One alternative effort to eliminate the accumulation of transuranic (TRU) elements generated from LWR spent fuel is by utilizing TRU as nuclear fuel in pebble bed reactor. This work was aimed to ...investigate the neutronic behavior of a pebble bed reactor used for TRU transmutation. The reactor geometry is adopted from the HTR-Modul. MCNP6 multipurpose radiation transport code with ENDF/B-VII.1 neutron library was used to calculate the neutronic aspects. The results indicated that the initial effective multiplication factor (keff) values decrease as the amount of TRU fuel pebbles within the core increases, concurrent with the fuel burnup. Increasing the TRU fuel pebble ratio resulted in weakened Doppler temperature coefficients (DTCs) and improvement in moderator temperature coefficients (MTCs). The calculated total mass of plutonium and minor actinide transmutation demonstrated that a pebble bed reactor with 100% TRU pebble can reduce TRU mass up to 55.47% of its initial mass load.
Center for Nuclear Reactor Safety and Technology (PTRKN) has tasks to perform research and development on the conceptual design of gas cooled reactor (RDE) with small power level of 10 MWt. In line ...with the development of RDE conceptual design, an analysis on the effects of thermal flow on TRISO fuel temperature distribution in RDE core has been performed. The objective of this research is to analyze the effects of variation of power level, coolant mass flow rate, and pressure on the temperature distribution of TRISO fuel in RDE core that meet safety requirements. This research employs VSOP code, by which the reactor core is divided into 50 components to represent the positions of various material compositions and to model TRISO fuel into 5 layers. The analysis is conducted at the following conditions: power level of 3 MW - 10 MW, flow rate of 1.5 kg/s - 4.32 kg/s, and system pressure of 1 MPa - 3 MPa. For the modelling, reactor parameters, core specification, and TRISO fuel specification as well as TRISO layer data of RDE are used. Based on the analysis results on the effect of helium gas flow, pressure, and reactor power variable on TRISO fuel temperature distribution, it can be concluded that, among these three variables, reactor power gives the most significant effect. The maximum temperatures at kernel and fuel outer layer is 740°C and 736.4°C, respectively. Compared to the temperature distribution resulted between VSOP'94 code and fuel temperature limitation as high as 1600 °C, there is enough safety margin from melting or disintegrating. Therefore, it can be concluded that TRISO fuel is able to contain all radioactive fission products.
Thorium-based nuclear fuel has become an interesting subject for a variety of research with a wide range of applications. Research focusing on thorium-based fuel is aimed to overcome the scarcity and ...limitation of natural uranium resources as an alternative nuclear fuel in a thermal reactor. As thorium has no naturally occurring fissile isotope, it requires other fissile isotopes in order to be converted into fissile 233U to produce energy. The isotopes 235U and 2359Pu are two of the few alternatives available as the fissile nuclei for a thorium-fueled reactor. The purpose of this paper is to investigate the criticality and burnup performance of pebble bed reactor using two options of thorium-based fuel–namely, UO2-ThO2 and PuO2-ThO2. The HTR-10 was chosen as the reactor model. A series of criticality calculations with various uranium contents in UO2-ThO2 fuel and various plutonium contents in PuO2-ThO2 fuel was conducted using the Monte Carlo transport code MCNP6 and continuous energy nuclear data library ENDF/B-VII. The calculation result shows that 35% plutonium content in PuO2-ThO2 fuel has comparable criticality with 80% uranium content in UO2-ThO2 fuel. The former is shown to be better to ensure longer reactor operational time. However, that is due to the higher fissile fraction compared to that of the latter fuel. Meanwhile, thorium played little part in prolonging reactor criticality, as 233U production is not particularly significant.
The reaktor daya eksperimental (RDE) is 10 MWth experimental power reactor HTR Pebble-bed type. Comparing with water pool research reactors, the distance between the core and the reactor vessel is ...very close while the medium mainly is graphite. Thus, the main role of the radiation shield relies on biological shields to obstruct radiation exposure coming from the reactor core. Meanwhile, the RDE reactor should be met with the operation standard safety limits of a nuclear reactor. Therefore, the purpose of this study is to estimate the radiation dose rates in the working area and outside the RDE reactor building. By analyzing the distribution of radiation dose rates can be estimated radiation safety levels for workers and society and the surrounding environment. The radiation dose rates calculation was done using MCNP. By determining the tally position, it can be determined the radiation dose rates inside and outside the RDE reactor building. The results show that the maximum dose rates on the outer surface of the biological shielding (working area) under normal operating conditions are 8 μSv/h. That dose rates are below the limit value determined by BAPETEN showing the radiation shielding design make the RDE safe for workers and the community and the surrounding environment from radiation hazards.
