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
