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Monte Carlo MSM correction factors for control rod worth estimates in subcritical and near-critical fast neutron reactors
The GUINEVERE project was launched in 2006, within the 6th Euratom Framework Program IP-EUROTRANS, in order to study the feasibility of transmutation in Accelerator Driven subcritical Systems (ADS).
6/18/2020 7:05:21 AM +00:00
Thermal hydraulic simulations of the Angra 2 PWR
In the present work, the thermal hydraulic RELAP5-3D code was used to develop a model of this reactor. The model was performed using geometrical and material data from the Angra 2 final safety analysis report (FSAR).
6/18/2020 7:05:14 AM +00:00
Heterogeneous world model and collaborative scenarios of transition to globally sustainable nuclear energy systems
INPRO has developed heterogeneous global model to capture countries’ different policies regarding the back end of the nuclear fuel cycle in regional and global scenarios of nuclear energy evolution and applied in a number of studies performed by participants of the project. This paper will highlight the model and major conclusions obtained in the studies.
6/18/2020 7:05:08 AM +00:00
Preliminary accident analysis of Flexblue® underwater reactor
The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.
6/18/2020 7:05:02 AM +00:00
Why nuclear energy is essential to reduce anthropogenic greenhouse gas emission rates
Reduction of anthropogenic greenhouse gas emissions is advocated by the Intergovernmental Panel on Climate Change. To achieve this target, countries have opted for renewable energy sources, primarily wind and solar.
6/18/2020 7:04:56 AM +00:00
Flexblue® core design: optimisation of fuel poisoning for a soluble boron free core with full or half core refuelling
In this paper, these two topics are dealt with, from a neutronic point of view. Firstly, an overview of the main challenges of operating without soluble boron is proposed (cold shutdown, reactivity swing during cycle, load following, xenon stability).
6/18/2020 7:04:50 AM +00:00
Cooling the intact loop of primary heat transport system using Shutdown Cooling System in case of LOCA events
The purpose of this paper is to model the operation of the Shutdown Cooling System (SDCS) for CANDU 6 nuclear power plants in case of LOCA accidents, using Flowmaster calculation code, by delimiting models and setting calculation assumptions, input data for hydraulic analysis and input data for calculating thermal performance check for heat exchangers that are part of this system.
6/18/2020 7:04:44 AM +00:00
Evaluation of relevant information for optimal reflector modeling through data assimilation procedures
The goal of this study is to look after the amount of information that is mandatory to get a relevant parameters optimisation by data assimilation for physical models in neutronic diffusion calculations, and to determine what is the best information to reach the optimum of accuracy at the cheapest cost.
6/18/2020 7:04:38 AM +00:00
Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident
The radio nuclei present in the damaged fuel are supposed to be released into the main heat transport system and after that into the containment building in the worst case scenario. Assessing the radioactive nuclei maximum release is the purpose of the present paper.
6/18/2020 7:04:32 AM +00:00
Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges
This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility.
6/18/2020 7:04:24 AM +00:00
Helium behaviour in implanted boron carbide
When boron carbide is used as a neutron absorber in nuclear power plants, large quantities of helium are produced. To simulate the gas behaviour, helium implantations were carried out in boron carbide. The samples were then annealed up to 1500 °C in order to observe the influence of temperature and duration of annealing.
6/18/2020 7:04:18 AM +00:00
Safety operation of chromatography column system with discharging hydrogen radiolytically generated
In this study, behaviors of gas inside the extraction chromatography column were investigated through experiments and Computation Fluid Dynamics (CFD) simulation. N2 gas once accumulated as bubbles in the packed bed was hardly discharged by the flow of mobile phase.
6/18/2020 7:04:12 AM +00:00
A new MV bus transfer scheme for nuclear power plants
Fast bus transfer method is the most popular and residual voltage transfer method that is used as a backup in medium voltage buses in general. The use of the advanced technology like open circuit voltage prediction and digital signal processing algorithms can improve the reliability of fast transfer scheme.
6/18/2020 7:04:05 AM +00:00
Possible in-vessel corium progression way in the Unit 1 of Fukushima Dai-ichi nuclear power plant using a phenomenological analysis
In the field of severe accident, the description of corium progression events is mainly carried out by using integral calculation codes. However, these tools are usually based on bounding assumptions because of high complexity of phenomena.
6/18/2020 7:03:59 AM +00:00
Resonance parameter and covariance evaluation for 16O up to 6 MeV
The intent of this paper is to describe the procedures used in the evaluation of the RP and RPC, the use of the RPC in benchmark calculations and to assess the impact of the 16O nuclear data uncertainties in the calculate dkeff for critical benchmark experiments.
