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SIGMA EPSILON - Majalah Ilmiah Teknologi Keselamatan Nuklir

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Published by Badan Tenaga Nuklir Nasional

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Core Subject : Science,

Computer Science & IT Energy

SIGMA EPSILON adalah majalah ilmiah yang menyajikan makalah hasil kegiatan riset dan kegiatan teknis penunjang riset lainnya yang dilaksanakan di Pusat Reaktor dan Keselamatan Nuklir (PTRKN) Badan Tenaga Nuklir Nasional.

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Articles 6 Documents

Search results for , issue "Vol 19, No 1 (2015): Februari 2015" : 6 Documents clear

INHIBITION CHARACTER ANALYSIS OF CORROSION INHIBITOR ON CARBON STEEL MATERIALS IN 1M HCL SOLUTION USING THE EIS METHOD Rahayu Kusumastuti; Yustinus Purwamargapratala; Sofia Butarbutar; Sagino Sagino; Sriyono Sriyono; Abdul Hafidz
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 19, No 1 (2015): Februari 2015
Publisher : Badan Tenaga Nuklir Nasional

Research on the effect of the concentration of the inhibitor on the corrosion behavior of carbon-steel material has been done. The research was started by immersing the prepared carbon-steel plate in a 1 M HCl en-vironment. After that, corrosion inhibitor was added with several concentrations, which are 0, 100, 200, 300, and 400 ppm in to that environment, to be stirred using a magnetic stirrer at 300 rpm for 30 minutes under room temperatur condition. The effect of the added inhibitor was then analyzed using the Electrochemical Impedance Spectroscopies (EIS) method. The experiment results showed that the greater the concentration of the inhibitor, the greater the resistance, so that the metal is more pro-tected from corrosion attack. The calculation results showed that the inhibitor efficiency is directly proportional to the concentration of inhibitor that is achieved at a concentration of 400 ppm with an efficiency of 71.24%.

ANALYSIS OF THE EFFECT OF ELEVATION DIFFERENCE BETWEEN HEATER AND COOLER POSITION IN THE FASSIP-01 TEST LOOP USING RELAP5 Andi Sofrany Ekariansyah; Hendro Tjahjono; Mulya Juarsa; Surip Widodo
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 19, No 1 (2015): Februari 2015
Publisher : Badan Tenaga Nuklir Nasional

To understand the natural circulation phenomena on the passive residual heat removal system (PRHRS), development of a test section describing that phenomena in particular in the one phase condition is required. That test facility is named as FASSIP-01 in form of a vertically closed loop consisting of piping compo-nents, one cylinder tank featured with heater elements and one cooler. The heater tank will work as the heat source, and the cooler as the heat sink. This research is intended to support the experimental activity of the FASSIP-01 by conducting a simulation using the RELAP5/SCDAP/Mod3.4. Beside the standard loop configuration, the simulation is also conducted by varying the elevation of heater and cooler position to evaluate the best position resulting in the most optimal natural circulation. The results will be used as the comparison with the later performed experiment. The simulation result shows that for the case where the heater position is at the same level with the cooler position, the temperature distribution of the water after the heater and after the cooler are higher than the other two position. Looking at the natural circulation, that position results in the lowest mass flow. The position with the heater below the cooler will result in the best mass flow. On that position, only an optimiza-tion in the heat transfer surface area is needed to increase the heat transfer coefficient and secondary mass flow to remove the heat are needed to obtain more optimal performance of the water circulation caused by the density difference in the FASSIP-01 test loop.

PERFORMANCE ANALYSIS OF HELIUM INVENTORY CONTROL OF RGTT200K COOLING SYSTEM Sriyono Sriyono; Rahayu Kusumastuti; Sofia Butarbutar; Geni Rina Sunaryo
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 19, No 1 (2015): Februari 2015
Publisher : Badan Tenaga Nuklir Nasional

RGTT200K is a power reactor, designed based on HTGR tech-nology having capability to operate at high temperatures. RGTT200K features are 200 MWth power, helium-cooled, graphite moderator and reflector, pebble fuel type, and uses the Brayton direct cycle. Helium Inventory Control System (HICS) is one of its safety system which maintains the pressure, the helium coolant quality and quantity to meet safety requirements. The HICS consists of 3 subsys-tems, namely: Inventory Control System (ICS), Helium Purification System (HPS), and Helium Make-Up System (HMS). All of the systems have the function to maintain pressure, helium quality and quantity so that the reactor can operate reliable and safely. This paper discusses the performance of the ICS, which is integrated to the reactor coolant. The research objective was to determine the helium storage tank response rate, when primary coolant is overpressured and depressurized. The methodology used in this research is modeling and simulation by using ChemCAD. In previous re-search, the HPS, ICS and HMS have been modeled but have not been integrated yet in to the primary coolant. The simulation results showed that the time required for the injection tank back to the cool-ant normal pressure of 52 bars, due to depressurization up to 5 % was 160 seconds. While the time required for bleeding / blowdown to the storage tanks due to overpressurization up to 5 % was 186 seconds.

