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
