Nuclear Science and Data Bank Publications


Alphabetical list of titles
3-D Deterministic Radiation Transport Computer Codes (1997)
Features, Applications and Perspectives - Proceedings of an OECD Meeting held on 2-3 December 1996, OECD Château la Muette, Paris, France
3-D Deterministic Radiation Transport Computer Programs (1997)
Features, Applications and Perspectives
Actinide and Fission Product Partitioning and Transmutation (2007)
Ninth Information Exchange Meeting, Nîmes, France, 25-29 September 2006
Actinide and Fission Product Partitioning and Transmutation (2012)
Eleventh Information Exchange Meeting, San Francisco, California, USA, 1-4 November 2010
Actinide and Fission Product Partitioning and Transmutation (2003)
Seventh Information Exchange Meeting, Jeju, Republic of Korea, 14-16 October 2002
Actinide and Fission Product Partitioning and Transmutation + CD-ROM (2010)
Tenth Information Exchange Meeting, Mito, Japan, 6-10 October 2008
Advanced Reactors with Innovative Fuels (2002)
Workshop Proceedings, Chester, United Kingdom, 22-24 October 2001
Advanced Reactors with Innovative Fuels (1999)
Workshop Proceedings, Villigen, Switzerland, 21-23 October 1998
Analytical Benchmarks for Nuclear Engineering Applications (2008)
Case Studies in Neutron Transport Theory
Assessment of Fission Product Decay Data for Decay Heat Calculations (2007)
International Evaluation Co-operation, Volume 25
Basic Studies in the Field of High-temperature Engineering (2004)
Third Information Exchange Meeting, Ibaraki-ken, Japan, 11-12 September 2003
Basic Studies on High-Temperature Engineering (2000)
First Information Exchange Meeting, Paris, France, 27-29 September 1999
Basic Studies on High-temperature Engineering (2002)
Second Information Exchange Meeting, Paris, France, 10-12 October 2001
Benchmark Calculations of Power Distribution Within Fuel Assemblies (2000)
Phase II: Comparison of Data Reduction and Power Reconstruction Methods in Production Codes
Benchmark on Beam Interruptions in an Accelerator-driven System (2004)
Final Report on Phase II Calculations
Benchmark on Beam Interruptions in an Accelerator-driven system (2003)
Final Report on Phase I Calculations
Boiling Water Reactor Turbine Trip (TT) Benchmark - Vol. IV (2010)
Volume IV: Summary Results of Exercise 3
Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume II (2005)
Volume II: Summary Results of Exercise 1
Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume III (2006)
Volume III: Summary Results of Exercise 2
Burn-up Credit Criticality Benchmark - Phase II-C (2007)
Phase II-C: Impact of the Asymmetry of PWR Axial Burn-up Profiles on the End Effect
Burn-up Credit Criticality Benchmark - Phase II-D (2006)
PWR-UO2 Assembly - Study of Control Rod Effects on Spent Fuel Composition
Burn-up Credit Criticality Benchmark - Phase IV-A (2003)
Phase IV-A: Reactivity Prediction Calculations for Infinite Arrays of PWR MOX Fuel Pin Cells
Burn-up Credit Criticality Benchmark - Phase IV-B (2003)
Phase IV-B: Results and Analysis of MOX Fuel Depletion Calculations
Burn-up Credit Criticality Safety Benchmark – Phase VII (2012)
UO2 Fuel: Study of Spent Fuel Compositions for Long-term Disposal
CD-CINDA 2000 (2000)
Index to Literature and Computer Files on Microscopic Neutron Data
CINDA 2003 (2003)
The Index to Literature and Computer Files on Microscopic Neutron Data
Calculations of Different Transmutation Concepts (2000)
An International Benchmark Exercise
Chemical Thermodynamics of Tin (2012)
Chemical Thermodynamics Volume 12
Computing Radiation Dosimetry - CRD 2002 (2004)
Workshop Proceedings, Sacavém, Portugal, 22-23 June 2002
Core Monitoring for Commercial Reactors: Improvements in Systems and Methods (2000)
Workshop Proceedings, Stockholm, Sweden, 4-5 October 1999
Evaluated Data Library for the Bulk of Fission Products (Volume 23) (2009)
International Evaluation Co-operation, Volume 23
Evaluation and Processing of Covariance Data (1993)
Proceedings of a Specialists' Meeting Oak Ridge National Laboratory, U.S.A., 7-9 October 1992
Evaluation of Speciation Technology (2001)
Workshop Proceedings, Tokai-mura, Ibaraki, Japan, 26-28 October 1999
Fission Gas Behaviour in Water Reactor Fuels (2002)
Workshop Proceedings, Cadarache, France, 26-29 September 2000
Fission Product Nuclear Data (1992)
Proceedings of a Specialists' Meeting Tokai, Japan, May 1992
Forsmark 1 & 2 Boiling Water Reactor Stability Benchmark (2001)
Time Series Analysis Methods for Oscillations During BWR Operation: Final Report
In-Core Instrumentation and Reactor Core Assessment (1997)
Proceedings of a Specialist Meeting, Mito-shi, Japan, 16-17 October 1996
Independent Evaluation of the MYRRHA Project (2009)
Report by an International Team of Experts
Inter-code Comparison Exercise for Criticality Excursion Analysis (2009)
Benchmarks Phase I: Pulse Mode Experiments with Uranyl Nitrate Solution Using the TRACY and SILENE Experimental Facilities
International Evaluation Co-operation (2000)
Processing and Validation of Intermediate Energy Evaluated Data Files (Volume 14)
International Evaluation Co-operation (1996)
Comparison of Evaluated Data for Chromium-52, Iron-56 and Nickel-58 [Volume 1]
International Evaluation Co-operation (2002)
Delayed Neutron Data for the Major Actinides (Volume 6)
International Evaluation Co-operation (1998)
Nuclear Models to 200 MeV for High-Energy Data Evaluations [Volume 12]
International Evaluation Co-operation (1996)
Cross-section Fluctuations and Self-shielding Effects in the Unresolved Resonance Region [Volume 15]
International Evaluation Co-operation (1996)
Plutonium-239 Fission Cross-section Between 1 and 100 keV [Volume 5]
International Evaluation Co-operation (1996)
Actinide Data in the Thermal Energy Range [Volume 3]
International Evaluation Co-operation (1996)
Generation of Covariance Files for Iron-56 and Natural Iron [Volume 2]
International Evaluation Co-operation (1998)
Status of Pseudo-Fission-Product Cross-Sections for Fast Reactors [Volume 17]
International Evaluation Co-operation (1998)
Effects of Shape Differences in the Level Densities of Three Formalisms on Calculated Cross-Sections [Volume 16]
International Evaluation Co-operation (1998)
Intermediate Energy Data [Volume 13]
International Evaluation Co-operation (2001)
Evaluation Method of Inelastic Scattering Cross-sections for Weakly Absorbing Fission-product Nuclides (Volume 10)
International Evaluation Co-operation (1999)
238U Capture and Inelastic Cross-Sections (Volume 4)
International Evaluation Co-operation (1999)
Epithermal Capture Cross-Section of 235U (Volume 18)
International Evaluation Co-operation (1999)
Present Status of Minor Actinide Data (Volume 8)
International Evaluation Co-operation (Vol. 19) (2005)
Neutron Activation Cross-section Measurements from Threshold to 20 MeV for the Validation of Nuclear Models and Their Parameters (Volume 19)
International Evaluation Co-operation (Vol. 20) (2006)
Covariance Matrix Evaluation and Processing in the Resolved/Unresolved Resonance Regions (Volume 20)
International Evaluation Co-operation (Vol. 21) (2005)
Assessment of Neutron Cross-section Evaluations for the Bulk of Fission Products (Volume 21)
International Evaluation Co-operation (Vol. 22) (2006)
Nuclear Data for Improved LEU-LWR Reactivity Predictions (Volume 22)
International Evaluation Co-operation (Vol. 26) + CD-ROM (2008)
Uncertainty and Target Accuracy Assessment for Innovative Systems Using Recent Covariance Data Evaluations (Volume 26)
International Evaluation Co-operation (Vol. 33) (2013)
Methods and Issues for the Combined Use of Integral Experiments and Covariance Data (Volume 33)
International Evaluation Co-operation (Vol. 7) (2006)
Nuclear Data Standards (Volume 7)
International Evaluation Co-operation (Vol. 9) (2003)
Fission Neutron Spectra of Uranium-235 (Volume 9)
International Handbook of Evaluated Criticality Safety Benchmark Experiments (2001)
A Project by the NEA Nuclear Science Committee
International Nuclear Data Evaluation Co-operation - CD-ROM (2003)
Complete Collection of Published Reports as of October 2003
International Nuclear Data Evaluation Co-operation - CD-ROM (2010)
Complete Collection of Published Reports as of January 2010
Ion and Slow Positron Beam Utilisation (1999)
Workshop Proceedings, Costa da Caparica, Portugal, 15-17 September 1998
JANIS 3 (2010)
A Java-based Nuclear Data Display Program
JEFF 3.1.2 (2012)
Joint Evaluated Nuclear Data Library for Fission and Fusion Applications February 2012
JEFF Reports (2001)
Complete Collection of JEFF Reports - Numbers 1-18
JEFF Reports CD-ROM (2010)
Complete Collection of JEFF Reports 1-22
Janis 3.4 (2012)
A Java-based Nuclear Data Display Program
Long-Lived Radionuclide Chemistry in Nuclear Waste Treatment (1998)
Workshop Proceedings, Villeneuve-lès-Avignon, France, 18-20 June 1997
Minor Actinide Burning in Thermal Reactors (2013)
A Report by the Working Party on Scientific Issues of Reactor Systems
Mixed-oxide (MOX) Fuel Performance Benchmark (2007)
Summary of the Results for the Halden Reactor Project MOX Rods
Mixed-oxide (MOX) Fuel Performance Benchmark (PRIMO) (2009)
Summary of the Results for the PRIMO BD8 MOX Rod
Mobile Fission and Activation Products in Nuclear Waste Disposal (2009)
Workshop Proceedings, La Baule, France, 16-19 January 2007
NEA Nuclear Model and Code Comparisons (2001)
Complete Collection of the Report 1982-1998
NJOY and THEMIS Nuclear Data (1994)
Proceedings of a Seminar Saclay, France, 7-8 April 1992
NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark (Vol. II) (2010)
Volume II: Uncertainty and Sensitivity Analyses of Void Distribution and Critical Power - Specification
Neutron Cross-section Standards for the Energy Region Above 20 MeV (1991)
Proceedings of a Specialists' Meeting, Uppsala, Sweden, May 1991
Neutronics/Thermal-hydraulics Coupling in LWR Technology, Vol. 1 (2004)
CRISSUE-S - WP1: Data Requirements and Databases Needed for Transient Simulations and Qualification - 5th EURATOM Framework Programme (1998-2002)
Neutronics/Thermal-hydraulics Coupling in LWR Technology, Vol. 2 (2004)
CRISSUE-S - WP2: State-of-the-art Report - 5th EURATOM Framework Programme (1998-2002)
Neutronics/Thermal-hydraulics Coupling in LWR Technology, Vol. 3 (2004)
CRISSUE-S - WP3: Achievements and Recommendations Report - 5th EURATOM Framework Programme (1998-2002)
Nuclear Production of Hydrogen (2001)
First Information Exchange Meeting, Paris, France, 2-3 October 2000
Nuclear Production of Hydrogen (2010)
Fourth Information Exchange Meeting, Oakbrook, Illinois, United States, 13-16 April 2009
Nuclear Production of Hydrogen (2006)
Third Information Exchange Meeting, Oarai, Japan, 5-7 October 2005
Nuclear Production of Hydrogen (2004)
Second Information Exchange Meeting, Argonne, Illinois, USA, 2-3 October 2003
Nucleon Nucleus Optical Model up to 200 MeV (1997)
Proceedings of a Specialist Meeting, Bruyères-le-Châtel, France, 13-15 November 1996
PENELOPE 2001 - A Code System for Monte Carlo Simulation of Electron and Photon Transport (2002)
Workshop Proceedings, Issy-les-Moulineaux, France, 5-7 November 2001
PENELOPE 2003 - A Code System for Monte Carlo Simulation of Electron and Photon Transport (2003)
Workshop Proceedings, Issy-les-Moulineaux, France, 7-10 June 2003
PENELOPE-2006: A Code System for Monte Carlo Simulation of Electron and Photon Transport (2006)
Workshop Proceedings, Barcelona, Spain, 4-7 July 2006
PENELOPE-2008: A Code System for Monte Carlo Simulation of Electron and Photon Transport (2009)
Workshop Proceedings, Barcelona, Spain, 30 June-3 July 2008
Pellet-clad Interaction in Water Reactor Fuels (2005)
Seminar Proceedings, Aix-en-Provence, France, 9-11 March 2004
Perspectives on Nuclear Data for the Next Decade (2006)
Workshop Proceedings, Bruyères-le-Châtel, France, 26-28 September 2005
Physics and Fuel Performance of Reactor-Based Plutonium Disposition (1999)
Workshop Proceedings, Paris, France, 28-30 September 1998
Physics of Plutonium Recycling (2002)
Multiple Pu Recycling in Advanced PWRs - Volume VI
Physics of Plutonium Recycling - Vol. VII (2003)
Volume VII: BWR MOX Benchmark - Specification and Results
Physics of Plutonium Recycling - Volume IX (2007)
Volume IX: Benchmark on Kinetic Parameters in the CROCUS Reactor
Physics of Plutonium Recycling - Volume VIII (2007)
Volume VIII: Results of a Benchmark Considering a High-temperature Reactor (HTR) Fuelled with Reactor-grade Plutonium
Plutonium Management in the Medium Term (2003)
A Review by the OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR)
Pressured Water Reactor Main Steam Line Break (MSLB) Benchmark (2003)
Volume IV: Results of Phase III on Coupled Core-plant Transient Modelling
Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark (2000)
Volume II: Results of Phase I on Point Kinetics
Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark (2002)
Volume III: Results of Phase 2 on 3-D Core Boundary Conditions Modelling
Pyrochemical Separations (2001)
Workshop Proceedings, Avignon, France, 14-16 March 2000
SATIF-3 - Shielding Aspects of Accelerators, Targets and Irradiation Facilities (1998)
Proceedings of 3rd Specialists Meeting Sendai, Japan, 12-13 May 1997
SMORN-VII (1996)
Proceedings of the Seventh Symposium on Surveillance and Diagnostics in Nuclear Reactors
Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 4 (1999)
Workshop Proceedings, Knoxville, Tennessee, USA, 17-18 September 1998
Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 5 (2001)
Workshop Proceedings, Paris, France, 18-21 July 2000
Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 6 (2004)
Workshop Proceedings, Stanford, California, USA, 10-12 April 2002
Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 7 (2005)
Workshop Proceedings, Lisbon, Portugal, 17-18 May 2004
Shielding Aspects of Accelerators, Targets and Irradiation Facilities -- SATIF-11 (2013)
Workshop Proceedings, Tsukuba, Japan, 11-13 September 2012
Shielding Aspects of Accelerators,Targets and Irradiation Facilities - SATIF-10 (2010)
Workshop Proceedings, Geneva, Switzerland, 2-4 June 2010
Source Convergence in Criticality Safety Analyses (2006)
Phase I: Results for Four Test Problems
Speciation Techniques and Facilities for Radioactive Materials at Synchrotron Light Sources (2006)
Workshop Proceedings, Berkeley, California, USA, 14-16 September 2004
Speciation Techniques and Facilities for Radioactive Materials at Synchrotron Light Sources (2007)
Workshop Proceedings, Karlsruhe, Germany, 18-20 September 2006
Speciation, Techniques and Facilities for Radioactive Materials at Synchroton Light Sources (2002)
Workshop Proceedings, Grenoble, France, 10-12 September 2000
Speciation, Techniques and Facilities for Radioactive Materials at Synchrotron Light Sources (1999)
Workshop Proceedings, Grenoble, France, 4-6 October 1998
Structural Materials for Innovative Nuclear Systems (SMINS) (2008)
Workshop Proceedings, Karlsruhe, Germany, 4-6 June 2007
Structural Materials for Innovative Nuclear Systems (SMINS-2) (2012)
Workshop Proceedings, Daejon, Republic of Korea, 31 August-3 September 2010
Technology and Components of Accelerator-driven Systems (2011)
Workshop Proceedings, Karlsruhe, Germany, 15-17 March 2010
The JEF-2.2 Nuclear Data Library (2000)
JEFF Report 17
The JEFF-3.0 Nuclear Data Library (Reprint) (2005)
JEFF Report 19 - Synopsis of the General Purpose File
The JEFF-3.1 Nuclear Data Library (2006)
JEFF Report 21
The JEFF-3.1.1 Nuclear Data Library (2009)
JEFF Report 22 - Validation Results from JEF-2.2 to JEFF-3.1.1
The Need for Integral Critical Experiments with Low-moderated MOX Fuels (2004)
Workshop Proceedings, Paris, France, 14-15 April 2004
The Use of Thermodynamic Databases in Performance Assessment (2002)
Workshop Proceedings, Barcelona, Spain, 29-30 May 2001
Thermal Performance of High Burn-Up LWR Fuel (1998)
Seminar Proceedings, Cadarache, France, 3-6 March 1998
Utilisation and Reliability of High Power Proton Accelerators (2005)
Workshop Proceedings, Daejeon, Republic of Korea, 16-19 May 2004
Utilisation and Reliability of High Power Proton Accelerators (2001)
Workshop Proceedings, Aix-en-Provence, France, 22-24 November 1999
Utilisation and Reliability of High Power Proton Accelerators (2003)
Workshop Proceedings, Santa Fe, New Mexico, USA, 12-16 May 2002
Utilisation and Reliability of High Power Proton Accelerators (1999)
Workshop Proceedings, Mito, Japan, 13-15 October 1998
Utilisation and Reliability of High Power Proton Accelerators (HPPA5) (2008)
Workshop Proceedings, Mol, Belgium, 6-9 May 2007
VVER-1000 Coolant Transient Benchmark (2010)
Phase 2 (V1000CT-2) Summary Results of Exercise 1 on Vessel Mixing Simulation
VVER-1000 Coolant Transient Benchmark (Vol. I) (2002)
PHASE 1 (V1000CT-1), Vol. I: Main Coolant Pump (MCP) Switching On - Final Specifications
VVER-1000 Coolant Transient Benchmark (Vol. II) (2006)
Phase 1 (V1000CT-1), Vol. 2: Summary Results of Exercise 1 on Point Kinetics Plant Simulation
VVER-1000 Coolant Transient Benchmark - Phase 1 (Vol. 3) (2007)
Phase I (V1000CT-1), Vol. 3: Summary Results of Exercise 2 on Coupled 3-D Kinetics/Core Thermal-hydraulics
VVER-1000 MOX Core Computational Benchmark (2006)
Specification and Results

