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.

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 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 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 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.

Assessment of Irradiation Conditions in WWER-440 (213) RPV Surveillance Location

2004 ◽  
Author(s):  
A. Ballesteros ◽  
J. Bros ◽  
L. Debarberis ◽  
F. Sevini ◽  
D. Erak ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Fuel Cycle ◽  
Pressure Vessels ◽  
Irradiation Temperature ◽  
The Arctic ◽  
Charpy Specimen ◽  
Exposure Level ◽  
Reactor Operation ◽  
Surveillance Specimens ◽  
Reactor Pressure

The key component of WWER is the Reactor Pressure Vessel (RPV). The evaluation and prognosis of RPV material embrittlement and the allowable period of its safe operation are performed on the basis of impact test results of irradiated surveillance specimens (SS). The main problem is that the SS irradiation conditions (temperature of SS, neutron flux and neutron spectrum) have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV condition. In particular, the key issue is the possible difference between the irradiation temperature of the SS and the actual RPV temperature. It is recognized that the direct measurement of temperature by thermocouples during reactor operation is the only way for receiving reliable information. In addition, the neutron field’s parameters for surveillance specimens have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level. The COBRA project (http://ie.jrc.cec.eu.int/ames/), which started in August 2000 and had a duration of three years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contain state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during the reactor operation. The selected reactor was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of the capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5±4°C. The results obtained show that there is not need in temperature correction when data of surveillance specimens studies are used for assessment of WWER-440/213 reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals ∼2.7·1012 cm−2s−1 with E>0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.

scholarly journals Modelling of the anti-neutrino production and spectra from a Magnox reactor

2018 ◽  
Vol 170 ◽  
pp. 07008
Author(s):  
Robert W Mills ◽  
David J Mountford ◽  
Jonathon P. Coleman ◽  
Carl Metelko ◽  
Matthew Murdoch ◽  
...  
Keyword(s):  
Daily Basis ◽  
Natural Uranium ◽  
Fissile Material ◽  
Reactor Operation ◽  
Neutrino Source ◽  
Fission Reactor ◽  
Neutrino Production ◽  
Product Distributions ◽  
The Individual ◽  
The University

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.

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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