Tài liệu miễn phí Năng lượng

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Multiobjective optimization for nuclear fleet evolution scenarios using COSI

This paper is about a first optimization study of a transition scenario from the current French nuclear fleet to a Sodium Fast Reactors fleet as defined in the frame of the 2006 French Act for waste management.

6/18/2020 7:01:15 AM +00:00

Modelling of as-fabricated porosity in UO2 fuel by MFPR code

On this base, the mechanistic MFPR code, including physically-grounded models for the fuel porosity evolution in UO2 fuel under various irradiation and thermal regimes, is refined. These modifications complete the consistent description of the fuel porosity evolution in the MFPR code and result in a notable improvement of the code predictions.

6/18/2020 7:01:09 AM +00:00

Physical and economical aspects of Pu multiple recycling on the basis of REMIX reprocessing technology in thermal reactors

The basic strategy of Russian nuclear energy is propagation of a closed fuel cycle on the basis of fast breeder and thermal reactors, as well as the solution of the spent nuclear fuel accumulation and resource problems.

6/18/2020 7:01:03 AM +00:00

Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes.

6/18/2020 7:00:57 AM +00:00

Statistical model of global uranium resources and long-term availability

In this paper, we describe a methodology based on “geological environments”. It provides a more detailed resource estimation and it is more flexible regarding cost modelling. The global uranium resource estimation introduced in this paper results from the sum of independent resource estimations from different geological environments.

6/18/2020 7:00:44 AM +00:00

Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches.

6/18/2020 7:00:38 AM +00:00

An attempt for a unified description of mechanical testing on Zircaloy-4 cladding subjected to simulated LOCA transient

The present study is an attempt to provide a unified description of the failure not directly depending on the tested geometry. This effort aims at providing a better understanding of the link between several existing safety criteria relying on very different mechanical testing.

6/18/2020 7:00:31 AM +00:00

Lock-in thermography for characterization of nuclear materials

A simplified procedure of lock-in thermography was developed and applied for characterization of nuclear materials. The possibility of thickness and thermal diffusivity measurements with the accuracy better than 90% was demonstrated with different metals and Zircaloy-4 claddings.

6/18/2020 7:00:25 AM +00:00

Modelling of powder die compaction for press cycle optimization

A new electromechanical press for fuel pellet manufacturing was built last year in partnership between CEA-Marcoule and ChampalleAlcen. This press was developed to shape pellets in a hot cell via remote handling. It has been qualified to show its robustness and to optimize the compaction cycle, thus obtaining a better sintered pellet profile and limiting damage.

6/18/2020 7:00:19 AM +00:00

2D simulation of hydride blister cracking during a RIA transient with the fuel code ALCYONE

This paper presents 2D generalized plain strain simulations of the thermo-mechanical response of a pellet fragment and overlying cladding during a RIA transient. A fictitious hydride blister of increasing depth (25 to 90% of the clad thickness) is introduced at the beginning of the calculation.

6/18/2020 7:00:12 AM +00:00

Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly Part A: under tensile loading condition

This study has been made in an attempt to find the structural integrity of the fuel assembly (FA) of Chashma Nuclear Power Plant-1 (CHASNUPP-1) at room temperature in air.

6/18/2020 7:00:06 AM +00:00

Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

In this paper, an analysis of the Integral Inherently Safe LWR design (I2S-LWR), a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered.

6/18/2020 6:59:59 AM +00:00

Influence of the austenitic stainless steel microstructure on the void swelling under ion irradiation

To understand the role of different metallurgical parameters on the void formation mechanisms, various austenitic stainless steels were elaborated and irradiated with heavy ions. Two alloys, in several metallurgical conditions (15Cr/15Ni–Ti and 15Cr/25Ni–Ti), were irradiated in the JANNUS-Saclay facility at 600 °C with 2 MeV Fe2+ ions up to 150 dpa.

6/18/2020 6:59:52 AM +00:00

Effect of heat transfer correlations on the fuel temperature prediction of SCWRs

In this paper, we present a numerical analysis of the effect of different heat transfer correlations on the prediction of the cladding wall temperature in a supercritical water reactor at nominal operating conditions. The neutronics process with temperature feedback effects, the heat transfer in the fuel rod, and the thermal-hydraulics in the core were simulated with a three-pass core design.

6/18/2020 6:59:46 AM +00:00

Selection of a tool to decision making for site selection for high level waste

The aim of this paper is to create a panel comparing some of the key decision-making support tools used in situations with the characteristics of the problem of selecting suitable areas for constructing a final deep geologic repository.

