THERMAL PARAMETER TRANSIENT ANALYSIS OF DROPLETS IN NUCLEAR POWER PLANT COOLING TOWER
Hendro Tjahjono
Pusat Teknologi Reaktor dan Keselamatan Nuklir BATAN
ABSTRACT
THERMAL PARAMETER TRANSIENT ANALYSIS OF DROPLETS IN NUCLEAR POWER PLANT COOLING TOWER. In Nuclear Power Plant using fresh water from river as condenser cooling, a cooling tower still used for decreasing the amount of fresh water used so that could reduce the negative impact to the environment. Inside a cooling tower, warm water coming from condenser drops from a certain level of height in a form of droplet and being cooled by the air. The heat transfer between droplets and the air determines significantly the effectiveness of cooling tower. The heat transfer process involves latent heat transfer owing to vaporization of small portion of water and sensible heat transfer owing to difference in temperature of water and air. The objective of this research is to determine the influence of droplets size and its fall height to temperature transient during the fall. The analysis is performed explicitly using finite difference method in spherical coordinate to resolve the transient conduction equation. The air temperature is supposed constant as 30�C and the sensible heat transfer is performed by convection and radiation. As independent variable in this analysis are droplets size and fall height. The result shows that the heat transfer effectiveness is higher as droplet size is small and fall height is high. For NPP of 1000 MWe, with the fall height of 20 m and the droplet diameter of 2 mm, the final average temperature of droplets is 31.6�C for 40�C initially and the volume rate of cooling water is 58.3 m3/s.
Keywords: cooling tower, NPP, heat transfer effectiveness, droplets.
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
SAFETY EVALUATION OF REACTOR CORE FOR PWR BASED ON INITIATING EVENT AND DESIGN ASPECT
D. T. Sony Tjahyani
PTRKN - BATAN
ABSTRACT
SAFETY EVALUATION OF REACTOR CORE FOR PWR BASED ON INITIATING EVENT AND DESIGN ASPECT. Safety evaluation for NPP is important to determine frequency and consequence of fission product released to public and environmental. Those condition is caused by core damage and containment system failure. Core damage is caused initiating events and safety system failure. Safety system failure is dependent by 6 items that is single failure criteria, redundancy, independency, diversity, fail-safe concept, system interaction and dependencies. The objective of the evaluation is to determine those items to system failure and initiating events contribution to core damage. PWR for generation II and III (III+) are used as object of study for this assessment. The analysis was carried out by collecting initiating event and core damage data also to assess design configuration of PWR for generation II and III (III+). The evaluation results showed that system modification of generation II is significant to core safety level for generation III (III+) PWR, so it is to reduce initiating events and core damage frequency.
Keywords: PWR, Core Damage, Initiating Event, PWR for Generation II and III
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
ANALYSIS ON EARLY PHASE OF SEVERE ACCIDENT IN NUCLEAR POWER PLANT
Sugiyanto
PTRKN - BATAN
ABSTRACT
ANALYSIS ON EARLY PHASE OF SEVERE ACCIDENT IN NUCLEAR POWER PLANT. Analysis on early phase (during100 minutes after accident initiated) of severe accident in the nuclear power plant has been conducted. The objective of this analysis is to understand the progress of core condition from core heat-up, core uncovery, until core melting. This phenomena is interesting to understand because as based for mitigation action by operator. Two scenarios were assumed for analysis, the first scenario, accident is initiated by loss of coolant accident (LOCA) and the second scenario, accident initiated by loss of electric power then each sequence was followed by emergency core cooling system (ECCS) failure. The analysis was conducted using THALES-2 computer code. This analysis showed that, in the first scenario core uncovery occurred at about 14 minutes after accident and core melt started at about 42 minutes. In the second scenario, core uncovery occurred at about 27 minutes after accident and core melt started at about 52 minutes. From this analysis can be concluded that severe accident with initiated LOCA core uncovery will be occur faster.
Keywords: Severe Accident, Nuclear Power Plant, THALES-2
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
MULTIOBJECTIVE SIMULATED ANNEALING METHOD IMPLEMENTATION FOR PWR FUEL LOADING PATTERN OPTIMIZATION USING COREBN CODE
Christina Novila Soewono, Alexander Agung, Sihana
Jurusan Teknik Fisika Fakultas Teknik - Universitas Gadjah Mada
ABSTRACT
MULTIOBJECTIVE SIMULATED ANNEALING METHOD IMPLEMENTATION FOR PWR FUEL LOADING PATTERN OPTIMIZATION USING COREBN CODE. Optimizing loading/reloading pattern design is one of nuclear fuel management activities in order to reduce fuel cycle costs while satisfying safety constraints and operational targets. Multiplication factor at the end of cycle and maximum power peaking factors are the parameters to define the optimal LP design. This optimization initial fuel loading pattern study is based on multiobjective simulated annealing algorithm which is coupled to COREBN code for core burn up calculation. Optimization is implemented on � core model (52 fuel assemblies) which represent the whole core. The result will then be compared to standard model in order to observe the improvement.
