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EFFECTS OF FUEL DENSITY ON REACTIVITY COEFFICIENTS AND KINETIC PARAMETERS OF PEBBLE BED REACTOR Suwoto; R. Andika Putra Dwijayanto; Wahid Luthfi; Zuhair
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 1 (2022): 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.2022.28.1.6625

Few decades ago, a large number of nuclear reactors were designed to use HEU as the fuel. But the use of HEU is being discouraged since it can be used as a nuclear explosive material which makes its proliferation potential. Most of the HEU-fueled nuclear reactors in the world are either closed or converted into other types that have LEU fuel with 235U enrichment below 20%. To extend lifetime, LEU fuel with high density is developed. The change in fuel density from low to high will also change core neutronics and thermal hydraulics of the reactor, and as a result, it affects the transient response of the reactor. This paper studies the effects of fuel density on reactivity coefficients and kinetic parameters of pebble bed reactor through several calculations with MCNP6 code combine with ENDF/B-VII.1 continuous energy cross-section nuclear data library. The overall calculation results show that the Doppler temperature coefficient (DTC) increases with increasing fuel density, but the moderator temperature coefficient (MTC) decreases due to hardening of neutron spectrum. Kinetic parameters such as effective delayed neutron fraction (βeff), prompt neutron lifetime(ℓ) and neutron generation time (Ʌ) which significantly reduced with increasing fuel density will strongly affect the reactor control and safety. The results of this work conclude that the selection of 9–15 g/cm3 fuel density should be considered carefully given that its effect on reactor controllability.

CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL R. Andika Putra Dwijayanto; Ihda Husnayani; Zuhair
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 2 (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.2019.25.2.5525

There are several options of nuclear fuel utilization in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental/RDE). Although mainly RDE utilizes low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilize other fuel, for example thorium-based and possibly even plutonium-based fuel. Different fuel yields different spent fuel characteristics, so it is necessary to identify the characteristics to understand and evaluate their handling and interim storage. This paper provides the study on the characteristics of thorium-fueled RDE spent fuel, assuming typical operational cycle. ORIGEN2.1 code is employed to determine the spent fuel characteristics. The result showed that at the end of the calculation cycle, each thorium-based spent fuel pebble generates around 0.627 Watts of heat, 28 neutrons/s, 8.28x1012 photons/s and yield 192.53 curies of radioactivity. These higher radioactivity and photon emission possibly necessitate different measures in spent fuel management, if RDE were to use thorium-based fuel. Tl-208 activity, which found to be emitting potentially non-negligible strong gamma emission, magnified the requirement of proper spent fuel handling especially radiation shielding in spent fuel cask.

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY Zuhair; Wahid Luthfi; R. Andika Putra Dwijayanto; Suwoto
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 3 (2021): 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.2021.27.3.6509

NuScale, a typical Pressurized Water Reactor (PWR) Small Modular Reactors (SMRs), offers a new opportunity for the future of nuclear industry. With 160 MW thermal power, NuScale has several advantages such as flexibility due to its modularity in construction. This work is focused on the study of criticality and neutronic safety parameters of NuScale fuel assembly using MCNP6 code and ENDF/B-VII library. The calculation results shows that criticality of fuel assembly type D is the highest among other assembly types because it has a fuel pin with pure UO2 without Gd2O3 concentration. The Doppler temperature coefficient (DTC) of fuel assembly type C is the most negative among other assemblies due to Doppler broadening effect on resonance region of capture cross section of 238U which is the highest concentration. The moderator temperature coefficient (MTC) of fuel assembly type D is the most negative among the other assembly types. The effective delayed neutron fraction (βeff) does not reflect a consistent trend among fuel assembly types. Fuel assembly type D shows the highest prompt neutron lifetime (ℓ) while the highest neutron generation time (Ʌ) is shown in assembly type C. It can be concluded that this study provides adequate results that can be used as a first step to carry out the neutronic computation and analysis of the NuScale full core.

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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