Nuclear Science and Data Bank Publications

Accelerator and Spallation Target Technologies for ADS Applications A Status Report (2005)

Actinide and Fission Product Partitioning and Transmutation Ninth Information Exchange Meeting, Nîmes, France, 25-29 September 2006 (2007)

Actinide and Fission Product Partitioning and Transmutation + CD-ROM Tenth Information Exchange Meeting, Mito, Japan, 6-10 October 2008 (2010)

Analytical Benchmarks for Nuclear Engineering Applications Case Studies in Neutron Transport Theory (2008)

Assessment of Fission Product Decay Data for Decay Heat Calculations International Evaluation Co-operation, Volume 25 (2007)

Benchmark on Deterministic Transport Calculations Without Spatial Homogenisation MOX Fuel Assembly 3-D Extension Case (2005)

Benchmark on the KRITZ-2 LEU and MOX Critical Experiments Final Report (2006)

Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume II Volume II: Summary Results of Exercise 1 (2005)

Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume III Volume III: Summary Results of Exercise 2 (2006)

Burn-up Credit Criticality Benchmark - Phase II-C Phase II-C: Impact of the Asymmetry of PWR Axial Burn-up Profiles on the End Effect (2007)

Burn-up Credit Criticality Benchmark - Phase II-D PWR-UO2 Assembly - Study of Control Rod Effects on Spent Fuel Composition (2006)

Chemical Thermodynamics of Organic Ligands (Volume 9) (2005)

Chemical Thermodynamics of Solid Solutions of Interest in Nuclear Waste Management - Volume 10 A State-of-the-art Report (2007)

Chemical Thermodynamics of Thorium - Volume 11 (2009)

Computer Simulation of MASURCA Critical and Subcritical Experiments MUSE-4 Benchmark - Final Report (2006)

Evaluated Data Library for the Bulk of Fission Products (Volume 23) International Evaluation Co-operation, Volume 23 (2009)

Evaluation of Proposed Integral Critical Experiments with Low-moderated MOX Fuel (2005)

Fuels and Materials for Transmutation A Status Report (2005)

Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-hydraulics and Technologies + CD-ROM (2007)

Independent Evaluation of the MYRRHA Project Report by an International Team of Experts (2009)

Inter-code Comparison Exercise for Criticality Excursion Analysis Benchmarks Phase I: Pulse Mode Experiments with Uranyl Nitrate Solution Using the TRACY and SILENE Experimental Facilities (2009)

International Evaluation Co-operation (Vol. 19) Neutron Activation Cross-section Measurements from Threshold to 20 MeV for the Validation of Nuclear Models and Their Parameters (Volume 19) (2005)

International Evaluation Co-operation (Vol. 20) Covariance Matrix Evaluation and Processing in the Resolved/Unresolved Resonance Regions (Volume 20) (2006)

International Evaluation Co-operation (Vol. 21) Assessment of Neutron Cross-section Evaluations for the Bulk of Fission Products (Volume 21) (2005)

International Evaluation Co-operation (Vol. 22) Nuclear Data for Improved LEU-LWR Reactivity Predictions (Volume 22) (2006)

International Evaluation Co-operation (Vol. 26) + CD-ROM Uncertainty and Target Accuracy Assessment for Innovative Systems Using Recent Covariance Data Evaluations (Volume 26) (2008)

International Evaluation Co-operation (Vol. 7) Nuclear Data Standards (Volume 7) (2006)

International Nuclear Data Evaluation Co-operation - CD-ROM Complete Collection of Published Reports as of January 2010 (2010)

JANIS 3 - DVD A Java-based Nuclear Data Display Program - 2010 (2010)

JEFF 3.1 (CD-ROM) (2005)

JEFF Reports CD-ROM Complete Collection of JEFF Reports 1-22 (2010)

Mixed-oxide (MOX) Fuel Performance Benchmark Summary of the Results for the Halden Reactor Project MOX Rods (2007)

Mixed-oxide (MOX) Fuel Performance Benchmark (PRIMO) Summary of the Results for the PRIMO BD8 MOX Rod (2009)

Mobile Fission and Activation Products in Nuclear Waste Disposal Workshop Proceedings, La Baule, France, 16-19 January 2007 (2009)

NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark Volume I: Specifications (2006)

National Programmes in Chemical Partitioning A Status Report (2010)

Nuclear Fuel Cycle Synergies and Regional Scenarios for Europe (2009)

Nuclear Fuel Cycle Transition Scenario Studies Status Report (2009)

Nuclear Production of Hydrogen Third Information Exchange Meeting, Oarai, Japan, 5-7 October 2005 (2006)

Nuclear Production of Hydrogen Fourth Information Exchange Meeting, Oakbrook, Illinois, United States, 13-16 April 2009 (2010)

PENELOPE-2006: A Code System for Monte Carlo Simulation of Electron and Photon Transport Workshop Proceedings, Barcelona, Spain, 4-7 July 2006 (2006)

PENELOPE-2008: A Code System for Monte Carlo Simulation of Electron and Photon Transport Workshop Proceedings, Barcelona, Spain, 30 June-3 July 2008 (2009)

Pellet-clad Interaction in Water Reactor Fuels Seminar Proceedings, Aix-en-Provence, France, 9-11 March 2004 (2005)

Perspectives on Nuclear Data for the Next Decade Workshop Proceedings, Bruyères-le-Châtel, France, 26-28 September 2005 (2006)

Physics and Safety of Transmutation Systems A Status Report (2006)

Physics of Plutonium Recycling - Volume IX Volume IX: Benchmark on Kinetic Parameters in the CROCUS Reactor (2007)

Physics of Plutonium Recycling - Volume VIII Volume VIII: Results of a Benchmark Considering a High-temperature Reactor (HTR) Fuelled with Reactor-grade Plutonium (2007)

Pressurised Water Reactor MOX/UO2 Core Transient Benchmark Final Report (2006)

Reference Values for Nuclear Criticality Safety + CD-ROM (2006)

Research and Test Facilities Required in Nuclear Science and Technology (2009)

Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 7 Workshop Proceedings, Lisbon, Portugal, 17-18 May 2004 (2005)

Source Convergence in Criticality Safety Analyses Phase I: Results for Four Test Problems (2006)

Speciation Techniques and Facilities for Radioactive Materials at Synchrotron Light Sources Workshop Proceedings, Berkeley, California, USA, 14-16 September 2004 (2006)

Speciation Techniques and Facilities for Radioactive Materials at Synchrotron Light Sources Workshop Proceedings, Karlsruhe, Germany, 18-20 September 2006 (2007)

Structural Materials for Innovative Nuclear Systems (SMINS) Workshop Proceedings, Karlsruhe, Germany, 4-6 June 2007 (2008)

The JEFF-3.0 Nuclear Data Library (Reprint) JEFF Report 19 - Synopsis of the General Purpose File (2005)

The JEFF-3.1 Nuclear Data Library JEFF Report 21 (2006)

The JEFF-3.1.1 Nuclear Data Library JEFF Report 22 - Validation Results from JEF-2.2 to JEFF-3.1.1 (2009)