Keywords: optimization, loading pattern, multiplication factor at end of cycle, power peaking factors, multiobjective simulated annealing, COREBN
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
IMPLEMENTATION OF GENETIC ALGORITHM METHOD FOR PWR FUEL LOADING PATTERN OPTIMIZATION USING COREBN CODE
Petrus, Alexander Agung, Sihana
Jurusan Teknik Fisika, Fakultas Teknik, Universitas Gadjah Mada
ABSTRACT
IMPLEMENTATION OF GENETIC ALGORITHM METHOD FOR PWR FUEL LOADING PATTERN OPTIMIZATION USING COREBN CODE. Since the large number of possible combination for the fuel assembly loading in the core at the beginning of reactor operation, the core configuration optimized to find an optimal core configuration that will achieve maximum keff at end of cycle and minimum power peaking factor (PPF). This optimization has 2 Genetic Algorithm methods, the first method uses single objective and the second method uses multi objective. The optimization uses � symmetry reactor core model (52 fuel assemblies position), with 3 types of fuel assemblies consists 13 assemblies of 1,5%, 15 assemblies of 2,5% and 24 assemblies of 3% U-235 enrichment without burnable poisson rod. Neutronic calculation of fuel assembly using PIJBurn code and core calculation using COREBN code. From the single objective optimization is obtained the optimum configuration with 8,9% (60 days) cycle length extension and 23,31% decrease in PPF compared to standard model. For multi objective optimization obtained a set pareto front containing 47 non-dominated solutions. By using standard deviation of the crowding distances method, a single final solutions is obtained. The solution gives 10,45% (70 days) cycle length extension and 27,7 % decrease in PPF compared to standard model. Both of optimization method success to obtain optimum solution and fulfill the safety standard.
Keywords: fuel assembly, keff, PPF, Genetic Algorithm, cycle length.
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
ANALYSIS FUEL BURN-UP DISTRIBUTION OF 1000 MWE PWR-NPP CORE FUELLED BY UO2 3,4 WT%.
Jati Susilo1, Tukiran Surbakti2, Iman Kuntoro3
1,2Pusat Teknologi Reaktor Dan Keselamatan Nuklir (PTRKN)
3Pusat Teknologi Bahan Industri Nuklir
ABSTRACT
ANALYSIS FUEL BURN-UP DISTRIBUTION OF 1000 MWE PWR-NPP CORE FUELLED BY UO2 3,4 WT%. To support utilization of nuclear energy programme, therefore preliminary research about characteristic neutronic for PWR-NPP of core has been done. Some neutronic characteristic that related to core safety is limitation value of discharge burn-up maximum produced by fuel assembly in the core. In this research, to know value of discharge burn-up each fuel assembly in the core, then calculation of fuel burn-up distribution at the PWR core fuelled UOBBB2BBB with 3.4wt% enrichment and Zr-4 for cladding. Those PWR core can produce about 3411 MWth power heat, so that it is classified into PWR 1000 MWe class power of NPP. To analysis fuel burn-up distribution, then use 2 different method of fuel loading pattern as follow. In the PWR-A core, fuel group divided to 2 class of fuel burn-up (2 batch) or each � part of the core. And, fuel group in the PWR-B core distributed in the 3 class of fuel burn-up (3 batch) or each 1/3 part of the core. Core burn-up calculation done using ASMBURN module of SRAC computer code in the 2 dimension geometry with � model of the core. The macroscopic cross section table get by calculation of fuel cell using module PIJ of SRAC with JEND.3.3. As public library data. Temperature of the fuel pellet, cladding and moderator are 900 K, 600 K, and 600 K, respectively. From the calculation result knew that PWR-A core produce average discharge burn-up of fuel assembly (34.55 GWd/t) smaller then PWR-B core (38.01 GWd/t). Discharge burn-up maximum of fuel assembly at the PWR-A core and the PWR-B core are 38.77 GWd/t and 42.36 GWd/t, respectively. So that, with limitation of fuel assembly discharge burn-up maximum in the PWR core fuelled UOBBB2BBB with 3.4 wt% is about 39 GWd/t, then those PWR more exactly using 2 batch fuel loading pattern.
