CATHARE-3: A new system code for thermal-hydraulics in the context of the NEPTUNE project

After a thorough analysis of the industrial needs and of the limitations of current simulation tools, EDF and CEA (Commissariat à l’Energie Atomique) launched the NEPTUNE Project in 2001 (see Guelfi et al., 2007) with the support of AREVA-NP and IRSN. The NEPTUNE activities include software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques and new experimental programs. Four different simulation scales were addressed including DNS (Direct Numerical Simulation), CFD in open medium (Computational Fluid Dynamics), component (subchannel-type analysis) and system (reactor modeling) scales.In 2006 CEA, EDF, AREVA-NP and IRSN defined the strategy for the system scale of NEPTUNE and the CATHARE-3 development was launched. The main objectives are:

•

advanced physical modeling of two-phases flows, mainly by using multi-field and turbulence models,

•

improved 3D modeling by the use of fine and non conforming structured meshes,

•

generalized coupling capabilities with other thermal-hydraulic scales and with other disciplines (core physics, structural mechanics, …),

•

extension of the applicability to new Gen IV reactors (Sodium Cooled Fast Breeder Reactors, Gas Cooled Reactors, Supercritical Light Water Reactors),

•

a true object-oriented code architecture.

At the same time CATHARE-3 is in continuity with the CATHARE-2 code which is the current industrial version of CATHARE and internationally used for nuclear power plant safety analysis, in simulators and in coupled simulation tools. The road map of these two codes will allow a smooth transition from CATHARE-2 to CATHARE-3 for all users.This paper gives an overview of the choices made for the development of CATHARE-3 including new physical models, validation strategy and experimental programs, numerical improvements, enhanced coupling capability and software architecture evolution. The current status of the project as well as the overall schedule will be presented.     

Measurement of secondary emissions during laser cutting of steel equipments

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:

•

quantifying and characterizing the secondary emissions produced by Nd-YAG laser cutting of Uranus 65 steel pieces and examining the influence of different parameters,

•

qualifying a prefiltration technique and particularly an electrostatic precipitator,

•

comparing the Nd-YAG laser used with other cutting tools previously studied especially on aerosol production and aerosol size distribution.

     

Deposition of ZrC coats on UO2 particles using the chloride process

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

•

to determine the conditions of ZrC deposition on fuel kernels;

•

to assess its viability as a fuel coating before and after irradiation including interactions between ZrC and key fission products.

The main goal of the present study is to investigate influence of conditions of zirconium carbide deposition on uranium dioxide kernel in a fluidized bed on composition of coats and some their characteristics. Results of such investigations are given.     

Comparison of several resuspension models against measured data

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:

•

Firstly, a method of implementation of a resuspension into a system code is presented.

•

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.

The following conclusions are drawn from the performed exercise:

•

The implementation of resuspension model is performed in such a way that it has a general validity for both steady state and transient conditions.

•

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.

•

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.

•

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.

     

Sustainable integration of EU research in severe accident phenomenology and management

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:

-

Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents;

-

Harmonizing and re-orienting the research programmes, and defining new ones;

-

Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena;

-

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;

-

Developing scientific databases, in which the results of research experimental programmes are stored in a common format;

-

Developing a common methodology for probabilistic safety assessment of NPPs;

-

Developing short courses and writing a text book on severe accidents for students and researchers;

-

Promoting personnel mobility amongst various European organizations.

This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvements of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various partners.Most initial objectives were reached but the continuation of the SARNET network, co-funded by EC in the 7th Framework Programme (SARNET2 project that started in April 2009 for 4 years), will consolidate the first assets and focus mainly on the highest priority pending issues as determined during the first period. The objective will be also to make the network evolve towards a complete self-sustainability.     

