Surip Widodo

Simulation of Heat Flux Effect in Straight Heat Pipe as Passive Residual Heat Removal System in Light Water Reactor Using RELAP5 Mod 3.2

Heat pipe is considered being used as a passive system to remove residual heat that generated from reactor core when incident occur or from spent fuel pool. The present research is aimed to studying the characteristics of straight heat pipe as passive residual heat removal system. As an initial step, a numerical simulation was conducted to simulate the best experimental design set up being prepared for the next step of the research. The objective is to get the thermal hydraulic characteristic due to variation of heat flux of heat source. The thermal hydraulic RELAP5 MOD 3.2 code is used to simulate and analyze the straight heat pipe characteristics. Variations of heat flux are 1567 Watt/m2, 3134 Watt/m2, 4701 Watt/m2, 6269 Watt/m2, and 7837 Watt/m2. Water as working fluid is heated on evaporation section with filling ratio 60%. Environmental air with variation 5 m/s and 10 m/s of velocity are used as external cooler. Straight heat pipe used in the simulation is wickless with 0.1 m of diameter and 6 m of length. The results show that higher heat flux given to the evaporator section will lead to more rapid heat transfer and achievement of steady state condition. The increasing of heat flux leads to an increase of evaporation of the working fluid and of pressure built in the heat pipe affecting higher saturation temperature of working fluid. Heat flux loading must consider the velocity of air as heat removal in the condenser to prevent dry out phenomenon in the evaporator. Based on the results, given the experimental set-up, the optimum range of experimental parameters could be determined.

Preliminary investigation of natural circulation stability in FASSIP-01 experimental facility using RELAP5 code

Natural circulation has an important role in the safety system of a nuclear reactor. It is part of passive system, which is operated without any external prime mover. The behavior of natural circulation shall be known to be effectively incorporated in the safety system of a nuclear reactor. This paper describes the preliminary results of the study on natural circulation conducted at a vertical rectangular experimental facility FASSIP-01. The RELAP5 code is used to simulate the experiment. The goal is to study the validity of the RELAP 5 model to analyze the natural circulation in FASSIP-01. The simulation could provide a reasonable good results describing the natural circulation established in the rectangular loop of FASSIP-01. However, there are still discrepancies with the experimental results, especially with regard to mass flow rate and temperature distribution in the loop. The model needs be improved and further comprehensive data from experiments are necessitated for better validation, as well.Natural circulation has an important role in the safety system of a nuclear reactor. It is part of passive system, which is operated without any external prime mover. The behavior of natural circulation shall be known to be effectively incorporated in the safety system of a nuclear reactor. This paper describes the preliminary results of the study on natural circulation conducted at a vertical rectangular experimental facility FASSIP-01. The RELAP5 code is used to simulate the experiment. The goal is to study the validity of the RELAP 5 model to analyze the natural circulation in FASSIP-01. The simulation could provide a reasonable good results describing the natural circulation established in the rectangular loop of FASSIP-01. However, there are still discrepancies with the experimental results, especially with regard to mass flow rate and temperature distribution in the loop. The model needs be improved and further comprehensive data from experiments are necessitated for better validation, as well.

Numerical study on natural circulation characteristics in FASSIP-02 experimental facility using RELAP5 code

As in many other energy generating system, in the nuclear power plant, the use of natural circulation principle for the safety system is increasingly considered. The natural circulation could play an important role in providing emergency cooling of the nuclear reactor. The reliability of the use of natural circulation in the nuclear power plant should be demonstrated in order to assure the level of its safety. FASSIP-02 is a large scale test facility to study the natural circulation in the safety system of a nuclear power plant. To study the characteristics of the natural circulation and to help validating the design of FASSIP-02, a numerical study using RELAP5 code is undertaken. Based on the existing design of FASSIP-02, the numerical simulation is done with two variables, i.e. the heat flux and the pipe diameter. The effect of heat transfer surface area for dissipating the heat is also studied. The results show the natural circulation established in the FASSIP-02. The characteristics of the natural circulation with the values of several important parameters such as temperature, mass flow rate and pressure in the loop are obtained. The RELAP 5 calculations have provided the results that could be used to support the design and future operation of FASSIP-02.