Keywords : burn-up, PWR, UOBBB2BBB, SRAC
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
ANALYSIS OF THE TEMPERATURE COEFFICIENT OF THE PWR 1000 MWe FUEL ASSEMBLY OF ENRICHMENT FUNCTION USING MCNP CODE
Rokhmadi dan Tukiran
PUSAT TEKNOLOGI REAKTOR DAN KESELAMATAN NUKLIR-BATAN
Kawasan PUSPIPTEK Gd. No. 80 Serpong 15310
ABSTRACT
ANALYSIS OF THE TEMPERATURE COEFFICIENT OF THE PWR 1000 MWe FUEL ASSEMBLY OF ENRICHMENT FUNCTION USING MCNP CODE. As a part of preparation for the first Nuclear Power Plant, NPP, in Indonesia, it is necessary to assess the safety of the NPP. One of the safety parameters of an NPP reactor is temperature coefficient parameters. The parameters must be determined with high accuracy because those are important values to analyze the stability and transient control of the reactor. In this paper, the moderator, cladding and fuel temperature coefficients were calculated for the PWR 1000MWe fuel assembly with enrichment of 3%, 2,5% and 2%. The calculations were carried out using the Monte Carlo method code of MCNP5 version of 1.3. The nuclear data of ENDF/B-VI.2 is used as a main nuclear data. In hot condition, some neutron cross-section materials were taken from the ENDF/B-V nuclear data. The cold condition with temperature of 293.6K is used as a reference. The calculations showed that the temperature coefficient for fuel on 3%, 2.5%, 2% enrichment are -1.16 pcm �k/k/K, -1.47 pcm �k/k/K and -1.61 pcm �k/k/K respectively. The fuel is the most sensitive materials if the change of temperature occurred, while the effect on cladding material can be avoided. However, all values of the temperature coefficient are negative.
Keywords : PWR fuel assembly, enrichment, kinf, temperature coeficient, MCNP5
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
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
NUMERICAL ASSESSMENT OF CHARACTERISTIC PASSIVE COOLING SYSTEM WITH AIR AT CONTAINMENT AP1000 MODEL
Widi Laksmono1), Ari Darmawan Pasek1), Efrizon Umar2)
1)Fakultas Teknik Mesin dan Dirgantara � ITB
2)Pusat Teknologi Nuklir Bahan dan Radiometri � BATAN
ABSTRACT
NUMERICAL ASSESSMENT OF CHARACTERISTIC PASSIVE COOLING SYSTEM WITH AIR AT CONTAINMENT AP1000 MODEL. Nuclear power plant technology has been growing rapidly. Nowadays, research and development had been taken place especially passive utilization of safety feature. The purpose of this research is to assess a natural convective heat transfer characteristic at containment AP1000 model by using natural air circulation. The analysis method used in this research is finite volume by using computational fluid dynamic (CFD) code. Based on numerical analysis result, the containment of AP1000 model becomes cooler with existence of baffle. This result indicated that the baffle as air director work properly and a better cooling system is achieved. The new concentric cylinder heat transfer correlation derives from containment model with baffle is proposed in the form of : Nu = f (Ra)*
Keywords: natural convection, passive containment cooling system, AP1000
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
ANALYSIS OF VOID PERCENTAGE IMPACT ON NEUTRONIC AND TERMOHYDRAULIC CONDITION OF BOILING WATER REACTOR
Nanang Triagung Edi Hermawan dan Catur Febriyanto Sutopo
Program Magister Rekayasa Energi Nuklir � Institut Teknologi Bandung
ABSTRACT
ANALYSIS OF VOID PERCENTAGE IMPACT ON NEUTRONIC AND THERMOHYDRAULIC CONDITION OF BOILING WATER REACTOR. Analysis of neutronic and thermohydraulic condition of Boiling Water Reactor have been done by void percentage changes. The analysis did by Matlab software modeling. Neutron distribution approach in reactor core modeling by two energy groups and one dimension neutron diffusion equation solved numerically. By known of neutron flux distribution, temperature in the center of fuel element could be known. The analysis continued with temperature distribution in the surface of fuel element and cooler. The calculation neutron distribution was done radially by assumption the dimension of fuel element infinite in axially. Maximum thermal neutron distribution was happened on 75% void, and for fast neutron on 100%. The power distribution relatively isn�t influenced by void percentage changes. The temperature distribution on fuel element will little decrease relatively by raising of void percentage.
Keywords: void percentage, neutronic, thermohydraulic, BWR, neutron distribution, temperature distribution.
Proceeding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
VERIFICATION OF THERMAL-HYDRAULIC DESIGN OF PWR CORE 1000 MWe CLASS.
Muh. Darwis Isnaini dan Pudjijanto MS
PTRKN � BATAN, Kawasan Puspiptek Gd.80 Serpong, Tangerang, 15310
ABSTRACT
VERIFICATION OF THERMAL-HYDRAULIC DESIGN OF PWR CORE 1000 MWe CLASS. Verification of thermal-hydraulics design of PWR core 1000 MWe class (reactor power of 900 � 1,100 MWe) was carried out. The reasoning of this research is, the decision of nuclear power plant (NPP) type has not been selected yet, because there were not enough technical data about any kinds of NPP�s characteristics owned. Therefore, a verification was carried out, constraint for PWR type only, by the objective, comparing the PWR�s thermal-hydraulic characteristics, in order that the result be usable as opinion input in deciding what NPP�s type will be selected. Verifications were carried out for two types of PWR, i.e., PWR 2nd Generation (PWR G2) made by Mitsubishi that contains of 157 fuel element assemblies for 2,660 MWt and Typical PWR that made by Westinghouse that contains of 193 fuel element assemblies for 3,411 MWt. The calculations were performed using THAL program (Thermal- Hydraulics Assigned for LWR) in which the program is useful for thermal-hydraulics calculation in light water typed reactor of BWR or PWR. The program is capable of calculation of one fuel rod or one fuel assembly or one core in time. For reactor power of 2,660 MWt with flow rate of 45,400 ton/h and inlet temperature of 288 �C, the verification result of Mitsubishi PWR G2 design shows that outlet temperature is 340 �C (different is 4.62%), maximum cladding temperature and meat temperature are 360.71oC and 1,943.83oC, and the safety margin for DNBR is 2.15. Whereas the verification result for Westinghouse Typical PWR design for reactor power of 3,411 MWt with flowrate 60,000 ton/h and inlet temperature 292.6oC shows that the outlet temperature is 344.7oC, the maximum cladding and meat temperature are 372.17 �C and 2,036.06 �C, and safety margin for DNBR is 1.45. Referring to the maximum meat temperature limit is 2,594 �C to avoid fuel melting and safety margin for DNBR is 1.24, both PWR typed nuclear power plant can be operated safely. Even the calculation used global inputs of one fuel assembly, the program results axial temperature distribution for coolant, cladding, and fuel meat including safety margin of DNBR as well.