Detailed publication list

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Chemical Thermodynamics of Iron, Part I, Volume 13a
English, 1124 pages, published: 12/13/13
NEA#6355
Available online at: http://www.oecd-nea.org/dbtdb/pubs/6355-vol13a-iron.pdf
This volume is the 13th in the OECD Nuclear Energy Agency (NEA) "Chemical Thermodynamics" series. It is the first part of a critical review of the thermodynamic properties of iron, 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 iron, but also among 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.
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International Evaluation Co-operation (Vol. 33)
Methods and Issues for the Combined Use of Integral Experiments and Covariance Data (Volume 33)
English, 178 pages, published: 12/20/13
NEA#7171
Volume of the series: Nuclear Science
Available online at: http://www.oecd-nea.org/science/wpec/volume33/volume33.pdf
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International Handbook of Evaluated Reactor Physics Benchmark Experiments (DVD)
March 2013
English, published: 05/13/13
NEA#7140
Available online at: http://www.oecd-nea.org/science/wprs/irphe/irphe-handbook/handbook.html
The International Reactor Physics Experiments Evaluation Project (IRPhEP) was launched in 1999 by the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). While co-ordination and administration of the IRPhEP is managed at the international level by the NEA, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis.
This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at various nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; they do not, however, constitute validation or endorsement of the codes or cross-section data.
The 2013 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 130 experimental series performed at 47 reactor facilities. One hundred twenty-six of the 130 evaluations are published as approved benchmarks; the remaining four are published as draft documents only.
New to the handbook are benchmark specifications for selected measurements on the very-high-temperature reactor critical assembly (VHTRC) which were performed at the Japan Atomic Energy Agency (JAEA) Tokai Research Establishment in Japan between 1985 and 1996.
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Minor Actinide Burning in Thermal Reactors
A Report by the Working Party on Scientific Issues of Reactor Systems
English, 82 pages, published: 11/18/13
NEA#6997
Available online at: http://www.oecd-nea.org/science/pubs/2013/6997-minor-actinide.pdf
This publication provides an introduction to minor actinide nuclear properties and discusses some of the arguments in favour of minor actinide recycling, as well as the potential role of thermal reactors in this regard. Various technical issues and challenges are examined from the fuel cycle, operations, fuel designs, core management and safety/dynamics responses to safety and economics. The focus of this report is on the general conclusions of recent research that could be applied to thermal reactors. Further research and development needs are also considered, with summaries of findings and recommendations for the direction of future R&D efforts.
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Shielding Aspects of Accelerators, Targets and Irradiation Facilities -- SATIF-11
Workshop Proceedings, Tsukuba, Japan, 11-13 September 2012
English, 202 pages, published: 10/08/13
NEA#7157
Available online at: http://www.oecd-nea.org/science/pubs/2013/7157-satif-11.pdf
Particle accelerators have evolved over the last decades from simple devices to powerful machines, and are having an increasingly important impact on research, technology and daily life. Today they have a wide range of applications in many areas including material science and medical applications. In recent years, new technological and research applications have helped to define requirements while the number of accelerator facilities in operation, being commissioned, designed or planned has grown significantly. Their parameters, which include the beam energy, currents and intensities, and target composition, can vary widely, giving rise to new radiation shielding aspects and problems.
Particle accelerators must be operated in safe ways to protect operators, the public and the environment. As the design and use of these facilities evolve, so must the analytical methods used in the safety analyses. These workshop proceedings review the state of the art in radiation shielding of accelerator facilities and irradiation targets. They also evaluate progress in the development of modelling methods used to assess the effectiveness of such shielding as part of safety analyses.
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Status Report on Structural Materials for Advanced Nuclear Systems
English, 107 pages, published: 10/21/13
NEA#6409
Available online at: http://www.oecd-nea.org/science/pubs/2013/6409-sr-smans.pdf
Materials performance is critical to the safe and economic operation of any nuclear system. As the international community pursues the development of Generation IV reactor concepts and accelerator-driven transmutation systems, it will be increasingly necessary to develop advanced materials capable of tolerating the more challenging environments of these new systems. The international community supports numerous materials research programmes, with each country determining its individual focus on a case-by-case basis. In many instances, similar alloys of materials systems are being studied in several countries, providing the opportunity for collaborative and cross-cutting research that benefits different systems.

This report is a snapshot of the current materials programmes supporting the development of advanced concepts. The descriptions of the research are grouped by concept, and national programmes are described within each concept. The report provides an overall sense of the importance of materials research worldwide and the opportunities for synergy among the countries represented in this overview.
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Validation of the JEFF-3.1 Nuclear Data Library
JEFF Report 23
English, 76 pages, published: 02/14/13
NEA#7079
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-23/nea7079-jeff23.pdf
The Joint Evaluated Fission and Fusion (JEFF) Project is a collaborative effort among OECD Nuclear Energy Agency (NEA) Data Bank member countries to develop a reference nuclear data library for use in different energy applications. These data can be used to help improve the safety and economy of existing installations, as well as to design advanced nuclear reactors and their associated fuel cycles, including radioactive waste management. The JEFF-3.1 library contains several different data types, including neutron and proton interaction data, neutron activation data, radioactive decay data, fission yield data and thermal scattering data. This report describes the initial validation of the complete JEFF-3.1 library for thermal reactors, fuel cycle, storage and reprocessing, fusion technology and intermediate energy applications. It will be useful for scientists and engineers in national laboratories, universities and industry who use basic nuclear data, and is particularly suitable for those who work with application libraries based on JEFF-3.1.

The JEF/DOC and EFFDOC working documents cited in the report are available online at www.oecd-nea.org/dbdata/nds_jefreports/jefreport-23/.

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Actinide and Fission Product Partitioning and Transmutation
Eleventh Information Exchange Meeting, San Francisco, California, USA, 1-4 November 2010
English, 404 pages, published: 06/01/12
NEA#6996, ISBN: 978-92-64-99174-3
Available online at: http://www.oecd-nea.org/science/reports/2012/nea6996-11thPandT.pdf
In order to provide experts with a forum to present and discuss developments in the field of partitioning and transmutation (P&T), the OECD Nuclear Energy Agency (NEA) has been organising, since 1990, a series of biennial information exchange meetings on actinide and fission product P&T.

These proceedings contain all the technical papers presented at the 11th Information Exchange Meeting, which was held on 1-4 November 2010 in San Francisco, California, USA. The meeting covered national programmes on P&T; fuel cycle strategies and transition scenarios; waste forms and geological disposal; transmutation fuels and targets; pyro and aqueous processes; transmutation physics and materials; and transmutation system design, performance and safety.
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Burn-up Credit Criticality Safety Benchmark – Phase VII
UO2 Fuel: Study of Spent Fuel Compositions for Long-term Disposal
English, 180 pages, published: 02/21/12
NEA#6998, ISBN: 978-92-64-99172-9
Available online at: http://www.oecd-nea.org/science/docs/2012/burn-up-credit-phaseVII.pdf
After spent nuclear fuel (SNF) is discharged from a nuclear reactor, fuel composition and reactivity continue to vary as a function of time due to the decay of unstable nuclides. Accurate predictions of the concentrations of long-lived radionuclides in SNF, which represent a significant potential hazard to human beings and to the environment over a very long period, are particularly necessary for radiological dose assessments.
This report assesses the ability of existing computer codes and associated nuclear data to predict isotopic compositions and their corresponding neutron multiplication factor (keff) values for pressurised-water-reactor (PWR) UO2 fuel at 50 GWd/MTU burn-up in a generic spent fuel cask configuration. Fuel decay compositions and keff values have been calculated for 30 post-irradiation time steps out to one million years.
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Chemical Thermodynamics of Tin
Chemical Thermodynamics Volume 12
English, 644 pages, published: 12/31/12
NEA#6354, ISBN: 978-92-64-99206-1
Available online at: http://www.oecd-nea.org/dbtdb/pubs/tin.pdf
This volume is the 12th in the OECD Nuclear Energy Agency (NEA) "Chemical Thermodynamics" series. It is based on a critical review of the thermodynamic properties of tin, its solid compounds and aqueous complexes, carried out 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 tin, but also among 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.
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Homogeneous versus Heterogeneous Recycling of Transuranics in Fast Nuclear Reactors
English, 92 pages, published: 12/31/12
NEA#7077, ISBN: 978-92-64-99177-4
Available online at: http://www.oecd-nea.org/science/docs/2012/7077-hvh-recycling-transuranics-fnr.pdf
Fuel transuranics (TRU) multi-recycling is a mandatory feature if both the resource sustainability and the waste minimisation objectives for future fuel cycles are to be pursued. The resulting TRU transmutation can be implemented in fast neutron spectrum reactors according to two main options commonly referred to as the homogeneous and heterogeneous modes.

In this study, the two alternatives have been compared in terms of reactor core feasibility, fuel development and impact on the fuel cycle. The multi-criteria analysis indicates that there are major challenges in minor actinide-loaded fuel development, its experimental validation and possibly in its reprocessing. Both modes of recycling have an impact on the overall fuel cycle, even if at different stages, for example complex target fabrication and handling in the case of heterogeneous recycling and full core fuel fabrication in the case of homogeneous recycling. The study finds that an economic evaluation according to specific implementation scenarios should still be undertaken.
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International Handbook of Evaluated Criticality Safety Benchmark Experiments
September 2011
English, published: 03/26/12
NEA#7038, ISBN: 978-92-64-99163-7
Available online at: http://www.oecd-nea.org/science/wpncs/icsbep/handbook.html
The International Criticality Safety Benchmark Evaluation Project (ICSBEP), originally initiated at the national level by the US Department of Energy in 1992, became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) in 1995.

This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; these calculations do not, however, constitute a validation of the codes or cross-section data.

The evaluated criticality safety benchmark data are presented in nine volumes, containing over 58 000 pages and 533 evaluations with benchmark specifications for 4 552 critical, near-critical or subcritical configurations, 24 criticality alarm placement/shielding configurations with multiple dose points for each, and 200 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications.

New to the handbook are benchmark specifications for the GROTESQUE: Complex Geometric Arrangement of Unreflected HEU (93.15) Metal Pieces experiment (see front cover) that was performed by John T. Mihalczo at the Oak Ridge National Laboratory Critical Experiment Facility in June 1964.
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International Handbook of Evaluated Criticality Safety Benchmark Experiments (DVD)
English, published: 12/31/12
NEA#7080, ISBN: 978-92-64-99192-7
Available online at: http://www.oecd-nea.org/science/wpncs/icsbep/handbook.html
The International Criticality Safety Benchmark Evaluation Project (ICSBEP), originally initiated at the national level by the US Department of Energy in 1992, became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) in 1995.
This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; these calculations do not, however, constitute a validation of the codes or cross-section data.
The evaluated criticality safety benchmark data are presented in nine volumes, containing over 65 000 pages and 549 evaluations with benchmark specifications for over 4 700 critical, near-critical or subcritical configurations, 24 criticality alarm placement/shielding configurations with multiple dose points for each, and 200 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications.
New to the handbook are benchmark specifications for the Water-moderated Square-pitched U(6.90)O2 Fuel Rod Lattices with 0.67 Fuel-to-water Ratio experiments (see front cover) that were performed by a team of experimenters at Sandia National Laboratories between 2009 and 2012.
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International Handbook of Evaluated Reactor Physics Benchmark Experiments (DVD)
English, published: 05/15/12
NEA#7081, ISBN: 978-92-64-99168-2
Available online at: http://www.oecd-nea.org/science/wprs/irphe/irphe-handbook/handbook.html
The International Reactor Physics Experiments Evaluation Project (IRPhEP) was launched in 1999 by the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). While co-ordination and administration of the IRPhEP is managed at the international level by the NEA, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis.

This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at various nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; they do not, however, constitute validation or endorsement of the codes or cross-section data.

The 2012 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 56 experimental series performed at 32 reactor facilities. Fifty-four of the 56 evaluations are published as approved benchmarks; the remaining two are published as draft documents only.

New to the handbook are benchmark specifications for selected configurations from the HTR-PROTEUS Pebble Bed Experimental Program which were performed at the Paul Scherrer Institute’s PROTEUS zero-power research reactor in Villigen, Switzerland between 1992 and 1996.
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JEFF 3.1.2
Joint Evaluated Nuclear Data Library for Fission and Fusion Applications February 2012
English, published: 04/19/12
NEA#7111
Available online at: http://www.oecd-nea.org/dbdata/jeff
The Joint Evaluated Fission and Fusion File is an evaluated library produced through international collaboration among Data Bank member countries co-ordinated by the NEA Data Bank. As of February 2012, JEFF 3.1.2 is the latest update of the general purpose neutron data library.

