scholarly journals Development of a MOX equivalence Python code package for ANICCA

2021 ◽  
Vol 7 ◽  
pp. 25
Author(s):  
Bart Vermeeren ◽  
Hubert Druenne
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Supply And Demand ◽  
Calculated Data ◽  
Analysis Tool ◽  
Fuel Cycle Analysis ◽  
Multidimensional Interpolation ◽  
Energy Equivalence ◽  
Fuel Storage ◽  
Data Tables

The basis of the MOX (Mixed OXide) energy equivalence principle is keeping the in-core fuel management characteristics (cycle length, feed size, etc.) of a nuclear reactor unchanged when replacing UOX (Uranium OXide) fuel assemblies by MOX. If the effect of the loading pattern is neglected, such an equivalence is obtained by tuning the Pu content in the MOX fuel, while considering the specific Pu isotopic vector at the time of the core reload to obtain a crossing of the reactivity curves of UOX and MOX at the end-of-cycle core average burnup. It is proposed in this work to extend the fuel cycle analysis tool ANICCA (Advanced Nuclear Inventory Cycle Code) with a MOX equivalence Python code package, which automatically governs the supply and demand of Pu vector isotopes required to obtain MOX equivalence. This code package can determine the reactivity evolution for any given Pu vector by means of a multidimensional interpolation on a directive grid of pre-calculated data tables generated by WIMS10, covering the physically accessible Pu vector space. A fuel cycle scenario will be assessed for a representative evolution of the Pu vector inventory available in spent UOX fuel as a demonstration case, defining the interim fuel storage building dimensional requirements for different reprocessing strategies.

Comparative Fuel Cycle Analysis of Critical and Subcritical Fast Reactor Transmutation Systems

2003 ◽  
Vol 144 (1) ◽  
pp. 83-106 ◽  
Author(s):  
Edward A. Hoffman ◽  
Weston M. Stacey
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Fuel Cycle Analysis

scholarly journals Modeling of Closed Internal Fuel Cycle of a Nuclear Reactor

2021 ◽  
Vol 2021 (1) ◽  
pp. 108-121
Author(s):  
Sergey Nikolaevich Stolbov ◽  
Ilya Mikhailovich Anfimov ◽  
Valery Aleksandrovich Varlachev ◽  
Svetlana Petrovna Kobeleva ◽  
Sergey Aleksandrovich Nekrasov ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle

CAREM X INPRO Case Study

2004 ◽  
Author(s):  
Pablo C. Florido ◽  
Dari´o Delmastro ◽  
Daniel Brasnarof ◽  
Osvaldo E. Azpitarte
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
High Level Waste ◽  
Fuel Cycles ◽  
Surface Intermediate ◽  
Geological Repository ◽  
Enriched Uranium ◽  
Intermediate Storage ◽  
High Level

Argentina is performing CAREM X Nuclear System Case Study based on CAREM nuclear reactor and Once Through Fuel Cycle, using SIGMA for enriched uranium production, and a deep geological repository for final disposal of high level waste after surface intermediate storage in horizontal natural convection silos, to verify INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) methodology. Projections show that developing countries could play a crucial role in the deployment of nuclear energy, in the next fifty years. This case study will be highly useful for checking INPRO methodology for this scenario. In this contribution to ICONE 12, the preliminary findings of the Case Study are presented, including proposals to improve the INPRO methodology.

Thermal Evaluations for Next Generation Nuclear Plant Fuel Testing

Volume 4 ◽  
2004 ◽  
Author(s):  
Richard G. Ambrosek ◽  
Debbie J. Utterbeck ◽  
Brandon Miller
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Nuclear Plant ◽  
Light Water Reactors ◽  
Next Generation ◽  
Preliminary Information ◽  
Advanced Fuel ◽  
Water Reactors ◽  
Test Reactor ◽  
Test Fuel

The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility.

scholarly journals Neutronical aspects of a decay heat measurement: the PRESTO experiment

2020 ◽  
Vol 225 ◽  
pp. 04005
Author(s):  
Francesco Muratori ◽  
Frédéric Nguyen ◽  
Christian Gonnier ◽  
Christophe Le Niliot ◽  
Romain Eschbach
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Thermal Power ◽  
Cooling Time ◽  
Monte Carlo Code ◽  
Heat Measurement ◽  
Power Sources ◽  
Decay Heat ◽  
Human Operators ◽  
Nuclear Reactor Safety

Decay heat is the thermal power released by radioactive decays of unstable isotopes after the nuclear reactor shutdown, and delayed fission reactions. It constitutes a key parameter for the nuclear reactor safety and the nuclear fuel cycle; for this reason, design codes have to be qualified by comparison with experimental measurements. The CEA’s package DARWIN2.3 has been qualified for the calculation of PWR decay heat with two integral measurements: the MERCI experience and the CLAB laboratory’s experiments; performed respectively on the following cooling time intervals: 40 min – 40 days and 12 years – 25 years. A lack of validation in the first hour of cooling time requires to consider large margins on the calculated decay heat value. As a result, delays in core unloading, intervention of human operators and safety systems dimensioning may occur. The PRESTO experiment, under conception at CEA, deals with a decay heat measurement between 1 and 40 minutes of cooling time for a PWR fuel sample irradiated in the Jules Horowitz Reactor (JHR). A previous thermal study showed that measurements could be sensitive to the decay heat 1 minute after the beginning of the cooling time. Now, a more precise estimation of power sources was performed with the Monte-Carlo code TRIPOLI. In this framework, four different device configurations were considered. Our results show that the irradiation power is not enough elevated in configurations where a tungsten shield is present.

Development of Experimental Setup of High Temperature Helium Loop and Preparation for In-Pile Operation

2013 ◽  
Author(s):  
Marija Miletić ◽  
Rostislav Fukač ◽  
Igor Pioro ◽  
Alexey Dragunov
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Fossil Fuels ◽  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Strong Dependence ◽  
Nuclear Industry ◽  
Research Centre ◽  
Experimental Setup ◽  
Creep Tests

Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuels supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow’s needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts which will utilize thermal neutron spectrum is a Very High Temperature Reactor (VHTR). This reactor concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behaviour within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Centre Rez, Ltd. The purpose of the High Temperature Helium Loop (HTHL) is to simulate technical and chemical conditions of VHTR’s coolant. The loop is intended to serve an as experimental device for fatigue and creep tests of construction metallic materials for gas-cooled reactors and it should be also employed for research in field of gaseous coolant chemistry. The loop will serve also for tests of nuclear graphite, dosing and Helium purification systems. Because the VHTR is a new reactor concept, major technical uncertainties remain relative to helium-cooled advanced reactor systems. This paper summarizes the concept of the HTHL in the Research Centre Rez Ltd., its design, utilization and future plans for experimental setup.

Fuel cycle analysis of TRU or MA burner fast reactors with variable conversion ratio using a new algorithm at equilibrium

2009 ◽  
Vol 239 (10) ◽  
pp. 2160-2168 ◽  
Author(s):  
Massimo Salvatores ◽  
Christine Chabert ◽  
Concetta Fazio ◽  
Robert Hill ◽  
Yannick Peneliau ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Conversion Ratio ◽  
Fast Reactors ◽  
Fuel Cycle Analysis
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