The JEFF-3.1/-3.1.1 Radioactive Decay Data and Fission Yields Sub-libraries JEFF Report 20 (2009)

Utilisation and Reliability of High Power Proton Accelerators Workshop Proceedings, Daejeon, Republic of Korea, 16-19 May 2004 (2005)

Utilisation and Reliability of High Power Proton Accelerators (HPPA5) Workshop Proceedings, Mol, Belgium, 6-9 May 2007 (2008)

VENUS-2 MOX-fuelled Reactor Dosimetry Calculations Final Report (2006)

VVER-1000 Coolant Transient Benchmark (Vol. II) Phase 1 (V1000CT-1), Vol. 2: Summary Results of Exercise 1 on Point Kinetics Plant Simulation (2006)

VVER-1000 Coolant Transient Benchmark - Phase 1 (Vol. 3) Phase I (V1000CT-1), Vol. 3: Summary Results of Exercise 2 on Coupled 3-D Kinetics/Core Thermal-hydraulics (2007)

VVER-1000 MOX Core Computational Benchmark Specification and Results (2006)

Very High Burn-ups in Light Water Reactors (2006)

Detailed publications list by year

Synopsis

Hydrogen has the potential to play an important role as a sustainable and environmentally acceptable energy carrier in the 21st century. This report describes the scientific and technical challenges associated with the production of hydrogen using heat and/or electricity from nuclear power plants, with special emphasis on recent developments in high-temperature electrolysis and the use of different chemical thermodynamic processes. Economics and market analysis as well as safety aspects of the nuclear production of hydrogen are also discussed.

Synopsis

Many countries have been performing a wide range of research on the partitioning and transmutation (P&T) of minor actinides and fission products. The aim is to provide greater flexibility in terms of radioactive waste management strategies and deploying advanced nuclear fuel cycles. This report describes recent and ongoing national research programmes related to chemical partitioning in the Czech Republic, France, Italy, Japan, Korea, the Russian Federation, Spain, the United Kingdom and the United States. European Commission research programmes are also included.

Synopsis

For the successful deployment of the advanced fuel cycle, it is important to apply partitioning and transmutation (P&T) technologies to radioactive waste management. In order to provide experts with a forum to present and to discuss the latest developments in partitioning and transmutation, the NEA has organised, since 1990, a series of biennial information exchange meetings on actinide and fission product P&T.

These proceedings contain all the technical papers and posters presented at the 10th Information Exchange Meeting, which was held on 6-10 October 2008 in Mito, Japan. The meeting addressed the following technical issues: the impact of P&T on waste management and geological disposal; transmutation fuels and targets; partitioning, waste forms and management; materials, spallation targets and coolants; transmutation physics experiments and nuclear data; and transmutation systems design, performance and safety.

Synopsis

The renewed interest in nuclear energy – to a large extent stimulated by concerns about global climate change, high volatility of fossil fuel prices and security of energy supply – has also revived discussions on advanced reactor concepts with the potential to reduce significantly the long-term radioactivity of nuclear waste. One of these concepts is an accelerator-driven system (ADS) which combines a particle accelerator with a subcritical reactor core. The Belgian research centre SCK•CEN at Mol has launched a project aiming to construct an ADS consisting of a high energy proton, linear accelerator combined with a lead-bismuth-cooled, subcritical reactor. The project is called MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications).

The Belgian government asked the OECD Nuclear Energy Agency (NEA) to organise an international peer review of the MYRRHA project to provide an independent evaluation as part of the decision-making process. This report presents the findings from the review, which was conducted by a team of seven high-level experts from seven countries, assisted by the NEA Secretariat.

This publication is also available in French as: Besoins d'installations de recherche et d'expérimentation en sciences et technologies nucléaires -

Synopsis

Experimental facilities are essential research tools both for the development of nuclear science and technology and for testing systems and materials which are currently being used or will be used in the future. As a result of economic pressures and the closure of older facilities, there are concerns that the ability to undertake the research necessary to maintain and to develop nuclear science and technology may be in jeopardy.

An NEA expert group with representation from ten member countries, the International Atomic Energy Agency and the European Commission has reviewed the status of those research and test facilities of interest to the NEA Nuclear Science Committee. They include facilities relating to nuclear data measurement, reactor development, neutron scattering, neutron radiography, accelerator-driven systems, transmutation, nuclear fuel, materials, safety, radiochemistry, partitioning and nuclear process heat for hydrogen production.

This report contains the expert group’s detailed assessment of the current status of these nuclear research facilities and makes recommendations on how future developments in the field can be secured through the provision of high-quality, modern facilities. It also describes the online database which has been established by the expert group which includes more than 700 facilities.

Synopsis

Regional strategies can provide a useful framework for implementing innovative nuclear fuel cycles. The appropriate sharing of efforts and facilities among different countries is necessary in today’s context, as is taking into account proliferation concerns and resource optimisation. The preliminary studies examined in this report show that the expected benefits deriving from partitioning and transmutation (P&T), notably the reduction of radiotoxicity and heat load in a shared repository, can bring advantages to all countries of the region concerned, even when different nuclear energy policies are pursued. The studies also demonstrate that regional strategies tend to favour a nuclear “renaissance” in some countries.

A regional approach is proposed in order to implement the innovative fuel cycles associated with partitioning and transmutation in Europe. The impact of different deployment strategies and policies in various countries is addressed. Regional facilities’ characteristics and potential deployment schedules are also discussed. Further studies should be undertaken to investigate practical issues (fuel transport in particular) and institutional issues which will, without doubt, be very challenging.

Synopsis

This publication reports the conclusions from the work undertaken by Subgroup 23 of the NEA Working Party on International Nuclear Data Evaluation Co-operation (WPEC), whose mission was to produce an international library of neutron cross-section evaluations for the most important fission products.

These fission products are important in the operation of nuclear reactors because some of them contribute delayed neutrons that are useful for reactor control, whereas others have a very high neutron capture cross-section, thus inhibiting the nuclear reaction. The build-up of the fission product poisons determines the maximum duration a given fuel element can be kept in a reactor.

Synopsis

The Joint Evaluated Fission and Fusion (JEFF) Project is a collaborative effort among NEA Data Bank member countries to develop a reference nuclear data library for use in different energy applications. Radioactive decay data forms an integral part of the nuclear data requirements for nuclear applications. In 2005, a completely revised library, JEFF-3.1, was made available. The updated JEFF-3.1.1 Radioactive Decay Data and Fission Yields Sub-libraries were released in 2007.

This report describes the development, contents and initial validation of the JEFF-3.1 Radioactive Decay Data and Fission Yields Sub-libraries, including the 2007 update, JEFF-3.1.1, of these sub-libraries.

Synopsis

The NEA Working Party on Nuclear Criticality Safety established an Expert Group on Criticality Excursion Analysis in 2001 to explore the performance of various transient codes to evaluate criticality accidents in a fissile solution. Inter-code comparison exercises among four transient codes (AGNES, CRITEX, INCTAC and TRACE) have been carried out with typical transient experiments using uranyl nitrate fuel solution.