This DVD contains:
• General purpose incident neutron data in ENDF-6 and ACE formats
• Activation data
• Thermal scattering data
• Incident proton data
• Radioactive decay data
• Neutron-induced fission yields data
• Spontaneous fission yields data
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Janis 3.4
A Java-based Nuclear Data Display Program
English, published: 08/21/12
NEA#7116
Available online at: http://www.oecd-nea.org/tools/abstract/detail/nea-1760/
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Structural Materials for Innovative Nuclear Systems (SMINS-2)
Workshop Proceedings, Daejon, Republic of Korea, 31 August-3 September 2010
English, 444 pages, published: 12/31/12
NEA#6896, ISBN: 978-92-64-99209-2
Available online at: http://www.oecd-nea.org/science/docs/2012/6896-smins-korea-proceedings.pdf
Materials research is a field of growing relevance for innovative nuclear systems, such as Generation IV reactors, critical and sub-critical transmutation systems and fusion devices. For these different systems, structural materials are selected or developed taking into account the specificities of their foreseen operational environment. Since 2007, the OECD Nuclear Energy Agency (NEA) has begun organising a series of workshops on Structural Materials for Innovative Nuclear Systems (SMINS) in order to provide a forum to exchange information on current materials research programmes for different innovative nuclear systems. These proceedings include the papers of the second workshop (SMINS-2) which was held in Daejon, Republic of Korea on 31 August-3 September 2010, and hosted by the Korea Atomic Energy Research Institute (KAERI).

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Potential Benefits and Impacts of Advanced Nuclear Fuel Cycles with Actinide Partitioning and Transmutation
English, 74 pages, published: 09/29/11
NEA#6894, ISBN: 978-92-64-99165-1
Available online at: http://www.oecd-nea.org/science/reports/2011/6894-benefits-impacts-advanced-fuel.pdf
This report provides a comparative analysis of different studies performed to assess the potential impact of partitioning and transmutation (P&T) on different types of geological repositories for radioactive waste in various licensing and regulatory environments. Criteria, metrics and impact measures have been analysed and compared with the goal of providing an objective comparison of the state of the art to help shape decisions on options for future advanced fuel cycles.
P&T allows a reduction of the inventory of the emplaced materials which can have a significant impact on the repository. Such a reduction can also make the uncertainty about repository performance less important both during normal evolution and in the case of disruptive scenarios. While P&T will never replace the need for waste repositories, it has the potential to significantly improve public perception regarding the ability to effectively manage radioactive waste by largely reducing the transuranic (TRU) waste masses to be stored and, consequently, to improve public acceptance of the geological repositories. Both issues are important for the future sustainability of nuclear power.
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Technology and Components of Accelerator-driven Systems
Workshop Proceedings, Karlsruhe, Germany, 15-17 March 2010
English, 442 pages, published: 06/28/11
NEA#6897, ISBN: 978-92-64-11727-3
Available online at: http://www.oecd-nea.org/science/pubs/2011/6897-technology-components.pdf
The accelerator-driven system (ADS) is a potential transmutation system option as part of partitioning and transmutation strategies for radioactive waste in advanced nuclear fuel cycles. These proceedings contain all the technical papers presented at the workshop on Technology and Components of Accelerator-driven Systems held on 15-17 March 2010 in Karlsruhe, Germany. The workshop provided experts with a forum to present and discuss state-of-the-art developments in the field of ADS and neutron sources. It included a special session on the EUROTRANS as well as four technical sessions covering current ADS experiments and test facilities, accelerators, neutron sources and subcritical systems.

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Actinide and Fission Product Partitioning and Transmutation + CD-ROM
Tenth Information Exchange Meeting, Mito, Japan, 6-10 October 2008
English, 454 pages, published: 02/08/10
NEA#6420, ISBN: 978-92-64-99097-5
Available online at: http://www.oecd-nea.org/science/reports/2010/nea6420-actinide10th.html
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.
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Boiling Water Reactor Turbine Trip (TT) Benchmark - Vol. IV
Volume IV: Summary Results of Exercise 3
English, 276 pages, published: 10/08/10
NEA#6050, ISBN: 978-92-64-99137-8
Available online at: http://www.oecd-nea.org/science/reports/2010/nea6050-tt-benchmark-vol4.pdf
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 of the 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 that purpose.

The present volume is the last in a series of four and summarises the results of the third benchmark exercise, which analyses a turbine trip (TT) in a BWR in its entirety, involving pressurisation events in which the coupling between core phenomena and system dynamics plays an important role. Exercise 3 also analyses four extreme scenarios which allowed participants to test the capabilities of their code(s) in terms of coupling and feedback modelling. The data made available from experiments carried out at the plant make the present benchmark particularly valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4).
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International Nuclear Data Evaluation Co-operation - CD-ROM
Complete Collection of Published Reports as of January 2010
English, published: 03/19/10
NEA#6942
Free on request
The NEA International Nuclear Data Evaluation Co-operation programme brings together evaluation projects being carried out in Japan (JENDL), the United States (ENDF), Europe (JEFF) and non-OECD countries (BROND, CENDL and FENDL). The Nuclear Data Section of the International Atomic Energy Agency (IAEA) sponsors the participation of evaluation projects from non-OECD countries.

The Co-operation programme was established to promote the exchange of information on nuclear data evaluations, measurements, nuclear model calculations, validation and related topics, as well as to provide a framework for co-operative activities among the participating projects. The Co-operation programme assesses needs for nuclear data improvements and addresses those needs by initiating joint evaluation and/or measurement efforts. Expert groups are established to solve specific common nuclear data problems. Each expert group produces a final report of its findings.

This CD-ROM contains the full collection of the expert group reports as of January 2010.
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JANIS 3
A Java-based Nuclear Data Display Program
English, published: 07/02/10
NEA#6907
Free on request
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JEFF Reports CD-ROM
Complete Collection of JEFF Reports 1-22
English, published: 03/19/10
NEA#6941
Free on request
The Joint Evaluated File (JEF) project was started in 1982 as a collaborative project among NEA Data Bank member countries. The main objective is to provide participating countries with a common and unique source of nuclear data for the calculation and prediction of different nuclear applications. The first version of the JEF file was issued in 1985, and was followed in spring 1993 by a second version (JEF-2.2). An improved, third version was developed in collaboration with the European Fusion File (EFF) project and released in 2005 as the Joint Evaluated Fission and Fusion file (JEFF-3.1). Further updates of the radioactive decay data and neutron data sub-libraries were successively released in 2007 and 2009 as JEFF-3.1.1.

This CD-ROM contains the complete collection of JEF(F) Reports as of January 2010. Among the various JEF(F) publications, reports and documents, only the JEF(F) reports should be used as an official reference.
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NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark (Vol. II)
Volume II: Uncertainty and Sensitivity Analyses of Void Distribution and Critical Power - Specification
English, 44 pages, published: 07/30/10
NEA#6343, ISBN: 978-92-64-99124-8
Available online at: http://www.oecd-nea.org/science/docs/2010/NUPEC-BWR-VOL2.pdf
The government of Japan and the Japanese Nuclear Power Engineering Corporation (NUPEC) have released high-quality data, based on a series of void measurements using full-size mock-up tests for boiling water reactors (BWRs), with the aim of assisting the scientific community to advance its understanding of the two-phase flow (a system containing both gas and liquid) in BWR fuel bundles.

An international benchmark, based on the NUPEC data, has been defined to encourage advancement in the development of two-phase flow theory, which is of importance, for example, for the evaluation of the safety margins in a reactor. The benchmark specifications are being designed so that it systematically assesses and compares the capability of the numerical models to predict detailed void distributions and critical powers.

This report is the second in a series and describes the specification of the sensitivity and uncertainty analysis exercises undertaken to assess the accuracy of the results obtained when modelling basic thermal-hydraulics in a single channel relative to void fraction and critical power. Further volumes will be published, with a synthesis showing to what extent the most recent models are capable of predicting two-phase flow in BWR fuel bundles.
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National Programmes in Chemical Partitioning
A Status Report
English, 120 pages, published: 03/10/10
NEA#5425, ISBN: 978-92-64-99096-8
Available online at: http://www.oecd-nea.org/science/reports/2010/nea5425-National-Prog.pdf
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.
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Nuclear Production of Hydrogen
Fourth Information Exchange Meeting, Oakbrook, Illinois, United States, 13-16 April 2009
English, 464 pages, published: 06/24/10
NEA#6805, ISBN: 978-92-64-08713-2
Available online at: http://www.oecd-nea.org/science/pubs/2010/6805-production-hydrogen.pdf
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.
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Shielding Aspects of Accelerators,Targets and Irradiation Facilities - SATIF-10
Workshop Proceedings, Geneva, Switzerland, 2-4 June 2010
English, 444 pages, published: 12/15/10
NEA#6898, ISBN: 978-92-64-03467-9
Available online at: http://www.oecd-nea.org/science/pubs/2010/6898-satif-10.pdf
Particle accelerators have evolved over the last decades from simple devices to powerful machines, and are having an increasingly important impact on research, technology and daily life. Today they cover a wide range of applications including material science and medical applications. In recent years, requirements from new technological and research applications have emerged while the number of accelerator facilities in operation, being commissioned, designed or planned has significantly grown. Their parameters (such as the beam energy, beam currents and intensities, and target composition) vary widely, giving rise to new radiation shielding aspects and problems.

Particle accelerators must be operated in safe ways to protect operators, the public and the environment. As the design and use of these facilities evolve, so must the analytical methods used in the safety analyses. These workshop proceedings review the state of the art in radiation shielding of accelerator facilities and irradiation targets. They also evaluate progress on the development of modelling methods used to assess the effectiveness of such shielding as part of safety analyses.
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VVER-1000 Coolant Transient Benchmark
Phase 2 (V1000CT-2) Summary Results of Exercise 1 on Vessel Mixing Simulation
English, 144 pages, published: 10/05/10
NEA#6964, ISBN: 978-92-64-99152-1
Available online at: http://www.oecd-nea.org/science/reports/2010/nea6964-ex-l-vessel-mixing.pdf
Recently developed best-estimate computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear reactors need to be validated against results from experiments and compared with each other to help understand how the different modelling methods adopted affect the accuracy of the simulation. This benchmark was set up for that purpose.

This report is one of a series covering benchmarks designed to test modelling methods for a range of transient scenarios in a VVER-1000 reactor. In this case, the transient is initiated by isolation of one steam generator causing asymmetric loop heat-up. The benchmark is based on experiments conducted at the Kozloduy nuclear power plant.

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Chemical Thermodynamics of Thorium - Volume 11
English, 942 pages, published: 01/22/09
NEA#6254, ISBN: 978-92-64-05667-1
Available online at: http://www.oecd-nea.org/science/pubs/2007/6254-DB-chemical-thermodyn-11.pdf
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.
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Evaluated Data Library for the Bulk of Fission Products (Volume 23)
International Evaluation Co-operation, Volume 23
English, 44 pages, published: 09/27/09
NEA#6283, ISBN: 978-92-64-99092-0
Available online at: http://www.oecd-nea.org/science/wpec/volume23/volume23.pdf
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.
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Independent Evaluation of the MYRRHA Project
Report by an International Team of Experts
English, 44 pages, published: 12/16/09
NEA#6881, ISBN: 978-92-64-99114-9
Available online at: http://www.oecd-nea.org/science/reports/2009/nea6881-MYRRHA.pdf
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.
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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
English, 172 pages, published: 07/17/09
NEA#6285, ISBN: 978-92-64-99073-9
Available online at: http://www.oecd-nea.org/science/reports/2009/6285_CriticalityComparison.pdf
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.
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Mixed-oxide (MOX) Fuel Performance Benchmark (PRIMO)
Summary of the Results for the PRIMO BD8 MOX Rod
English, 40 pages, published: 07/13/09
NEA#6291, ISBN: 978-92-64-99085-2
Available online at: http://www.oecd-nea.org/science/reports/2009/6291-MOX.pdf
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.
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Mobile Fission and Activation Products in Nuclear Waste Disposal
Workshop Proceedings, La Baule, France, 16-19 January 2007
English, 264 pages, published: 05/25/09
NEA#6310, ISBN: 978-92-64-99072-2
Available online at: http://www.oecd-nea.org/science/reports/2009/nea6310-MOFAP.pdf
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.
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Nuclear Fuel Cycle Synergies and Regional Scenarios for Europe
English, 36 pages, published: 09/27/09
NEA#6857, ISBN: 978-92-64-99086-9
Available online at: http://www.oecd-nea.org/science/reports/2009/nea6857-Regional-Scenarios.pdf
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.
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Nuclear Fuel Cycle Transition Scenario Studies
Status Report
English, 124 pages, published: 02/03/09
NEA#6194, ISBN: 978-92-64-99068-5
Available online at: http://www.oecd-nea.org/science/reports/2009/nea6194_transition_scenario_studies.pdf
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.
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PENELOPE-2008: A Code System for Monte Carlo Simulation of Electron and Photon Transport
Workshop Proceedings, Barcelona, Spain, 30 June-3 July 2008
English, 336 pages, published: 02/20/09
NEA#6416, ISBN: 978-92-64-99066-1
Available online at: http://www.oecd-nea.org/science/pubs/2009/nea6416-penelope.pdf
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.
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Research and Test Facilities Required in Nuclear Science and Technology
English, 156 pages, published: 10/09/09
NEA#6293, ISBN: 978-92-64-99070-8
Available online at: http://www.oecd-nea.org/science/reports/2009/6293-Research-Test-Facilities.pdf

Other language(s):
- Japanese: Research and Test Facilities Required in Nuclear Science and Technology (Japanese version) 原子力の科学技術で必要とされる試験研究施設 
- Français: Besoins d'installations de recherche et d'expérimentation en sciences et technologies nucléaires 
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.
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The JEFF-3.1.1 Nuclear Data Library
JEFF Report 22 - Validation Results from JEF-2.2 to JEFF-3.1.1
English, 62 pages, published: 05/05/09
NEA#6807, ISBN: 978-92-64-99074-6
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-22/nea6807-jeff22.pdf
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.
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The JEFF-3.1/-3.1.1 Radioactive Decay Data and Fission Yields Sub-libraries
JEFF Report 20
English, 148 pages, published: 09/27/09
NEA#6287, ISBN: 978-92-64-99087-6
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-20/nea6287-jeff-20.pdf
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.

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Analytical Benchmarks for Nuclear Engineering Applications
Case Studies in Neutron Transport Theory
English, 296 pages, published: 09/01/08
NEA#6292, ISBN: 978-92-64-99056-2
Available online at: http://www.oecd-nea.org/databank/docs/2008/db-doc2008-1.pdf
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.
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International Evaluation Co-operation (Vol. 26) + CD-ROM
Uncertainty and Target Accuracy Assessment for Innovative Systems Using Recent Covariance Data Evaluations (Volume 26)
English, 196 pages, published: 09/01/08
NEA#6410, ISBN: 978-92-64-99053-1
Available online at: http://www.oecd-nea.org/science/wpec/volume26/volume26.pdf
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.
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Structural Materials for Innovative Nuclear Systems (SMINS)
Workshop Proceedings, Karlsruhe, Germany, 4-6 June 2007
English, 544 pages, published: 07/10/08
NEA#6260, ISBN: 978-92-64-04806-5
Available online at: http://www.oecd-nea.org/science/pubs/2008/6260-smins2008.pdf
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.
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Utilisation and Reliability of High Power Proton Accelerators (HPPA5)
Workshop Proceedings, Mol, Belgium, 6-9 May 2007
English, 456 pages, published: 04/03/08
NEA#6259, ISBN: 978-92-64-04478-4
Available online at: http://www.oecd-nea.org/science/pubs/2008/6259-HPPA-Belgium.pdf
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.

