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.
