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 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 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.

Thorium Cycle in High Temperature Gas-Cooled Gas Turbine Reactors (HTG-GT) Using Highly Enriched Uranium Obtained From the Dismantling of Nuclear Weapons

1994 ◽  
Author(s):  
Nicola Cerullo ◽  
Giovanni Guglielmini ◽  
A. Di Pietro
Keyword(s):  
High Temperature ◽  
Gas Turbine ◽  
Nuclear Weapons ◽  
Fuel Cycle ◽  
Fissile Material ◽  
High Investment ◽  
Enriched Uranium ◽  
The Cold War ◽  
Thorium Cycle ◽  
High Temperature Gas

The closed thorium fuel cycle is based on the use of fissile U-233 produced by the thorium fertilization in the original fuel element without any refabrication action, which is very difficult, due to the high activity of Thorium activated products. The need of a consistent amount of fissile material for beginning the U-Th cycle activity, in order to sustain the Thorium conversion reactions, requires an high initial U-235 enrichment. This condition, due to high investment costs, stopped, in the last years, any initiative in this field. The end of the cold war and the disarmament agreements pose the problem of the use of military grade fissile materials resulting from the dismantling of nuclear weapons both Russian and American. In this paper the problem is analyzed and a High Temperature Gas-cooled Gas Turbine (HTG-GT) reactor, using a nuclear U-Th fuel cycle utilizing military grade highly enriched uranium, is proposed.

scholarly journals Optimization of the self-sufficient thorium fuel cycle for CANDU power reactors

2008 ◽  
Vol 23 (1) ◽  
pp. 3-10
Author(s):  
Boris Bergelson ◽  
Alexander Gerasimov ◽  
Georgy Tikhomirov
Keyword(s):  
New Technologies ◽  
Fuel Cycle ◽  
Fission Products ◽  
The Self ◽  
Fissile Material ◽  
Heavy Nuclei ◽  
Candu Reactors ◽  
Mode Of Operation ◽  
Enriched Uranium ◽  
Thorium Cycle

The results of optimization calculations for CANDU reactors operating in the thorium cycle are presented in this paper. Calculations were performed to validate the feasibility of operating a heavy-water thermal neutron power reactor in a self-sufficient thorium cycle. Two modes of operation were considered in the paper: the mode of preliminary accumulation of 233U in the reactor itself and the mode of operation in a self-sufficient cycle. For the mode of accumulation of 233U, it was assumed that enriched uranium or plutonium was used as additional fissile material to provide neutrons for 233U production. In the self-sufficient mode of operation, the mass and isotopic composition of heavy nuclei unloaded from the reactor should provide (after the removal of fission products) the value of the multiplication factor of the cell in the following cycle K>1. Additionally, the task was to determine the geometry and composition of the cell for an acceptable burn up of 233U. The results obtained demonstrate that the realization of a self-sufficient thorium mode for a CANDU reactor is possible without using new technologies. The main features of the reactor ensuring a self-sufficient mode of operation are a good neutron balance and moving of fuel through the active core.

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.

The Effect of Dolomite Addictive Ratio on Torrified Cassava Rhizome in the Biomass Combustion Process

2021 ◽  
Vol 407 ◽  
pp. 113-120
Author(s):  
Nat Thuchayapong ◽  
Nattawut Tharawadee
Keyword(s):  
Combustion Process ◽  
Fixed Bed ◽  
Air Pressure ◽  
Biomass Combustion ◽  
Optimum Ratio ◽  
Bomb Calorimeter ◽  
Useful Heat ◽  
Main Component ◽  
The Cost ◽  
Research Studies

This research studies on the effect of additive (Dolomite) on Biomass powder (Cassava rhizome) which passes Torrefied process and fixed bed at 250 degrees Celsius for one hour and a half. The gasifier with up-draft type was used in this experiment. Air pressure was fixed at 0.1 Bar. The useful heat (Quseful) and Low heating valves (LHV) was investigated by using an Automatic Bomb Calorimeter. Moreover, the dolomite was varied 0, 10 and 15% by weight mixed with Cassava rhizome achieved with Torrefied process. When Low heating valves (LHV) slightly decreases from 21.96±0.22 MJ/kg to 18.15±0.50 MJ/kg, Quseful heat from the burning from gasifier sharply increase when it is mixed with dolomite from 753.34±39.18 to 1,003.97±33.49KJ respectively. The loading of dolomite has significance affecting the useful heat. The present study reveals that low heating valves (LHV) decreases and Quseful heat increase result from dolomite which gives a clean gas product and the Tar molecule can be easily broken. The CO2 gas from the combustion process was absorbed by CaO, which is the main component in dolomite. The cost of mixing 8.9% of Dolomite with Cassava rhizome is the optimum ratio for the biomass combustion process.