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Actinide and Fission Product Partitioning and Transmutation
Ninth Information Exchange Meeting, Nîmes, France, 25-29 September 2006
English, 752 pages, published: 10/29/07
NEA#6282, ISBN: 978-92-64-99030-2
Available online at: http://www.oecd-nea.org/science/pubs/2007/nea6282-iempt9.pdf
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.
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Assessment of Fission Product Decay Data for Decay Heat Calculations
International Evaluation Co-operation, Volume 25
English, 60 pages, published: 11/14/07
NEA#6284, ISBN: 978-92-64-99034-0
Available online at: http://www.oecd-nea.org/science/wpec/volume25/volume25.pdf
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.
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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
English, 512 pages, published: 09/09/08
NEA#5435, ISBN: 978-92-64-99049-4
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea5435-burnup-IIC.pdf
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.
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Chemical Thermodynamics of Solid Solutions of Interest in Nuclear Waste Management - Volume 10
A State-of-the-art Report
English, 288 pages, published: 07/25/07
NEA#6255, ISBN: 978-92-64-02655-1
Available online at: http://www.oecd-nea.org/science/pubs/2007/6255-DB-chemical-thermodyn-10.pdf
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).
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Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-hydraulics and Technologies + CD-ROM
English, 692 pages, published: 05/31/07
NEA#6195, ISBN: 978-92-64-99002-9
Available online at: http://www.oecd-nea.org/science/reports/2007/nea6195-handbook.html
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.
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Mixed-oxide (MOX) Fuel Performance Benchmark
Summary of the Results for the Halden Reactor Project MOX Rods
English, 64 pages, published: 06/21/07
NEA#4450, ISBN: 978-92-64-99019-7
Available online at: http://www.oecd-nea.org/science/docs/2007/nsc-doc2007-6.pdf
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.
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Physics of Plutonium Recycling - Volume IX
Volume IX: Benchmark on Kinetic Parameters in the CROCUS Reactor
English, 94 pages, published: 06/21/07
NEA#4440, ISBN: 978-92-64-99020-3
Available online at: http://www.oecd-nea.org/science/pubs/2007/nea4440-crocus.pdf
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.
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Physics of Plutonium Recycling - Volume VIII
Volume VIII: Results of a Benchmark Considering a High-temperature Reactor (HTR) Fuelled with Reactor-grade Plutonium
English, 102 pages, published: 06/21/07
NEA#6200, ISBN: 978-92-64-99007-4
Available online at: http://www.oecd-nea.org/science/pubs/2007/nea6200-htr.pdf
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.
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Speciation Techniques and Facilities for Radioactive Materials at Synchrotron Light Sources
Workshop Proceedings, Karlsruhe, Germany, 18-20 September 2006
English, 336 pages, published: 05/09/07
NEA#6288, ISBN: 978-92-64-99006-7
Available online at: http://www.oecd-nea.org/science/pubs/2007/nea6288-speciation.pdf
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.
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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
English, 92 pages, published: 11/16/07
NEA#6201, ISBN: 978-92-64-99035-7
Available online at: http://www.oecd-nea.org/science/docs/2007/nsc-doc2007-18.pdf
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.

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Benchmark on the KRITZ-2 LEU and MOX Critical Experiments
Final Report
English, 232 pages, published: 07/07/06
NEA#3130, ISBN: 92-64-02298-8
Available online at: http://www.oecd-nea.org/science/docs/2005/nsc-doc2005-24.pdf
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.
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Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume III
Volume III: Summary Results of Exercise 2
English, 180 pages, published: 12/21/06
NEA#5437, ISBN: 92-64-02331-3
Available online at: http://www.oecd-nea.org/science/docs/2006/nsc-doc2006-23.pdf
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.
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Burn-up Credit Criticality Benchmark - Phase II-D
PWR-UO2 Assembly - Study of Control Rod Effects on Spent Fuel Composition
English, 184 pages, published: 12/20/06
NEA#6227, ISBN: 92-64-02316-X
Available online at: http://www.oecd-nea.org/science/nea6227-burnupIID.pdf
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.
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Computer Simulation of MASURCA Critical and Subcritical Experiments
MUSE-4 Benchmark - Final Report
English, 44 pages, published: 04/18/06
NEA#4439, ISBN: 92-64-01086-6
Available online at: http://www.oecd-nea.org/science/docs/2005/nsc-doc2005-23.pdf
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.
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International Evaluation Co-operation (Vol. 20)
Covariance Matrix Evaluation and Processing in the Resolved/Unresolved Resonance Regions (Volume 20)
English, 36 pages, published: 08/18/06
NEA#6198, ISBN: 92-64-02302-X
Available online at: http://www.oecd-nea.org/science/wpec/volume20/volume20.pdf
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.
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International Evaluation Co-operation (Vol. 22)
Nuclear Data for Improved LEU-LWR Reactivity Predictions (Volume 22)
English, 44 pages, published: 11/29/06
NEA#6199, ISBN: 92-64-02317-8
Available online at: http://www.oecd-nea.org/science/wpec/volume22/volume22.pdf
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International Evaluation Co-operation (Vol. 7)
Nuclear Data Standards (Volume 7)
English, 40 pages, published: 11/29/06
NEA#6197, ISBN: 92-64-02313-5
Available online at: http://www.oecd-nea.org/science/wpec/volume7/volume7.pdf
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.
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NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark
Volume I: Specifications
English, 136 pages, published: 07/07/06
NEA#6212, ISBN: 92-64-01088-2
Available online at: http://www.oecd-nea.org/science/docs/2005/nsc-doc2005-5.pdf
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.
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Nuclear Production of Hydrogen
Third Information Exchange Meeting, Oarai, Japan, 5-7 October 2005
English, 412 pages, published: 07/13/06
NEA#6122, ISBN: 92-64-02629-0
Available online at: http://www.oecd-nea.org/science/pubs/2006/6122-production-hydrogen.pdf
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.
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PENELOPE-2006: A Code System for Monte Carlo Simulation of Electron and Photon Transport
Workshop Proceedings, Barcelona, Spain, 4-7 July 2006
English, 296 pages, published: 06/26/06
NEA#6222, ISBN: 92-64-02301-1
Available online at: http://www.oecd-nea.org/science/pubs/2006/nea6222-penelope.pdf
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.
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Perspectives on Nuclear Data for the Next Decade
Workshop Proceedings, Bruyères-le-Châtel, France, 26-28 September 2005
English, 260 pages, published: 12/18/06
NEA#6121, ISBN: 92-64-02857-9
Available online at: http://www.oecd-nea.org/science/pubs/2006/6121-perspectives-next-decade.pdf
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.
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Physics and Safety of Transmutation Systems
A Status Report
English, 120 pages, published: 02/07/06
NEA#6090, ISBN: 92-64-01082-3
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea6090-transmutation.pdf
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.
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Pressurised Water Reactor MOX/UO2 Core Transient Benchmark
Final Report
English, 72 pages, published: 12/29/06
NEA#6048, ISBN: 92-64-02330-5
Available online at: http://www.oecd-nea.org/science/pubs/2006/nea6048-mox.pdf

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.
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Reference Values for Nuclear Criticality Safety + CD-ROM
English, 68 pages, published: 12/29/06
NEA#5433, ISBN: 92-64-02333-X
Available online at: http://www.oecd-nea.org/science/pubs/2006/nea5433-refvalues.pdf
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.
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Source Convergence in Criticality Safety Analyses
Phase I: Results for Four Test Problems
English, 200 pages, published: 09/11/06
NEA#5431, ISBN: 92-64-02304-6
Available online at: http://www.oecd-nea.org/science/pubs/2006/nea5431-source-convergence.pdf
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.
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Speciation Techniques and Facilities for Radioactive Materials at Synchrotron Light Sources
Workshop Proceedings, Berkeley, California, USA, 14-16 September 2004
English, 192 pages, published: 10/27/06
NEA#6046, ISBN: 92-64-02311-9
Available online at: http://www.oecd-nea.org/science/pubs/2006/nea6046-speciation.pdf
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.
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The JEFF-3.1 Nuclear Data Library
JEFF Report 21
English, 140 pages, published: 11/20/06
NEA#6190, ISBN: 92-64-02314-3
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-21/
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.
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VENUS-2 MOX-fuelled Reactor Dosimetry Calculations
Final Report
English, 228 pages, published: 04/11/06
NEA#6192, ISBN: 92-64-01084-X
Available online at: http://www.oecd-nea.org/science/docs/2005/nsc-doc2005-22.pdf
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.
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VVER-1000 Coolant Transient Benchmark (Vol. II)
Phase 1 (V1000CT-1), Vol. 2: Summary Results of Exercise 1 on Point Kinetics Plant Simulation
English, 96 pages, published: 06/30/06
NEA#6219, ISBN: 92-64-02295-3
Available online at: http://www.oecd-nea.org/science/docs/2006/nsc-doc2006-5.pdf
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.
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VVER-1000 MOX Core Computational Benchmark
Specification and Results
English, 88 pages, published: 01/20/06
NEA#6088, ISBN: 92-64-01081-5
Available online at: http://www.oecd-nea.org/science/docs/2005/nsc-doc2005-17.pdf
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.
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Very High Burn-ups in Light Water Reactors
English, 140 pages, published: 08/22/06
NEA#6224, ISBN: 92-64-02303-8
Available online at: http://www.oecd-nea.org/science/pubs/2006/nea6224-burn-up.pdf
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.

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Accelerator and Spallation Target Technologies for ADS Applications
A Status Report
English, 92 pages, published: 04/28/05
NEA#5421, ISBN: 92-64-01056-4
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea5421-accelerator.pdf
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.
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Benchmark on Deterministic Transport Calculations Without Spatial Homogenisation
MOX Fuel Assembly 3-D Extension Case
English, 160 pages, published: 09/16/05
NEA#5420, ISBN: 92-64-01069-6
Available online at: http://www.oecd-nea.org/science/docs/2005/nsc-doc2005-16.pdf
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.
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Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume II
Volume II: Summary Results of Exercise 1
English, 132 pages, published: 07/07/05
NEA#4448, ISBN: 92-64-01064-5
Available online at: http://www.oecd-nea.org/science/docs/2004/nsc-doc2004-21.pdf
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.
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Evaluation of Proposed Integral Critical Experiments with Low-moderated MOX Fuel
English, 124 pages, published: 07/07/05
NEA#6047, ISBN: 92-64-01049-1
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea6047-mox.pdf
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.
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Fuels and Materials for Transmutation
A Status Report
English, 240 pages, published: 07/29/05
NEA#5419, ISBN: 92-64-01066-1
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea5419_fuels_materials.pdf
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.
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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)
English, 258 pages, published: 12/12/05
NEA#5426, ISBN: 92-64-01070-X
Available online at: http://www.oecd-nea.org/science/wpec/volume19/volume19.pdf
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International Evaluation Co-operation (Vol. 21)
Assessment of Neutron Cross-section Evaluations for the Bulk of Fission Products (Volume 21)
English, 48 pages, published: 07/08/05
NEA#5428, ISBN: 92-64-01063-7
Available online at: http://www.oecd-nea.org/science/wpec/volume21/volume21.pdf
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.
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JEFF 3.1 (CD-ROM)
English, 65 pages, published: 06/02/05
NEA#6071
Available online at: http://www.oecd-nea.org/dbdata/JEFF/index.html
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Pellet-clad Interaction in Water Reactor Fuels
Seminar Proceedings, Aix-en-Provence, France, 9-11 March 2004
English, 550 pages, published: 07/20/05
NEA#6004, ISBN: 92-64-01157-9
Available online at: http://www.oecd-nea.org/science/pubs/2005/6004-pellet-clad.pdf
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.
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Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 7
Workshop Proceedings, Lisbon, Portugal, 17-18 May 2004
English, 260 pages, published: 05/13/05
NEA#6005, ISBN: 92-64-01042-4
Available online at: http://www.oecd-nea.org/science/pubs/2005/6005-SATIF-7.pdf
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.
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The JEFF-3.0 Nuclear Data Library (Reprint)
JEFF Report 19 - Synopsis of the General Purpose File
English, 136 pages, published: 04/01/05
NEA#6068, ISBN: 92-64-01046-7
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-19/jefreport-19.pdf
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.
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Utilisation and Reliability of High Power Proton Accelerators
Workshop Proceedings, Daejeon, Republic of Korea, 16-19 May 2004
English, 528 pages, published: 10/26/05
NEA#6003, ISBN: 92-64-01380-6
Available online at: http://www.oecd-nea.org/science/pubs/2005/6003-HPPA-Korea 2004.pdf
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.

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Basic Studies in the Field of High-temperature Engineering
Third Information Exchange Meeting, Ibaraki-ken, Japan, 11-12 September 2003
English, 280 pages, published: 06/01/04
NEA#5309, ISBN: 92-64-01601-5
Available online at: http://www.oecd-nea.org/science/pubs/2004/5309-basic-studies.pdf
In response to growing interest in high-temperature, gas-cooled reactors (HTGRs) in many countries and the need for improved materials for nuclear applications in high-temperature environments, the NEA organised the Third Information Exchange Meeting on Basic Studies in the Field of High-temperature Engineering. The proceedings of this meeting provide an overview of high-temperature research currently under way, including studies on the behaviour of irradiated graphite and improvements in material properties under high-temperature irradiation. These proceedings also contain recommendations for further international work in the areas of high-temperature engineering.
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Benchmark on Beam Interruptions in an Accelerator-driven System
Final Report on Phase II Calculations
English, 80 pages, published: 07/09/04
NEA#5422, ISBN: 92-64-02072-1
Available online at: http://www.oecd-nea.org/science/docs/2004/nsc-doc2004-7.pdf
In accelerator-driven system (ADS) development, it is important to evaluate temperature variations caused by beam trips, as this type of event in an ADS results in a temperature transient that can lead to thermal fatigue in the structural components of the subcritical system. A series of benchmarks is therefore being organised by the OECD Nuclear Energy Agency (NEA) for lead-bismuth-cooled and MOX-fuelled accelerator-driven systems.

This report provides a comparative analysis of the Phase II calculation results of the beam trip transients at different power densities. In subsequent phases of the benchmark, temperature transients under irradiated fuel conditions will also be investigated. This report and those to follow will be of particular interest to ADS designers, including subcritical system physicists and accelerator scientists.
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Benchmark on the Three-dimensional VENUS-2 MOX Core Measurements
Final Report
English, 204 pages, published: 02/09/04
NEA#4438, ISBN: 92-64-02160-4
Available online at: http://www.oecd-nea.org/science/docs/2003/nsc-doc2003-5.pdf
In order to validate the calculations methods and nuclear data used for the prediction of power in MOX-fuelled systems, the OECD Nuclear Energy Agency (NEA) has examined a series of theoretical physics benchmarks and multiple recycling issues of various MOX-fuelled systems. This had led to many improvements and clarifications in nuclear data libraries and calculation methods. The final validation requires linking those findings to data from experiments. Hence, the first experiment-based benchmarks using the VENUS-2 MOX core measurement data were undertaken in 1999. The two-dimensional benchmark was completed in 2000. A full three-dimensional benchmark using 3-D VENUS-2 MOX core experimental data was launched in 2001 for a more thorough investigation of the calculation methods.

This report provides details of the comparative analysis of the 3-D calculation results against experimental data. Results obtained with the latest nuclear data libraries and various modern 3-D calculation methods are analysed. The report will be of particular interest to reactor physicists and nuclear engineers as well as to nuclear data evaluators.
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Chemical Thermodynamics of Americium (reprint) (Volume 2)
Reprint of the 1995 Review
English, 392 pages, published: 06/02/04
NEA#3713, ISBN: 92-64-02168-X
Available online at: http://www.oecd-nea.org/dbtdb/pubs/americium.pdf
The present volume is a reprint of the 1995 edition of Chemical Thermodynamics of Americium by Robert J. Silva, Giovanni Bidoglio, Malcom H. Rand, Piotr B. Robouch, Hans Wanner and Ignasi Puigdomenech, which also contains an Appendix on the Chemical Thermodynamcis of Uranium by Ingmar Grenthe, M.C. Amaia Sandino, Ignasi Puidomenech and Malcom H. Rand.

As part of Phase II of the NEA Thermochemical Database Project (TDB), a new publication entitled Update on the Chemical Thermodynamics of Uranium, Neptunium, Plutonium, Americium and Technetium and authored by Robert Guillaumont, Thomas Fanghänel, Jean Fuger, Ingmar Grenthe, Volker Neck, Donald A. Palmer and Malcom H. Rand, was published by Elsevier in 2003. For americium (and for the topics dealt with in the 1995 Appendix on uranium), this Update contains a review of the literature published since the cut-off date for the literature reviewed in the 1995 edition cited above. As a consequence on this new TDB Review, some of the values selected in the earlier publication have been superseded while others have retained their validity. The 2003 Update is self-contained with respect to any new data selections, but the discussions leading to the retained selections can in most cases only be found in the 1995 publication. Since the latter is no longer available from its original publisher, the NEA is making the present reprint available to the scientific community.
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Chemical Thermodynamics of Uranium (reprint) (Volume 1)
Reprint of the 1992 Review
English, 738 pages, published: 06/02/04
NEA#3712, ISBN: 92-64-02167-1
Available online at: http://www.oecd-nea.org/dbtdb/pubs/uranium.pdf
The present volume is a reprint of the 1992 edition of Chemical Thermodynamics of Uranium by Ingmar Grenthe, Jean Fuger, Rudy J.M. Konings, Robert J. Lemire, Anthony B. Muller, Chinh Nguyen-Trung and Hans Wanner.