Two sets of benchmarks were carried out based on experimental programmes performed in the TRACY reactor in Japan, and the SILENE reactor in France. TRACY and SILENE have the same geometrical features: an annular cylinder with a central void tube for a transient rod and similar operational modes for reactivity insertion. The experiments selected are representative benchmarks for low- and high-enriched uranyl nitrate solution, about 10 wt% for TRACY and 93 wt% for the SILENE core.

This report provides an analysis of the benchmark results obtained with four different codes. It will be of particular interest to criticality safety practitioners developing transient codes, notably since little experimental data is available and the existing transient codes are presently unavailable to the public.

Synopsis

The plutonium produced during the operation of commercial nuclear power plants or that has become available from the dismantlement of nuclear weapons needs to be properly managed. One important contribution to the management process consists in validating the calculation methods and nuclear data used for estimates concerning power systems burning mixed-oxide (MOX) fuel. Another important contribution is the improved modelling of MOX fuel behaviour in such systems.

Within the framework of the NEA Expert Group on Reactor-based Plutonium Disposition, a fuel modelling code benchmark test was carried out for MOX fuel, with irradiation data on the BD8 MOX rod of the PRIMO programme provided by SCK•CEN and Belgonucléaire. This report summarises the data provided and the fuel characteristics for the irradiation, and presents the calculation results provided by the contributors.

Synopsis

Most experts worldwide agree that disposal of spent nuclear fuel in appropriate formations deep underground provides a suitable option. Most public discussions about these underground repositories concentrate on the radiological hazard associated with the potential leak of actinides to the biosphere. However, the radiotoxicity of the fission products dominates the total radiotoxicity of the spent nuclear fuel during the first 100 years. Thereafter, their radiotoxicity diminishes and the long-term radiotoxicity becomes dominated by the actinides, mainly by the plutonium and americium isotopes.

The aim of the international workshop on Mobile Fission and Activation Products in Nuclear Waste Disposal, MOFAP07, was to review and to identify the needs for further studies on the transport and chemical behaviour of fission products in the geosphere for the safety assessment of radioactive waste repositories. These proceedings contain 22 peer-reviewed papers from the workshop, which should be of particular interest to professionals in the radioactive waste management field.

Synopsis

The JEFF-3.1.1 library is an updated version of the JEFF-3.1 Joint Evaluated File for Fission and Fusion. It consists of sets of evaluated nuclear data for reactor applications. Reliable data of this sort are necessary to improve the safety and economy of existing installations, as well as for the design and efficient operation of advanced nuclear reactors. The improvements in this latest version of the JEFF-3.1.1 library are particularly noteworthy as regards light water reactor applications and the associated fuel cycle.

The present report provides detailed information on the analysis and incremental validation process employed with regard to the JEF-2.2 library, which has provided the basis for the JEFF-3.1.1 library.

Synopsis

Radiation is used in many applications of modern technology. However, its proper handling requires competent knowledge of the basic physical laws governing its interaction with matter. To ensure its safe use, appropriate tools for predicting radiation fields and doses, and subsequently establishing pertinent regulations, are required.

One area of radiation physics that has received much attention concerns electron-photon transport in matter. PENELOPE is a modern, general-purpose Monte Carlo tool for simulating the transport of electrons and photons, which is applicable for arbitrary materials and in a wide energy range. PENELOPE provides quantitative guidance for many practical situations and techniques, including electron and X-ray spectroscopies, electron microscopy and microanalysis, biophysics, dosimetry, medical diagnostics and radiotherapy, and radiation damage and shielding.

These proceedings contain the extensively revised teaching notes of the latest workshop/training course on PENELOPE (version 2008), along with a detailed description of the improved physics models, numerical algorithms and structure of the code system.

Synopsis

Future nuclear fuel cycles could effectively address radioactive waste issues with the implementation of partitioning and transmutation (P&T). Previous studies have defined the infrastructure requirements for several key technical approaches. While these studies have proven extremely valuable, several countries have also recognised the complex, dynamic nature of the infrastructure problem: severe new issues arise when attempting to transit from current open or partially closed cycles to a final equilibrium or burn-down mode. While the issues are country-specific when addressed in detail, it is believed that there exists a series of generic issues related only to the current situation and to the desired end point.

These issues are critical to implementing a sustainable nuclear energy infrastructure. The present report focuses on the definition of key issues, the assessment of technologies and national scenario assessments.

Synopsis

This volume is the eleventh in the OECD Nuclear Energy Agency (NEA) “Chemical Thermodynamics” series. It is based on a critical review of the thermodynamic properties of thorium, its solid compounds and aqueous complexes, initiated as part of the NEA Thermochemical Database Project Phase III (TDB III). The database system developed at the OECD/NEA Data Bank ensures consistency not only within the recommended data sets of thorium, but also amongst all the data sets published in the series. This volume will be of particular interest to scientists carrying out performance assessments of deep geological disposal sites for radioactive waste.

Synopsis

Since 1991, the OECD Nuclear Energy Agency (NEA) has conducted a number of scientific studies to examine nuclear fuel burn-up issues as applied to criticality safety in the transportation, storage and treatment of spent fuel. They have covered a wide range of fuel types, including UOX and MOX fuels for PWR, BWR and VVER reactors.

The objective of the current study was to examine the axial burn-up profiles of PWR UO2 spent fuel assemblies and specifically the fuel assembly end effects and the axial fission density distributions. The study was based on the evaluation of a database of experimentally measured axial burn-up profiles of the Siemens Convoy fuel assemblies, irradiated in the German nuclear power plant Neckarwestheim II.

The report analyses and summarises the solutions to the specified benchmark exercises provided by ten contributors from seven countries. It shows that there is a significant correlation between the asymmetry of axial fuel assembly burn-up profiles and both the end effect and the axial fission density distribution. The results also illustrate the importance of using accurate axial fuel burn-up profiles when designing transport/storage fuel casks.

Synopsis

Preservation of know-how in the nuclear field is promoted through the activities of the OECD Nuclear Energy Agency Data Bank. One area of importance concerns methods for solving radiation transport problems, especially with regard to neutrons. This handbook (in the form of a case study), prepared by Barry D. Ganapol, is the result of such an initiative. It is a compilation of solutions to the transport equation for which analytical representations can be found. It is designed for educational use in courses on analytical transport methods and numerical methods with application to reactor physics. In addition, it contains elements for the continuous improvement of transport methods and for computer code verification. The areas of neutron slowing down, thermalization and one-, two- and three-dimensional neutron transport theory are covered. A series of training courses, based on this compilation of solutions has recently begun.

Synopsis

This publication reports the conclusions from the work undertaken by Subgroup 26 of the NEA Working Party on International Nuclear Data Evaluation Co-operation (WPEC), which focused on the development of a systematic approach to define data needs for advanced reactor systems and to make a comprehensive study of such needs for Generation IV (Gen-IV) reactors. A comprehensive sensitivity and uncertainty study has been performed to evaluate the impact of neutron cross-section uncertainty on the most significant integral parameters related to the core and fuel cycle of a wide range of innovative systems. A compilation of preliminary “Design Target Accuracies” has been put together and a target accuracy assessment has been performed to provide an indicative quantitative evaluation of nuclear data improvement requirements by isotope, nuclear reaction and energy range, in order to meet the design target accuracies, as compiled in the present study. First priorities were formulated on the basis of common needs for fast reactors and, separately, thermal systems.

