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1.
A. Auvinen R. Dickson Y. Dutheillet M. Kunstar M. Mladin C. Séropian 《Nuclear Engineering and Design》2008,238(12):3418-3428
A particular concern in the event of a hypothetical severe accident is the potential release of highly radiotoxic fission product (FP) isotopes of ruthenium. The highest risk for a large quantity of these isotopes to reach the containment arises from air ingress following vessel melt-through. One work package (WP) of the source term topic of the EU 6th Framework Network of Excellence project SARNET is producing and synthesizing information on ruthenium release and transport with the aim of validating or improving the corresponding modelling in the European ASTEC severe accident analysis code. The WP includes reactor scenario studies that can be used to define conditions for new experiments.The experimental database currently being reviewed includes the following programmes:
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- AECL experiments conducted on fission product release in air; results are relevant to CANDU loss of end-fitting accidents;
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- VERCORS tests on FP release and transport conducted by CEA in collaboration with IRSN and EDF; additional tests may potentially be conducted in more oxidizing conditions in the VERDON facility;
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- RUSET tests by AEKI investigating ruthenium transport with and without other FP simulants;
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- Experiments by VTT on ruthenium transport and speciation in highly oxidizing conditions.
2.
Guy Pilot Sylvain Fauvel Guillaume de Dinechin 《Nuclear Engineering and Design》2008,238(8):2124-2134
In order to dismantle some equipments of an obsolete reprocessing plant in Marcoule, studies were carried out by IRSN (Institut de Radioprotection et de Sûreté Nucléaire)/DSU/SERAC in cooperation with CEA (power laser group) on the laser cutting of steel structures, on the request of AREVA NC/Marcoule (UP1 dismantling project manager) and CEA/UMODD (UP1 dismantling owner).These studies were aimed at:
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- quantifying and characterizing the secondary emissions produced by Nd-YAG laser cutting of Uranus 65 steel pieces and examining the influence of different parameters,
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- qualifying a prefiltration technique and particularly an electrostatic precipitator,
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- comparing the Nd-YAG laser used with other cutting tools previously studied especially on aerosol production and aerosol size distribution.
3.
D.C. Groeneveld 《Nuclear Engineering and Design》2011,241(11):4604-4611
This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes or bundle subassemblies. They include the following:
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- Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated channels. This phenomenon has been observed at high flows in both water and Freon.
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- Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30-90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment controlled and the deposition controlled region and causes problems for prediction methods of the form CHF = f(X).
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- Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, and diminishes the effects of flow regime/heat transfer regime transitions responsible for some of the unusual CHF trends, and results in a more gradually decreasing CHF vs. X curve.
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- Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a temperature excursion) to a positive slope.
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- Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the axial temperature distribution of a heated tube experiencing CHF when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical.
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- CHF fluid-to-fluid modelling: differences in CHF trends at certain conditions between refrigerants and water at equivalent conditions.
4.
The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a Lead Coolant Test Facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements are identified in this paper:
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- Develop and demonstrate feasibility of submerged heat exchanger
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- Develop and demonstrate open-lattice flow in electrically heated core
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- Develop and demonstrate chemistry control
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- Demonstrate safe operation
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- Provision for future testing
5.
The graphite dust that will be generated in an HTR/PBMR during normal reactor operation will be deposited inside the primary system and will become radioactive due to sorption of fission products. A significant amount of radioactive dust may be resuspended and released to the environment in case of LOCA. Therefore accurate particle resuspension models are required for HTR/PBMR safety analyses. Thermal-hydraulic safety analyses of HTR/PBMR type reactors are typically performed using computer codes such as FLOWNEX, MELCOR, or SPECTRA. A resuspension model has been implemented in the past into the system code SPECTRA.The purpose of the present paper is twofold:
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- Firstly, a method of implementation of a resuspension into a system code is presented.
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- Secondly, two new resuspension models are introduced and the results are compared with the existing Vainshtein and Rock’n Roll resuspension models. In contrast to the existing models which are valid for turbulent flows, the new models are applicable for both laminar and turbulent flow regimes.
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- The implementation of resuspension model is performed in such a way that it has a general validity for both steady state and transient conditions.
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- Relatively simple, quasi-static models, such as the NRG3 and NRG4 models are as useful as the more complicated dynamic models for resuspension calculation. Applicability to both laminar and turbulent flow is important for analyses of, for example, the PBMR recuperator, where the flow is largely laminar.
