bluuebunny/crossref_metadata_2025_split huggingface dataset:
Framework for Assessing Transition Scenarios to Sustainable Nuclear Energy Systems
Authors: Fesenko Galina , Kuznetsov Vladimir , Usanov Vladimir 
Published: November 2014

The International Atomic Energy Agency’s (IAEA’s) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was established in 2000 with the goal to ensure a sustainable nuclear energy supply to meet the global energy needs in the 21st century. The INPRO activities on global and regional nuclear energy scenarios provide newcomers and mature nuclear countries alike with better understanding of options for making a collaborative transition to future sustainable nuclear energy systems. Collaborative project GAINS (Global Architecture of Innovative Nuclear Energy Systems Based on Thermal and Fast Reactors Including a Closed Fuel Cycle) developed an internationally verified analytical framework for assessing such transition scenarios. The framework (hereafter, GAINS framework) is a part of the integrated services provided by IAEA to Member States considering initial development or expansion of their nuclear energy programmes. The paper presents major elements of the analytical framework and selected results of its application, including:
• Long-term nuclear energy demand scenarios based on the IAEA Member States’ high and low estimations of nuclear power deployment until 2030 and expected trends until 2050 and on forecasts of competent international energy organizations;
• Heterogeneous world model comprised of groups of non-personified non-geographical countries (NGs) with different policy regarding nuclear fuel cycle back end;
• Architectures of nuclear energy systems;
• Metrics and tools for the assessment of dynamic nuclear energy system evolution scenarios regarding sustainability, including a set of key indicators and evaluation parameters;
• An internationally verified database with best estimate material flow and economic characteristics of existing and advanced nuclear reactors and associated nuclear fuel cycles needed for material flow analysis and comparative economic analysis, extending the previously developed IAEA databases and taking into account preferences of different countries;
• Selected results of sample analysis for scenarios involving transition from the present fleets of nuclear reactors and fuel cycles to future sustainable nuclear energy system architectures involving innovative technological solutions.
Improvement of the System for Passive Heat Removal Through Steam Generators (SG PHRS) on NPP With VVER-1200 in the Light of “Fukushima” Accident
Authors: Bezlepkin Vladimir , Semashko Sergey , Alekseev Sergey , Ivanova Marina , Vardanidze Teimuraz , Petrov Yuriy 
Published: November 2014

As a result of catastrophic events on the nuclear power plant “Fukushima” the European organizations on regulation of nuclear power (ENSREG) initiated wide-scale measures for complex designs revision of already operating and under construction European and Russian NPPs. Inspection was made about resistance of power units to external influences of the natural character, being accompanied by multiple failures of safety systems.
Within these works stress tests for constructed power units of LAES-2 and the Baltic NPP were executed. The structure of these checks included the settlement analysis of a condition of NPP at accident with loss of all AC power supply sources which results are presented in report materials.
Accident calculations with a full blackout were executed on the best-estimated heat-hydraulics code KORSAR/GP for justification of power unit preservations in the intact condition within 72 hours from the accident beginning by means of SG PHRS.
The system is developed for feed of the SG PHRS tanks and the fuel pool for working capacity extension the SG PHRS and power unit preservation in a stable condition more than 72 hours from the accident beginning.
Use of system for feed of tanks the SG PHRS and the fuel pool allows to increase significantly resistance of the NPP to external influences of the natural character and to increase time of preservation of the blackout power unit in a stable condition more than 5 days.
Safety Feature of EU-ABWR for Fukushima Accident
Authors: Kamei Kazuhiro , Kuroda Masatomo , Kojima Yoshihiro , Yano Kazuki 
Published: November 2014