Synopsis

Structural materials research is a field of growing relevance in the nuclear sector, especially for the different innovative reactor systems being developed within the Generation IV International Forum (GIF), for critical and subcritical transmutation systems, and of interest to the Global Nuclear Energy Partnership (GNEP). Under the auspices of the NEA Nuclear Science Committee (NSC) the Workshop on Structural Materials for Innovative Nuclear Systems (SMINS) was organised in collaboration with the Forschungszentrum Karlsruhe in Germany. The objectives of the workshop were to exchange information on structural materials research issues and to discuss ongoing programmes, both experimental and in the field of advanced modelling. These proceedings include the papers and the poster session materials presented at the workshop, representing the international state of the art in this domain.

Synopsis

The accelerator-driven system (ADS) is one of the viable concepts for transmuting the long-lived isotopes contained in spent nuclear fuel and for this reason has been receiving considerable interest. In turn, attention must be given to the high power proton accelerators whose reliability and performance are key to the functioning of the ADS.

It is in this context that the NEA organised the fifth workshop on the Utilisation and Reliability of High Power Proton Accelerators (HPPA5) which was held on 6-9 May 2007 in Mol, Belgium. The workshop included a special session on the MEGAPIE programme as well as five technical sessions: accelerator programmes and applications; accelerator reliability; spallation target development and coolant technology; subcritical system design and ADS simulations; and ADS experiments and test facilities. These proceedings contain all the technical papers presented at the workshop and will be of particular interest to scientists working on ADS development.

Synopsis

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as current applications.

Recently developed best-estimate computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose.

The present volume is a follow-up to the first two volumes. While the first described the specification of the benchmark, the second presented the results of the first exercise that identified the key parameters and important issues concerning the thermal-hydraulic system modelling of the simulated transient caused by the switching on of a main coolant pump when the other three were in operation. Volume 3 summarises the results for Exercise 2 of the benchmark that identifies the key parameters and important issues concerning the 3-D neutron kinetics modelling of the simulated transient.

These studies are based on an experiment that was conducted by Bulgarian and Russian engineers during the plant-commissioning phase at the VVER-1000 Kozloduy Unit 6. The final volume will soon be published, completing Phase 1 of this study.

Synopsis

This publication presents the conclusions of the work undertaken by Subgroup 25 of the NEA Working Party on International Evaluation Co-operation, which focused on the assessment and improvement of the evaluated decay data sub-libraries in order to obtain more accurate estimations of decay heat. Recommendations have been prepared for total absorption gamma-ray spectroscopy (TAGS) measurements of specific fission product nuclides to be undertaken in close collaboration with experimentalists in Subgroup 25.

Synopsis

Partitioning and transmutation (P&T) has the potential of significantly reducing the radiotoxicity of nuclear waste and thus minimising the amount of it that needs to be stored in deep geological repositories. In order to provide experts with a forum to present and discuss developments in the field of P&T, since 1990 the OECD Nuclear Energy Agency (NEA) has been organising biennial information exchange meetings on actinide and fission product partitioning and transmutation.

These proceedings contain all the technical papers and posters presented at the Ninth Information Exchange Meeting, which was held on 25-29 September 2006 in Nîmes, France. The meeting covered such issues as progress in fuels and targets, partitioning and waste forms, spallation targets, dedicated transmutation systems, coolants, and physics and nuclear data. In addition, the integration of P&T programmes within different fuel cycle strategies was discussed, as well as the potential transmutation of waste in Generation IV reactors. The implications for waste management strategies, in particular for geological disposal, were also explored. More than 100 papers were presented during the meeting.

Synopsis

This volume provides a state-of-the-art report on the modelling of aqueous-solid solution systems by the combined use of chemical thermodynamics and experimental and computational techniques. These systems are ubiquitous in nature and therefore intrinsic to the understanding and quantification of radionuclide containment and retardation processes present in geological repositories of radioactive waste. Representative cases for study have been chosen from the radioactive waste literature to illustrate the application of the various approaches.

This report has been prepared by a team of four leading experts in the field under the auspices of the OECD/NEA Thermochemical Database (TDB) Project. The team comprised Jordi Bruno (Enviros, Spain), Dirk Bosbach (FZK, Germany), Dmitrii Kulik (PSI, Switzerland) and Alexandra Navrotsky (UC Davis, USA).

Synopsis

The NEA has studied multiple recycling issues associated with various reactor systems fuelled with mixed-oxide (MOX) and published a series of computational physics benchmarks. This has led to improvements in the nuclear data libraries and calculation methods. Several benchmarks were completed comparing those findings with data from experiments. Previous benchmarks have concentrated mainly on PWRs, BWRs, VVER-1000s and FRs. The present benchmark concerns a pebble bed modular reactor (PBMR) fuelled with reactor-grade plutonium.

Although the benchmark has been specifically designed to provide intercomparisons for plutonium and thorium fuels, phases of calculations for uranium fuel have also been included. The purpose of these phases is to identify any increased uncertainties, relative to uranium fuel, that are associated with plutonium and thorium fuel.

This report provides an analysis of the twelve sets of results supplied by seven experts from five countries. Participants have used nuclear data from three different evaluations having applied both Monte Carlo and deterministic methods of analysis. Participants using the same nuclear data report similar results, although some differences have been noted, particularly in relation to the fuel temperature coefficients and the whole-core xenon fission product poisoning effect. There is also evidence of good agreement between Monte Carlo and deterministic solutions for some of the participants despite the difficult nature of the problem with stochastic geometry.

The report will be of interest to reactor physicists and designers.

Synopsis

The plutonium produced during the operation of commercial nuclear power plants or that has become available from the dismantlement of nuclear weapons needs to be properly managed. One important contribution to the management process consists in validating the calculation methods and nuclear data used for the prediction of power in systems burning mixed-oxide (MOX) fuel. Another important contribution is the improved modelling of MOX fuel behaviour in such systems.

Within the framework of the NEA Expert Group on Reactor-based Plutonium Disposition, a fuel modelling code benchmark test for MOX fuel was initiated, with in-pile irradiation data on two short MOX rods provided by the OECD/NEA Halden Reactor Project. This report summarises the in-pile data and fuel characteristics, and presents the calculation results provided by the contributors.

Synopsis

The NEA has studied multiple recycling issues associated with various reactor systems fuelled with mixed-oxide (MOX) and published a series of computational physics benchmarks. This has led to improvements in the nuclear data libraries and calculation methods. Several benchmarks were completed comparing those findings with data from experiments. Previous benchmarks have concentrated mainly on PWRs, BWRs, VVER-1000s and FRs.

The present report provides an evaluation and analysis of the reactor period measurements carried out in the CROCUS reactor of the École polytechnique fédérale de Lausanne (EPFL) for several different delayed super-critical conditions. Two types of reactivity changes were measured employing an appropriate stable period technique in each case. The first series of experiments involved increasing the water level above the critical level. The second series was carried out by inserting/removing one of the absorber rods into/out of the core. The report also provides a benchmark model and the results obtained with different computer codes.