Keywords: Thermal-hydraulics design, PWR 2nd Generation, PWR Typical.
Prosiding Seminar Nasional Ke-15 "TEKNOLOGI DAN KESELAMATAN PLTN SERTA FASILITAS NUKLIR", Surakarta, 17 Oktober 2009
EVALUATION AND ANALYSIS OF COOLING STUDY OF HIGH TEMPERATURE HEATED ROD ON QUEEN-II TEST SECTION FOR BOTTOM REFLOODING PROCESS EXPERIMENTS.
Puradwi I.W., M. Juarsa, Hendro Tj.
Pusat Teknologi Reaktor dan Keselamatan Nuklir - BATAN
Kawasan PUSPIPTEK Gedung 80, Serpong 15310, Tangerang
ABSTRACT
EVALUATION AND ANALYSIS OF COOLING STUDY OF HIGH TEMPERATURE HEATED ROD ON QUEEN-II TEST SECTION FOR BOTTOM REFLOODING PROCESS EXPERIMENTS. Emergency cooling processes of LOCA in PWR is done by core reflooding especially to cooling fuel rod which is still hot, with bottom reflooding of water flow is injected from ECCS to reactor core. The bottom reflooding phenomenon is necessary to investigated and to understood through post LOCA boiling heat transfer phenomenons on fuel rod with boiling heat transfer experiments and the end results are heat flux and boiling curves. The Experimental simulation of heated rod which is cooled from bottom to up by water cooling flow, was done using water cooling flow rate variation G= 15 g/s, G= 59 g/s dan G= 140 g/s at 850-900 oC of temperature and initial heated rod temperature variation as 600 oC, 700 oC and 800 oC on 62 g/s flow rate. Analysis was done base on the measurement, visualization, and analytic calculations. The experiment results was showing that the heat transfer on bottom reflooding is occurred an film boiling heat transfer at bottom reflooding process which is initiating the cooling process of the bottom reflooding mainly on initial heated rod temperature higher than 800 oC. For initial heated rod temperature variations with same flow rate, MHF and CHF values are tend to increase for higher initial heated rod temperature. The evaluation of study results was saw that boiling curve is resulted through transient cooling of the high temperature heated rod and the boiling is a flow boiling. This flow boiling was saw that post burn out film boiling and flow rate variable and initial heated rod temperature were give important meaning in the cooling process of post LOCA mainly on MHF and CHF.
Keywords: boiling, film, heat transfer, bottom reflooding, LOCA.
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
Prosiding Seminar Nasional ke-15 Teknologi dan Keselamatan PLTN Serta Fasilitas Nuklir Surakarta, 17 Oktober 2009
INDIA SAFEGUARDS AGREEMENT: IS NUCLEAR NON-PROLIFERATION REGIME STRENGTHENING?
Eri Hiswara
Pusat Teknologi Keselamatan dan Metrologi Radiasi � BATAN
Kawasan Nuklir Pasar Jumat, Jl. Cinere Pasar Jumat, Jakarta 10270
ABSTRACT
INDIA SAFEGUARDS AGREEMENT: IS NUCLEAR NON-PROLIFERATION REGIME STRENGTHENING? The August 2008 approval by the International Atomic Energy Agency (IAEA) Board of Governors of an India-specific safeguards agreement was an important step toward implementing the July 2005 nuclear deal between the then U.S. President Bush and Indian Prime Minister Singh. Under this deal, President Bush pledged to seek an exemption for India from U.S. nonproliferation standards. The President also committed to seeking an exemption from similar international rules adopted by the Nuclear Suppliers Group (NSG), which was then granted on September 2008. Even though the IAEA Director General has voiced a support to this India-specific safeguards agreement, the meaning and legal requirements established by the agreement, particularly in terms of the conditions under which safeguards may be terminated, are still controversial and doubted by many. The contents of the NSG document which exempts India from its guidelines for international nuclear trade can also be variously interpreted. This international development on nuclear safeguards in turns bring about a big question, is nuclear non-proliferation regime strengthen with this new India safeguards agreement?