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- The framework of resuspension modeling built into SPECTRA, due to its flexibility and large amount of user-defined coefficients, may be used to perform a quick check of the newly developed theoretical models.
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- A key factor in successful resuspension predictions is a good knowledge of the adhesion force and its distribution for dust particles deposited on rough surfaces. Experimental data is needed that will allow to obtain adhesion force distribution for the materials and corresponding surfaces roughness of the components in an actual plant.
6.
Zirconium carbide is the most probable candidate for the replacement of silicon carbide as a force layer in the advanced TRISO fuel particles. To come to such decision in practice first of all it is necessary
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- to determine the conditions of ZrC deposition on fuel kernels;
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- to assess its viability as a fuel coating before and after irradiation including interactions between ZrC and key fission products.
7.
Influence of some variable parameters on bremsstrahlung dosimetric characteristic for electron linac
《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2008,266(10):2134-2137
The paper presents some important problems related to the practical aspect of employing the bremsstrahlung radiation generated by linear accelerators used in the irradiation technological processes, namely:
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- The optimization of the electron-bremsstrahlung conversion output by optimizing the target thickness and
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- the study on the influence of some variable parameters of the accelerator (i.e. the beam current, the magnetron frequency and the injection voltage) on the dosimeter characteristics of the bremsstrahlung radiation.
8.
An alternative way of reprocessing nuclear fuel by hydrometallurgy could be using treatment with molten salts, particularly fluoride melts. Moreover, one of the six concepts chosen for GEN IV nuclear reactors (Technology Roadmap - http://gif.inel.gov/roadmap/) is the molten salt reactor (MSR). The originality of the concept is the use of molten salts as liquid fuel and coolant. During the running of the reactor, fission products, particularly lanthanides, accumulate in the melt and have to be eliminated to optimise reactor operation. This study concerns the feasibility of the separation actinides-lanthanides-solvent by selectively electrodepositing the elements to be separated on an inert (Mo, Ta) or a reactive (Ni) cathodic substrate in molten fluoride media. The main results of this work lead to the conclusions that:
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- The solvents to be used for efficient separation must be fluoride media containing lithium as cation.
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- Inert substrates are suitable for actinide/lanthanide separation; nickel substrate is more suitable for the extraction of lanthanides from the solvent, owing to the depolarisation occurring in the cathodic process through alloy formation.
9.
Leon Cizelj Matja? Leskovar Marko ?epin Borut Mavko 《Nuclear Engineering and Design》2009,239(9):1641-1646
The blast loads have in most cases not been assumed as design basis loads of nuclear power plant buildings and structures. Recent developments however stimulated a number of analyses quantifying the potential effects of such loads.An effort was therefore made by the authors to revisit simple and robust structural analysis methods and to propose their use in the vulnerability assessment of blast-loaded structures. The leading idea is to break the structure into a set of typical structural elements, for which the response is estimated by the use of slightly modified handbook formulas. The proposed method includes provisions to predict the inelastic response and failure. Simplicity and versatility of the method facilitate its use in structural reliability calculations.The most important aspects of the proposed method are presented along with illustrative sample applications demonstrating:
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- results comparable to full scale dynamic simulations using explicit finite element codes and
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- the performance of the method in screening the existing structures and providing the structural reliability information for the vulnerability analysis.
10.
11.
Jean-Pierre Van Dorsselaere Thierry Albiol Tim Haste Leonhard Meyer Bernd Schwinges Alessandro Annunziato 《Nuclear Engineering and Design》2011,241(9):3451-3460
In order to optimise the use of the available means and to constitute sustainable research groups in the European Union, the Severe Accident Research NETwork of Excellence (SARNET) has gathered, between 2004 and 2008, 51 organizations representing most of the actors involved in severe accident (SA) research in Europe plus Canada. This project was co-funded by the European Commission (EC) under the 6th Euratom Framework Programme. Its objective was to resolve the most important pending issues for enhancing, in regard of SA, the safety of existing and future nuclear power plants (NPPs).SARNET tackled the fragmentation that existed between the national R&D programmes, in defining common research programmes and developing common computer codes and methodologies for safety assessment. The Joint Programme of Activities consisted in:
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- Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents;
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- Harmonizing and re-orienting the research programmes, and defining new ones;
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- Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena;
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- Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA) by capitalizing in terms of physical models the knowledge produced within SARNET;
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- Developing scientific databases, in which the results of research experimental programmes are stored in a common format;
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- Developing a common methodology for probabilistic safety assessment of NPPs;
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- Developing short courses and writing a text book on severe accidents for students and researchers;
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- Promoting personnel mobility amongst various European organizations.