As part of Phase II of the NEA Thermochemical Database Project (TDB), a new publication entitled Update on the Chemical Thermodynamics of Uranium, Neptunium, Plutonium, Americium and Technetium and authored by Robert Guillaumont, Thomas Fanghänel, Jean Fuger, Ingmar Grenthe, Volker Neck, Donald A. Palmer and Malcom H. Rand, was published by Elsevier in 2003. For uranium this Update contains a review of the literature published since the cut-off date for the literature reviewed in the 1992 edition cited above. As a consequence on this new TDB Review, some of the values selected in the earlier publication have been superseded while others have retained their validity. The 2003 Update is self-contained with respect to any new data selections but the discussions leading to the retained selections can in most cases only be found in the 1992 publication. Since the latter is no longer available from its original publisher, the NEA is making the present reprint available to the scientific community.
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Computing Radiation Dosimetry - CRD 2002
Workshop Proceedings, Sacavém, Portugal, 22-23 June 2002
English, 256 pages, published: 08/11/04
NEA#4311, ISBN: 92-64-10823-8
Available online at: http://www.oecd-nea.org/science/pubs/2004/4311-computing-radiation-dosimetry.pdf
Establishing reliable computational methods and tools for radiation dosimetry is of great importance today because of the increased use of radiation in a number of areas of science, technology and medical applications. Fields concerned include radiation protection, radiation shielding, radiation diagnostics and therapy, radiobiology, biophysics and radiation detection.

A series of lectures delivered by experts provides the content of these workshop proceedings. They are a valuable reference for those wishing to better understand the most advanced computational methods in radiation dosimetry.
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JANIS - Version 2.0 (A Java-based Nuclear Data Display Program)
English, 2 pages, published: 01/28/04
NEA#3728
Available online at: http://www.oecd-nea.org/janis/welcome.html
JANIS (Java-based nuclear information software) is a display program designed to facilitate the visualisation and manipulation of nuclear data. Its objective is to allow the user of nuclear data to access numerical values and graphical representations without prior knowledge of the storage format. It offers maximum flexibility for the comparison of different nuclear data sets. The first version of JANIS was used by more than 700 users around the world. Important feedback was accumulated and improvements were added to the software to produce JANIS-2.0.
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JANIS 2.1 (DVD)
English, 1 pages, published: 07/27/04
NEA#5688
Available online at: http://www.oecd-nea.org/janis/welcome.html
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Neutronics/Thermal-hydraulics Coupling in LWR Technology, Vol. 1
CRISSUE-S - WP1: Data Requirements and Databases Needed for Transient Simulations and Qualification - 5th EURATOM Framework Programme (1998-2002)
English, 104 pages, published: 11/19/04
NEA#4452, ISBN: 92-64-02083-7
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea4452-crissue-s-vol1.pdf
The interaction between system thermal-hydraulics and 3-D neutron kinetics is relevant for both the safety and the design and operation of existing nuclear reactors and reactor cores. Today, advanced coupled thermal-hydraulics/neutronics computer tools along with powerful computers can perform realistic best-estimate analyses of complex power plant transients. The results provide new insights into the conversatisms for the specification of relevant operational safety margins and can imply new optimisations of emergency operating procedures in existing plants. They also improve knowledge of the physical phenomena behind "old-fashioned" problems (critical issues) in light water reactor technology, and can specifically shed light on the interaction between thermal-hydraulics and neutronics that still can challenge the design and operation of nuclear power plants.

This is the first of three reports addressing the type of transients that are of interest in relation to reactivity initiated accidents in light water reactor power plants, and elaborating on the data needed for coupled 3-D neutron kinetics/thermal-hydraulic analysis and associated validations.
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Neutronics/Thermal-hydraulics Coupling in LWR Technology, Vol. 2
CRISSUE-S - WP2: State-of-the-art Report - 5th EURATOM Framework Programme (1998-2002)
English, 296 pages, published: 11/19/04
NEA#5436, ISBN: 92-64-02084-5
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea5436-crissue-s-vol2.pdf
The interaction between system thermal-hydraulics and 3-D neutron kinetics is relevant for both the safety and the design and operation of existing nuclear reactors and reactor cores. Today, advanced coupled thermal-hydraulics/neutronics computer tools along with powerful computers can perform realistic best-estimate analyses of complex power plant transients. The results provide new insights into the conversatisms for the specification of relevant operational safety margins and can imply new optimisations of emergency operating procedures in existing plants. They also improve knowledge of the physical phenomena behind "old-fashioned" problems (critical issues) in light water reactor technology, and can specifically shed light on the interaction between thermal-hydraulics and neutronics that still can challenge the design and operation of nuclear power plants.

This is the second of a series of three reports. The first is devoted to the assembly and structure of the existing database related to this subject. This publications provides the state-of-the-art report on the subject. The third report summarises the results, selects the most important findings and indicates the industry position on the related subjects.
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Neutronics/Thermal-hydraulics Coupling in LWR Technology, Vol. 3
CRISSUE-S - WP3: Achievements and Recommendations Report - 5th EURATOM Framework Programme (1998-2002)
English, 68 pages, published: 11/19/04
NEA#5434, ISBN: 92-64-02085-3
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea5434-crissue-s-vol3.pdf
The interaction between system thermal-hydraulics and 3-D neutron kinetics is relevant for both the safety and the design and operation of existing nuclear reactors and reactor cores. Today, advanced coupled thermal-hydraulics/neutronics computer tools along with powerful computers can perform realistic best-estimate analyses of complex power plant transients. The results provide new insights into the conversatisms for the specification of relevant operational safety margins and can imply new optimisations of emergency operating procedures in existing plants. They also improve knowledge of the physical phenomena behind "old-fashioned" problems (critical issues) in light water reactor technology, and can specifically shed light on the interaction between thermal-hydraulics and neutronics that still can challenge the design and operation of nuclear power plants.

This is the third of a series of three reports. The first is devoted to the assembly and structure of the existing database related to this subject. The second provides the state-of-the-art report on the subject. The present report summarises the results, selects the most important findings and indicates the industry position on the related subjects.
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Nuclear Production of Hydrogen
Second Information Exchange Meeting, Argonne, Illinois, USA, 2-3 October 2003
English, 316 pages, published: 04/28/04
NEA#5308, ISBN: 92-64-10770-3
Available online at: http://www.oecd-nea.org/science/pubs/2004/5308-production-hydrogen.pdf
Hydrogen has the potential to play an important role as a sustainable and environmentally acceptable source of energy in the 21st century. Present methods for producing hydrogen are mainly based on the reforming of fossil fuels with subsequent release of greenhouse gases. To avoid producing greenhouse gases, the possibility to use heat and surplus electricity from nuclear power plants to produce hydrogen by water cracking is being investigated. This report presents the state of the art in the nuclear production of hydrogen and describes the scientific and technical challenges associated with it.
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Pyrochemical Separations in Nuclear Applications
A Status Report
English, 180 pages, published: 06/25/04
NEA#5427, ISBN: 92-64-02071-3
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea5427-pyrochemical.pdf
The treatment of spent nuclear fuel is presently performed by the industry using different aqueous chemical processes. Alternative dry processes, using pyrochemical methods, are beginning to receive greater attention due to their potential advantages for more compact reprocessing plant designs, as well as for reduced criticality and radiation dose risks.

Effective transmutation of long-lived fission products and minor actinides will be based in future on multi-recycling of the fuel with very high burn-up and short cooling times, conditions for which pyrochemical methods offer various advantages over traditional aqueous processes. Closed nuclear fuel cycles, considered for the future generation of nuclear reactors, could also benefit from pyrochemical reprocessing methods. Studies of pyrochemical processes have so far been carried out at laboratory level. Much R&D work will still be required in order to upgrade these processes to the level of current industrial aqueous processing.

This publication describes ongoing national programmes, collaborative international activiities, present research needs and future applications for pyrochemical methods, used in the treatment of irradiated nuclear fuel. It will be of particular interest to nuclear scientists involved in the development of advanced fuel cyles.
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Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 6
Workshop Proceedings, Stanford, California, USA, 10-12 April 2002
English, 448 pages, published: 07/22/04
NEA#3828, ISBN: 92-64-01733-X
Available online at: http://www.oecd-nea.org/science/pubs/2004/3828-SATIF-6.pdf
Particle accelerators are used today for an increasing range of scientific and technological applications. They are very powerful tools for investigating the origin and structure of matter, and for improving understanding of the interaction of radiation with materials, including the transmutation of nuclides and the beneficial or harmful effects of radiation. Particle accelerators are used to identify properties of molecules that can be used in pharmacy, for medical diagnosis and therapy, and for biophysics studies.

Particle accelerators must be operated in safe ways that protect the operators, the population and the environment. New technological and research applications give rise to new issues in radiation shielding. These workshop proceedings review the state of the art in radiation shielding of accelerator facilities and irradiated targets. They also evaluate advancements and discuss the additional developments required to meet radiation protection needs.
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The Need for Integral Critical Experiments with Low-moderated MOX Fuels
Workshop Proceedings, Paris, France, 14-15 April 2004
English, 220 pages, published: 07/19/04
NEA#5668, ISBN: 92-64-02078-0
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea5668-mox.pdf
The use of MOX fuel in commercial reactors is a means of burning plutonium originating from either surplus weapons or reprocessed irradiated uranium fuel. This requires the fabrication of MOX assemblies on an industrial scale.

The OECD/NEA Expert Group on Experimental Needs for Criticality Safety has highlighted MOX fuel manufacturing as an area in which there is a specific need for additional experimental data for validation purposes. Indeed, integral experiments with low-moderated MOX fuel are either scarce or not sufficiently accurate to provide an appropriate degree of validation of nuclear data and computer codes. New and accurate experimental data would enable a better optimisation of the fabrication process by decreasing the uncertainties in the determination of multiplication factors of configurations such as the homogenisation of MOX powders.

This report contains the proceedings of a workshop organised by the OECD/NEA Nuclear Science Committee. Issues debated include the expression of research needs, proposals of experimental programmes and prospects for an international co-operative programme to address these needs.

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Actinide and Fission Product Partitioning and Transmutation
Seventh Information Exchange Meeting, Jeju, Republic of Korea, 14-16 October 2002
English, 176 pages, published: 08/28/03
NEA#4454, ISBN: 92-64-02125-6
Available online at: http://www.oecd-nea.org/pt/docs/iem/jeju02/welcome.html
During the last decade interest in partitioning and transmutation (P&T) has grown in many countries around the world. In the years to come, P&T is expected to be one of the key technologies for nuclear waste management, together with geological disposal. In order to provide experts a forum to present and discuss state-of-the-art developments in the P&T field, the OECD Nuclear Energy Agency (NEA) has been holding biennial information exchange meetings on actinide and fission product partitioning and transmutation since 1990.

This book and its enclosed CD-ROM contain the proceedings of the 7th Information Exchange Meeting held in Jeju, Republic of Korea, on 14-16 October 2002. The meeting covered the broad spectrum of developments in the field, such as the role of P&T in advanced nuclear fuel cycles; developments in partitioning; developments in accelerators, materials and fuels; the performance of transmutation systems and their safety; R&D needs, including benchmarks, data improvement and experiments; and the role of international collaboration. More than 100 papers were presented during the meeting. These proceedings also contain a summary of the panel discussion on perspectives for the future development of P&T.
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Benchmark on Beam Interruptions in an Accelerator-driven system
Final Report on Phase I Calculations
English, 28 pages, published: 08/08/03
NEA#3136, ISBN: 92-64-02138-8
Available online at: http://www.oecd-nea.org/science/docs/2003/nsc-doc2003-17.pdf
In accelerator-driven system (ADS) development, it is important to evaluate temperature variations caused by beam trips as they can result in a temperature transient that would lead to thermal fatigue in the structural components of the subcritical system. A series of benchmarks is therefore being organised by the OECD Nuclear Energy Agency (NEA) for a lead-bismuth-cooled and MOX-fuelled accelerator-driven system.

This report provides a comparative analysis of the Phase I calculation results of the beam trip transient benchmark. In subsequent phases of the benchmark, temperature transients in different power densities and under irradiated fuel conditions will also be investigated. This report and those to follow will be of particular interest to ADS designers, including subcritical system physicists as well as accelerator scientists.
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Benchmark on Deterministic Transport Calculations Without Spatial Homogenisation
A 2-D/3-D MOX Fuel Assembly Benchmark
English, 152 pages, published: 08/08/03
NEA#3135, ISBN: 92-64-02139-6
Available online at: http://www.oecd-nea.org/science/docs/2003/nsc-doc2003-16.pdf
One of the important issues regarding deterministic transport methods for whole core calculations is that homogenised techniques can introduce errors into results. On the other hand, with modern computation abilities, direct whole core heterogeneous calculations are becoming increasingly feasible.

This report provides an analysis of the results obtained from a challenging benchmark on deterministic MOX fuel assembly transport calculations without spatial homogenisation. A majority of the participants obtained solutions that were more than acceptable for typical reactor calculations. The report will be of particular interest to reactor physicists and transport code developers.
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Burn-up Credit Criticality Benchmark - Phase IV-A
Phase IV-A: Reactivity Prediction Calculations for Infinite Arrays of PWR MOX Fuel Pin Cells
English, 88 pages, published: 05/16/03
NEA#3694, ISBN: 92-64-02123-X
Available online at: http://www.oecd-nea.org/science/docs/2003/nsc-doc2003-3.pdf
The OECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs.

The report summarises and analyses the solutions to the specified exercises provided by 37 contributors from 10 countries. The exercises were based upon the calculation of infinite PWR fuel pin cell reactivity for fresh and irradiated MOX fuels with various MOX compositions, burn-ups and cooling times. In addition, several representations of the MOX fuel assembly were tested in order to check various levels of approximations commonly used in reactor physics calculations.
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Burn-up Credit Criticality Benchmark - Phase IV-B
Phase IV-B: Results and Analysis of MOX Fuel Depletion Calculations
English, 180 pages, published: 05/16/03
NEA#3709, ISBN: 92-64-02124-8
Available online at: http://www.oecd-nea.org/science/docs/2003/nsc-doc2003-4.pdf
The OECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide (UOX) fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs.

The exercises consisted of inventory calculations of MOX fuels for two initial plutonium compositions. The depletion calculations were carried out using three representations of the MOX assemblies and their interface with UOX assemblies. This enabled the investigation of the spatial and spectral effects during the irradiation of the MOX fuels.
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CINDA 2003
The Index to Literature and Computer Files on Microscopic Neutron Data
English, 616 pages, published: 10/27/03
NEA#4331, ISBN: 92-64-02144-2, ISSN: 1011-2545
Available online at: http://www.oecd-nea.org/cinda/cd-2003.html
CINDA, the Computer Index of Neutron Data, contains bibliographical references to measurements, calculations, reviews and evaluations of neutron cross-sections and other microscopic neutron data; it also includes index references to computer librarires of numerical neutron data available from four regional neutron data centres.

