scholarly journals The Effects of U-233 Impurity on U-232 and Tl-208 Buildup in Experimental Power Reactor with Thorium-Based Fuel

2021 ◽  
Vol 2072 (1) ◽  
pp. 012001
Author(s):  
R A P Dwijayanto ◽  
Suwoto ◽  
Zuhair ◽  
Z Su’ud
Keyword(s):  
Fuel Cycle ◽  
Power Reactor ◽  
High Energy ◽  
Spent Fuel ◽  
Thorium Cycle ◽  
Gamma Emitter

Abstract The existence of Tl-208 in thorium fuel cycle is a double-edged sword. Tl-208 is a high-energy 2.6 MeV gamma emitter, which acts as an effective proliferation barrier while simultaneously complicating the handling of the spent fuel. To ensure the safety of the latter, the buildup of both Tl-208 and its parent, U-232, are necessary to be understood. This paper attempts to analyse the buildup of U-232 and Tl-208 in the Reaktor Daya Eksperimental (Experimental Power Reactor/RDE) fuel based on thorium cycle, using various U-233 isotopic vectors. The simulation result shows that U-232-contaminated fresh fuels ended up with higher Tl-208 and U-232 activities at the end of cycle (EOC) compared with uncontaminated fresh fuel. However, their U-232 build-up rate are lower and even negative at one case. Then, lower U-233 purity caused a higher U-232 and Tl-208 activities at EOC. This result implies a considerable difference of isotope buildup between the various U-233 vectors. Consequently, the thorium cycle-based RDE spent fuel handling should consider the isotopic vector of U-233 used in fresh fuel.

scholarly journals The mode of operation of CANDU power reactor in thorium self-sufficient fuel cycle

2008 ◽  
Vol 23 (2) ◽  
pp. 16-21
Author(s):  
Boris Bergelson ◽  
Alexander Gerasimov ◽  
Georgy Tikhomirov
Keyword(s):  
Fuel Cycle ◽  
Power Reactor ◽  
The Self ◽  
Fissile Material ◽  
Reactor Operation ◽  
Active Core ◽  
Mode Of Operation ◽  
Neutron Power ◽  
Thorium Cycle ◽  
Maximum Content

This paper presents the results of calculations for CANDU reactor operation in the thorium fuel cycle. The calculations were performed to estimate feasibility of operation of a heavy-water thermal neutron power reactor in the self-sufficient thorium cycle. The parameters of the active core and the scheme of fuel reloading were considered to be the same as for the standard operation in the uranium cycle. Two modes of operation are discussed in the paper: the mode of preliminary accumulation of 233U and the mode of operation in the self-sufficient cycle. For calculations for the mode of accumulation of 233U, it was assumed that plutonium was used as the additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. The maximum content of 233U in the target channels was about 13 kg/t of ThO2. This was achieved by six year irradiation. The start of reactor operation in the self-sufficient mode requires content of 233U not less than 12 kg/t. For the mode of operation in the self-sufficient cycle, it was assumed that all the channels were loaded with the identical fuel assemblies containing ThO2 and a certain amount of 233U. It was shown that the non-uniform distribution of 233U in a fuel assembly is preferable.

scholarly journals CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL.

2019 ◽  
Vol 25 (2) ◽  
Author(s):  
R. Andika Putra Dwijayanto, S.T. ◽  
Ihda Husnayani ◽  
Zuhair Zuhair
Keyword(s):  
Power Reactor ◽  
Photon Emission ◽  
Radiation Shielding ◽  
Spent Fuel ◽  
Gamma Emission ◽  
Fuel Characteristics ◽  
Enriched Uranium ◽  
The Moment ◽  
Interim Storage ◽  
Thorium Cycle

CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL. There are several options of nuclear fuel utilisation in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental/RDE). Although mainly RDE utilises low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilise 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-fuelled 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.Keywords: RDE, spent fuel, thorium, HTGR, Tl-208.

Optimization of the Mode of the Uranium-233 Accumulation for Application in Thorium Self-Sufficient Fuel Cycle of CANDU Power Reactor

2006 ◽  
Author(s):  
Boris Bergelson ◽  
Alexander Gerasimov ◽  
Georgy Tikhomirov
Keyword(s):  
Fuel Cycle ◽  
Power Reactor ◽  
Elementary Cell ◽  
Fissile Material ◽  
Optimum Ratio ◽  
Accumulation Time ◽  
Production Parameters ◽  
Active Core ◽  
The Cost ◽  
Thorium Cycle

Results of calculation studies of the first stage of self-sufficient thorium cycle for CANDU reactor are presented in the paper. The first stage is preliminary accumulation of 233U in the CANDU reactor itself. Parameters of active core and scheme of fuel reloading were accepted the same as those for CANDU reactor. It was assumed for calculations, that enriched 235U or plutonium was used as additional fissile material to provide neutrons for 233U production. Parameters of 10 different variants of the elementary cell of active core were calculated for the lattice pitch, geometry of fuel channels, and fuel assembly of the CANDU reactor. The results presented in the paper allow to determine the time of accumulation of the required amount of 233U and corresponding number of targets going into processing for 233U extraction. Optimum ratio of the accumulation time to number of processed targets can be determined using the cost of electric power produced by the reactor and cost of targets along with their processing.