Keywords: safeguards agreement, non-proliferation, nuclear suppliers group
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009
STUDY ON REACTOR TRIP SYSTEMS PERFORMANCE OF EPR 1600
Nafi Feridian, Yuliastuti
Pusat Pengembangan Energi Nuklir (PPEN) - BATAN
Jl. Mampang Prapatan, Kuningan Barat, Jakarta
ABSTRACT
STUDY ON REACTOR TRIP SYSTEMS PERFORMANCE OF EPR 1600. The study on reactor trip systems of European Pressurized Reactor (EPR) 1600 has been carried out. EPR design has applied several sophisticated safety feature in order to enhance the overall plant safety margin. One of the outstanding features was the implementation of fully digitalized instrumentation and control (I&C) system which could reduce human error sensitivity. I&C system of EPR 1600 was divided into four levels namely process interface, system automation, supervision and control unit, and business management systems. Plant protection system is part of automation systems. This system could actuate the reactor trip system automatically or manually by operator. Reactor trip system actuation process on EPR 1600 was carried out by Acquisition and Processing Unit and also the Actuation Logic Unit. The study showed that the advantages of EPR reactor trip system was lay on the reliability system which have four fold 100% redundancy, room separating between each division, the self supporting of each division to do the actuation system, and the implementation of 2/4 logic system.
Keywords: EPR 1600, instrumentation and control system, reactor trip system.
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009
THE PROBABILISTIC ANALYSIS OF POWER REACTOR RADIATION SAFETY AT LOCA (LOSS OF COOLANT ACCIDENT) CONDITION
Pande Made Udiyani
Pusat Teknologi Reaktor dan Keselamatan Nuklir - BATAN
PTRKN-BATAN, Kawasan PUSPIPTEK Gd.80, Serpong, Tangerang, 15310
ABSTRACT
THE PROBABILISTIC ANALYSIS OF POWER REACTOR RADIATION SAFETY AT LOCA (LOSS OF COOLANT ACCIDENT) CONDITION. Operation of power reactor NPPs (Nuclear Power Plants) requires important document safety analysis. The objectives this paper is to get supporting data for Safety Analysis Report (SAR) document. The probabilistic analysis for radiologic and environment consequences done at accident condition with LOCA postulation. The assumption of LOCA is Large Break Loss of Coolant Accident ) based on the fourth level of DBA (Design Basis Accident),which was started double guillotine break in the primary pipes. The assumptions of fission product releases from core inventory to containment are: Emergency core cooling system (ECCS) injection cold and hot legs types, 3 % failed fuel fraction; by gap inventory; fraction of the core inventory present in the gap are: 7,5 % Kr; noble gas Xe 3,95 %; and Iodine is 0,65 %. Released core inventory to containment for Kr-85 is 0,23%, Xe-133 (0,07 %), I-131 (0,02 %) and Cs-137 (0,06 %). The containment is without spray system. Source term data are from PWR 1000 MWe generic power reactor with Muria Peninshula site study. The radiology consequences was estimated by PC Cosyma programme code with probabilistic calculating mode. From various pathway, the estimation results are: the maximum mean individual probability distributions dose is 2,98 x 10-4 mSv/year for 1 km radius reactor distances. This dose is under BAPETEN and IAEA dose limit for accident.
Key words: probabilistic, radiation dose, LOCA, safety
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009
ANALYSIS OF MAIN STEAM LINE BREAK ACCIDENT OUTSIDE CONTAINMENT ON PWR
Andi Sofrany Ekariansyah
Pusat Teknologi Reaktor dan Keselamatan Nuklir - BATAN
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
ANALYSIS OF MAIN STEAM LINE BREAK ACCIDENT OUTSIDE CONTAINMENT ON PWR. An analysis of main steam line break (MSLB) accident on PWR using RELAP/SCDAPSIM/Mod3.2 code has been done. Accident analysis is important to be performed because it is contained in the Safety Analysis Report (SAR) before the nuclear power plant construction for assessing the safety of reactor facility. The purposes of analysis are to know the accident sequence characteristics and any parameter changes important for safety. The Tsuruga Unit 2 with 1160 MWe output is used as reference. One main steam pipe located at one loop outside the containment is assumed to experience a double-ended steam pipe break. Analysis is performed for 2 cases, which are one MSIV of one secondary loop is failed to close (Case 1) and all MSIVs of 4 secondary loop are failed to close (Case 2). The results show that the excessive core cool down on Case 2 is higher than on Case1, shown by the higher decrease of primary pressure, decrease of pressurize water level into empty condition, and the higher decrease of average core temperature. There are no coolant condition at upper vessel and voids at core channel, which is disappeared along with the safety injection from ECCS. An increase of core reactivity and thermal power are indicated, but no differences between the two cases. From the safety point of view, the success of reactor protection system to trip the reactor and operability of safety injection from ECCS is able to maintain the core in a safe and cool condition according to the safety criteria.