User Requirements and Criteria for Proliferation Resistance in INPRO

2004 ◽  
Author(s):  
Kun-Mo Choi ◽  
Robert D. Hurt ◽  
Thomas E. Shea ◽  
Richard Nishimura
Keyword(s):  
Nuclear Weapons ◽  
Nuclear Weapon ◽  
Nuclear Energy ◽  
Energy Systems ◽  
Nuclear Material ◽  
User Requirements ◽  
Fissile Material ◽  
Nuclear Weapon Program ◽  
The Cost ◽  
Weapon Program

In designing future nuclear energy systems, it is important to consider the potential that such systems could be misused for the purpose of producing nuclear weapons. INPRO set out to provide guidance on incorporating proliferation resistance into innovative nuclear energy systems (INS). Generally two types of proliferation resistance measures are distinguished: intrinsic and extrinsic. Intrinsic features consist of technical design features that reduce the attractiveness of nuclear material for nuclear weapon program, or prevent the diversion of nuclear material or production of undeclared nuclear material for nuclear weapons. Extrinsic measures include commitments, obligations and policies of states such as the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) and IAEA safeguards agreements. INPRO has produced five basic principles and five user requirements for INS. It emphasizes that INS must continue to be an unattractive means to acquire fissile material for a nuclear weapon program. It also addresses as user requirements: 1) a balanced and optimised combination of intrinsic features and extrinsic measures, 2) the development and implementation of intrinsic features, 3) an early consideration of proliferation resistance in the development of INS and 4) the utilization of intrinsic features to increase the efficiency of extrinsic measures. INPRO has also developed criteria, consisting of indicators and acceptance limits, which would be used by a state to assess how an INS satisfies those user requirements. For the first user requirement, the most important but complex one, INPRO provides a 3-layer hierarchy of indicators to assess how unattractive a specific INS would be as part of a nuclear weapon program. Attributes of nuclear material and facilities are used as indicators to assess intrinsic features. Extrinsic measures imposed on the system are also assessed. Indicators to assess defence in depth for proliferation resistance include the number and robustness of barriers, and the redundancy or complementarity of barriers. The cost of incorporating proliferation resistant features is used to assess the cost-effectiveness of any particular INS in providing proliferation resistance. The stages in the development of an INS at which proliferation resistance is considered in the process are assessed. Awareness of extrinsic measures by designers and use of intrinsic features for verification illustrate how intrinsic features facilitate extrinsic measures. An INPRO-consistent methodology to assess the proliferation resistance of an INS is still under development, with feedback expected from the case studies undertaken by Argentina, India, Russia and the Republic of Korea.

scholarly journals Evaluation of the Breeding Performance of a NaCl-UCl-Based Reactor System

Energies ◽  
2019 ◽  
Vol 12 (20) ◽  
pp. 3853 ◽  
Author(s):  
Bruno Merk ◽  
Anna Detkina ◽  
Seddon Atkinson ◽  
Dzianis Litskevich ◽  
Gregory Cartland-Glover
Keyword(s):  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Salt System ◽  
Fissile Material ◽  
Spent Fuel ◽  
Breeding Performance ◽  
Closed Fuel Cycle ◽  
Term Operation ◽  
Start Up ◽  
Innovative Solution

The energy trilemma forms the key driver for the future of energy research. In nuclear technologies, molten salt reactors are an upcoming option which offers new approaches. However, the key would be closed fuel cycle operation which requires sufficient breeding for a self-sustained long term operation ideally based on spent fuel. To achieve these attractive goals two challenges have been identified: achieving of sufficient breeding and development of a demand driven salt clean up system. The aim is to follow up on previous work to create an initial approach to achieving sufficient breeding. Firstly, identification of a salt system with a high solubility for fertile material and sufficiently low melting point. Secondly, evaluation of the sensitivity of the breeding performance on the sort of fissile material, the fissile material loading, and the core dimension all based on a realistic salt system which provides the solubility for sufficient fertile material to achieve the required breeding in a homogeneous reactor without breeding blanket. Both points are essential to create an innovative solution to harvest the fruits of a closed fuel cycle without the penalty of the prohibitively huge investments. It is demonstrated that the identified and investigated NaCl-UCl based systems are feasible to deliver the requested in-core breeding within the given solubility limits of fertile material in the salt system using either uranium as start-up fissile component or plutonium. This result is enriched by the analysis of the achievable full power days per inserted mass of plutonium. These new insights support reactor optimization and lead to a first conclusion that systems with lower power density could be very attractive in the case of low fuel cost, like it would be given when operating on spent nuclear fuel.

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