Preliminary investigation of wickless-heat pipe as passive cooling system in emergency cooling tank

To improve the thermal safety of a nuclear power plant, a wickless-heat pipe is proposed as a passive cooling system on an emergency cooling tank. Heat pipe will keep the water in the emergency cooling tank on normal operating temperature. The objective of this preliminary research is to investigate the characteristics of the wickless-heat pipe. RELAP/SCDAPSIM/MOD3.4 thermal-hydraulic code was used to analyze the characteristics of the wickless-heat pipe. The simulation results will be used as a basis for designing the future experimental investigations of the proposed wickless-heat pipe. The influence of the hot water temperature in the emergency cooling tank and the mass flow rate of coolant on the water jacket were investigated. The hot water temperature in the emergency cooling tank was varied, i.e. 60, 70, and 80°C. The mass flow rate of condenser coolant in the water jacket was varied, i.e. 2, 4, and 8 L/min. The initial pressure of the heat pipe was -74 cm Hg. De-mineralized water, which serves as heat pipe working fluid, was charged with filling ratio of 80%. The working fluid in the heat pipe can circulate naturally in stable condition if cooling water could absorb the latent heat in condenser. The simulation results showed that for constant hot water temperature, the increasing of condenser coolant volumetric flow rate will decrease the temperature of evaporator and condenser. While for constant condenser coolant volumetric flow rate, the increasing of hot water temperature will increase the temperature of evaporator and condenser.To improve the thermal safety of a nuclear power plant, a wickless-heat pipe is proposed as a passive cooling system on an emergency cooling tank. Heat pipe will keep the water in the emergency cooling tank on normal operating temperature. The objective of this preliminary research is to investigate the characteristics of the wickless-heat pipe. RELAP/SCDAPSIM/MOD3.4 thermal-hydraulic code was used to analyze the characteristics of the wickless-heat pipe. The simulation results will be used as a basis for designing the future experimental investigations of the proposed wickless-heat pipe. The influence of the hot water temperature in the emergency cooling tank and the mass flow rate of coolant on the water jacket were investigated. The hot water temperature in the emergency cooling tank was varied, i.e. 60, 70, and 80°C. The mass flow rate of condenser coolant in the water jacket was varied, i.e. 2, 4, and 8 L/min. The initial pressure of the heat pipe was -74 cm Hg. De-mineralized water, which serves as ...

PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC). Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO) independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih rinci. Kata kunci : pemodelan, Generasi III+, RELAP5. Westinghouse’s AP1000 reactor design is the first Generation III+ nuclear power reactor to receive final design approval from the U.S. Nuclear Regulatory Commission (NRC). Currently, the China’s utilities are starting construction several units of AP1000 on two selected sites for scheduled operation in 2013–2015. The AP1000, based on proven technology of Westinghouse-designed PWR with enhancement on the passive safety system, could be considered to be built in Indonesia referring to the requirements of government regulation No. 43/2006 regarding the Nuclear Reactor Licensing. To be accepted by the regulation agency, the design needs to be verified by independent Technical Support Organization (TSO), which can be done using RELAP5 computer code as accident analyses. Currently, NPP safety accident analysis is performed for PWR 1000 MWe of generation II or conventional type. Considering that nowadays references about the technology of AP1000 that includes passive safety technology has been available and assessed, a modeling activity used for future accident analyzes is introduced. Method for developing the model refers to IAEA guide consisting of plant data collection, engineering data and input deck development, and verification and validation of input data. The model developed should be considered preliminary but has been generally representing the AP1000 systems as the basic model. The model has been verified and validated by comparing thermalhidraulic parameter responses with design data in references with ± 13% deviation except for core pressure drop with 13% lower than design. As a basic model, the input deck is ready for further development by integrating safety system, protection system and control system model specified for AP1000 for purposes of safety simulation in detailed way. Keywords : Modeling, Generation III+ , RELAP5.