Neutron Beam Characterization on the Beam Tubes of 30 MW G.A. Siwabessy Reactor Using Monte CarloRasito, ; Su’ud, Z; Permana, S
doi: 10.1088/1742-6596/2328/1/012001pmid: N/A
The neutron beam characterization had been conducted on six tubes of RSG-GAS reactor for BNCT application in a simulation using Monte Carlo method with MCNP computer code. The simulation was conducted by modeling the geometry and material of RSG-GAS reactor with a model of radiation source from 235U fission reaction in 40 fuel bundles type U3Si2Al with 235U levels of 19.75%. The distribution of neutron and gamma fluxes was simulated from the reactor core to the edge of beam tube of S1, S2, S3, S4, S5, and S6 with a tube length of 400 cm and a diameter of 30 cm. The highest neutron flux produced by the beam tube of S5 was 4.3×1010 cm−2s−1 with the highest gamma dose was 1362 Sv/j. The lowest neutron flux produced by the beam tube of S6 was 5.9×109 cm−2s−1 with a gamma dose was 51 Sv/j. Based on the characterization result, it was shown that the output neutron beam of each RSG-GAS beam tube had the potential to be used for many application.
An overview of the applicability of SNI IEC 61331-1:2016 on Lead apron for medical useDarmawati, Suzie; Sunarto, Sunarto; Yasmine, Hanna; Santosa, Sigit
doi: 10.1088/1742-6596/2328/1/012002pmid: N/A
The use of lead apron for radiation protection is regulated under the Indonesia Nuclear Regulatory Agency (BAPETEN) Decree no. 8 year 2011 about Radiation Safety and the Use of Diagnostic and Interventional Radiological X-Ray Machine. It listed the apron specifications are as follows: having thickness equivalent to 0.2 mm Pb or 0.25 mm Pb for diagnostic use and equivalent to 0.35 mm Pb or 0.5 mm Pb for interventional use. Further, National Standardization Agency (BSN) had issued SNI IEC 61331-1:2016, providing guidance for testing the plate materials on the apron using 400 kV x-ray machine and 1.3 MeV gamma exposure with narrow beam, to measure the attenuation ratio and air kerma rate. The method used is to determine the attenuation ratio, build-up factors, and equivalent attenuation coefficient. There were 4 different aprons (A, B, C, and D) with 9 measurement points. The results showed the air kerma rate without apron was 0.664 mGy/second, the air kerma rate with lead-equivalent layer was 0.0006 mGy/second, and the best result was produced using the apron C, with the attenuation ratio ranging from 17.2 to 29.1, showing the most homogeneity.
Study Analysis of Multiplication Factor on ADS Transmutation SystemSudarmono, ; Hastuti, E P; Rohanda, A; Ekariansyah, A S; Kasesaz, Y; Suwoto,
doi: 10.1088/1742-6596/2328/1/012008pmid: N/A
New research technology for partition and transmutation (P-T) of minor actinide (MA) contained in high-level waste (HLW) has been carried out to design accelerator-based transmutation system or commonly known as Accelerator Driven Subcritical (ADS) device. The objective of this system is to eliminate or minimize high-level radioactive waste that has long half-life and to develop long-term safety assurance in HLW management. Analysis done by using Origen2.1 and mCnP6 codes. As matrix, Th-232 with optimum weight 75% is used.. Comparing with , analysis results using MCNP shows that the use of PWR nuclear spent fuel for ADS device can be done by reprocessing to eliminate U-238 nuclide, which is the source of the formation of plutonium nuclide and minor actinides. From the results and discussion, it can be concluded that observation on transmutation of transuranic nuclides using ORIGEN2.1 Code on ADS device has been conducted successfully with LMFBR library. 15%-Th232 fuel has been justified as the most optimum ADS fuel based on K-inf reached at EOC on various fuel types. Mass reduction occurs at EOC for U-236, U-235, U-234, Th-232, Np-237, Pu-238, Am-241, Pu-242, and Cm 244. Radionuclides that experience mass changes during cooling time are Pu-238, Pu-240, Am-241, Pa-233, and Cm-244.