The CINDA bibliography allows its users to find the references to specific types of cross-section information or other microscopic data from neutron-induced reactions, for any given targets nucleus. In this publicaton CINDA entries are sorted first by element and mass number and then by cross-section or other quantity. Within these isotopes and quantity groups, the entries are sorted by date of publication.
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International Evaluation Co-operation (Vol. 9)
Fission Neutron Spectra of Uranium-235 (Volume 9)
English, 28 pages, published: 07/15/03
NEA#3693, ISBN: 92-64-02134-5
Available online at: http://www.oecd-nea.org/science/wpec/volume9/volume9.pdf
This report has been prepared by Subgroup 9 which was set up in 1998 with the aim of investigating discrepancies found between microscopic and macroscopic data for the uranium-235 fission neutron spectrum. In addition, it was noted that the most recent evaluation of this spectrum had been performed in 1988 and had been based on only one experiment. It was thus felt necessary to review the existing evaluations, taking into account new experimental data and improved calculations methods.
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International Nuclear Data Evaluation Co-operation - CD-ROM
Complete Collection of Published Reports as of October 2003
English, 1 pages, published: 10/10/03
NEA#4968
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea4968-data-evaluation.html
The NEA International Nuclear Data Evaluation Co-operation programme brings together evaluation projects being carried out in Japan (JENDL), the United States (ENDF), western Europe (JEFF) and non-OECD countries (BROND, CENDL and FENDL). The Nuclear Data Section of the International Atomic Energy Agency (IAEA) sponsors the participation of evaluation projects from non-OECD countries.

The Co-operation programme was established to promote the exchange of information on nuclear data evaluations, measurements, nuclear model calculations, validation, and related topics, and to provide a framework for co-operative activities between the participating projects. The Co-operation programme assesses needs for nuclear data improvements and addresses those needs by initiating joint evaluation and/or measurement efforts. Expert groups are established to solve specific common nuclear data problems. Each expert group produces a final report of its findings.

The present CD-ROM contains a full collection of the expert group reports as of October 2003.
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PENELOPE 2003 - A Code System for Monte Carlo Simulation of Electron and Photon Transport
Workshop Proceedings, Issy-les-Moulineaux, France, 7-10 June 2003
English, 254 pages, published: 10/20/03
NEA#4488, ISBN: 92-64-02145-0
Available online at: http://www.oecd-nea.org/dbprog/penelope-2003.pdf
Radiation is used in many applications of modern technology. 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, as well as radiation damage and shielding.

These proceedings contain the extensively revised teaching notes of the second workshop/training course on PENELOPE held in 2003, along with a detailed description of the improved physics models, numerical algorithms and structure of the code system.
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Physics of Plutonium Recycling - Vol. VII
Volume VII: BWR MOX Benchmark - Specification and Results
English, 150 pages, published: 03/10/03
NEA#3038, ISBN: 92-64-19905-5
Available online at: http://www.oecd-nea.org/science/pubs/2003/3038-physics-plutonium-recycling-vol.7.pdf
The commercial recycling of plutonium as PUO2/UO2 mixed-oxide (MOX) fuel is an established practice in pressurised water reactors (PWRs) in several countries, the main motivation being the consumption of plutonium arising from spent fuel reprocessing. Although the same motivating factors apply in the case of boiling water reactors (BWRs), they have lagged behind PWRs for various reasons, and MOX utilisation in BWRs has been implemented in only a few reactors to date. One of the reasons is that the nuclear design of BWR MOX assemblies (or bundles) is more complex than that of PWR assemblies. Recognising the need and the timeliness to address this issue at the international level, the OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR) conducted a physics code benchmark test for a BWR assembly. This volume reports on the benchmark results and conclusions that can be drawn from it.
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Plutonium Management in the Medium Term
A Review by the OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR)
English, 72 pages, published: 11/03/03
NEA#4451, ISBN: 92-64-02151-5
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea4451-plutonium.pdf
The decision to re-use plutonium generated in thermal reactors is a strategic one for a utility, and is closely tied to the spent fuel management strategy. One option is to reprocess the spent fuel in existing reprocessing plants and immediately re-use the plutonium. Another option is to postpone re-use of the plutonium by placing the irradiated fuel in interim storage. The availability of different types of reactors determines the timescales for the present, medium-term or long-term future re-use of plutonium.

Current commercial reprocessing plants are all designed to separate the remaining plutonium at discharge for re-use. Historically, the rationale was to recover sufficient plutonium to enable a build-up of fast reactors, which were expected to be deployed as uranium reserves became scarce and prices rose. For a variety of reasons, but principally that of the low price of uranium ore, fast reactors have not yet been deployed commercially and projected timescales for doing so have been postponed everywhere.

Fast reactors are nevertheless still judged by many to be the most promising for long-term sustainability. Until such time as fast reactors are deployed commercially, however, the issue of how best to manage plutonium arisings from existing reprocessing plants remains. This report reviews the technical options available for plutonium management during this interim period. Presenting the consensus views of experts in this field, it is intended to serve as a reference source for researchers as well as utilities.
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Pressured Water Reactor Main Steam Line Break (MSLB) Benchmark
Volume IV: Results of Phase III on Coupled Core-plant Transient Modelling
English, 184 pages, published: 11/19/03
NEA#3129, ISBN: 92-64-02152-3
Available online at: http://www.oecd-nea.org/science/docs/2003/nsc-doc2003-21.pdf
This benchmark is based on a well-defined problem concerning a pressurised water reactor (PWR) main steam line break, which may occur as a consequence of the rupture of one steam line upstream of the main steam isolation valves. This event is characterised by significant space-time effects in the core caused by asymmetric cooling and an assumed stuck-out control rod during reactor trip. It is based on reference design and data from Unit 1 of the Three Mile Island nuclear power plant (TMI-1). It includes a description of the event sequence with set points of all activated system functions and typical plant conditions during the transient.

This report summarises the results contributed by international participants to Phase III of the exercise addressing best-estimate, coupled core-plant transient modelling.
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Research and Development Needs for Current and Future Nuclear Energy Systems (REPRINT)
English, 152 pages, published: 12/31/03
NEA#5188, ISBN: 92-64-02159-0
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea5188-research-needs.pdf

Other language(s):
- Français: Besoins de R-D pour les systèmes nucléaires actuels et futurs 
- Japanese: Research and Development Needs for Current and Future Nuclear Energy Systems (Japanese translation) 
Research capability and technical expertise in the area of nuclear science are needed to maintain a high level of performance and safety of present nuclear installations, as well as to develop future-generation nuclear power programmes.

The NEA Nuclear Science Committee (NSC) has completed a study on future research and development needs in specific areas of nuclear science, covering nuclear data; reactor physics and systems behaviour; and reactor fuels, materials and coolants.

This report contains information on past and present international R&D activities conducted under the aegis of the NSC and on R&D needs for new nuclear systems in different NEA member countries. Recommendations for further work in the areas mentioned above are also given in the report. Possible follow-up actions to these recommendations will be considered by the NSC.
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Utilisation and Reliability of High Power Proton Accelerators
Workshop Proceedings, Santa Fe, New Mexico, USA, 12-16 May 2002
English, 432 pages, published: 05/22/03
NEA#4310, ISBN: 92-64-10211-6
Available online at: http://www.oecd-nea.org/science/pubs/2003/4310-HPPA-USA 2002.pdf
R&D activities and construction plans related to high power proton accelerators are being considered in various countries to promote basic and applied sciences. This includes plans for accelerator-driven nuclear energy systems (ADS) for the transmutation of nuclear 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 integration of the accelerator with spallation targets and subcritical systems.

Both accelerator scientists and reactor physicists gathered together at an NEA workshop to discuss, inter alia, the reliability of the accelerator and the impact of beam interruptions on the design and performance of the ADS; spallation target design characteristics and their impact on the subcritical system design; safety and operational characteristics of a subcritical system driven by a spallation source; and test facilities.

These proceedings contain all the technical papers presented at the workshop as well as summaries of the discussions held during each technical session.

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A VVER-1000 LEU and MOX Assembly Computational Benchmark
Specification and Results
English, 156 pages, published: 09/27/02
NEA#3133, ISBN: 92-64-18491-0
Available online at: http://www.oecd-nea.org/science/docs/2002/nsc-doc2002-10.pdf
The United States and the Russian Federation have each agreed to dispose of 34 tonnes of weapons-grade plutonium that are in surplus of their defence needs. One effective way to do this is to convert the plutonium into mixed-oxide (MOX) fuel, burn it in a nuclear reactor and produce electricity with it. The Russian Federation intends to use this MOX fuel in both fast (BN-600) and light water (VVER-1000) reactors.

This report describes a benchmark study that compared the results obtained for low-enriched uranium (LEU) and MOX fuel in a VVER-1000. It contributes to the computer code certification process and to the verification of calculation methods used in the Russian Federation.
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Advanced Reactors with Innovative Fuels
Workshop Proceedings, Chester, United Kingdom, 22-24 October 2001
English, 512 pages, published: 08/30/02
NEA#3056, ISBN: 92-64-19847-4
Available online at: http://www.oecd-nea.org/science/pubs/2002/3056-advanced-reactors-2001.pdf
A new generation of nuclear reactor designs is being developed in order to meet the needs of the 21st century. In the short term, the most important objective is to improve competitiveness in the deregulated market. For this purpose evolutionary light water reactors are being developed and promoted actively. In the longer term, other requirements related to long-term sustainability will emerge, including the need to minimise the environmental burden passed on to future generations, the need to establish sustainability of the fuel and the need to minimise stocks of separated plutonium and their accessibility.

At this workshop, information on R&D activities for advanced reactor systems was exchanged and research areas in which international co-operation could be strengthened were identified, in particular the roles that could be played by existing experimental facilities and the possible needs for new infrastructure.
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Basic Studies on High-temperature Engineering
Second Information Exchange Meeting, Paris, France, 10-12 October 2001
English, 360 pages, published: 05/29/02
NEA#3632, ISBN: 92-64-19796-6
Available online at: http://www.oecd-nea.org/science/pubs/2002/3632-basic-studies.pdf
In response to increasing interest in high-temperature, gas-cooled reactors (HTGRs) in many countries and the need for improved materials for nuclear applications in high-temperature environments, the NEA organised a Second Information Exchange Meeting on Basic Studies in the Field of High-temperature Engineering. These proceedings provide an overview of the activities being carried out in eight countries, the improvement of material properties for HTGR application, in-core monitoring methods and properties of irradiated graphite, and HTGR fuel fabrication and performance.
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Comparison Calculations for an Accelerator-driven Minor Actinide Burner
English, 200 pages, published: 02/21/02
NEA#3128, ISBN: 92-64-18478-3
Available online at: http://www.oecd-nea.org/science/docs/2001/nsc-doc2001-13.pdf
International interest in accelerator-driven systems (ADS) has recently been increasing in view of the important role that these systems may play as efficient minor actinide and long-lived fission-product (LLFP) burners and/or energy producers with an enhanced safety potential. However, the current methods of analysis and nuclear data for minor actinide and LLFP burners are not as well established as those for conventionally fuelled reactor systems. Hence, in 1999, the OECD/NEA Nuclear Science Committee organised a benchmark exercise for an accelerator-driven minor actinide burner to check the performances of reactor codes and nuclear data for ADS with unconventional fuel and coolant. The benchmark model was a lead-bismuth-cooled subcritical system driven by a beam of 1 GeV protons.

This report provides an analysis of the results supplied by seven participants from eight countries. The analysis reveals significant differences in important neutronic parameters, indicating a need for further investigation of the nuclear data, especially minor actinide data, as well as the calculation methods. This report will be of particular interest to reactor physicists and nuclear data evaluators developing nuclear systems for nuclear waste management.
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Fission Gas Behaviour in Water Reactor Fuels
Workshop Proceedings, Cadarache, France, 26-29 September 2000
English, 564 pages, published: 02/04/02
NEA#3053, ISBN: 92-64-19715-X
Available online at: http://www.oecd-nea.org/science/pubs/2002/3053-fission-gas-behaviour.pdf
During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets.

These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors.
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International Evaluation Co-operation
Delayed Neutron Data for the Major Actinides (Volume 6)
English, 132 pages, published: 10/11/02
NEA#3209
Volume of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume6/volume6.pdf
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PENELOPE 2001 - A Code System for Monte Carlo Simulation of Electron and Photon Transport
Workshop Proceedings, Issy-les-Moulineaux, France, 5-7 November 2001
English, 250 pages, published: 01/11/02
NEA#3388, ISBN: 92-64-18475-9
Available online at: http://www.oecd-nea.org/dbprog/penelope-2001.pdf
Radiation is used in many applications of modern technology. 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 teaching notes of a recent workshop/training course on PENELOPE, with a detailed description of the physics, numerical algorithms and structure of the code system.
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Physics of Plutonium Recycling
Multiple Pu Recycling in Advanced PWRs - Volume VI
English, 162 pages, published: 10/23/02
NEA#3037, ISBN: 92-64-19957-8
Available online at: http://www.oecd-nea.org/science/pubs/2002/3037-physics-plutonium-recycling vol.6.pdf
Although the recycling of plutonium as thermal mixed-oxide (MOX) fuel in pressurised water reactors (PWRs) is now well-established on a commercial scale, many physics questions remain. The main question addressed in this report is the number of times plutonium can effectively be recycled in a PWR.

This report describes in particular an exercise based on realistic, multiple-recycle scenario, which followed plutonium through five generations of recycling in a PWR. It considered both a standard PWR design currently in use and a highly moderated design. The latter is a possible option for a dedicated, MOX-fuelled PWR in which it would be possible to optimise the moderation for plutonium. The study of these two designs in parallel has provided a better understanding of their relative merits, as well as insight into the limitations of multiple recycling and the long-term toxicity of fission products and actinides.
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Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark
Volume III: Results of Phase 2 on 3-D Core Boundary Conditions Modelling
English, 172 pages, published: 11/22/02
NEA#3117, ISBN: 92-64-18495-3
Available online at: http://www.oecd-nea.org/science/docs/2002/nsc-doc2002-12.pdf
This benchmark is based on a well-defined problem concerning a pressurised water reactor (PWR) main steam line break, which may occur as a consequence of the rupture of one steam line upstream of the main steam isolation valves. This event is characterised by significant space-time effects in the core caused by asymmetric cooling and an assumed stuck-out control rod during reactor trip. It is based on reference design and data from the Three Mile Island Unit 1 Nuclear Power Plant (TMI-1). It includes a description of the event sequence with set points of all activated system functions and typical plant conditions during the transient.

This report summarises the results contributed by international participants concerning Phase II of the exercise: a coupled 3-D neutronics/core thermal-hydraulics response evaluation using inlet and outlet core transient boundary conditions.
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Speciation, Techniques and Facilities for Radioactive Materials at Synchroton Light Sources
Workshop Proceedings, Grenoble, France, 10-12 September 2000
English, 380 pages, published: 06/20/02
NEA#3054, ISBN: 92-64-18485-6
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea3054.html
This NEA Workshop and Euroconference was the second in a series devoted to the application of synchrotron-based techniques to radionuclide and actinide sciences. The unique properties of synchrotron radiation allow one to obtain information about the molecular structure of radionuclides and actinide species, which is essential 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, and development of effective separation technologies, as well as radiopharmaceutical chemistry.

These proceedings contain the abstracts and some of the full papers presented at the meeting. In addition to presenting the latest experimental and theoretical results, the meeting was aimed at providing opportunities for learning and scientific discussions between experts in the field and young scientists.
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The Use of Thermodynamic Databases in Performance Assessment
Workshop Proceedings, Barcelona, Spain, 29-30 May 2001
English, 216 pages, published: 08/29/02
NEA#3055, ISBN: 92-64-19846-6
Available online at: http://www.oecd-nea.org/science/pubs/2002/3055-use-thermodynamic-databases.pdf
Performance assessment of repository concepts for the geological disposal of long-lived radioactive waste relies on the availability of thermodynamic data for many radionuclides and other elements under a wide range of physico-chemical conditions. For the past ten years, the OECD Nuclear Energy Agency (NEA) has been co-ordinating a multinational effort to produce a database of selected thermochemical values that would satisfy the requirements of the various national programmes in Member countries. This project is known as the NEA thermochemical Database (TDB) Project.