•Investigating neutronic behavior of VVER-1000 fuel assembly with Duplex fuel rod using ThO2.•The VVER-1000 with a duplex fuel rod slightly improves its safety characteristics through DTC and MTC.•As ...predicted, Plutonium and Minor Actinide slightly reduced by increasing ThO2 Duplex fuel rod.•Increasing fissile material content may be important for optimizing VVER-1000 fuel management, but further calculation might be needed to optimize the whole core model.
Various studies suggested that duplex fuel rods have better neutronic characteristics than ordinary single-oxide fuel pellets. A duplex fuel rod consists of an inner uranium dioxide (UO2) layer and an outer thorium dioxide (ThO2) layer. This paper aims to study the neutronic behavior of the VVER-1000 fuel assembly with duplex fuel rod design. The duplex fuel rods were loaded into the fuel assembly in three configurations: outermost region, mid-region, and central region of the assembly. Three compositions with volume fractions of 75% UO2-25% ThO2, 70% UO2-30% ThO2, and 65% UO2-35% ThO2 were considered. A series of numerical calculations were conducted using the MCNP6 code and ENDF/B-VII.1 library. In general, there is an improvement in DTC and MTC value for fuel assembly using Duplex fuel rods. However, βeff and ℓ values tend to decrease while the value of Λ tends to increase which may affect fuel assembly response during transient scenarios. The results show that Configuration I can maintain its criticality for up to 15 MWd/kg for all fractions of UO2-ThO2 being used in duplex fuel, while Configuration II and Configuration III only reach below 15 MWd/kg when using duplex fuel rods. The Doppler temperature coefficient (DTC) and the moderator temperature coefficient (MTC) become stronger with the increasing ThO2 volume fraction within the VVER fuel assembly. Compared with non-duplex fuel, the amount of 239Pu was lower when using duplex fuel rods, but minor actinides was slightly higher at 40 MWd/kg. It can be concluded that the VVER-1000 with a duplex fuel rod could improve its safety characteristics besides that increasing fissile material content might be crucial to optimize VVER-1000 fuel management.
STUDY AND ASSESSMENT OF GENERATION IV REACTOR NUCLEAR DATA WITH FAST NEUTRON SPECTRA. Generation IV International Forum (GIF) has evaluated and assessed NES of Gen- IV and selected six potential ...types of reactors to be deployed in the next decade. Those include GFR, LFR, SFR, MSR, SCWR and VHTR. The first three reactors were fast neutron spectrum applied and the rest reactors were thermal neutron spectrum used. The study and assessment focused on the nuclear data characteristic parameter and nuclear data uncertainties of Gen-IV reactor with fast neutron spectrum. Until 2008, the accuracy target of nuclear data cross-sections used it in fast reactor spectrum calculation are relatively significant especially for σ-capture, σ-fission, and σ-inelastic. Several differences of nuclear data cross-sections on minor actinide isotopes between expected and targeted parameters are observed such as σ-fission of Cm-244 isotope up to 10 times larger and σ-capture of 92-U-238 isotope around 1.5-2 times higher than targeted parameters. Uncertainty and accuracy of minor actinide cross-sections for fast spectrum Gen-IV reactors provide relatively significant discrepancies (1.3 to 10 times higher) in term of accuracy between expected and targeted parameters. Some differences of provided results from any experimental and assessment data with several evaluated nuclear data files for Pb are found. Some discrepancies on integral parameter of fast spectrum Gen-IV reactors between expected and targeted such keff, void reactivity and Doppler effects, peak power and burn-up are clearly observed. Accurate and precise cross-sections data of radiation captured and threshold reaction cross sections such as (n,2n), (n,3n), (n,p), (n,α) are necessary for fast reactors. Keywords: cross-sections, fast neutron spectrum, GFR, LFR, SFR, uncertainty, target accuracy
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