The report will be of interest to reactor physicists and designers.

Synopsis

As part of the development of advanced nuclear systems, including accelerator-driven systems (ADS) proposed for high-level radioactive waste transmutation and generation IV reactors, heavy liquid metals such as lead (Pb) or lead-bismuth eutectic (LBE) are under evaluation as reactor core coolant and ADS neutron target material. Heavy liquid metals are also being envisaged as target materials for high-power neutron spallation sources. The objective of this handbook is to collate and publish properties and experimental results on Pb and LBE in a consistent format in order to provide designers with a single source of qualified properties and data and to guide subsequent development efforts. The handbook covers liquid Pb and LBE properties, materials compatibility and testing issues, key aspects of the thermal-hydraulics and system technologies, existing test facilities, open issues and perspectives.

Synopsis

This workshop was the fourth in a series devoted to the application of synchrotron radiation techniques for studying actinide species. The unique properties of synchrotron radiation allow the elucidation of the molecular and electronic structure of radionuclide samples. Since 2004 when the previous workshop was held, worldwide experimental capabilities for carrying out such studies have expanded. Synergy is developing with advanced theoretical and simulation tools, and it is expected that this progress will contribute significantly to developments in areas such as radioactive waste management, site environmental remediation and separation technologies, as well as in the radiopharmaceutical industry.

The Actinide-XAS-2006 workshop brought together experts in solution, co-ordination and solid state chemistry of the actinides, actinide physics and environmental and life sciences. Workshop sessions were organised on cutting-edge experimental techniques, theoretical and modelling tools and reports on experimental facilities. These proceedings contain abstracts and peer-reviewed papers for 24 presentations as well as 33 poster session contributions, representing the current state of the art in speciation techniques and facilities for radioactive materials at synchrotron light sources.

Synopsis

Computational benchmarks based on well-defined problems with a complete set of input and a unique solution are often used as a means of verifying the reliability of numerical solutions. The problems usually employ some simplifications in order to make the analysis manageable and to enable the consistent comparison of several different models, yet complex enough to make the problem applicable to actual reactor core designs.

The present benchmark has been designed to provide the framework to assess the ability of modern reactor kinetic codes to predict the transient response of a core partially loaded with mixed-oxide (MOX) fuel. It is a follow-up to a pressurised water reactor (PWR) benchmark designed to assess the ability of spatial kinetics codes to model rod ejection transients in a core with uranium-dioxide (UO2) fuel. The current problem adds the complexity of modelling a rod eject in a core fuelled partially with weapons-grade MOX. The core chosen for the simulation is based on a four-loop Westinghouse PWR power plant similar to the reactor chosen for plutonium disposition in the United States.

This report provides an analysis of the results supplied by experts. The report will be of interest to reactor physicists and designers as well as to nuclear power plant utilities.

Synopsis

Access to accurate and reliable information is of prime importance in all nuclear energy applications. This is especially true in the area of nuclear criticality safety for the front- and back-end of the fuel cycle, including transport and storage of spent fuel. The data needed in this area comprises reference values for minimum critical mass, concentration and geometry, as well as the maximum critical moderation of well-defined systems. The accuracy of such data influences both the safety and economy of the fuel cycle.

In 1999, the NEA Working Party on Nuclear Criticality Safety (WPNCS) established an expert group to study the status of nuclear criticality safety reference values (minimum and maximum critical values), following the detection of large deviations in existing reference values between different criticality safety handbooks and guidelines.

The present report represents the outcome of the NEA study and contains a compilation and evaluation of nuclear criticality safety reference values from various sources. Some of the values were taken from published reports, while others were calculated specifically for this study. Many discrepancies have been identified and resolved, thus reinforcing the importance of data verification and validation as essential tools in this field.

Synopsis

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current applications. Recently developed “best-estimate” computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose.

The present volume is the third in a series of four and summarises the results of the second benchmark exercise, which identifies the key parameters and important issues concerning the coupled neutronics/thermal-hydraulic core modelling with provided core inlet and outlet boundary conditions. The transient addressed is a turbine trip in a boiling water reactor, involving pressurisation events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable.

Synopsis

The objective of the Phase II-D Burn-up Credit Criticality Benchmark was to study the impact of control rod (CR) insertion on spent fuel composition and on reactivity for a PWR-UO2 assembly. For this purpose, a range of CR insertion profiles during irradiation were defined, and participants were asked to calculate the spent fuel inventory and the neutron multiplication factor for each case. To assist in the evaluation of the benchmark results, the sensitivity of the neutron multiplication factor to a variation of isotope concentration was performed.

The large effect of CR insertion (9 000 pcm when the CRs are inserted from 0 to 45 GWd/t) is due in part to the fact that the CRs are axially fully inserted in this benchmark. A more “typical” CR insertion profile would not consider CRs fully inserted throughout the irradiation, particularly over three cycles. An additional benchmark has been initiated to study the effect of CR insertion when considering partial axial CR insertion and an axial burn-up profile.

Synopsis

With a declining number of nuclear data evaluators in the world and an increasing demand for high-quality data, there is a risk that evaluators will concentrate on producing new nuclear data to the detriment of developing new models and methods for evaluating existing data. In this context, it is essential to identify the basic physics issues that are going to be important for future nuclear data evaluation processes. At the same time, demand for new types of data, which will be needed in emerging nuclear applications, could warrant new evaluation techniques that are presently only used in the context of fundamental research and not in nuclear data production.

These proceedings present the main findings of the "Perspectives on Nuclear Data for the Next Decade" workshop, which explored innovative approaches to nuclear data evaluation with the aim of opening new perspectives, building new research programmes and investigating prospects for international collaboration.

Synopsis

Subgroup 7 of the NEA Working Party on International Nuclear Data Co-operation (WPEC) was established to re-evaluate the nuclear data standards cross-section. These cross-section data are the basis for the evaluated nuclear data libraries, as most of the underlying experimental data are measured relative to these standards. The incentive to undertake this re-evaluation work was based on the fact that significant improvements to the experimental database have been made since the standard data were last evaluated some 20 years ago.

The work of the subgroup was performed in close collaboration with an IAEA Co ordinated Research Project (CRP) and a task force of the US Cross-section Evaluation Working Group (CSEWG). This report provides a brief overview of the work accomplished, outlining the main findings and providing a full list of references. A more extensive report will be issued by the above-mentioned IAEA Co-ordinated Research Project.

Synopsis

The safe and economical operation of nuclear energy technologies requires detailed and reliable calculations. While simulation calculations are becoming more and more economical thanks to rapid advances in computer technology, the accuracy of these calculations is largely determined by the accuracy of the atomic and nuclear input data.

The Joint Evaluated Fission and Fusion (JEFF) Project is a collaborative effort among NEA Data Bank member countries to produce high-quality evaluated nuclear data. These data can be used to help improve the safety and economics of existing installations as well as to design advanced nuclear reactors and their associated fuel cycles. Such data may also be useful in the area of radioactive waste management.

The present report describes the contents of the general purpose file of the JEFF-3.1 data library. The library contains a number of different data types, including neutron and proton interaction data, radioactive decay data, fission yield data, thermal scattering law data and photo-atomic interaction data.