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash, severe accident mitigation systems, the N+2 principle in safety systems, the diversity principle and a large output of 1600 MWe. These features enable EU-ABWR’s design objectives and principles to be consistent with the safety requirements of Western European Nuclear Regulators’ Association (WENRA) and Finnish YVL guides.
By designing safety features of the EU-ABWR based on Defense in Depth Principle, EU-ABWR has a capability to mitigate Design Extension Conditions like Fukushima Accident by using a combination of passive and active systems. Furthermore, a severe accident is mitigated passively without containment venting for at least 72 hours.
Compatibility of Ferritic/Martensitic Steel T91 With Flowing LBE Containing 10−7 Mass% Dissolved Oxygen at 450 and 550°C
Authors: Tsisar Valentyn , Schroer Carsten , Wedemeyer Olaf , Skrypnik Aleksandr , Konys Jürgen 
Published: November 2014

Corrosion behavior of ferritic/martensitic (F/M) steel T91 with 9 % nominal chromium content was investigated in flowing (2 m/s) Pb-Bi eutectic (LBE) with 10−7 mass% dissolved oxygen at 450 and 550°C for up to 8766 and 2011 h, respectively. The corrosion process and material loss were analyzed. The steel generally shows oxidation consisting of the formation of an Fe-Cr spinel layer. Oxidation resulted in metal recession which does not exceed ∼ 10 μm after 8766 h at 450°C and ∼15 μm after 2011 h at 550°C. Local liquid-metal attack in the form of pits occurs at both temperatures. At 450°C, remarkable liquid-metal attack is observed after 5015 h, while, at 550°C, it is found already after 1007 h. The maximum depth of local attack reaches 960 μm after 8766 h at 450°C, and 190 μm after 1007 h at 550°C.
The obtained results are compared with previous experiments at similar testing conditions, except for a higher oxygen concentration of 10−6 mass%.
Analysis of Effects of Buoyancy Force on Steam Condensation With Non-Condensable Gases in a Rectangular Channel
Authors: Wang Xian-Mao , Chang Hua-Jian , Han Li-Yong 
Published: November 2014

With the development of science and technology, some important passive features have been used in nuclear reactors, one of which is passive containment cooling system (PCCS). In the system, steam condensation plays an important role in removing heat from the containment atmosphere during a postulated accident. It has been found that during most time of an accident, the gas regime in the containment will be under natural and mixed convection. Advanced pressurized water reactor (CAP1400), designed by State Nuclear Power Technology Corporation (SNPTC) in China, is one of Chinese national science and technology projects. Since the PCCS has been applied in CAP1400, the study of condensation with non-condensable gases under natural and mixed convection becomes necessary.
To have a deeper understanding on the phenomenon of condensation with non-condensable gases under natural and mixed convection, an experiment facility was set up by State Nuclear Power Technology Research & Development Centre (SNPTRD). The test section of the facility is a rectangular channel with one of the walls acting as a condensing plate. The effects of buoyancy force on steam condensation with non-condensable gases are investigated. Also, a CFD model is set up to simulate the process.
Thermal-Hydraulic and Neutronic Analysis of a Re-Entrant Fuel Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors
Authors: Peiman W. , Pioro I. , Gabriel K. 
Published: November 2014

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts one of which is the SuperCritical Water-cooled Reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as the coolant. The SCWR concept is considered to be an evolution of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), which comprise 81% of the current fleet of operating nuclear reactors and are categorized under Generation II nuclear reactors. The latter water-cooled reactors have thermal efficiencies in the range of 30–35% while the evolutionary SCWR will have a thermal efficiency of about 40–45%.
In terms of a pressure boundary SCWRs are classified into two categories, namely, Pressure Vessel (PV) SCWRs and Pressure Channel (PCh) SCWRs. A generic pressure channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350 and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve the thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel channel design (for PCh SCWR). Second, a nuclear fuel and fuel cycle should be selected. Third, materials for core components and other key components should be selected based on material testing and experimental results.
Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and re-entrant channel concepts. The objective of this paper is to study thermal-hydraulic and Neutronic aspects of a re-entrant fuel channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB which calculates the fuel centerline temperature, sheath temperature, coolant temperature and heat transfer coefficient profiles.
A lattice code and a diffusion code were used in order to determine the power distribution inside the core. Then, the heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents the fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature and the sheath temperature exceed the temperature limits of 1850°C and 850°C for fuel and sheath, respectively.
Investigation of MHD RANS Modeling Base on DNS Database Under the Advanced Blanket Design Conditions Utilized Molten Salt
Authors: Osawa Naoki , Yamamoto Yoshinobu , Kunugi Tomoaki 
Published: November 2014