12.
A stakeholder needs assessment, carried out under the EU-EURAC and EU-ENEN-II projects, clearly showed that, at the European level, there are a significant and constant need for post-graduates with skills in radiochemistry, radioecology, radiation dosimetry and environmental modelling and a smaller, but still important, demand for radiobiologists and bio-modellers. Most of these needs are from government organizations. If only the nuclear industry is considered, then the largest demand is for radio chemists and radiation protection dosimetry experts. Given this spectrum of need and existing capacity in the areas of radiobiology it was concluded that the needs identified would be most efficiently met by three new degree programs:
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- European MSc Radiation Protection,
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- European MSc Analytical Radiochemistry,
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- European MSc Radioecology.
13.
Michael A. Fütterer Gerard Berg Enrique Toscano Klaas Bakker 《Nuclear Engineering and Design》2008,238(11):2877-2885
The irradiation experiment HFR-EU1bis was performed by the European Commission's Joint Research Centre-Institute for Energy (JRC-IE) in the HFR Petten to test five spherical High Temperature Reactor (HTR) fuel pebbles of former German production with TRISO coated particles for their potential for very high temperature performance and high burn-up. The irradiation started on 9 September 2004 and was terminated on 18 October 2005 after 10 reactor cycles totaling 249 efpd and a maximum burn-up of 11.07% FIMA.The objective of the HFR-EU1bis test was to irradiate five HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, better use of fuel) and thus reduced waste production. The central temperature of all pebbles was kept as closely as possible at 1250 °C and held constant during the entire irradiation, with the exception of HFR downtime and power transients. This is the expected maximum central fuel temperature of a pebble bed VHTR with a coolant outlet temperature of 1000 °C.HFR-EU1bis should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically:
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- increased central fuel temperature of 1250 °C compared to 1000-1200 °C in earlier irradiation tests;
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- irradiation to a burn-up close to 16% FIMA, which is double the license limit of the HTR-Modul; due to a neutronics data processing error, the experiment was prematurely terminated at 11.07% FIMA maximum so that this objective was not fully achieved;
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- confirmation of low coated particle failure fractions due to temperature, burn-up and neutron fluence.
14.
15.
V. Bhatnagar P. Manolatos K. Ptackova G. Van Goethem S. Webster 《Nuclear Engineering and Design》2011,241(9):3376-3388
This paper is an introduction to the research and training activities carried out under the Euratom 7th Framework Programme (FP7, 2007-2011) in the field of nuclear fission science and technology, covering in particular nuclear systems and safety, and including innovative reactor systems and partitioning and transmutation. It is based on the more than 40 invited lectures that were delivered by Euratom project coordinators and keynote speakers at the FISA-2009 Conference (FISA, 2009), organised by the European Commission DG Research, 22-24 June 2009, Prague, Czech Republic.The Euratom programme must be considered in the context of current and future nuclear technology and the respective research effort:
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- Generation-II (i.e. yesterday, NPP construction 1970-2000): safety and reliability of nuclear facilities and energy independence in order to ensure security of supply worldwide;
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- Generation-III (i.e. today, construction 2000-2040+): continuous improvement of safety and reliability, and increased industrial competitiveness in a growing energy market;
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- Generation-IV (i.e. tomorrow, construction from 2040) for increased sustainability though optimal utilisation of natural resources and waste minimisation, and increased proliferation resistance.
16.
Novak Zuber 《Nuclear Engineering and Design》2010,240(8):1986-1996
This paper has three objectives. The first objective is to show how the Einstein-de Broglie equation (EdB) can be extended to model and scale, via fractional scaling, both conservative and dissipative processes ranging in scale from quanta to nuclear reactors. The paper also discusses how and why a single equation and associated fractional scaling method generate for each process of change the corresponding scaling criterion. The versatility and capability of fractional scaling are demonstrated by applying it to:
- (a)
- particle dynamics,
- (b)
- conservative (Bernoulli) and dissipative (hydraulic jump) flows,
- (c)
- viscous and turbulent flows through rough and smooth pipes, and
- (d)
- momentum diffusion in a semi-infinite medium.