Synopsis

This NEA workshop is the third in a series devoted to the application of synchrotron accelerator-based techniques to radionuclide and actinide sciences. As synchrotron radiation is particularly well-suited for obtaining information about the molecular structure of radionuclides and actinide species, it is useful for understanding and predicting the behaviour of these hazardous elements in the environment. Application areas include risk assessment of nuclear waste storage, remediation of contaminated sites, development of effective separation technologies and radiopharmaceutical chemistry.

These proceedings contain all of the abstracts and some of the full papers presented at the workshop. In addition to presenting the latest experimental and theoretical results, the workshop also provided opportunities for knowledge transfer between established experts in the field and young scientists.

Synopsis

The NEA Working Party on Nuclear Criticality Safety established an Expert Group on Source Convergence in Criticality Safety Analysis to explore the problems of slow convergence and statistical fluctuations that can combine to produce unreliable source distributions and fission rates as well as underestimates of keff and its uncertainty. Aimed at fostering improved robustness of criticality safety analyses with respect to source convergence, the group's first task was to assemble four test problems that represent cases previously encountered in criticality safety analyses. They are intended to be used as a basis for comparison of source convergence performance rather than comparison of physics results. The problems include a reactor fuel storage array, a spent fuel pin array, an aqueous processing system and an array of small fissile components. The results of the four test problems are described herein.

Synopsis

Average fuel burn-up in light water reactors (LWRs) has steadily increased with time as technological advances have been made. The practical limit is currently in the region of 50 GWd/t. The main driving forces behind this increase have been to reduce the cost of the nuclear fuel cycle and to benefit from the increased operational flexibility that high burn-ups allow. One of the main questions at this stage is whether this historic trend will continue, or whether there are scientific and technological limits to current LWR fuel burn-ups.

This publication investigates the limitations and potential benefits of very high fuel burn-up (60-100 GWd/t) in light water reactors. It covers technical aspects, such as fuel fabrication, thermal-hydraulic design limits and fuel performance, as well as economic aspects. The report provides several recommendations regarding scientific and technological areas in which further development is required to achieve these very high burn-ups.

Synopsis

This document serves as a summary of the work of Subgroup 20 (SG20) on covariance matrix evaluation and processing in the resolved/unresolved resonance regions, organised under the auspices of the NEA's Nuclear Science Committee Working Party on International Evaluation Co-operation (WPEC).

The work described in this report focuses on: summarising the issues related to covariance evaluation in the resonance region; discussing the retroactive method used in the SAMMY code; describing the compact format for storing huge covariance matrices in ENDF-6 files; recent developments and upgrades of processing codes to generate a multi-group covariance matrix from resonance parameter covariance data.

Synopsis

Hydrogen has the potential to play an important role as a sustainable and environmentally acceptable energy carrier in the 21st century. Since natural sources of pure hydrogen are extremely limited, it is necessary to develop technologies to produce large quantities of hydrogen economically. The currently dominant technology for producing hydrogen is based on reforming fossil fuels, a process which releases greenhouse gases. Hydrogen produced by water cracking, using heat and surplus electricity from nuclear power plants, requires no fossil fuels and results in lower greenhouse gas emissions. This report presents the state of the art in the nuclear production of hydrogen and describes its associated scientific and technical challenges.

Synopsis

The plutonium produced during the operation of commercial nuclear power plants or that has become available from the dismantlement of nuclear weapons needs to be properly managed. One important contribution to the management process consists in validating the calculation methods and nuclear data used for the prediction of power in systems using mixed-oxide (MOX) fuel. A series of computational physics benchmarks and issues regarding multiple recycling in various MOX-fuelled systems have been studied and published by the NEA. This has led to improvements in the nuclear data libraries and calculation methods. Full validation requires comparing those findings with data from experiments. The experiment at the KRITZ research reactor in Sweden is being used for this purpose.

This report provides an analysis of the 12 sets of results supplied by 16 experts from 7 countries, together with the comparison against the KRITZ evaluated experimental data. The report concludes that the computer codes and cross-sections used by the participants, which are presently in widespread use, can adequately predict the multiplication factor and pin-power distributions of the MOX cores.

This report will be of particular interest to reactor physicists and designers as well as to nuclear power plant utilities.

Synopsis

Refined models for best-estimate calculations based on good-quality experimental data can improve the understanding of phenomena and the quantification of margins for operating nuclear power reactors. According to experts, refinements should not be limited to currently available macroscopic approaches but should be extended to next-generation approaches that focus on more microscopic processes. Multi-scale/multi-physics approaches are the way forward in this respect.

This report describes the specification of an international benchmark based on high-quality fine mesh data, released through the government of Japan and the Nuclear Power Engineering Corporation (NUPEC), with the aim of advancing the insufficiently developed field of two-phase flow theory. It has been designed for systematically assessing and comparing different numerical models used for predicting detailed void distributions and critical powers.

Additional volumes concerning this benchmark are planned and are intended to show to what extent the most recent approaches are capable of predicting two-phase flow phenomena.

Synopsis

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as current applications.

Recently developed best-estimate computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose.

The present volume, a follow-up to the first volume describing the specification of the benchmark, presents the results of the first exercise that identifies the key parameters and important issues concerning the thermal-hydraulic system modelling of the simulated transient. This exercise aims to achieve the correct initialisation and testing of the system code models. The transient chosen for the exercise is caused by the switching on of a main coolant pump while the other three are in operation. It is based on an experiment that was conducted by Bulgarian and Russian engineers during the plant-commissioning phase at the VVER-1000 Kozloduy Unit 6.

Synopsis

Radiation is used in many applications of modern technology. However, its proper handling requires competent knowledge of the basic physical laws governing its interaction with matter. To ensure its safe use, appropriate tools for predicting radiation fields and doses, as well as pertinent regulations, are required.

One area of radiation physics that has received much attention concerns electron-photon transport in matter. PENELOPE is a modern, general-purpose Monte Carlo tool for simulating the transport of electrons and photons, which is applicable for arbitrary materials and in a wide energy range. PENELOPE provides quantitative guidance for many practical situations and techniques, including electron and X-ray spectroscopies, electron microscopy and microanalysis, biophysics, dosimetry, medical diagnostics and radiotherapy, and radiation damage and shielding.

These proceedings contain the extensively revised teaching notes of the latest workshop/training course on PENELOPE (version 2006), along with a detailed description of the improved physics models, numerical algorithms and structure of the code system.

Synopsis

The efficient and safe management of spent fuel arising from the operation of commercial nuclear power plants is an important issue. In this context, the partitioning and transmutation (P&T) of minor actinides and long-lived fission products can play an important role, reducing significantly the burden on geological repositories of radioactive waste and enabling their more effective use.

International interest in accelerator-driven systems (ADS) has been expressed due to their potential use in the transmutation of minor actinides. However, much R&D work is still required in order to demonstrate the desired capability of the system as a whole, and the current methods of analysis and nuclear data for minor actinide burners are not as well established as those for conventionally fuelled systems.