Lead Coolant Test Facility—Design concept and requirements

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:

•

Develop and demonstrate feasibility of submerged heat exchanger

•

Develop and demonstrate open-lattice flow in electrically heated core

•

Develop and demonstrate chemistry control

•

Demonstrate safe operation

•

Provision for future testing

Across these five broad areas are supported by twenty-one specific requirements. The purpose of this facility is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420 °C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7 M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.     

Anomalies and other concerns related to the critical heat flux

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:

•

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.

•

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).

•

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.

•

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.

•

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.

•

CHF fluid-to-fluid modelling: differences in CHF trends at certain conditions between refrigerants and water at equivalent conditions.

The mechanisms responsible for these trends and the implications for bundle geometries are discussed.Concerns regarding the reported uncertainty of predicted CHF values and the range of application of CHF prediction methods are also discussed.     

Influence of some variable parameters on bremsstrahlung dosimetric characteristic for electron linac

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:

•

The optimization of the electron-bremsstrahlung conversion output by optimizing the target thickness and

•

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.

The measurements have been performed with the 3.5 and 10 MeV linear electron accelerators in NILPRP-Electron Accelerator Laboratory.     

Feasibility of the electrochemical way in molten fluorides for separating thorium and lanthanides and extracting lanthanides from the solvent

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:

•

The solvents to be used for efficient separation must be fluoride media containing lithium as cation.

•

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.

     

Results of AVR fuel pebble irradiation at increased temperature and burn-up in the HFR Petten

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:

•

increased central fuel temperature of 1250 °C compared to 1000-1200 °C in earlier irradiation tests;

•

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;

•

confirmation of low coated particle failure fractions due to temperature, burn-up and neutron fluence.

This paper provides the irradiation history of the experiment including data on fission gas release. Post-irradiation examinations at NRG Petten and JRC-ITU Karlsruhe included the verification of the received neutron fluences, burn-up and spectrum. They will be followed shortly by safety-relevant heating tests at JRC-ITU to verify fission product retention by out-of-pile heating tests beyond 1600 °C.     

FISA-2009 Conference on Euratom Research and Training Activities: Nuclear Fission - Past, Present and Future (Generation-II, -III and -IV + Partitioning and Transmutation)

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:

•

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;

•

Generation-III (i.e. today, construction 2000-2040+): continuous improvement of safety and reliability, and increased industrial competitiveness in a growing energy market;

•

Generation-IV (i.e. tomorrow, construction from 2040) for increased sustainability though optimal utilisation of natural resources and waste minimisation, and increased proliferation resistance.

Consequently, the focus of the lectures devoted to Generation-II and -III is on the major scientific challenges and technological developments needed to guarantee safety and reliability, in particular issues associated with plant lifetime extension and operation.The focus of the lectures devoted to Generation-IV is on the design objectives and associated research issues that have been agreed upon internationally, in particular the ambitious criteria and technology goals established at the international level by the Generation-IV International Forum (GIF). In the future, electricity must continue to be produced competitively, and in addition high temperature process heat may also be required, while exploiting a maximum of fissile and fertile material and recycling all actinides, both safely and reliably. Scientific viability studies and technological performance tests for each Generation-IV system are now being carried out in many European Union (EU) Member States, in collaboration with other laboratories worldwide as part of the inter-governmental GIF agreement. The ultimate phase of commercial deployment will be from 2040, but no one can predict accurately when industry and investors will make firm, often difficult decisions regarding the construction of these very innovative Generation-IV systems. However, to be deployed commercially, it must first be demonstrated that Generation-IV technology can be a beneficial, responsible and sustainable response to the long-term challenges faced by society to establish a low-carbon economy.     

A method for rapid vulnerability assessment of structures loaded by outside blasts

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:

•

results comparable to full scale dynamic simulations using explicit finite element codes and

•

the performance of the method in screening the existing structures and providing the structural reliability information for the vulnerability analysis.