PENGKAJIAN METODA PENGOLAHAN DATA NUKLIR UNTUK PERHITUNGAN NEUTRONIK HTGR Suwoto Suwoto
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 19, No 1 (2015): Februari 2015
Publisher : Badan Tenaga Nuklir Nasional

Kajian terhadap metoda pengolahan data nuklir yang digunakan dalam perhitungan neutronik teras High Temperature Gas-cooled Reactor (HTGR) perlu dilakukan karena data tampang lintang nuklir yang digunakan dalam perhitungan neutronik memegang peranan penting dalam analisis keselamatan kritikalitas. Metoda pengolahan dalam generasi tampang lintang data nuklir yang biasa digunakan selama ini adalah metode deterministik yang biasa digunakan dalam program deterministik seperti WIMS/D5B dan yang menggunakan metode probabilistik seperti pada program Monte Carlo MCNP5v1.2. Kedua metode tersebut mempunyai keunggulan dan kelemahan masing-masing. Program pengolah data nuklir NJOY, berguna dalam me-nyelesaikan persoalan pengolahan data nuklir dalam format ENDF (Evaluated Nuclear Data File) yang akan digunakan dalam perhitungan fisika neutronik teras reaktor HTGR, baik yang menggunakan tampang lintang multi-kelompok seperti pada program WIMS/D5B dengan memanfaatkan modul WIMSR maupun yang menggunakan tampang lintang energi kontinu pada program MCNP/MCNPX dengan memanfaatkan modul ACER. Data hasil kajian dengan kedua metoda dalam pengolahan dan penyiapan data tampang lintang nuklir digunakan dalam perhitungan neutronik bahan bakar pebble teras HTGR. Hasil perhitungan neutronik bahan bakar pebble HTGR dengan UO2 dengan pengkayaan 10% dan fraksi packing TRISO 10% untuk variasi tem-peratur 900K, 1200K dan 1500K dengan metode probabilistik MCNP5v1.2 menggunakan tampang lintang energi kontinu dari file ENDF/B-VII menghasilkan perbedaan nilai multiplikasi tak hingga (k¥) masing-masing 7,42%, 5,7% dan 4,36% lebih besar dibanding dengan program deterministik WIMS/D5B. Nilai perbedaan tersebut dikarenakan adanya perbedaan pendekatan geometri dan juga pendekatan energi tampang lintang data nuklir yang digunakan. Dengan demikian metode probabilistik dengan MCNP5v1.2 lebih disukai karena dinilai lebih dan teliti dalam perhitungan neutronik teras reaktor HTGR.

PERFORMANCE ANALYSIS OF RECUPERATOR OF RGTT200K CONCEPTUAL DESIGN USING CHEMCAD Piping Supriyatna; Sriyono Sriyono
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 19, No 1 (2015): Februari 2015
Publisher : Badan Tenaga Nuklir Nasional

RGTT200K is a high temperature gas cooled reactor with 200 MW thermal powers, designed with cogeneration concept to produce hydrogen, electricity generation and potable water by desalination process. RGTT200K uses helium gas as a coolant with core inlet tem-perature of 615 °C and outlet temperature about 950 °C. The coolant is circulated at 120 kg/sec mass flow rate at initial pressure of 5 MPa. To keep material integrity of RGTT200K structure, the recu-perator performance of RGTT200K must be maintained due to its double function. Those main func-tions are to reduce the output temperature coolant from the turbine and transfer it back to the main primary circuit using a compressor and to increase the coolant gas from the compressor before ente-ring the core again. This paper describes an analysis to evaluate the recuperator performance by mo-delling using ChemCAD computer code. The calculation results showed that to obtain the core inlet temperature of 615 °C with the recuperator effectiveness of 0.95, the value of the logarithmic mean temperature difference (LMTD) should be 2.51, and the recuperator heat load (BPR) of 264.7 and the heat exchanger coefficient and heat exchange (UA) of 10.546 are needed. Based on those values, the difference between the inlet and outlet temperature of reactor core is not so big and still in stable con-dition to maintain the material structure integrity of the core.

CALCULATION OF RADIONUCLIDE CONTENT OF NUCLEAR MATERIALS USING ORIGEN2.1 COMPUTER CODE Ihda Husnayani
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 19, No 1 (2015): Februari 2015
Publisher : Badan Tenaga Nuklir Nasional

Nuclear materials contain a number of radionuclides produced from radioactive decay process. The composition of these radionuclides which are accumulated in-side the nuclear materials changes over the time. The calculation of radionuclide composition inside nuclear materials is very important especially in the aspect of nuclear reactor safety evaluation, nu-clear fuel behavior evaluation, and radioactive waste management. One method to calculate radionu-clide content of nuclear materials is by using ORIGEN2.1 computer code. Beside radionuclide com-position, this code can also calculate some characteristics related to decay process such as total radio-activity, decay heat, and neutron flux. This paper is a literature study about ORIGEN2.1 computer code. A brief description of ORIGEN2.1 and its use for calculating radionuclide content of nuclear materials are presented. Radionuclide content produced from californium-252 decay was chosen as a simple case solved by ORIGEN2.1. Californium-252 was simulated to undergo decay for 10 years. The variables which are calculated by ORIGEN2.1 in this case are radionuclide composition, total radioactivity, total alpha radioactivity, and neutron flux. From the results of this simulation, it is shown that small amount of californium-252 produces high neutron intensity so that it can be used as a reliable neutron source for many applications.

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