In this study, validations of Reynolds Averaged Navier-Stokes Simulation (RANS) based on Kenjeres & Hanjalic MHD turbulence model (Int. J. Heat & Fluid Flow, 21, 2000) coupled with the low-Reynolds number k-epsilon model have been conducted with the usage of Direct Numerical Simulation (DNS) database.
DNS database of turbulent channel flow imposed wall-normal magnetic field on, are established in condition of bulk Reynolds number 40000, Hartmann number 24, and Prandtl number 5.
As the results, the Nagano & Shimada model (Trans. JSME series B. 59, 1993) coupled with Kenjeres & Hanjalic MHD turbulence model has the better availability compared with Myong & Kasagi model (Int. Fluid Eng, 109, 1990) in estimation of the heat transfer degradation in MHD turbulent heat transfer.
Application of 3D Coupled Code SPARKLE-2 to PWR Anticipated Transient Without Scram
Authors: Asano Koji , Sakamoto Hikaru , Imura Satoshi , Ogawa Junto 
Published: November 2014

An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) followed by failure of the automatic reactor trip function of the reactor protection system. The failure of the reactor to shut down during the certain AOOs can lead to increase in reactor coolant system (RCS) pressure and decrease in departure from nucleate boiling ratio (DNBR) margin for a pressurized water reactor (PWR). Japanese standard PWRs are equipped with ATWS mitigation system which consists of a diverse mitigation system which is independent from the reactor trip system. The ATWS mitigation system automatically initiates isolation of the main steam line flow and the auxiliary feed water system under condition indicative of an ATWS.
Mitsubishi Heavy Industries, Ltd. (MHI) applies 3D coupled code, SPARKLE-2 [1] [2], to the ATWS evaluation. SPARKLE-2 is a 3D coupled code developed by MHI and consist of the PWR system transient analysis code M-RELAP5, the 3D neutron kinetics code COSMO-K [3] and the 3D core thermal-hydraulics code MIDAC [4]. SPARKLE-2 implements the 3D characteristics such as local moderator feedback and change in 3D power distribution during transient.
Thanks to gain from the 3D calculation, the analysis results show that the plant transients are effectively mitigated by the ATWS mitigation system and the RCS pressure and the minimum DNBR meet the safety criteria. These results also show that operational margins are increased, which enables more flexible design of the reload core.
Scaling Application of a Hot Leg SBLOCA Scenario to a Nuclear Power Plant
Authors: Querol Andrea , Gallardo Sergio , Verdú Gumersindo 
Published: November 2014

Several experimental facilities, such as the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA), have been built to reproduce some accidental scenarios because full-scale testing is usually impossible to perform.
One of the objectives of these Integral Test Facilities (ITFs) is to obtain measured data to be compared to simulations in order to test the capability of the thermalhydraulic codes to reproduce experimental conditions.
The applicability of these experimental results to a full-size power plant system depends on the scaling criteria adopted.
The present paper is focused on the simulation and the scaling of the Test 1-2 in the frame of the OECD/NEA ROSA Project to a Nuclear Power Plant (NPP). This test simulates a hot leg 1% Small Break Loss-Of-Coolant Accident (SBLOCA) in a Pressurized Water Reactor (PWR) under the actuation of High Pressure Injection (HPI) system and Accumulator Injection System (AIS).
A scaled-up NPP TRACE5 input has been developed from a LSTF TRACE5 model validated by authors in previous works. The scaled-up model has been developed conserving the power-to-volume scaling ratios of LSTF components, initial and boundary conditions. Lengths and diameters of hot legs have been scaled from LSTF model trying to conserve Froude number.
A comparison between both TRACE5 models (LSTF and scaled-up NPP) is performed (system pressures, discharged inventory and collapsed liquid levels). Special TRACE5 models such as Choked flow model and OFFTAKE model have been tested. A 3D VESSEL component has been tested in comparison to 1D TEE component to simulate the hot leg where the SBLOCA is located and varying the break orientation (downwards and upwards). Finally, a sensitivity analysis has been made to determine the effect of the break size in the SBLOCA range.
Characterisation of FeCrAlY Coatings for Applications in Lead Environment
Authors: Di Gabriele Fosca , Houdkova Sarka , Soukup Ondrej , Karnik Dalibor 
Published: November 2014