6/18/2020 7:02:53 AM +00:00
Measurements of the effective cumulative fission yields of 143Nd, 145Nd, 146Nd, 148Nd and 150Nd for 235U in the PHENIX fast reactor
The effective Neodymium cumulative fission yields for 235U have been measured in the fast reactor PHENIX relatively to the 235U fission cross-section. The data were derived from isotope-ratio measurements obtained in the frame of the PROFIL-1, PROFIL-2A and PROFIL-2B programs.
6/18/2020 7:02:47 AM +00:00
Cross check of the new economic and mass balance features of the fuel cycle scenario code TR_EVOL
This work is intended to cross check the new capabilities of the fuel cycle scenario code TR_EVOL.This process has been divided in two stages. The first stage is dedicated to check the improvements in the nuclear fuel mass balance estimation using the available datafor the Spanish nuclear fuel cycle.
6/18/2020 7:02:41 AM +00:00
The impact of metrology study sample size on uncertainty in IAEA safeguards calculations
Quantitative conclusions by the International Atomic Energy Agency (IAEA) regarding States' nuclear material inventories and flows are provided in the form of material balance evaluations (MBEs). MBEs use facility estimates of the material unaccounted for together with verification data to monitor for possible nuclear material diversion.
6/18/2020 7:02:34 AM +00:00
Statistical sampling applied to the radiological characterization of historical waste
The present article discusses the application of linear regression, scaling factors (SF) and the so-called “mean activity method” to estimate the activity of DTM nuclides on metallic waste produced at the European Organization for Nuclear Research (CERN).
6/18/2020 7:02:28 AM +00:00
Initial economic appraisal of nuclear district heating in France
This paper focuses on the use of cogeneration for district heating and its possible development perspectives within the French energy transition. After recapping some common assumptions about nuclear cogeneration, we will describe the techno-economic model that we built to evaluate the characteristics of introducing cogeneration into an already operating power plant.
6/18/2020 7:02:22 AM +00:00
Impact of the thermal scattering law of H in H2O on the isothermal temperature reactivity coefficients for UOX and MOX fuel lattices in cold operating conditions
The contribution of the thermal scattering law of hydrogen in light water to isothermal temperature reactivity coefficients for UOX and MOX lattices was studied in the frame of the MISTRAL critical experiments carried out in the zero power reactor EOLE of CEA Cadarache (France).
6/18/2020 7:02:16 AM +00:00
Multiphysics simulation of fast transients with the FINIX fuel behaviour module
In this work, FINIX is used as an internal fuel behaviour module both in reactor physics and in reactor dynamics codes to simulate coupled behaviour in fast transient scenarios.
6/18/2020 7:02:10 AM +00:00
Lessons learned from a review of international approaches to spent fuel management
The review surveyed spent fuel storage and disposal practices, standards, trends and recent developments in 16 countries and carried out more detailed studies into the evolution of spent fuel storage and disposal strategies in four countries.
6/18/2020 7:02:04 AM +00:00
Storage of thermal reactor fuels – Implications for the back end of the fuel cycle in the UK
This paper presents potential fuel storage scenarios for two options: the current nuclear power replacement strategy, which will see 16 GWe of new capacity installed by 2030 and a median strategy, intended to ensure implementation of the UK’s carbon reduction target, involving 48 GWe of nuclear capacity installed by 2040.
6/18/2020 7:01:57 AM +00:00
Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code
The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding.
6/18/2020 7:01:51 AM +00:00
Evaluation of corrosion on the fuel performance of stainless steel cladding
This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4) under the same power history.
6/18/2020 7:01:45 AM +00:00
Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium
The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand.
6/18/2020 7:01:39 AM +00:00
Thermal decomposition analysis of simulated high-level liquid waste in cold-cap
In this study, reaction rates of the thermal decomposition reaction of 13 kinds of nitrates, which are main constituents of simulated HLLW (sHLLW), were investigated using thermogravimetrical instrument in a range of room temperature to 1000 °C.
6/18/2020 7:01:33 AM +00:00
Dose and temperature distribution in spent fuel containing material
The paper deals with dose and temperature characteristics inside the SFCM after transition of the molten mixture to solid state. Calculations were made on simplified spherical models, without connection to some specific nuclear accident.
6/18/2020 7:01:27 AM +00:00
Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors
This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom.
6/18/2020 7:01:21 AM +00:00