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STUDI PERFORMA TERAS INISIAL HTR PEBBLE BED DENGAN BAHAN BAKAR PLUTONIUM OKSIDA Zuhair Zuhair; Suwoto Suwoto; Hery Adrial
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 21, No 1 (2017): Februari 2017
Publisher : Badan Tenaga Nuklir Nasional

STUDI PERFORMA TERAS INISIAL HTR PEBBLE BED DENGAN BAHAN BAKAR PLU-TONIUM OKSIDA. Konsep desain reaktor nuklir Generasi IV yang paling menjanjikan adalah reaktor temperatur tinggi (high temperature reactor, HTR) jenis pebble bed karena karakteristik keselamatan melekat dan temperatur pendinginnya yang tinggi. Desain reaktor pebble bed memiliki puluhan bahkan ratusan ribu bahan bakar pebble dengan moderator grafit dan pendingin helium. Dalam studi ini, teras reaktor memiliki volume 8⅓-33⅓ m3 dan densitas daya 3 MW/m3 yang menghasilkan daya termal 25-100 MW. Beberapa tahap yang dikerjakan dimulai dari pemodelan bahan bakar dan teras reaktor, optimasi teras inisial HTR pebble bed sebagai fungsi rasio tinggi per diameter (H/D) hingga optimasi teras inisial HTR pebble bed sebagai fungsi pemuatan logam berat (heavy metal). Seluruh perhitungan dikerjakan dengan memanfaatkan program transport Monte Carlo MCNPX dan pustaka data nuklir energi kontinu ENDF/B-VII. Hasil analisis menyimpulkan bahwa, rasio H/D yang rendah dan massa Pu/pebble yang tinggi merupakan opsi yang paling ideal dalam teras HTR pebble bed dari sudut pandang ekonomi neutron.Kata kunci: TRISO, kernel, plutonium oksida, teras inisial, HTR pebble bed

EFEK IMPURITAS BORON PADA KERNEL BAHAN BAKAR HTGR PEBBLE BED Hery Adrial
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 21, No 1 (2017): Februari 2017
Publisher : Badan Tenaga Nuklir Nasional

EFEK IMPURITAS BORON PADA KERNEL BAHAN BAKAR HTGR PEBBLE BED. Boron merupakan material yang sangat efektif berfungsi menyerap neutron. Adanya impuritas boron pada teras bahan bakar, khususnya bahan bakar disinyalir akan dapat mempengaruhi performa sifat-sifat neutronik maupun keselamatan kritikalitas reaktor. Oleh karena itu penelitian tentang efek impuritas boron pada bahan bakar pebble penting dilakukan, agar dapat diketahui pengaruh impuritas terhadap kritikalitas teras HTGR. Tahap awal penelitian, dilakukan validasi program MCNP6.1 yang dibuat dengan cara membandingkan hasil tinggi teras kritis awal reaktor HTGR dengan tinggi kritis awal HTR-10 secara eksperimen maupun berdasarkan program VSOP kemudian dilakukan perhitungan kritikalitas terhadap impuritas boron pada kernel pebble. Hasil yang dicapai menunjukan program yang dibuat mempunyai tingkat keakuratan yang sangat baik yaitu 1,58 % terhadap hasil ekperimen. Hasil perhitungan kritikalitas terhadap impuritas boron menunjukkan tidak adanya pengaruh akibat impuritas boron pada teras awal.Kata kunci: impuritas boron, MCNP6.1, kritikalitas, HTGR, pebble bed

GAMMA DOSE RATE ANALYSIS IN BIOLOGICAL SHIELDING OF HTGR-10 MWth PEBBLE-BED REACTOR Hery Adrial; Amir Hamzah; Entin Hartini
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 1 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2019.25.1.4923

HTGR-10 MWth is a high-temperature gas-cooled reactor. The fuel and moderator are pebble shaped with a radius of 3 cm. One fuel pebble consists of thousands of UO2 kernels with a density of 10.4 gram/cc and the enrichment rate of 17%. The core of HTGR-10 MWth is the center of origin of neutrons and gamma radiation resulting from the interaction of neutrons with pebble fuel, moderator and biological shield. The various types of radiations generated from such nuclear reactions should be monitored to ensure the safety of radiation workers. This research was conducted using MCNP-6 Program package with the aim to calculate and analyze gamma radiation dose in biological shield of HTGR-10 MWth. In this study, the biological shield is divided into 10 equal segments. The first step of the research is to benchmark the created program against the critical height of HTR-10. The results of the benchmarking show an error rate of ± 1.1327%, while the critical core height of HTGR 10 MWth for the ratio of pebble fuel and pebble moderator (F:M) of 52:48 occurs at a height of 134 cm. The rate of gamma dose at the core is 3.0052E+05 mSv/hr. On the biological shield made of regular concrete with a density of 2.3 grams/cc, the rate of gamma dose decreases according to an equation y = 0.0042 e-0.03x. Referring to Perka Bapeten no 4 of 2013, the safe limits for workers and radiation protection officers will be achieved if the minimum thickness of biological shield is 115 cm with gamma dose rate of 0 mSv/hour.