This publication contains the full papers and summary discussion records of a workshop attended by scientists active in the field of chemical thermodynamics and experts in repository performance assessment who use the thermochemical databases for their evaluations. During the workshop, participants discussed current experimental and theoretical standpoints, new data requirements and the peculiarities of their application in performance assessment.
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VVER-1000 Coolant Transient Benchmark (Vol. I)
PHASE 1 (V1000CT-1), Vol. I: Main Coolant Pump (MCP) Switching On - Final Specifications
English, 172 pages, published: 12/31/02
NEA#4128, ISBN: 92-64-18496-1
Available online at: http://www.oecd-nea.org/science/docs/2002/nsc-doc2002-6.pdf
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 reactor 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 describes the specification of such a benchmark. The transient addressed is caused by the switching on of a main coolant pump when 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.

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Boiling Water Reactor Turbine Trip (TT) Benchmark
Volume I: Specifications
English, 96 pages, published: 07/20/01
NEA#3115, ISBN: 92-64-18470-8
Available online at: http://www.oecd-nea.org/science/docs/2001/nsc-doc2001-1.pdf
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 nuclear 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 (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for the purpose.

The present volume describes the specification of such a benchmark. The transient addressed is a turbine trip (TT) in a BWR 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 plant make the present benchmark very valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4).
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Chemical Thermodynamics of Neptunium and Plutonium (Volume 4)
Published by Elsevier
English, published: 06/06/01
NEA#3114
Ex-sale now free
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Evaluation of Speciation Technology
Workshop Proceedings, Tokai-mura, Ibaraki, Japan, 26-28 October 1999
English, 436 pages, published: 05/09/01
NEA#3168, ISBN: 92-64-18667-0
Available online at: http://www.oecd-nea.org/science/pubs/2001/3168-evaluation-speciation.pdf
It has been widely recognised among researchers that speciation data are essential for proper and reliable modelling of radionuclide behaviour, which is studied inter alia in the context of radioactive waste management. Participants at the OECD/NEA workshop on "Evaluation of Speciation Technology" reviewed the various techniques used to identify different species of actinide and fission product elements present in nuclear waste and nuclear reprocessing streams. The review takes into account the advantages, disadvantages and limitations of the various methods in relation to their field of application. Recommendations for future R&D are also provided. These proceedings will primarily be of interest to chemists specialised in separation techniques and radioactive waste management experts.
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Forsmark 1 & 2 Boiling Water Reactor Stability Benchmark
Time Series Analysis Methods for Oscillations During BWR Operation: Final Report
English, 150 pages, published: 06/20/01
NEA#3116, ISBN: 92-64-18469-4
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea3116-forsmark.pdf
Events involving unnoticed power oscillations have occurred at different boiling water reactors (BWRs) in the past, and have led to the implementation of interim corrective actions to avoid their repetition. Despite these measures, power oscillations continue to occur. In response to this situation, a great deal of research and analytical activities have been undertaken to improve the knowledge of the underlying phenomenology, and to define final solutions to handle this type of event.

An OECD/NEA expert group has carried out studies in which the predictive capability of the codes and models for stability analysis are compared. This report provides the results of a specific study investigating the possibility of determining the main stability parameters from the neutronic signals time series with sufficient reliability and accuracy. It is based on a series of six complex cases derived from measurements carried out at the Forsmark nuclear power plant in Sweden.
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International Evaluation Co-operation
Evaluation Method of Inelastic Scattering Cross-sections for Weakly Absorbing Fission-product Nuclides (Volume 10)
English, 100 pages, published: 09/21/01
NEA#3119
Available online at: http://www.oecd-nea.org/science/wpec/volume10/volume10.pdf
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International Handbook of Evaluated Criticality Safety Benchmark Experiments
A Project by the NEA Nuclear Science Committee
English, published: 10/19/01
NEA#3328
Available online at: http://www.oecd-nea.org/science/wpncs/icsbep/
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JANIS - A Java-based Nuclear Data Display Program
English, published: 10/04/01
NEA#3268
Available online at: http://www.oecd-nea.org/janis/welcome.html
JANIS (Java-based nuclear information software) is a display program designed to facilitate the visualisation and manipulation of nuclear data. Its objective is to allow the user of nuclear data to access numerical values and graphical representations without prior knowledge of the storage format. It offers maximum flexibility for the comparison of different nuclear data sets.
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JANIS Leaflet
English, published: 09/07/01
NEA#3488
Available online at: http://www.oecd-nea.org/janis/welcome.html
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JEFF Reports
Complete Collection of JEFF Reports - Numbers 1-18
English, published: 09/07/01
NEA#3249
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/nea3249-jeff-report-cd.html
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NEA Nuclear Model and Code Comparisons
Complete Collection of the Report 1982-1998
English, published: 09/07/01
NEA#3248
Available online at: http://www.oecd-nea.org/databank/nea3248-nuclear-model-codecomp.html
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Nuclear Production of Hydrogen
First Information Exchange Meeting, Paris, France, 2-3 October 2000
English, 244 pages, published: 06/05/01
NEA#3188, ISBN: 92-64-18696-5
Available online at: http://www.oecd-nea.org/science/pubs/2001/3188-production-hydrogen.pdf
Hydrogen has the potential to play an important role as a sustainable and environmentally acceptable energy source in the 21st century. However, hydrogen does not exist as a gas on earth and thus has to be produced from, for example, water or natural gas by different separation techniques. One way to do so would be to use nuclear-produced energy or heat in this separation process. The present publication gives an overview of the advancements in the scientific and technological fields related to the nuclear production of hydrogen.
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Pyrochemical Separations
Workshop Proceedings, Avignon, France, 14-16 March 2000
English, 332 pages, published: 02/14/01
NEA#2988, ISBN: 92-64-18443-0
Available online at: http://www.oecd-nea.org/science/pubs/2001/2988-pyrochemical-separations.pdf
The industrial treatment of spent nuclear fuel is presently performed using different wet chemical processess. Alternative dry processes, using pyrochemical methods, have received some attention due to their potential advantages in terms of plant design and criticality safety, as well as radiation dose.

Recent progress in the transmutation of long-lived fission products and minor actinides has brought renewed interest in pyrochemical methods, as effective transmutation will be based on multi-recycling of the fuel with very high burn-up and short cooling times, conditions under which pyrochemical methods offer various advantages over wet processes.

Studies of pyrochemical processes have so far been carried out at laboratory level. Considerable R&D work is still required in order to upgrade these processes to the current level of industrial aqueous processing.
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Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 5
Workshop Proceedings, Paris, France, 18-21 July 2000
English, 426 pages, published: 05/30/01
NEA#3052, ISBN: 92-64-18691-3
Available online at: http://www.oecd-nea.org/science/pubs/2001/3052-SATIF-5.pdf
Over the last 50 years particle accelerators have evolved from simple devices to powerful machines, and will continue to have an important impact on research, technology and lifestyle. Today, they cover a wide range of applications, from television and computer displays in households to investigating the origin and structure of matter. It has become common practice to use particle accelerators for material science and medical applications.

In recent years, requirements from new technological and research applications have emerged, giving rise to new radiation shielding aspects and problems. These workshop proceedings review recent progress in radiation shielding of accelerator facilities, evaluating advancements and discussing further developments needed with respect to international co-operation in this field.
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Utilisation and Reliability of High Power Proton Accelerators
Workshop Proceedings, Aix-en-Provence, France, 22-24 November 1999
English, 476 pages, published: 10/25/01
NEA#3051, ISBN: 92-64-18749-9
Available online at: http://www.oecd-nea.org/science/pubs/2001/3051-HPPA-France 1999.pdf
High power proton accelerators are being studied for their potential use in the transmutation of nuclear waste. The Second Workshop on Utilisation and Rerliability of High Power Proton Accelerators, organised by the NEA Nuclear Science Committee, placed special emphasis on accelerator-driven system (ADS) concepts comprising a sub-critical reactor coupled with a high power accelerator.

The information provided in these proceedings will primarily be of interest to scientists working on accelerator-driven systems, but also to those involved in the construction of high power accelerators.

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3-D Radiation Transport Benchmarks for Simple Geometries with Void Regions
English, 38 pages, published: 12/08/00
NEA#2828, ISBN: 92-64-18274-8
Available online at: http://www.oecd-nea.org/science/docs/2000/nsc-doc2000-4.pdf
Industry requires well-validated computation methods and computer codes for its nuclear applications. The predictive power of such tools must be established and users must be confident of their results. Model refinement requires that increasingly sophisticated tools be used. Moreover, the computing power available today no longer justifies a number of geometrical simplifications.

This report describes the results of challenging international benchmarks in three-dimensional radiation transport that contribute to the evaluation and validation of state-of-the-art computation methods and computer codes. It will be of particular interest to reactor physicists and radiation shielding specialists.
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Basic Studies on High-Temperature Engineering
First Information Exchange Meeting, Paris, France, 27-29 September 1999
English, 401 pages, published: 01/01/00
NEA#2408, ISBN: 92-64-17695-0
Available online at: http://www.oecd-nea.org/science/docs/pubs/hightemp.pdf
In response to increasing interest in high-temperature, gas-cooled reactors (HTGRs) and the need for improved knowledge of materials for nuclear applications that resist high temperatures, the NEA organised a first information exchange meeting on basic studies in the field of high-temperature engineering. The proceedings of the meeting cover studies on irradiation effects on advanced materials, safety-related behaviour of HTGRs and in-pile reactor instrumentation development. They also include recommendations for further promotion of international collaboration.
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Benchmark Calculations of Power Distribution Within Fuel Assemblies
Phase II: Comparison of Data Reduction and Power Reconstruction Methods in Production Codes
English, 234 pages, published: 01/01/00
NEA#2751, ISBN: 92-64-18275-6
Available online at: http://www.oecd-nea.org/science/docs/2000/nsc-doc2000-3.pdf
Systems loaded with plutonium in the form of mixed-oxide (MOX) fuel show somewhat different neutronic characteristics compared with those using conventional uranium fuels. In order to maintain adequate safety standards, it is essential to accurately predict the characteristics of MOX-fuelled systems and to further validate both the nuclear data and the computation methods used.

A computation benchmark on power distribution within fuel assemblies to compare different techniques used in production codes for fine flux prediction in systems partially loaded with MOX fuel was carried out at an international level. It addressed first the numerical schemes for pin power reconstruction, then investigated the global performance including cross-section data reduction methods. This report provides the detailed resutls of this second phase of the benchmark. The analysis of the results revealed that basic data still need to be improved, primarily for highter plutonium isotopes and minor actinides.
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Benchmark on the VENUS-2 MOX Core Measurements
English, 192 pages, published: 12/22/00
NEA#2848, ISBN: 92-64-18276-4
Available online at: http://www.oecd-nea.org/science/docs/2000/nsc-doc2000-7.pdf
The plutonium produced during the operation of commercial power plants and made 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 MOX-fuelled systems. A series of theoretical physics benchmarks and multiple recycling issues of various MOX-fuelled systems have been studied by the NEA. This led to many improvements and clarifications in nuclear data libraries and calculation methods. The final validation requires linking those findings to data from experiments. Hence, the first experiment-based benchmark using the two-dimensional VENUS-2 MOX core measurement data was launched in May 1999.

This report provides an analysis of the results supplied by 12 participants from 10 countries. The comparison of the latest nuclear data libraries and of different calculation methods - including stochastic Monte Carlo and deterministic transport/diffusion methods - is presented. The report will be of particular interest to reactor physicists and designers as well as to nuclear power plant utilities.
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CD-CINDA 2000
Index to Literature and Computer Files on Microscopic Neutron Data
English, published: 12/31/00
NEA#2929
Available online at: http://www.oecd-nea.org/cinda/cd.html
The Nuclear Energy Agency (NEA) operates as a special agency of the Organisation for Economic Co-operation and Development (OECD), an intergovernmental body based in Paris. The main objective of the NEA is to assist its Member countries in maintaining and further developing, through international co-operation, the scientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclear energy for peaceful purposes. The NEA Data Bank deals more specifically with nuclear science aspects and data needs covering the whole fuel cycle and some non-energy applications. The Data Bank is a member of the worldwide nuclear data exchange network and supplies nuclear data and computer programs to its participating countries.
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Calculations of Different Transmutation Concepts
An International Benchmark Exercise
English, 157 pages, published: 01/01/00
NEA#2148, ISBN: 92-64-17638-1
Available online at: http://www.oecd-nea.org/science/docs/2000/nsc-doc2000-6.pdf
In April 1996, the NEA Nuclear Science Committee (NSC) Expert Group on Physics Aspects of Different Transmutation Concepts launched a benchmark exercise to compare different transmutation concepts based on pressurised water reactors (PWRs), fast reactors, and an accelerator-driven system. The aim was to investigate the physics of complex fuel cycles involving reprocessing of spent PWR reactor fuel and its subsequent reuse in different reactor types. The objective was also to compare the calculated activities for individual isotopes as a function of time for different plutonium and minor actinide transmutation scenarios in different reactor systems.

This report gives the analysis of results of the 15 solutions provided by the participants: six for the PWRs, six for the fast reactor and three for the accelerator case. Various computer codes and nuclear data libraries were applied.
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Core Monitoring for Commercial Reactors: Improvements in Systems and Methods
Workshop Proceedings, Stockholm, Sweden, 4-5 October 1999
English, 291 pages, published: 01/01/00
NEA#2208, ISBN: 92-64-17659-4
Available online at: http://www.oecd-nea.org/science/pubs/2000/2208-core-monitoring.pdf
The opening of energy markets is leading to increased competition, and the nuclear power industry must adapt if it is to meet this challenge. Internationally discussions are taking place among government authorities and electric utilities and vendors on how to deal with the rapid technical development and optimisation of nuclear fuel and its utilisation under new, more aggressive fuel management strategies. Improving reactor core monitoring systems is an important part of this process.

Participants in a recent NEA workshop discussed how instrumentation, methods and models used in core monitoring can be validated or, if needed, improved and further developed to provide more reliable and detailed information on local power in the core and on other parameters indirectly affecting fuel duty.

This book shows how the core monitoring system can be used to support reactor operation in normal and anticipated transient modes and to supply data used to derive initial key core parameters for transient and accident analysis.
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Evaluation and Analysis of Nuclear Resonance Data
JEFF Report 18
English, 126 pages, published: 01/01/00
NEA#2748, ISBN: 92-64-18272-1
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-18/jeff18.pdf
Nuclear Data are fundamental to the development and application of all nuclear sciences and technologies. Preserving nuclear data knowledge in a field from which a large number of specialists have recently retired is also important, and this report aims to help the preservation effort.

The report provides a comprehensive presentation of the nuclear data evaluation process in the resonance energy range. The mathematical basis and the physical theories necessary for the experimental data analysis are presented in detail.

This report will be useful for experimentalists and evaluators involved in the preparation of nuclear data. It will also be of value for nuclear data users who are interested in understanding the process of preparation of these data.
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International Evaluation Co-operation
Processing and Validation of Intermediate Energy Evaluated Data Files (Volume 14)
English, 34 pages, published: 12/28/00
NEA#3208
Volume of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume14/volume14.pdf
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Prediction of Neutron Embrittlement in the Reactor Pressure Vessel
VENUS-1 and VENUS-3 Benchmarks
English, 265 pages, published: 01/01/00
NEA#2128, ISBN: 92-64-17637-3
Available online at: http://www.oecd-nea.org/science/docs/2000/nsc-doc2000-5.pdf
The OECD/NEA Task Force on Computing Radiation Dose and Modelling of Radiation-Induced Degradation of Reactor Components (TFRDD) launched two international blind intercomparison exercises to examine the current computation techniques used in NEA Member countries for calculating neutron and gamma doses to reactor components. Various methodologies and different nuclear data were applied to predict dose rates in the Belgian VENUS-1 and three-dimensional VENUS-3 configurations for comparison with measured data.