Keywords: Main steam line break, PWR, RELAP/SCDAPSIM/Mod3.2
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009
SAFETY ASSESSMENT Of SMALL AND MEDIUM SIZE GAS COOLED REACTOR
Suharno
Pusat Teknologi Reaktor dan Keselamatan Nuklir - BATAN
PTRKN-BATAN, Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
SAFETY ASSESSMENT Of SMALL AND MEDIUM SIZE GAS COOLED REACTOR. Water cooled reactor which in operation currently has large rating power of 600 MWe to 1400 MWe. The large rating power reactor is not suitable for supplying small load needs or variation load needs. Therefore it is only suitable for supporting based load. For supporting small load, it can be supplied by gas cooled reactor that designed in modular system with small to medium size reactor under 600 MWe. The status of gas cooled reactor is still in design development, therefore it must be proved that the safety level is higher than safety level of water cooled reactor. That case to be the aim of this assessment to formulate qualitatively the safety of gas cooled reactor is higher. The methode used in this assessment is by reviewing and comparing qualitatively the characteristic and the features of gas cooled reactor to light water cooled reactor PWR-Mitsubishi type. From the point of comparison it shall be concluded the level of safety of gas cooled reactor. The characteristic and the features of gas cooled reactor which contribute to the high safety level that are the reduce probability and severity of core damage and radioactive release are deeply assessed and compared to the characteristic and feature of that light water cooled reactor. The result which is also as conclusion shows that qualitatively the safety level of gas cooled reactor is higher than safety level of light water cooled reactor PWR-Mitsubishi type.
Key words : gas cooled reactor, water cooled reactor, safety feature, safety level.
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009
THE EFFECT OF FUEL VOLUME VARIATION TO THE BREEDING CHARACTERISTIC OF PASSIVE COMPACT MOLTEN SALT REACTOR (PCMSR)
Andang Widi Harto
Jurusan Teknik Fisika, Fakultas Teknik, Universitas Gadjah Mada
Jl. Grafika 2. Yogyakarta � 55281
ABSTRACT
THE EFFECT OF FUEL VOLUME VARIATION TO THE BREEDING CHARACTERISTIC OF PASSIVE COMPACT MOLTEN SALT REACTOR (PCMSR). Passive Compact Molten Salt Reactor (PCMSR) is a nuclear reactor using molten salt fuel (UF4-ThF4-LiF) and graphite moderator. The using of U and Th fuel makes a possibility to design the reactor core that has breeding capability at a thermal neutron spectrum. The using of molten salt fuel makes a possibility to operate the reactor at high temperature at low pressure in order to increase the thermal efficiency and reduce possibility of accident due to overpressure. The breeding capability of PCMSR is depend on neutronics characteristics of its fuel assembly design. The neutronics characteristic of the fuel assembly is depend on the fuel volume fraction of the fuel assembly. The burn up and criticality calculation of this paper shows that the breeding capability of the PCMSR fuel assembly is achieved at the fuel volume fraction of 15%. Below this value, plutonium must be injected continually to maintain the reactor criticality. It means that the reactor has no breeding capability. At the fuel volume fraction more than 15%, the plutonium injection is needed only for the first 2.5 years of the PCMSR operation and at the next years, the PCMSR can maintain its criticality merely with Th-232 fuel input. At the fuel volume fraction higher than 15%, the reactor criticality increases with time due to the U-233 bulid up. It means that the breeding capability increases by increasing the fuel volume fraction.
Key words: PCMSR, breeding capability, fuel volume fraction
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009
FINANCE ASPECT CALCULATION ON THE ESTABLISHMENT OF NUCLEAR FUEL
ELEMENT PLANT TYPE OF PWR IN INDONESIA THROUGH CONVERSION LINE OF
AMMONIUM URANYL CARBONATE (AUC)
Bambang G. Susanto
Pusat Teknologi Bahan Bakar Nuklir (PTBBN), BATAN
Kawasan Puspiptek Serpong, Tangerang 12440
ABSTRACT
FINANCE ASPECT CALCULATION ON THE ESTABLISHMENT OF NUCLEAR FUEL ELEMENT PLANT TYPE OF PWR IN INDONESIA THROUGH CONVERSION LINE OF AMMONIUM URANYL CARBONATE (AUC). The calculation of finance aspect on establishment of nuclear fuel element plant through conversion lane of ammonium uranyl carbonate ( AUK) having capacity of 710 tons UO2/year has been conducted. From finance aspect that has been calculated, is obtained the data or information that the establishment of nuclear fuel element plant type PWR requires a number of high investment costs at early construction stage as well as operating cost. The plant is still interesting enough to be built even though at early stage requires high costs of investment. By using 'PROFITABILITY ANALYSIS - 1,1 xls program, the results of calculation finance aspect are obtained the following data�s: total permanent investments equal to US $ 151,081,900,-, working capital US $ 283,432,500,- production cost per year US $ 1,280,759,451; cash flow projection in year 20th after operation US $ 403,127,900; number of net incomes obtained during 20 years is equal to US $ 2,204,463,300,-; break even point equal to 17.42 %; the expense of decommissioning US $ 141,559,900; the price of fuel element is US $ 1,061,025,-/assemblies.
