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.

INPRO Assessment of an INS in the Area of Safety of Fuel Cycle Installations

2006 ◽  
Author(s):  
B. Raj ◽  
Y. Busurin ◽  
F. Depisch
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Early Stage ◽  
Sol Gel ◽  
Safety Issues ◽  
Stage Of Development ◽  
Energy Needs ◽  
Fuel Fabrication ◽  
Specific Assessment ◽  
Acceptance Limits

INPRO has defined requirements organized in a hierarchy of Basic Principles, User Requirements and Criteria (consisting of an indicator and an acceptance limit) to be met by innovative nuclear reactor systems (INS) in six areas, namely: economics, safety, waste management, environment, proliferation resistance, and infrastructure. If an INS meets all requirements in all areas it represents a sustainable system for the supply of energy, capable of making a significant contribution to meeting the energy needs of the 21st century. Draft manuals have been developed, for each INPRO area, to provide guidance for performing an assessment of whether an INS meets the INPRO requirements in a given area. The manuals set out the information that needs to be assembled to perform an assessment and provide guidance on selecting the acceptance limits and, for a given INS, for determining the value of the indicators for comparison with the associated acceptance limits. Each manual also includes an example of a specific assessment to illustrate the guidance. This paper discusses the manual for performing an INPRO assessment in the area of safety of fuel cycle installations. The example, chosen solely for the purpose of illustrating the INPRO methodology, describes an assessment of an MOX fuel fabrication plant based on sol-gel technology and illustrates an assessment performed for an INS at an early stage of development. The safety issues and the assessment steps are presented in detail in the paper.

scholarly journals Fingerprinting commercial nuclear reactor forensics using the ORIGEN-S simulation code

2018 ◽  
Vol 48 ◽  
pp. 1860126
Author(s):  
Iyabo Usman ◽  
David Vermillion ◽  
Howard Hall ◽  
Steve Skutnik
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Reference Database ◽  
Cycle Period ◽  
Isotopic Ratios ◽  
Spent Fuel ◽  
Concentration Data ◽  
Simulation Code

The ability to determine the origin of a specific spent-fuel sample from a commercial nuclear reactor was studied using the Origen-S simulation code by calculating the plutonium and uranium isotopic concentration data for a range of nuclear power reactors. This range of reactors is based on a typical Westinghouse PWR fuel assembly with a fuel type of W17 X 17, having individual operating histories. Isotopic ratios of plutonium in nuclear reactors during the fuel-cycle period provide information on how the plutonium grows into the fuel as a function of burnup, as well as its attractiveness to proliferators. Using the results from the calculation of uranium and plutonium isotopic ratios, the origin of each spent-fuel assembly for a particular reactor can be predicted and documented for a future nuclear forensics reference database.

scholarly journals A new tool for the simulation of different nuclear fleets at equilibrium

2022 ◽  
Vol 8 ◽  
pp. 1
Author(s):  
Heddy Barale ◽  
Camille Laguerre ◽  
Paul Sabatini ◽  
Fanny Courtin ◽  
Kévin Tirel ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactor ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Main Tool ◽  
Simulation Code ◽  
Simulation Tools ◽  
Fuel Cycles ◽  
Cycle Simulation ◽  
The Impact

Scenario simulations are the main tool for studying the impact of a nuclear reactor fleet on the related fuel cycle facilities. This equilibrium preliminary study aims to present the functionalities of a new tool and to show the wide variety of reactors/cycles/strategies that can be studied in steady state conditions and validated with more details thanks to dynamic code. Different types of scenario simulation tools have been developed at CEA over the years, this study focuses on dynamic and equilibrium codes. Dynamic fuel cycle simulation code models the ingoing and outgoing material flow in all the facilities of a nuclear reactor fleet and their evolutions through the different nuclear processes over a given period of time. Equilibrium fuel cycle simulation code models advanced nuclear fuel cycles in equilibrium conditions, i.e. in conditions which stabilize selected nuclear inventories such as spent nuclear fuel constituents, plutonium or some minor actinides. The principle of this work is to analyze different nuclear reactors (PWR, AMR) and several fuel types (UOX, MOX, ERU, MIX) to simulate advanced nuclear fleet with partial and fully plutonium and uranium multi-recycling strategies at equilibrium. At this first stage, selected results are compared with COSI6 simulations in order to evaluate the precision of this new tool, showing a significant general agreement.

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