The use of heavy liquid metals (HLM), such as Lead, Pb, is being considered as a coolant in Generation IV fast Reactors. However, structural materials suffer significant damage when in contact with HLMs. Both austenitic and ferritic-martensitic steels are considered and are susceptible to corrosion and/or degradation of mechanical properties. One of the approaches to mitigate erosion/corrosion problems is the use of coatings. This is a layer of a different chemical composition between the base material and the liquid metal coolant in order to prevent the bulk metal from corrosive effects. Coatings are proposed as a valid protection against high temperature damage in this environment. Their capability to grow more stable and protective oxides, by introducing the oxide forming elements (e.g. Al, Si,…) in higher amount, is proven to be an effective alternative to material engineering.
For this specific application, several coatings deposition techniques and compositions have been proposed and tested. In this work the High Velocity Oxygen Fuel, HVOF, combined with laser melting was selected for deposition of FeCrAlY coatings. The combination of the two technologies lead to a compact and adherent coating with an enriched content of Al. The method was evaluated in terms of the corrosion resistance of the coating and also its effect on the microstructure of the substrate alloy. Several attempts were carried out to modify spraying and laser parameters in order to minimise the effect on the substrate and keep the protective properties.
Results are discussed in terms of deposition parameters and protection characteristics.
Delayed Equilibrium Model (DEM) of Flashing Choked Flows Relevant to LOCA and Implementation in System Codes
Authors: Bartosiewicz Yann , Seynhaeve Jean-Marie 
Published: November 2014

In the context of nuclear reactor safety, a pipe breach in the primary circuit is the initiator of a Loss of Coolant Accident (LOCA). The calculation of leak rates involving the discharge of water and steam mixtures plays an important role in the modeling of LOCA’s for both GEN II and GEN III reactors, and also for the Supercritical Water Reactor of GEN IV. Indeed, the flow though the breach determines the depressurisation rate of the system and the time to core uncover which in turn are of major concern for when and how different mitigation auxiliary systems will be initiated and be efficient.
This paper deals with the new development of the DEM model focused on thermodynamic non-equilibrium conditions, which prevail in the flashing flow process near the critical section. This model, developed at the University of Louvain (UCL), is the 1-D Delayed Equilibrium Model (DEM) for choked or critical flow rate in steady state or quasi-steady state conditions. The DEM are assessed against experimental data such as Super Moby-Dick and BETHSY experiments done in CEA during the eighties. The DEM model has been recently implemented in the WAHA code, which is based on a two fluid 1D six equations model. The methodology can be applied to other system code (CATHARE, RELAP, etc.).
Implementation of an Approach for Characterisation of Radiological Inventory for Decommissioning at Chinon A2 NPP
Authors: Traichel A. , Tardy F. , Mummert M. 
Published: November 2014