Keywords: Finance aspect, investment, production cost, cash flow projection, nuclear fuel element plant
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
ECONOMIC RISK ANALYSIS OF NPP
Suparman, Elok S. Amitayani
Pusat Pengembangan Energi Nuklir (PPEN), BATAN
Jl. Kuningan Barat, Mampang Prapatan, Jakarta Selatan, 12710
ABSTRACT
ECONOMIC RISK ANALYSIS OF NPP. Along with technology risks, economic risks are major consideration in a country�s nuclear power plant (NPP) program, due to its large investment cost if compared to conventional plants. At least there are two kinds of economic risks to pay attention to. Firstly, the costs escalation and the delay of construction work and secondly, the low capacity factor and the short lifetime. The parameter under consideration in this paper is the generation cost, measured in US$/kWh. Generation cost by definition is all the costs spent during the plant life time, takes into account the fixed and variable costs. Those economic risks mentioned above give a direct impact to the plant generation cost as they will be a burden to owner�s balance of payment. The parameters to be tested in this paper will be construction time, capacity factor, and plant lifetime. The risk of costs escalation will not be discussed further. The calculation results from IAEA�s DEEP program show that there is a relation between the three parameters and the generation cost. The delay of construction time will add up the generation cost, the high capacity factor will lower the generation cost, while the long lifetime of the plant will give an interesting cheaper generation cost unless the plant is extended over its economic lifetime. A comprehensive understanding on NPP economic risks will be a helpful tool for the decision makers.
Keywords: economic risks, NPP, generation cost
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
PSA APPROACH TO PREVENT EXTERNAL INCIDENT ON THE SAFETY OF NPP SITE
Basuki Wibowo, Imam Hamzah, Yarianto SBS
Pusat Pengembangan Energi Nuklir, BATAN
Jl. Kuningan Barat, Mampang Prapatan, Jakarta Selatan
ABSTRACT
PSA APPROACH TO PREVENT EXTERNAL INCIDENT ON THE SAFETY OF NPP SITE. PSA Methodology Approach Assessment for NPP Site Safety from External Events Hazards. Assessment for the application of PSA methodology of NPP site safety from external events hazards have done for the purpose of the effectiveness of those methodologies. The way of methodologies are: evaluation from IAEA and US-NRC correlated references. Base on those assessments, the contributions of external events hazards to NPP site safety design base will increase significantly after the IAEA standards criteria full applied for new NPP generation in the future. The IAEA standards criteria documents are: guideline and technical document of NPP site safety evaluation from external events hazards. Safety design base for existing NPP only considered the contribution from internal events hazards. Generally, the criteria applications for NPP external events hazards start from screening stage, where only significant hazards considered for design base. The next stage is taking detailed characterized of those hazards for specific site. After considering external events hazards for NPP safety design base, the probabilistic safety margin increase significantly from 10-4 to 10-8 per year. Uncertainty factor for PSA methodology for NPP site safety from external events hazards can be controlled by synchronizing of internal events and external events hazards.
Key Words: PSA (probabilistic safety assessment), site safety, external and internal hazards, safety design base.
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
STUDY OF PLANT LIFE EXTENSION FOR THE FIRST NUCLEAR POWER PLANT IN INDONESIA
Sri Nitiswati
Pusat Teknologi Reaktor dan Keselamatan Nuklir - BATAN
Kawasan PUSPIPTEK Gd. 80, Serpong, Tangerang, 15310
ABSTRACT
STUDY OF PLANT LIFE EXTENSION FOR THE FIRST NUCLEAR POWER PLANT IN INDONESIA. Generally, nuclear power plants originally had a nominal design lifetime of 30 years up to 40 years. Engineering assessment of many nuclear power plants over the last decade has established that many nuclear power plants can operate longer than its nominal design life. Many countries have been operated successful nuclear power plant longer than its nominal design lifetime by applying of license renewal for 20 years or longer. Study of plant life extension for the first nuclear power plant in Indonesia needs to be done since Indonesia has nuclear option as mix energy for the future. Aims of this study are to obtain procedure and requirements need to have plant life extension. Its method based on the IAEA document on �Plant Life Management for Long Term Operation of Light Water Reactor�, Technical Report Series No. 448, Vienna, (2006), with compares with other countries experience such as : Korea, Canada, USA, Russia, and Spain. As the conclusion that procedure and requirement needs for plant life extension are: PLEX organization, ageing management program document, time limited ageing analysis, research and development document, radiation impact document, and etc. A propose of plant life extension procedure has been obtained but is still in preliminary.
Keywords: ageing, plant life management, plant life extension, license renewal
Published : Proceeding "SEMINAR NASIONAL PENGEMBANGAN ENERGI NUKLIR II 2009", Jakarta, 25 Juni 2009.
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
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
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.
