIENCY OF HYDROGEN PRODUCTION
Tumpal Pandiangan
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
STUDY OF SILICA MEMBRANE PERFORMANCE FOR SEPARATION HYDROGEN GAS FROM THE MIXTURE OF H2-H2O-HI TO SUPPORT EFFICIENCY OF HYDROGEN PRODUCTION. The membrane pores Sizing can be controlled by the value of time and CVD process. That's parameters were represented by the values of selected power of the (He/N2) gas. The membranes that have controlled their pores sizing by it's parameters were tested the power selected and permeation of H2 gas from the gaseous mixtures and singular system. Related on it's observation, the value of hydrogen permeation both in mixtures and singular have the similar value that are about of 10-7mol.Pa-1.s-1. This value was generated from the S3 membrane silica type where that membranes were stopped in modification at the value of power selection gas (He/N2) was 2,8. That's fact say that the best selectivity and permeation of H2 both in gas mixtures and singular are not generated from the smallest pore size but it was generated from the S3 membrane type which the selective power is 2.8. The permeation of H2 gas is relative same for all of type membrane and this reality was predicted because the size difference of He and N2 gas is relative higher so it is not very sensitive for looking on the best permeation of membrane. The propose of this study is to add the sophisticated knowledge in a silica membrane synthesis.
Key words : Membrane, permeation, selective power, CVD, TEOS, pores
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
STUDY ON HELIUM TURBINE FOR SECONDARY COOLANT SYSTEM OF MOLTEN SALT REACTOR
Sri Sudadiyo
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
STUDY ON HELIUM TURBINE FOR SECONDARY COOLANT SYSTEM OF MOLTEN SALT REACTOR. From the viewpoint of energy system and environment, concept for molten salt reactor (MSR) is one of advanced nuclear reactors which have good potential for electricity generation device. Within MSR, molten salt fuel flows through graphite core channels, to produce thermal neutron. The obtained heat of nuclear fuel was transferred to secondary coolant system through the heat exchanger using closed cycle of helium turbine. The resulted hot helium gas was expanded to the turbine for getting power. This study purposed to determine the performance of helium turbine as main components of secondary coolant cycle in MSR. The applied parameter was pressure ratio, specific heat ratio, and temperature. By placing both of helium turbine and compressor at single shaft, it was obtained approximate 49 % from turbine power output for driving compressor and the residual power to turn on electricity generator. The yielded turbine adiabatic efficiency is 85 % and it is able to improve thermal efficiency for secondary coolant system of MSR.
Keywords : Turbine, efficiency, helium
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
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.
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 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 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
ANALYSIS OF HELIUM GAS FLOW THROUGH TURBINE NOZZLE FOR MOLTEN SALT POWER REACTOR
Sri Sudadiyo
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
ANALYSIS OF HELIUM GAS FLOW THROUGH TURBINE NOZZLE FOR MOLTEN SALT POWER REACTOR. From the viewpoint of energy system and environment, concept for Molten Salt Reactor (MSR) is one type of advanced generation nuclear power reactors which have good potential for electricity generation device. Within MSR, molten salt fuel flows through graphite core channels, to produce thermal neutron. The obtained heat of nuclear fuel was transferred to secondary coolant system through the heat exchanger using closed cycle of helium turbine. The resulted hot helium gas was expanded to the nozzle for running blade at turbine rotor. At the nozzle, crossed area constitutes very critical section, if crossed area was too small then the helium flow will be choked, and if crossed area was too large then turbine cannot yield its best efficiency. This study purposed to determine the characteristic of helium flow with speed of supersonic through nozzle as most important component within gas turbine system in secondary coolant cycle for giving safety on MSR installation operation. The applied solution method was by employed the equations of energy, mass, momentum, state, process. From the obtained results, it can be known that helium flow rate on critical crossed area had the speed of 1 M, critical pressure ratio of 0,49, and critical temperature ratio of 0,75, so that the flow via nozzle had the good characteristic and it could be used to helium turbine at secondary coolant cycle in MSR installation.
Keywords : Turbine, nozzle, helium
Published : Prosiding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
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