A series of theoretical ADS physics benchmarks has thus been organised by the NEA. Many improvements and clarifications in nuclear data and calculation methods have been achieved. However, following an initial series of benchmarks, some significant discrepancies in important parameters were not fully understood and still required clarification. Hence, the first experiment-based benchmark using MASURCA critical and subcritical experiments (called MUSE-4 experiments) was launched.

This report provides an analysis of the benchmark results supplied by 16 institutions from 14 countries. The calculated results were compared against experimental data, whenever available. This report will be of particular interest to reactor physicists and nuclear data evaluators developing nuclear systems, especially ADS, for radioactive waste management.

Synopsis

It is essential to calculate the structural integrity of reactor components with a high degree of accuracy in order to make correct decisions regarding plant lifetime at the design stage, safety margins and potential plant life extensions. The OECD Nuclear Energy Agency (NEA) is therefore organising a series of benchmarks to verify the current international level of accuracy in pressure vessel fluence calculations and to clarify the relative merits of various methodologies. By extension, this enables the identification of areas for possible improvements in the various calculation schemes.

As a follow-up to the previous UO2-fuelled VENUS-1 two-dimensional (2-D) and VENUS-3 three-dimensional (3-D) benchmarks, and given that many commercial nuclear power plants in Europe and in Japan use MOX fuel and that the use of MOX fuel in LWRs presents different neutron characteristics, the present benchmark was launched in 2004 using the measured data of the VENUS-2 MOX-fuelled critical experiments. This report provides an analysis of the results supplied by 12 participants from 7 countries. The results have revealed that the computer codes and nuclear data currently used for MOX-fuelled systems in OECD/NEA member countries appear able to produce results with a sufficiently high level of accuracy in dosimetry calculations. This report will be of particular interest not only to reactor physicists and nuclear data evaluators, but also to nuclear utilities.

Synopsis

The safe and efficient management of spent fuel from the operation of commercial nuclear power plants is an important issue. Worldwide, more than 250 000 tons of spent fuel from currently operating reactors will require disposal. These numbers account for only high-level radioactive waste generated by present-day power reactors.

Nearly all issues related to risks to future generations arising from the long-term disposal of such spent nuclear fuel is attributable to only about 1% of its content. This 1% is made up primarily of plutonium, neptunium, americium and curium (called transuranic elements) and the long-lived isotopes of iodine and technetium.When transuranics are removed from discharged fuel destined for disposal, the toxic nature of the spent fuel drops below that of natural uranium ore (that which was originally mined for the nuclear fuel) within a period of several hundred to a thousand years. This significantly reduces the burden on geological repositories and the problem of addressing the remaining long-term residues can thus de done in controlled environments having timescales of centuries rather than millennia stretching beyond 10 000 years.

Transmutation is one of the means being explored to address the disposal of transuranic elements. To achieve this, advanced reactors systems, appropriate fuels, separation techniques and associated fuel cycle strategies are required.

This status report begins by providing a clear definition of partitioning and transmutation (P&T), and then describes the state of the art concerning the challenges facing the implementation of P&T, scenario studies and specific issues related to accelerator-driven systems (ADS) dynamics and safety, long-lived fission product transmutation and the impact of nuclear data uncertainty on transmuation system design. The report will be of particular interest to nuclear scientists working on P&T issues as well as advanced fuel cycles in general.

Synopsis

The United States and the Russian Federation have each agreed to dispose of 34 tonnes of weapons-grade plutonium that are beyond their defence needs. One effective way to dispose of this plutonium is to convert it into mixed-oxide (MOX) fuel, burn it in a nuclear reactor and use it to produce electricity.

This report describes an international benchmark study that compared the results obtained for six different states in a VVER-1000 reactor core loaded with one-third MOX fuel. This NEA activity contributes to the computer code certification process and to the verification of calculation methods used in the Russian Federation.

Synopsis

Accelerator-driven systems (ADS) are being considered for their potential use in the transmutation of radioactive waste. The performance of such hybrid nuclear systems depends to a large extent on the specification and reliability of high power accelerators, as well as the integraton of the accelerator with spallation targets and sub-critical systems. At present, much R&D work is still required in order to demonstrate the desired capability of the system as a whole.

Accelerator scientists and reactor physicists from around the world gathered at an NEA workshop to discuss issues of common interest and to present the most recent achievements in their research. Discussions focused on accelerator reliability; target, window and coolant technology; sub-critical system design and ADS simulatons; safety and control of ADS; and ADS experiments and test facilities. These proceedings contain the technical papers presented at the workshop as well as summaries of the working group discussions held. They will be of particular interest to scientists working on ADS development as well as on radioactive waste management issues in general.

Synopsis

An important issue regarding deterministic transport methods for whole core calculations is that homogenised techniques can introduce errors into results. In addition, with modern computational abilities, direct whole core heterogeneous calculations are becoming increasingly feasible.

Following a previous benchmark in this series in 2003, this 3-D extension case was designed to simulate three core configurations with different levels of axial heterogeneity utilising control rods. A majority of the participants obtained solutions that were more than acceptable for typical nuclear reactor calculations, showing that modern deterministic transport codes and methods can calculate the flux distribution reasonably well without relying upon special homogenisation techniques. The report will be of particular interest to reactor physicists and transport code developers.

Synopsis

The safe and efficient management of spent fuel from the operation of commercial nuclear power plants is an important issue. Worldwide, more than 250 000 tons of spent fuel from reactors currently operating will require disposal. These numbers account for only high-level radioactive waste generated by present-day power reactors.

Nearly all issues related to risks to future generations arising from the long-term disposal of such spent nuclear fuel is attributable to only about 1% of its content. This 1% is made up primarily of plutonium, neptunium, americium and curium (called transuranic elements) and the long-lived isotopes of iodine and technetium. When transuranics are removed from discharged fuel destined for disposal, the toxic nature of the spent fuel drops below that of natural uranium ore (that which was originally mined for the nuclear fuel) within a period of several hundred to a thousand years. This significantly reduces the burden on geological repositories and the problem of addressing the remaining long-term residues can thus be done in controlled environments having timescales of centuries rather than millennia stretching beyond 10 000 years.

Transmutation is one of the means being explored to address the disposal of transuranic elements. To achieve this, advanced reactor systems, appropriate fuels, separation techniques and associated fuel cycle strategies are required.

This report describes the current status of fuel and material technologies for transmutation and suggests technical R&D issues that need to be resolved. It will be of particular interest to nuclear fuel and material scientists involved in the field of partitioning and transmutation (P&T), and in advanced fuel cycles in general.

Synopsis

This report communicates the results of an international seminar which reviewed recent progress in the field of pellet-clad interaction in light water reactor fuels. It also draws a comprehensive picture of current understanding of relevant phenomena and their impact on the nuclear fuel rod, under the widest possible conditions. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels.

Synopsis

Subgroup 21 of the NEA Nuclear Science Committee Working Party on International Evaluation Co-operation was charged with the task of assessing neutron cross-section evaluations for fission products. The undertaking of the taskgroup was considerable: the review and assessment of neutron-induced cross-sections in all major evaluated nuclear data libraries. As a result, the subgroup provided recommendations for the best evaluations for 218 fission products, as set out in this report.