17.
G. Strazzulla 《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2011,269(9):842-851
A large number of experiments have been performed in many laboratories in the world with the aim to investigate the physico-chemical effects induced by fast ions irradiating astrophysical relevant materials. The laboratory in Catania (Italy) has given a contribution to some experimental works. In this paper I review the results of two class of experiments performed by the Catania group, namely implantation of reactive (H+, C+, N+, O+ and S+) ions in ices and the ion irradiation induced synthesis of molecules at the interface between water ice and carbonaceous or sulfurous solid materials. The results, discussed in the light of some questions concerning the surfaces of the Galilean moons, contribute to understand whether minor molecular species (CO2, SO2, H2SO4, etc.) observed on those objects are endogenic i.e. native from the satellite or are produced by exogenic processes, such as ion implantation.The results indicate that:
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- C-ion implantation is not the dominant formation mechanism of CO2 on Europa, Ganimede and Callisto.
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- Implantation of sulfur ions into water ice produces hydrated sulfuric acid with high efficiency such to give a very important contribution to the sulfur cycle on the surface of Europa and other satellites.
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- Implantation of protons into carbon dioxide produces some species containing the projectile (H2CO3, and O-H in poly-water).
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- Implantation of protons into sulfur dioxide produces SO3, polymers, and O3 but not H-S bonds.
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- Water ice has been deposited on refractory carbonaceous materials: a general finding is the formation of a noteworthy quantity of CO2. We suggest that this is the primary mechanism to explain the presence of carbon dioxide on the surfaces of the Galilean satellites.
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- Water ice has been deposited on refractory sulfurous materials originating from SO2 or H2S irradiation. No evidence for an efficient synthesis of SO2 has been found.
18.
In the context of more and more demanding reactor managements, the fuel assembly discharge burn-up increases and raises the question of the current safety criteria relevance. In order to assess new safety criteria for reactivity initiated accidents, the IRSN is developing a consistent and original approach to assess safety. This approach is based on:
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- A thorough understanding of the physical mechanisms involved in each phase (PCMI and post-boiling phases) of the RIA, supported by the interpretation of the experimental database. This experimental data is constituted of global test outcomes, such as CABRI or Nuclear Safety Research Reactor (NSRR) experiments, and analytical program outcomes, such as PATRICIA tests, intending to understand some particular physical phenomena;
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- The development of computing codes, modelling the physical phenomena. The physical phenomena observed during the tests mentioned above were modelled in the SCANAIR code. SCANAIR is a thermal-mechanical code calculating fuel and clad temperatures and strains during RIA. The CLARIS module is used as a post-calculation tool to evaluate the clad failure risk based on critical flaw depth. These computing codes were validated by global and analytical tests results;
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- The development of a methodology. The first step of this methodology is the identification of all the parameters affecting the hydride rim depth. Besides, an envelope curve resulting from burst tests giving the hydride rim depth versus oxidation thickness is defined. After that, the critical flaw depth for a given energy pulse is calculated then compared to the hydride rim depth. This methodology results in an energy or enthalpy limit versus burn-up.
19.
Gianni Petrangeli 《Nuclear Engineering and Design》2010,240(4):886-890
Since the event of September 11, 2001 in New York City, many people started to speculate that the same type of attack could in future be brought against other installations. Indeed, the U.S. Nuclear Regulatory Commission decided to require for future plants to assess their resistance to the impact of a large civil airliner. Nuclear plant control authorities of other countries decided in a similar direction. The solutions to the technical problem is usually pursued in the direction of a reinforcement of external plant structures and, in some case, they may not be sufficient. Other solutions of more psychological nature have also been adopted. This paper aims at the demonstration that the use of barrage balloons, already adopted with success in both World Wars and also occasionally after these events, can afford a satisfactory solution to the protection problem at a reasonable cost. This solution is also applicable to existing plants. The history of barrage balloons is summarized. Modern technology offers electronic devices capable to detect in time an approaching threat and the paper describes a new barrage system based also on such new possibilities. If the aircraft crash problem is a real one or not for the next years, nobody knows for sure; however some considerations should be kept in mind:
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- The fact that an accident of this kind “anywhere” is an accident “everywhere” as usual;
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- The extremely uncertain political outlook worldwide, the peculiarities of the oil market and the possible nuclear renaissance.