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