scholarly journals On the Dimensions Required for a Molten Salt Zero Power Reactor Operating on Chloride Salts

2021 ◽  
Vol 11 (15) ◽  
pp. 6673
Author(s):  
Bruno Merk ◽  
Anna Detkina ◽  
Seddon Atkinson ◽  
Dzianis Litskevich ◽  
Gregory Cartland-Glover
Keyword(s):  
Heavy Metal ◽  
Molten Salt ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Power Reactor ◽  
Modelling And Simulation ◽  
Advanced Technology ◽  
Electricity Production ◽  
Eutectic System ◽  
Low Carbon

Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive expansion in low-carbon electricity production, which will be required for a future net-zero society. The importance of a zero-power reactor for the process of developing a new, innovative rector concept, such as that required for the molten salt fast reactor based on iMAGINE technology, which operates directly on spent nuclear fuel, is described here. It is based on historical developments as well as the current demand for experimental results and key factors that are relevant to the success of the next step in the development process of all innovative reactor types. In the systematic modelling and simulation of a zero-power molten salt reactor, the radius and the feedback effects are studied for a eutectic based system, while a heavy metal rich chloride-based system are studied depending on the uranium enrichment accompanied with the effects on neutron flux spectrum and spatial distribution. These results are used to support the relevant decision for the narrowing down of the configurations supported by considerations on cost and proliferation for the follow up 3-D analysis. The results provide for the first time a systematic modelling and simulation approach for a new reactor physics experiment for an advanced technology. The expected core volumes for these configurations have been studied using multi-group and continuous energy Monte-Carlo simulations identifying the 35% enriched systems as the most attractive. This finally leads to the choice of heavy metal rich compositions with 35% enrichment as the reference system for future studies of the next steps in the zero power reactor investigation. An alternative could be the eutectic system in the case the increased core diameter is manageable. The inter-comparison of the different applied codes and approaches available in the SCALE package has delivered a very good agreement between the results, creating trust into the developed and used models and methods.

scholarly journals The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 991-1003 ◽  
Author(s):  
Evaristo J. Bonano ◽  
Elena A. Kalinina ◽  
Peter N. Swift
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
United States Of America ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Spent Fuel ◽  
Dry Storage ◽  
The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

TRU Recycling Options for Environmentally Friendly and Proliferation-Resistant Nuclear Fuel Cycle of FBR

2010 ◽  
Author(s):  
Sidik Permana ◽  
Mitsutoshi Suzuki
Keyword(s):  
High Performance ◽  
Fuel Cycle ◽  
Operation Time ◽  
Nuclear Nonproliferation ◽  
High Level Waste ◽  
Minor Actinide ◽  
Spent Fuel ◽  
Closed Fuel Cycle ◽  
The Core ◽  
High Level

The embodied challenges for introducing closed fuel cycle are utilizing advanced fuel reprocessing and fabrication facilities as well as nuclear nonproliferation aspect. Optimization target of advanced reactor design should be maintained properly to obtain high performance of safety, fuel breeding and reducing some long-lived and high level radioactivity of spent fuel by closed fuel cycle options. In this paper, the contribution of loading trans-uranium to the core performance, fuel production, and reduction of minor actinide in high level waste (HLW) have been investigated during reactor operation of large fast breeder reactor (FBR). Excess reactivity can be reduced by loading some minor actinide in the core which affect to the increase of fuel breeding capability, however, some small reduction values of breeding capability are obtained when minor actinides are loaded in the blanket regions. As a total composition, MA compositions are reduced by increasing operation time. Relatively smaller reduction value was obtained at end of operation by blanket regions (9%) than core regions (15%). In addition, adopting closed cycle of MA obtains better intrinsic aspect of nuclear nonproliferation based on the increase of even mass plutonium in the isotopic plutonium composition.

scholarly journals Cooling Period Calculation of Evolutionary Power Reactor Spent Fuel for Dry Management Safety

2016 ◽  
pp. 17-21
Author(s):  
M. I. Youssef ◽  
G. F. Sultan ◽  
F. Morsi Hassan
Keyword(s):  
Heat Load ◽  
Power Reactor ◽  
Computer Code ◽  
Normal Operation ◽  
Spent Fuel ◽  
Cooling Period ◽  
Decay Heat ◽  
Safety Margins ◽  
Epr Parameters

The calculation of the evolutionary power reactor (EPR) spent fuel (SF) cooling period (CP) was performed. The CP was determined by comparing the heat load of a cask with the calculated value of EPR decay heat (DH). The EPR DH was calculated by the ORIGEN computer code based on the EPR parameters. For conservatively study, the EPR and ORIGEN parameters that lead to higher DH values were selected and safety margins were considered. The fitting tool was utilized in the calculation of CP to overcome the ORIGEN limitation. The resultant values of CP will maintain the peak cladding temperature (PCT) of SF lower than 400°C during storage, transport, and disposal. The results show that -for normal operation- the SF of EPR should stay in the pool at least 4.75 years before it is loaded to the passively cooled dry casks.

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