Synopsis

Athough the fabrication of mixed-oxide (MOX) fuel is well-established with appropriate safety margins, it would still be beneficial to optimise the process by further investigating and possibly reducing these margins. It is also important to demonstrate that all operations involving plutonium and MOX fuels adhere to strict safety standards, and that these standards are based upon the most reliable tools and data.

An NEA workshop, organised in April 2004, confirmed that even though existing unpublished experiments could partially address the need for more accurate experimental data, the need for additional experiments remained. An ad hoc expert group was therefore established to define a framework and method for the selection and performance of new experimental programme(s) of interest. The present publication describes the selection criteria and methodology that were used to compare experimental proposals and makes recommendations on which experimental programme(s) should be pursued.

Synopsis

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed "best-estimate" computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose.

The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermal-hydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurisation events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable.

Synopsis

Particle accelerators are used today for an increasing range of scientific and technological applications. They are very powerful tools to investigate the origin and structure of matter, and to improve understanding of the interaction of radiation with materials, including transmutation of nuclides and beneficial effects of risks from radiation. They are used to identify properties of molecules that can be used in pharmacy, for medical diagnosis and therapy, or for biophysics studies.

Particle accelerators must be operated in safe ways that protect operators, the population and the environment. New technological and research applications give rise to new aspects in radiation shielding. These workshop proceedings review the state of the art in radiation shielding of accelerator facilities and of irradiated targets. They also evaluate progress made and discuss the additional developments required to meet radiation protection needs.

Synopsis

The efficient and safe management of spent fuel produced during the operation of commercial nuclear power plants is an important issue. Worldwide, more than 250 000 tons of spent fuel from reactors currently operating will require disposal. These numbers account for only high-level radioactive waste generated by present-day power reactors.

Nearly all issues related to risks to future generations arising from the long-term disposal of such spent nuclear fuel is attributable to only about 1% of its content. This 1% is made up primarily of plutonium, neptunium, americium and curium (called transuranic elements) and the long-lived isotopes of iodine and technetium. When transuranics are removed from discharged fuel destined for disposal, the toxic nature of the spent fuel drops below that of natural uranium ore (that which was originally mined for the nuclear fuel) within a period of several hundred years. This significantly reduces the burden on geological repositories and the problem of addressing the remaining long-term residues can thus be done in controlled environments having timescales of centuries rather than millennia.

To address the disposal of transuranics, accelerator-driven systems (ADS), i.e. a sub-critical system driven by an accelerator to sustain the chain reaction, seem to have great potential for transuranic transmutation, though much R&D work is still required in order to demonstrate their desired capability as a whole system.

This report describes the current status of accelerator and spallation target technologies and suggests technical issues that need to be resolved for ADS applications. It will be of particular interest to nuclear scientists involved in ADS development and in advanced fuel cycles in general.

Synopsis

To master the technology and the economics of nuclear energy, deep insight is needed into the physical and chemical phenomena at work in nuclear reactors and all parts of the associated fuel cycle. Scientific knowledge should be constantly updated in order to:

- improve the safety and the economics of existing installations and anticipate possible problems;

- optimise the design of future installations;

- develop satisfactory techniques for radioactive waste storage and disposal.

One of the most important basic tools needed for accomplishing the above is accurate nuclear data. NEA Data Bank member countries have long supported the development of the Joint Evaluated Fission and Fusion (JEFF) library, which is used as reference data for nuclear applications in many European countries.

The third, improved version of the data library (JEFF-3.0) was recently issued. The present report describes the contents of this library.

This publication is also available in English as: Research and Test Facilities Required in Nuclear Science and Technology -

Synopsis

Les installations expérimentales sont des outils de recherche indispensables au développement des sciences et technologies nucléaires et à l’expérimentation des systèmes et matériaux utilisés actuellement ou qui le seront à l’avenir. Compte tenu des pressions économiques et de la fermeture des installations anciennes, il est à craindre que la capacité à entreprendre les recherches nécessaires au maintien et au développement de la science et de la technologie nucléaires ne soit menacée.

Un groupe d’experts de l’AEN comprenant des représentants de dix pays membres, de l’Agence internationale de l’énergie atomique et de la Commission européenne a examiné la situation des installations de recherche et d’expérimentation opérant dans les domaines d’intérêt du Comité des sciences nucléaires de l’AEN, à savoir : la mesure des données nucléaires, le développement des réacteurs, la diffusion de neutrons, la neutronographie, les systèmes hybrides, la transmutation, le combustible nucléaire, les matériaux, la sûreté, la radiochimie, la séparation et l’utilisation de la chaleur des réacteurs nucléaires pour la production d’hydrogène.

Ce rapport contient l’évaluation détaillée du groupe d’experts sur la situation actuelle des installations de recherche nucléaire, ainsi que des recommandations sur la façon de garantir le développement de ce secteur grâce à la mise en place d’installations modernes de qualité. Il décrit aussi la base de données en ligne établie par le groupe d’experts, qui recense plus de 700 installations.

This publication is also available in English as: Research and Development Needs for Current and Future Nuclear Energy Systems -

Synopsis

Des capacités de recherche et l'expertise technique dans le domaine des sciences nucléaires sont nécessaires pour maintenir un niveau élevé de performance et de sûreté des installations nucléaires actuelles ainsi que pour développer les programmes électronucléaires de la prochaine génération.

Le Comité des sciences nucléaires (CSN) de l'AEN a réalisé une étude sur les futurs besoins de recherche et développement dans des domaines spécifiques des sciences nucléaires : les données nucléaires, la physique des réacteurs et le comportement des systèmes, et enfin les combustibles, matériaux et caloporteurs des réacteurs.

Ce rapport comporte des informations sur les activités internationales de R-D passées et actuelles, réalisées sous l'égide du CSN, et sur les besoins en R-D des nouveaux systèmes nucléaires dans les différents pays membres de l'AEN. Des recommandations sur les travaux à effectuer dans les domaines mentionnés ci-dessus sont également présentées. Les éventuelles actions de suivi à ces recommandations seront étudiées par le CSN.

Synopsis

原子力の科学技術の発展及び、現在または将来用いられるシステムや材料の評価のためには、それらの研究に利用可能な試験研究施設が必要不可欠である。しかしながら、経済的な負担や老朽化施設の閉鎖等により、原子力の科学技術の維持、発展に必要な研究を行う能力が現在、危機にさらされている。

OECD/NEA加盟国のうちの十ヵ国と国際原子力機関(IAEA)、欧州委員会(EC)からの代表者によって組織されたNEA専門家会合は、NEA原子力科学委員会(NSC)の所掌範囲に関係する試験研究施設の状況についてレビューを行った。対象とした試験研究施設は、核データ測定、原子炉開発、中性子散乱、中性子ラジオグラフィ、ADS、核変換、燃料、材料、安全性、放射化学、分離技術、水素製造のための核熱利用等に関係するものである。

このレポートにはNEA専門家会合による試験研究施設の現状に関する評価の詳細が述べられており、また高性能で近代的な施設の整備により、将来の発展がいかに担保できるかについて勧告がなされている。さらに専門家会合が整備した、約800の試験研究施設の情報を掲載したオンライン・データベースについても解説がなされている。

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