     

Current status and future plans of Advanced ORIENT Cycle strategy

For minimization of the ecological risks inherent in nuclear fuel recycling, a new fuel cycle paradigm was proposed and its key technology developments have been carried out as a part of the Advanced Optimization by Recycling Instructive Elements (Adv.-ORIENT) Cycle strategy. The basic concept of the Adv.-ORIENT Cycle uses a three-pronged approach, separation, transmutation and utilization of nuclides and elements, based on the FBR fuel cycle. Fundamental research studies done in Adv.-OEIENT Cycle [Phase-1: 2006-2010] have led to the following findings.

1.

Cs and Sr separation and its utilization: Silica gel loaded with ammonium molybdophosphate (AMP) and hybrid organic microcapsules with crown ether D18C6 were investigated to chromatographically separate Cs and Sr, respectively. In particular, uptake experiments of Cs from solution simulating the spent fuel solution obtained were carried out by a batch method, and the uptake rates achieved were more than 90%.

2.

Minor actinide (MA)/lanthanide/fission product (FP) separation: A tertiary pyridine type resin (TPR) can be used to recover Am, Cm and lantanide elements with a high separation factor by a chromatographic method from spent fuel solution. The TPR can be used with hydrochloric acid (HCl) as well as nitric acid (HNO₃).

3.

Ru, Ph, Pd and Tc separation: A catalytic electrolytic extraction (CEE) method can effectively separate Ru, Ph, Pd and Tc. High recovery ratios of Ru, Rh, Pd, Tc, Se, etc. were achieved using HCl solutions. Rh co-deposition significantly accelerated reduction of Ru, Tc and Re using HNO₃ solutions.

4.

Ru, Ph, Pd and Tc utilization: Based on the mixed deposit obtained from the CEE experiments, Ru/Rh/Pd/Tc(Re)-Pt electrodes provided the highest catalytic reactivity in the electrolytic production of hydrogen in an alkali solution.

5.

Basic engineering research: Results of corrosion experiments showed that Hastelloy-B and Ta had a good anti-corrosive property in a wide range of HCl concentrations. The basic thermo-chemical stability of TPR and tri-butyl phosphorous (TBP, as a reference) was also experimentally studied, and the process safety conditions to be specified for practical use of TPR could be identified.

     

European MSc Programs in Nuclear Sciences—To meet the need of stakeholders

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:

•

European MSc Radiation Protection,

•

European MSc Analytical Radiochemistry,

•

European MSc Radioecology.

All three master programs would be developed using the framework provided by the Bologna Convention and the lecturing could be shared among specialist Scientists within a network of collaborating universities. Therefore, educational plans have been developed for the above MSc degrees. These plans envisage each degree comprising three modules that are common to all the degrees (3 × 10 ECTS credits), three specialist modules (3 × 10 ECTS credits) and a research project (1 × 60 ECTS credits). The courses should be aimed, not only to fill the identified European post-graduate education gap in radiological sciences, but also to provide a modular structure that is easily accessed by stakeholders for CPD training. It is anticipated that the European Masters will meet the academic training requirements of qualified experts”, as defined by the European Commission and the IAEA. At the Norwegian University of Life Sciences (UMB) a pilot MSc in Radioecology has successfully been initiated in collaboration with UK and France.     

A consistent approach to assess safety criteria for reactivity initiated accidents

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:

-

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;

-

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;

-

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.

This approach is planned to be followed for each phase of the RIA. An example of application is presented to evaluate a PCMI limit for a zircaloy-4 cladding UO₂ rod at Hot Zero Power.     

Cosmic ion bombardment of the icy moons of Jupiter

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:

-

C-ion implantation is not the dominant formation mechanism of CO₂ on Europa, Ganimede and Callisto.

-

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.

-

Implantation of protons into carbon dioxide produces some species containing the projectile (H₂CO₃, and O-H in poly-water).

-

Implantation of protons into sulfur dioxide produces SO₃, polymers, and O₃ but not H-S bonds.

-

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.

-

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.