Gamma Heating Evaluation of Silicide RSG-GAS Multipurpose ReactorRohanda, A; Waris, A; Kurniadi, R; Bakhri, S
doi: 10.1088/1742-6596/2328/1/012004pmid: N/A
In research reactor, gamma heating deposited in the samples is an important issue because it is related to the samples and the reactor operational safety. Multi Purpose Reactor (MPR) 30 MWth or its called Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) is a research reactor that play a role as a place to irradiate various target material type. RSG-GAS has been operating since 1987 and the last measurement of gamma heating in the core was conducted around twenty years ago in oxide core. In 1996, RSG-GAS core was converted from oxide fuel type to silicide fuel type. The purpose of the conversion are to improve the performance and efficiency of RSG-GAS. As part of RSG-GAS irradiation facilities safety analysis, the gamma heating measurement was re-conducted in order to obtain latest data as benchmark data in silicide core. This paper presents the results of gamma heating measurement of LEU silicide RSG-GAS core in Central Irradiation Position (CIP) at 15 MW and 30 MW power level using gamma calorimeter. There were four types of calorimeter used, which were calorimeter with graphite (C) sample, iron (Fe) sample, aluminum (Al) sample and zirconium (Zr) sample. Gamma heating calculations using GAMSET code were performed to verify the measurement results. The measurement results are lower than the GAMSET results and the gamma heating value increases in proportion to the increase of calorimeter sample atomic number. This results are corresponds to gamma heating benchmarking results of RSG-GAS oxide core. Several optimization efforts both measuring and modeling with GAMSET were conducted as an evaluation and justification of the results. The best optimization results are achieved using the maximum value of the measurement and adjusting the power peaking factor (PPF) distribution. The calculated gamma heating value optimization results at 15 MW power are 2.78 W/g (C sample), 2.74 W/g (Al sample), 3.36 W/g (Fe sample) and 4.60 W/g (Zr sample) while at 30 MW power level are 5.57 W/g (C sample), 5.49 W/g (Al sample), 6.75 W/g (Fe sample) and 9.23 W/g (Zr sample). The best optimization results serve as a benchmark data for developing new gamma heating calculation programs based on 18 gamma energy groups.
Simulation of Erosion Behavior of a Paraffin Plate with Particle MethodIfthacharo, M; Mustari, A P A; Permana, S; Nuril, A
doi: 10.1088/1742-6596/2328/1/012006pmid: N/A
There are many mechanisms in a reactor shutdown function of MSR and inherent self-stabilization. One of those mechanisms is the fuel-salt drain system. The present study focused on the melting and solidification phenomenon that occurs in the freeze valve. An experiment was performed to investigate the erosion behavior of a solid plate by an impinging liquid to time and the effects of fluid viscosity. In addition, numerical modeling based on the MPS method to visualize the heat distribution in the plate will also be carried out. The experiment will be conducted by varying the parameters such as the liquids, temperature, and diameter. Hot water (90°C), molten paraffin, and cooking oil will be used with molded pure paraffin wax will serve as the target plates. The dimension of the target plate is cylindrical, with 44 mm in thickness and 144 mm in length for paraffin wax. The data will then be compared to the MPS simulation. The radial dispersion of the heated liquid and the temperature of the impinged liquid will affect the penetration time, hence making the formation of a mushy zone more likely and promoting the pool effect. The noticeable difference of penetration time between simulation and experiment is likely caused by the changing value of kinematic viscosity of the liquids used in different temperatures. The kinematic viscosity is set to be a constant value in the simulation.
Noise experiments in BRAHMMA subcritical system using isotopic Poisson source and accelerator based neutron sourceRay, N K; Kumar, R; Patel, T; Sarkar, P S; Pant, L M
doi: 10.1088/1742-6596/2328/1/012009pmid: N/A
Reactivity measurement using noise methods have been carried out in zero power subcritical system BRAHMMA, installed in BARC, India. During noise experiments, an Am-Be neutron source (1Ci) has been placed at the centre of the subcritical system. However, the proposed external neutron source in ADS, based on particle accelerator, is different from isotopic neutron source and the inherent fluctuations in beam current and accelerating voltage makes the source non-Poisson. In this context, Degweker and Rana had postulated a noise theory based on exponentially correlated Gaussian source characteristics for ADS. In noise experiments based on this postulate, a D-T neutron generator has been used in pulsed mode and source characterisation has been carried out to determine the distribution function, source correlation factor and D+ beam pulse shape. It has been observed that the source is Gaussian in nature, the source correlation factor is very large compared to the prompt neutron decay constant and the pulsed D+ beam is rectangular in shape. In noise experiments, the time stamped data have been analysed using various noise methods. The measured prompt neutron decay constant using Poisson source and correlated Gaussian source are in good agreement with theoretical value and amongst them.