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
MODEL OF THE PHYSICAL PROTECTION SYSTEM DESIGN FOR NUCLEAR FACILITIES IN INDONESIA
Ign. Djoko Irianto
Center For Safeguards Technology � Batan
Kawasan Puspiptek Gedung 90 Lantai 3, Serpong � 15310
ABSTRACT
MODEL OF THE PHYSICAL PROTECTION SYSTEM DESIGN FOR NUCLEAR FACILITIES IN INDONESIA. The design of physical protection system (PPS) for nuclear facilities must be arranged in accordance national and international regulations. The international regulation is outlined on INFCIRC/225 Rev.4 while the national regulations are issued by National Nuclear Energy Agency (BATAN) and Nuclear Energy Control Board (BAPETEN). The design of physical protection system consists of several steps: determine the PPS objectives, make a preliminary design of PPS, analyze the PPS design and probably redesign the PPS. In determining the PPS objectives, first collect information on the condition of the nuclear facilities, including all features of the facilities, operating procedures and elements of physical protection systems. Then, conduct a detailed study of a range of adversaries that the PPS may encounter. Finally design PPS by arranging the best combination of the PPS elements: fences, barriers, sensors and monitor equipment, entry control equipment, communication equipment, procedures and security forces.
Keywords : Physical Protection, Nuclear Materials, Nuclear Facilities,
Published : Proceeding "Seminar Teknologi Pengamanan Bahan Nuklir Ke:1", Jakarta, 21 Nopember 2000.
DESIGN BASIS THREAT (DBT) APPROACH FOR THE FIRST NUCLEAR POWER PLANT SECURITY SYSTEM IN INDONESIA
Ign. Djoko Irianto
Center For Safeguards Technology � Batan
Kawasan Puspiptek Gedung 90 Lantai 3, Serpong � 15310
ABSTRACT
DESIGN BASIS THREAT (DBT) APPROACH FOR THE FIRST NUCLEAR POWER PLANT SECURITY SYSTEM IN INDONESIA. Design Basis Threat (DBT) is one of the main factors to be taken into account in the design of physical protection system of nuclear facility. In accordance with IAEA's recommendations outlined in INFCIRC/225/Rev.4 (Corrected), DBT is defined as: attributes and characteristics of potential insider and/or external adversaries, who might attempt unauthorized removal of nuclear material or sabotage against the nuclear facilities. There are three types of adversary that must be considered in DBT, such as adversary who comes from the outside (external adversary), adversary who comes from the inside (internal adversary), and adversary who comes from outside and colludes with insiders. Current situation in Indonesia, where many bomb attacks occurred, requires serious attention on DBT in the physical protection design of NPP which is to be built in Indonesia. This paper is intended to describe the methodology on how to create and implement a Design Basis Threat in the design process of NPP physical protection in Indonesia.
Keywords : Physical Protection, Design Basis Threat, Nuclear Materials, Nuclear Facilities,
Published : Proceeding "Seminar Teknologi Pengamanan Bahan Nuklir Ke:5", Jakarta, 29 September 2004.
IMPLEMENTATION OF PHYSICAL PROTECTION SYSTEMS IN BATAN'S NUCLEAR FACILITIES
Ign. Djoko Irianto
Center For Safeguards Technology � Batan
Kawasan Puspiptek Gedung 90 Lantai 3, Serpong � 15310
Email: igndjoko@yahoo.com
ABSTRACT
IMPLEMENTATION OF PHYSICAL PROTECTION SYSTEMS IN BATAN'S NUCLEAR FACILITIES. Physical protection systems for nuclear facilities have 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 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. The plan for physical protection system for nuclear facilities should also consider the categorization of nuclear materials as the subject of security. This paper discusses some of the implementation of physical protection system in BATAN, assessment on it efficiency and effectiveness, also possible improvement.
Keywords : Physical Protection System, Nuclear Materials, Nuclear Facilities,
Published : Proceeding "Seminar Teknologi Pengamanan Bahan Nuklir Ke:4", Jakarta, 13-14 Oktober 2003.
VALIDATION OF NUCLEAR DATA FOR STRUCTURAL MATERIALS BY INTEGRAL METHODS
Sahala Lumbanraja, Masdin, Ign.Djoko Irianto
Center for Nuclear Technology Assessment - BATAN
ABSTRACT
VALIDATION OF NUCLEAR DATA FOR STRUCTURAL MATERIALS BY INTEGRAL METHODS. Nuclear data for structural materials is needed to ensure the operation safety of nuclear reactors. To avoid some accident of nuclear reactors, so validation of nuclear data is needed continuously. Some methods have developed to increase accuracy of nuclear data, like differential and integral methods. In this paper, we will describe the integral method by energy independent adjoint flux that based only at measurement of reactivity absorption. Different structural material such as Fe, Ni, Mn, Cu and Cr, their nuclear data have been evaluated.
Keywords : Nuclear data, Structural materials, Integral methods,
Published : Proceeding "Seminar Ke-4 Reaktor Temperatur Tinggi dan Teknologi Nuklir", Jakarta, 15-16 February 1999, ISSN 0854-8803
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
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