A general overview of the existing radioactive inventory in the plant is necessary for the decommissioning of nuclear power plants. Based on the knowledge about radiological inventory, appropriate decommissioning techniques and procedures can be specifically used. In order to derive the existing radiological activity in the facility a study was carried out to obtain a representative overview of the total radiological situation at the NPP. Within a study a generic methodology for the radiological characterization was developed. This methodology has been applied on the CO2-circuit of the gas-cooled, graphite-moderated reactor Chinon A2 (MAGNOX type). This paper covers the implementation of an approach for characterisation of radiological inventory for decommissioning.
The approach aims at the definition of the number and distribution of local sampling, required measurements as well as suitable measurement systems leading to a confident result with minimized effort in sampling. The paper covers two main objectives: 1. Methodology at and 2. Determination of radiological inventory based on measured data. The proposed methodology is a stepwise procedure which offers the possibility for minimizing the number of required measurements/sample analyses. At the first step the underlying system is an “as-simple-as-possible”-example with homogeneous contamination. In a second step the methodology is expanded to a more realistic and complex system, for which additional investigations have to be performed. The determination of the radiological inventory using the methodology has to consider a given confidence level and maximum allowed error. Therefore statistical assessment is widely used in estimations.
The result of this first part of study generates the basis for further investigation. This comprises application of methodology to the mentioned technical system. Therefore corresponding measurement and analysis data have to be delivered and proven regarding adequacy for the proposed methodology. From the dataset various measurement systematic and retained parameters could be derived. The accuracy of given measured data was checked by further examination. The result of the performed analysis leads to a statement about the activity in the primary circuit. The result of this study is an comprehensive estimation of the activity by defined statistical processing of analysed data. The result consists moreover of the analysis of the measurement plan and of distribution and deviation within the measured data. Suggestions for further measurement campaigns are provided based on the deviations and inconsistencies of the data. With the help of these suggestions it should be possible to decrease the number of samples and measuring data as well as improve the comparability of separate measurement processes. Particular potential for improvement of the result for inventory can be seen in a deeper analysis of uncertainties, this was realised and will be explained in the paper.
Fast Neutron-Flux Monitoring Instrumentation for Lead Fast Reactors: A Preliminary Study on Fission Chamber Performances
Authors: Lepore Luigi , Remetti Romolo , Cappelli Mauro 
Published: November 2014

Although Sodium Fast Reactors (SFRs) are the most investigated solutions for the future fast-flux facilities so far, Lead Fast Reactors (LFRs) promise to be a very competitive alternative thanks to their peculiarity concerning coolant-safety, fuel cycle and waste management.
Nevertheless, the development of LFRs presents today some drawbacks still to be solved. Due to the harder neutron flux, the current instrumentation developed for SFRs is likely to be extended to LFRs as a first attempt. Otherwise, new monitoring instrumentation could be developed in order to assure more tailored results. Different measurement technologies can be considered for fast flux monitoring and flux absolute measurements in order to provide a reliable and quick calibration of the overall reactor neutron instrumentation. The goal of this paper is to study the validity of typical fast reactor fission chamber designs (e.g. SuperPhénix fission chambers), indicating which are the limitations when used in a LFR environment. Afterwards, alternative detector solutions with enhanced sensitivity and response will be proposed.
Experimental Measurements and CFD Evaluation of the Onset Flow Boiling at Low Pressure and High Subcooling Flow of R-11 Within Vertical Annulus
Authors: Bouaichaoui Y. , Kibboua R. , Matkovič M. 
Published: November 2014

The knowledge of the onset of subcooled boiling in forced convective flow at high liquid velocity and subcooling is of importance in thermal hydraulic studies. Measurements were performed under various conditions of mass flux, heat flux, and inlet subcooling, which enabled to study the influence of different boundary conditions on the development of local flow parameters. Also, some measurements have been compared to the predictions by the three-dimensional two-fluid model of subcooled boiling flow carried out with the computer code ANSYS-CFX-13.
A computational method based on theoretical studies of steady state two phase forced convection along a test section loop was released. The calculation model covers a wide range of two phase flow conditions. It predicts the heat transfer rates and transitions points such as the Onset of Critical Heat Flux.