EVALUATION OF RADIATION DOSE RATE OF RSG-GAS REACTOR Amir Hamzah; Hery Adrial; Subiharto
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2018.24.3.4909

The RSG-GAS reactor has been operated for 30 years. Since the nuclear reactor has been operated for a long time, aging process on its components may occur. One important parameter for maintaining the safety level of the RSG-GAS reactor is to maintain radiation exposure as low as possible, especially in the working area. The evaluation results should be able to demonstrate that the radiation exposure of the RSG-GAS is still safe for workers, communities and the surrounding environments. The purpose of this study is to evaluate radiation exposure in the working area to ensure that the operation of RSG-GAS is still safe for the next 10 years. The scope of this work is confirming the calculation results with the measured radiation dose in the RSG-GAS reactor working area. Measurement of radiation exposure is done by using the installed equipments at some points in the RSG-GAS working area and a portable radiation exposure measurement equipment. The calculations include performance of a modeling and analysis of dose rate distribution based on the composition and geometry data of RSG-GAS by using MCNP. The analysis results show that the maximum dose rate at Level 0 m working area of RSG-GAS reactor is 3.0 mSv/h with a deviation of 6%, which is relatively close to the measurement value. The evaluation results show that the dose rate in RSG-GAS working area is below the limit value established by the Nuclear Energy Regulatory Agency of Indonesia (BAPETEN) of 10 mSv/h (for the average effective dose of 20 mSv/year). Therefore, it is concluded that the dose rate in RSG-GAS working area is safe for personnel.

ANALISIS KUAT SUMBER NEUTRON DAN PERHITUNGAN LAJU DOSIS NEUTRON TERAS AWAL RDE Suwoto; Hery Adrial; Zuhair
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2017.23.1.3119

The reactor core RDE (Experimental Power Reactor) has non-annular cylindrical shape with coated fuel particles TRISO kernel in the spherical form (pebble) and cooled by helium gas. The RDE reactor core design is very safe, because it adopts the high-temperature reactor technology (HTGR) with inherent passive safety. The designed input helium temperature of RDE core is about 250 °C and the output of helium gas temperature helium of the reactor core of 700 °C. Beside of generating electricity, the RDE reactor is designed to generate high temperature gas that can be used for any purposes such as cogeneration (heat process application research). The spherical fuel called as pebble containing thousand of kernel TRISO-coated fuel particles of uranium dioxide 17 % (235U) enriched. TRISO coating comprises four layers, namely: porous carbon buffer layer, inner part pyrolytic carbon layer (IPyC, Inner Pyrolitic Carbon), silicon carbide layer (SiC) and a layer of pyrolytic carbon outer part (OPyC, Outer Pyrolitic Carbon). Analysis of neutron source strength and preliminary calculation of neutron dose rate on the RDE core were performed using Monte Carlo MCNP5v1.2 code. Double heterogeneity treatment in TRISO-coated fuel particles kernel and the pebble of the core are applied. Utilization of EGS99304 code to determine 640 amount of energy group structures (SAND-II neutron group structures) are used in the neutron spectrum calculation in the reactor. The reactor core is divided into 100 zones (10 in radial and 10 in axial directions). An analyses calculation result shows that the neutron source strength of the core is about 8.47027X1017neutron/sec. The neutron dose rate distributions are determined using a conversion factor of flux-to-dose taken from International Commission on Radiological Protection (ICRP) and National Council on Radiation Protection and Measurements (NCRP). The preliminary calculations result of neutrons dose rate using ICRP-21 and NCRP-38 conversion factor for radiation workers in the radial direction in the shield biological are already weakened as low as 6.69915 and 6.9964 μSv/h, respectively, at a position of 215 cm from the center part of the core. This result analysis indicate that radiation workers are safe and protected from neutron radiation sources in accordance with the regulatory requirements. Overall, from the calculation and analysis, it appears that the radiation and biological shielding model meets to the standards of radiation safety required. The results of this study can be used to complete the design data thickness of shielding of the reactor.

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