6/18/2020 6:59:40 AM +00:00

PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

In this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination.

6/18/2020 6:59:33 AM +00:00

The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

This paper shows basic features of different fuel assembly (FA) application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water–fuel ratio in the VVER FA affects on the fuel characteristics produced by REMIX technology during multiple recycling.

6/18/2020 6:59:27 AM +00:00

Thermal-hydraulics/thermal-mechanics temporal coupling for unprotected loss of flow accidents simulations on a SFR

In this paper, we propose an analysis of the impact of spatial variation and temporal evolution of thermal properties of fuel pins on a CFV-like core behavior during an ULOF accident. These effects are usually neglected under some a priori conservative assumptions.

6/18/2020 6:59:21 AM +00:00

Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2) salt Temperature Reactivity Coefficient (TRC).

6/18/2020 6:59:15 AM +00:00

Comparison of CATHARE results with the experimental results of cold leg intermediate break LOCA obtained during ROSA-2/ LSTF test 7

The paper presents the modeling of the Test Facility ROSA-2/LSTF in the calculation code CATHARE 2.V2.5. OECD/NEA ROSA-2 Project Test 7 was conducted with the Large Scale Test Facility on June 14, 2012. The experiment simulated the thermal-hydraulic responses during a PWR 13% cold leg Intermediate Break Loss Of Coolant Accident (IBLOCA).

6/18/2020 6:59:09 AM +00:00

Thermodynamic exergy analysis for small modular reactor in nuclear hybrid energy system

The paper will present background information on exergy theory; identify the core subsystems in an SMR plant coupled with storage systems in support of renewable energy and hydrogen production; perform a thermodynamic exergy analysis; determine the cost allocation among these subsystems; and calculate unit exergetic costs, unit exergoeconomic costs, and first and second law efficiencies.

6/18/2020 6:59:03 AM +00:00

Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior

The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models.

6/18/2020 6:58:56 AM +00:00

Testing of high temperature materials within HTR program in Czech Republic

Research institutes and also industrial companies in Czech Republic are involved in High Temperature Gas Cooled Reactor (HTGR) program and activities related to the study of advanced materials and HTGR technologies. These activities are supported by EC (within international projects, e.g. FP7-ARCHER, ALLIANCE, GoFastR can be mentioned) and also by Technology Agency of Czech Republic.

6/18/2020 6:58:50 AM +00:00

Investigation of the relationships between mechanical properties and microstructure in a Fe-9%Cr ODS steel

In this paper, a feasibility study concerning the generation of tensile specimens using a quenching dilatometer is presented. The ODS steel investigated contains 9%Cr and exhibits a phase transformation between ferrite and austenite around 870 °C.

6/18/2020 6:58:43 AM +00:00

A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

The paper covers the results of VVER core reflooding studies in fuel assembly (FA) mockup of 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed) FA mockup.

6/18/2020 6:58:37 AM +00:00

Comparison of SERPENT and SCALE methodology for LWRs transport calculations and additionally uncertainty analysis for cross-section perturbation with SAMPLER module

In nuclear safety research, the quality of the results of simulation codes is widely determined by the reactor design and safe operation, and the description of neutron transport in the reactor core is a feature of particular importance.

6/18/2020 6:58:30 AM +00:00

Growth of micrometric oxide layers to explore laser decontamination of metallic surfaces

In this paper we propose a method for the creation of oxide layers on stainless steel 304L with europium (Eu) as contaminant. This technique consists in spraying an Eu-solution on stainless steel samples.

6/18/2020 6:33:20 AM +00:00

Trends in severe accident research in Europe: SARNET network from Euratom to NUGENIA

SARNET (Severe Accident Research Network) was set up under the aegis of the Framework Programmes of the European Commission from 2004 to 2013 and coordinated by IRSN to perform R&D on severe accidents in water-cooled nuclear power plants.

6/18/2020 6:33:14 AM +00:00

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2-zirconium alloy system.

6/18/2020 6:33:07 AM +00:00

RELAP/SCDAPSIM/MOD3.5 analysis of KIT’s QUENCH-14 experiment

The test conditions used in the QUENCH-14 were comparable to the QUENCH-6 experiment that used Zircaloy-4. Simulations presented in the article were performed with MATPRO Zircaloy-4 properties and both QUENCH-6 and QUENCH-14 experimental conditions.

6/18/2020 6:33:01 AM +00:00