Graded Approach Establishment for the HTGR Maintenance Activities Using Modified Fuzzy FMEA & Expert Judgement MethodologyNgarayana, W; Murakami, K
doi: 10.1088/1742-6596/2328/1/012005pmid: N/A
Grading is an important step of the Nuclear Power Plant (NPP) operation & maintenance activities. However, there are several grading difficulties for the High Temperature Gas Reactor (HTGR) as well as the other type of NPPs causing by the lack of operational experiences and availability of the reliability data. Failure Mode & Effects Analysis (FMEA) is one of the mature techniques that are commonly used to solve such kinds of difficulties. Nevertheless, traditional FMEA has several issues and possibly become an obstacle in the grading process. The modified FMEA by utilizing expert judgment elicitation techniques combined with the fuzzy logic theory is proposed to solve those issues. As a study practice, the proposed methodology is applied by examining Japanese’s HTGR, Gas Turbine High Temperature Reactor 300 for Cogeneration (GTHTR300C) design carefully. This study establishing good practice especially for the future advanced NPP maintenance activities development.
Preface: 19th International Conference on Emerging Nuclear Energy Systems (ICENES 2019)doi: 10.1088/1742-6596/2328/1/011001pmid: N/A
This is an exclusively prepared special issue containing selected papers from well-established events, namely, International Conference on Emerging Nuclear Energy Systems (ICENES) and some invited papers to enrich and broaden the novelty of nuclear energy technologies and its applications. The 19th International Conference on Emerging Nuclear Energy Systems (ICENES 2019) is one of the international conference on scientific, engineering, education and other technical aspects of innovative nuclear reactor design, advanced nuclear technology, energy related technology and its applications.The conference was held in Holiday Inn, Bali, Indonesia (6-9 October 2019), organized by the Bandung Institute of Technology (ITB) and in cooperation with the International Atomic Energy Agency (IAEA). The participants come from several 14 countries and from many institutions from universities, governments, companies, society and some other organizations that shared their ideas and research results on emerging nuclear energy technologies and applications, which covered by keynote speakers, invited and contributed oral talks and poster presentations. Some selected presented paper in the conference have been elected as selected papers after reviewing process to be submitted to the Institute of Physics (IoP), Journal of Physics: Conference Series.Nuclear energy recently is recognized as secure, sustain and green energy source as an ultimate energy resource to secure the future of the mankind and its civilization. Hence, considerable research activities and international collaboration are continuing on innovative nuclear energy systems, reactor physics, radiations and its application, nuclear computational system, including fusion energy technology, fusion-fission hybrids systems, GEN-IV reactors technology, small and modular reactor (MSR) technology, space nuclear reactors, and power systems and accelerator-driven systems technologies. Some related topics are also covered related to nuclear power production; nuclear hydrogen production; hydrogen energy, energy efficiency, and management; solar energy; wind energy; hydrogen production and storage; renewable energy; fuel cells; bio-energy, etc.Finally, on behalf of the organizer and advisory board, we would like to express my sincere appreciation and gratitude to all of authors during the conference and publication processes for their valuable contributions and to the members of the committee, reviewers, and advisors for their excellent works in preparing and finalizing this document. We apologize for any inconveniences for this long process of publication due to our conditions and some restrictions as well as some difficulties during COVID19 pandemicList of Organizer, Editorial Board are available in this Pdf.
Simulation of Neutron and Gamma Dose Rate of The TRIGA 2000 Reactor Using Monte Carlo MethodRakotovao, L.O.; Permana, Sidik; Oetami, R.H.; Rasito,
doi: 10.1088/1742-6596/2328/1/012007pmid: N/A
The neutron and gamma radiation doses were calculated from the operation of the 1 MW TRIGA 2000 Reactor in a simulation using the Monte Carlo method with MCNPX and PHITS program. Simulation is done by modelling the geometry of the reactor component materials and running it on a computer. The radiation source in the form of a fission reaction in the reactor core has been simulated using MCNPX to produce a dose of neutron and gamma radiation in the TRIGA 2000 core. Attenuation of neutron and gamma radiation by the reactor building is simulated using the PHITS code so that the neutron and gamma dose rates are obtained on the source, y, and z from reactor core. Interpolation of dose rate curves on large material thicknesses was carried out with the TVL neutron and gamma values of the simulation results for each reactor material. The simulation results shows that the gamma neutron dose rate outside the TRIGA 2000 reactor building is still below the dose limit value for radiation workers.