This report provides the detailed results from the two benchmarks. The exercise revealed that three-dimensional neutron fluence calculations provide results that are significantly more accurate than those obtained from two-dimensional calculations. Performing three-dimensional calculations is technically feasible given the power of today's computers.
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Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark
Volume II: Results of Phase I on Point Kinetics
English, 136 pages, published: 12/22/00
NEA#2868, ISBN: 92-64-18280-2
Available online at: http://www.oecd-nea.org/science/docs/2000/nsc-doc2000-21.pdf
The benchmark is based on a well-defined problem concerning a PWR main steam line break, which may occur as a consequence of the rupture of one steam line upstream of the main steam isolation valves. This event is characterised by significant space-time effects in the core caused by asymmetric cooling and an assumed stuck-out control rod during reactor trip. It is based on reference design and data from the Three Mile Island Unit 1 Nuclear Power Plant (TMI-1). It includes a description of the event sequence with set points of all activated system functions and typical plant conditions during the transient.

This report summarises the results contributed by international participants concerning Phase I of the exercise: point kinetics simulation to test the primary and secondary system model responses.
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The JEF-2.2 Nuclear Data Library
JEFF Report 17
English, 253 pages, published: 01/01/00
NEA#2348, ISBN: 92-64-17686-1
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-17/
The JEF-2.2 library, the latest version in the Joint Evaluated Files series, is composed of sets of evaluated nuclear data, mainly for fission reactor applications. It contains a number of different data types, including neutron interaction data, radioactive decay data, fission yield data, thermal scattering law data and photo-atomic interaction data.

This report gives detailed information on JEF-2.2, including the origin of evaluations, measures of typical biases (calculation-experiment discrepancies) for different applications and indications on the changes needed to nuclear data in order to improve the predictive power of the file. The feedback contained herein will be used to prepare JEFF-3 (the Joint Evaluated Fission and Fusion file).

This report will be useful for scientists and engineers in national laboratories, universities and industry who use nuclear data constants. It is particularly suitable for those who work with computer codes utilising application libraries based on JEF-2.2.

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Advanced Reactors with Innovative Fuels
Workshop Proceedings, Villigen, Switzerland, 21-23 October 1998
English, 459 pages, published: 01/01/99
NEA#1747, ISBN: 92-64-17117-7
Available online at: http://www.oecd-nea.org/science/pubs/1999/1747-advanced-reactors-1998.pdf
Plutonium and minor actinide burning or recycling in thermal and fast reactors is being studied in many countries with the aim of maintaining and developing fuel cycle options which can be adjusted to changing demands and constraints. The challenge is to move towards an economically and socially sustainable nuclear energy system based on advanced reactors ? advanced water-cooled reactors, fast reactors and perhaps accelerator-based, hybrid reactors ? and new types of fuel cycles which help to minimise the waste arising. An additional issue concerns the availability of resources for the long-term future.

This workshop introduced new ideas on R&D activities and identified areas and research tasks relevant for the deployment of new systems and in which international co-operation can be strengthened. The roles played by existing experimental facilities as well as possible needs for new ones are discussed. The conclusions of the technical sessions are synthesised and the results of a round table discussion on international co-operation are presented.
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Data Bank Leaflet
English, published: 01/01/99
NEA#4208
Available online at: http://www.oecd-nea.org/dbdata/pubs/db-brochure.pdf
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Intercomparisons of Calculations Made for GODIVA and JEZEBEL
JEFF Report 16
English, 28 pages, published: 01/01/99
NEA#1988
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-16.pdf
This report gives details of intercomparisons studies undertaken to investigate the accuracy of calculation methods used for reactor physics studies. The same nuclear data library, JEF-2.2, was used in all the codes. The objective of the study was to estimate the uncertainties arising from approximations in the cross-section processing and neutronics methods, as well as to provide guidance on the sources of the differences. The intercomparisons studied were the small Los Alamos fast spectrum critical spheres, GODIVA and JEZEBEL.
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International Evaluation Co-operation
238U Capture and Inelastic Cross-Sections (Volume 4)
English, 48 pages, published: 01/01/99
NEA#1948
Available online at: http://www.oecd-nea.org/science/wpec/volume4/volume4.pdf
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International Evaluation Co-operation
Epithermal Capture Cross-Section of 235U (Volume 18)
English, 40 pages, published: 01/01/99
NEA#1907
Volume 18 of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume18/volume18.pdf
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International Evaluation Co-operation
Present Status of Minor Actinide Data (Volume 8)
English, 116 pages, published: 01/01/99
NEA#1947
Volume 8 of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume8/volume8.pdf
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Ion and Slow Positron Beam Utilisation
Workshop Proceedings, Costa da Caparica, Portugal, 15-17 September 1998
English, 236 pages, published: 01/01/99
NEA#1427, ISBN: 92-64-17025-1
Available online at: http://www.oecd-nea.org/science/pubs/1999/1427-ion-slow-positron.pdf
The use of ion beams in nuclear research is well established, with many facilities and networks of experts active in the field. Applications for ion beams are expanding, in particular in the development of new materials, biotechnology and the creation of new isotopes. Positron beams are likewise a very powerful tool for observing and influencing microscopic material structures, as well as for medical diagnosis.

The combined utilisation of ion and positron beams is expected to open up new horizons in the areas of material science and biotechnology. These proceedings provide an overview of the latest developments in this field, and highlight areas for future international co-operation.
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Light Water Reactor (LWR) Pin Cell Benchmark Intercomparisons
JEFF Report 15
English, 44 pages, published: 01/01/99
NEA#1867
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-15.pdf
Benchmark intercomparison studies are being undertaken to investigate the accuracy of calculation methods. The same nuclear data library, JEF-2.2, is used in all the codes. The objective has been to estimate the uncertainties arising from approximations in the cross-section processing and neutronics methods and to provide guidance on the sources of the differences. The intercomparisons studied in this phase have been simple models, LWR pin cells without leakage and also with leakage treated by means of a buckling.
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Physics and Fuel Performance of Reactor-Based Plutonium Disposition
Workshop Proceedings, Paris, France, 28-30 September 1998
English, 231 pages, published: 01/01/99
NEA#1467, ISBN: 92-64-17050-2
Available online at: http://www.oecd-nea.org/science/pubs/1999/1467-physics-fuel-performance.pdf
Following recent disarmament agreements, the Russian Federation and the USA have declared part of their stockpiles of weapons-grade plutonium as a surplus to their national defence needs. This material needs to be disposed of, and one of the suggested means of doing so is burning it in existing reactors and transforming the material into spent fuel. The experience in these two countries with mixed oxide fuel (MOX) is either dated or scarce. Several European countries and Japan, however, have acquired much experience in using MOX fuel in reactors which was shared at this important workshop.

This publication presents the workshop results which reviewed existing technical information from the civil nuclear power programmes that are benefical to weapons-grade plutonium disposition. It also proposes concrete actions that could help expedite this process in the near future.
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Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark
Volume I: Final Specifications
English, 101 pages, published: 01/01/99
NEA#1587
Available online at: http://www.oecd-nea.org/science/docs/1999/nsc-doc99-8.pdf
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Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF 4
Workshop Proceedings, Knoxville, Tennessee, USA, 17-18 September 1998
English, 305 pages, published: 01/01/99
NEA#1468, ISBN: 92-64-17044-8
Available online at: http://www.oecd-nea.org/science/pubs/1999/1468-SATIF-4.pdf
Over the last 50 years particle accelerators have evolved from simple devices to powerful machines, and will continue to have an important impact on research, technology and lifestyle. Today, they cover a wide range of applications, from television and computer displays in households to investigating the origin and structure of matter. It has become common practice to use particle accelerators for materials science and medical applications.

In recent years, requirements from new technological and research applications have emerged, giving rise to new radiation shielding aspects and problems.These workshop proceedings review recent progress in radiation shielding of accelerator facilities, evaluating advancements and discussing further developments needed with respect to international co-operation in this field.
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Speciation, Techniques and Facilities for Radioactive Materials at Synchrotron Light Sources
Workshop Proceedings, Grenoble, France, 4-6 October 1998
English, 343 pages, published: 01/01/99
NEA#1727
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea1727-speciation.pdf
The NEA Workshop and Euroconference on this subject was held in order to introduce the applications of synchrotron-based analytical techniques to scientists working in the environmental field or working with radionuclides. It was also aimed at providing a forum for teaching and for scientific discussion, as well as for establishing a possible co-operative scientific network.

These proceedings contain the abstracts and a selection of the papers presented at the meeting. They include introductions to synchrotron radiation techniques, results in the field of actinide chemistry and physics obtained at synchrotron light sources. Status reports on current and planned experimental activities at these installations are also provided.
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Utilisation and Reliability of High Power Proton Accelerators
Workshop Proceedings, Mito, Japan, 13-15 October 1998
English, 443 pages, published: 01/01/99
NEA#1627, ISBN: 92-64-17068-5
Available online at: http://www.oecd-nea.org/science/pubs/1999/1627-HPPA-Japan 1998.pdf
The use of high power particle accelerators in various areas of applied nuclear science is presented with special emphasis on accelerator driven reactor systems (ADS) for transmutation of nuclear waste. National programmes for the development of spallation neutron sources are presented and the performance and reliability of existing or planned accelerators for use in ADS are discussed. Effects, such as thermal shocks and material resistance, on the reactor part of an ADS from loss of accelerator beam are discussed more in greater detail.

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Evaluation of 242Pu Data for the Incident Neutron Energy Range 5-20 MeV
Bilingual, 129 pages, published: 01/01/98
NEA#1307
Available online at: http://www.oecd-nea.org/science/docs/1998/sen-nsc-wppr98-1.pdf
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International Evaluation Co-operation
Nuclear Models to 200 MeV for High-Energy Data Evaluations [Volume 12]
English, 20 pages, published: 01/01/98
NEA#1248
Volume of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume12/volume12.pdf
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International Evaluation Co-operation
Status of Pseudo-Fission-Product Cross-Sections for Fast Reactors [Volume 17]
English, 120 pages, published: 01/01/98
NEA#1251
Volume of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume17/volume17.pdf
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International Evaluation Co-operation
Effects of Shape Differences in the Level Densities of Three Formalisms on Calculated Cross-Sections [Volume 16]
English, 36 pages, published: 01/01/98
NEA#1250
Volume of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume16/volume16.pdf
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International Evaluation Co-operation
Intermediate Energy Data [Volume 13]
English, 22 pages, published: 01/01/98
NEA#1249
Volume of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume13/volume13.pdf
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JEF-PC 2.0 - A Personal Computer Program for Displaying Nuclear Data from the Joint Evaluated File Library
English, 113 pages, published: 01/01/98
NEA#666, ISBN: 92-64-15130-3
Ex-sale now free
The Joint Evaluated File (JEF) is a library of evaluated nuclear data, mainly for fission reactor applications.

JEF-PC provides, in a user-friendly manner, the ability to display selected data from the JEF-2.2 library in both numerical and graphical form.

No experience or familiarity with evaluated data formats is required; a display of the "Chart of the Nuclides" is used as the interface for the selection of specific nuclides and the corresponding data.
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Long-Lived Radionuclide Chemistry in Nuclear Waste Treatment
Workshop Proceedings, Villeneuve-lès-Avignon, France, 18-20 June 1997
English, 272 pages, published: 01/01/98
NEA#1106, ISBN: 92-64-16148-1
Out of print
The future management of nuclear wastes will possibly be safer and more economical if long-lived radionuclides are removed from the wastes before their conditioning and disposal. Such a separation strategy will minimise the volume of conditioned wastes requiring disposal in geologic repositories and, consequently, the cost of managing the waste will be reduced drastically. After their separation, the long-lived radionuclides could either be conditioned into suitable solid matrixes or converted into stable or short-lived radionuclides by nuclear means.

The objective of the workshop on "Long-Lived Radionuclide Chemistry in Nuclear Waste Treatment" was to provide up-to-date information on the chemistry of these hazardous elements to the community of chemists and chemical engineers in OECD Member countries concerned with the design of long-lived radionuclide separation processes. The workshop was attended by 82 paticipants from 11 countries. These proceedings include 24 presentations.
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SATIF-3 - Shielding Aspects of Accelerators, Targets and Irradiation Facilities
Proceedings of 3rd Specialists Meeting Sendai, Japan, 12-13 May 1997
English, 365 pages, published: 01/01/98
NEA#826, ISBN: 92-64-16071-X
Available online at: http://www.oecd-nea.org/science/pubs/1998/826-SATIF-3.pdf
Particle accelerators have evolved over the last 50 years from simple devices to powerful machines, and will continued to have an important impact on research, technology and lifestyle. Today, they cover a wide range of applications, from television and computer displays in households to the investigation of the origin and structure of matter. It has become common practice to use them for material science and medical applications.

In recent years, requirements from new technological and research applications have emerged, ginving rise to new radiation shielding aspects and problems. These proceedings review recent progress in radiation shielding of accelerator facilities, evaluate advancements and discuss further developments needed with respect to international co-operation in this field.
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Thermal Performance of High Burn-Up LWR Fuel
Seminar Proceedings, Cadarache, France, 3-6 March 1998
English, 389 pages, published: 01/01/98
NEA#1247, ISBN: 92-64-16957-1
Available online at: http://www.oecd-nea.org/science/pubs/1998/1247-thermal-cadarache-1998.pdf
Proper heat removal from the fuel is essential for the safe operation of nuclear reactors used for electricity production. As nuclear fuel is burnt it undergoes important changes, including a degradation of its thermal conductivity. This important phenomenon needs to be reliably predicted in order to make better use of the fuel, a factor which can help to achieve the economic competitiveness required by today's markets.

This report communicates the results of an international seminar which reviewed recent progress in the field of nuclear fuel thermal conductivity and sought to improve the models used in computer codes predicting thermal performance. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors.

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3-D Deterministic Radiation Transport Computer Codes
Features, Applications and Perspectives - Proceedings of an OECD Meeting held on 2-3 December 1996, OECD Château la Muette, Paris, France
English, published: 01/01/97
NEA#259, ISBN: 92-64-16020-5
Ex-sale now free
Improved prediction of the behaviour of systems we design and then set into operation is attained by science and technology through model refinement. Computer codes containing these models are essential for exploring and better exploiting the knowledge we have accumulated so far. Situations that are impossible or too costly to study experimentally can thus be analysed and systems optimised before they are set into operation. Three-dimensional deterministic radiation transport codes are tools that will increasingly contribute to the optimisation of complext systems in nuclear science as well as technology applications. These proceedings describe the state of the art in this field.
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3-D Deterministic Radiation Transport Computer Programs
Features, Applications and Perspectives
English, 427 pages, published: 01/01/97
NEA#526, ISBN: 92-64-16020-5
Ex-sale now free
Improved prediction of the behaviour of systems we design and then set into operation is attained by scinece and technology through model refinement. Comlputer codes containing these models are essential for exploring and better exploiting the knowledge we have accumulated so far. Situations that are impossible or too costly to study experimentally can thus be analysed and systems optimised before they are set into operation.

Three-dimensional deterministic radiation transport codes are tools that will increasingly contribute to the optimisation of complex systems in nuclear science as well as technololgy applications. These proceedings describe the state of the art in this field.
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Computing Radiation Dose to Reactor Pressure Vessel and Internals
State-of-the-art Report
English, 76 pages, published: 01/01/97
NEA#320
Available online at: http://www.oecd-nea.org/science/docs/1996/nsc-doc96-05.pdf
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Evaluation of Pu-242
English, published: 01/01/97
NEA#300
Available online at: http://www.oecd-nea.org/science/docs/1996/sen-nsc-wppr96-5.pdf
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In-Core Instrumentation and Reactor Core Assessment
Proceedings of a Specialist Meeting, Mito-shi, Japan, 16-17 October 1996
English, 426 pages, published: 01/01/97
NEA#347, ISBN: 92-64-15616-X
Available online at: http://www.oecd-nea.org/science/rsd/ic96/
Information on the conditions in the reactor core is essential for the safe and economic operation of nuclear reactors. These proceedings review improvements in the methods used to gather and interpret this information. An international selection of contributions from industry and research laboratories cover radiation sensors, safety analysis, system and validation, other miscellaneous measurements, core monitoring and core performance.
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Intermediate Energy Nuclear Data Benchmark Report
English, published: 01/01/97
NEA#305
Available online at: http://www.oecd-nea.org/science/docs/1997/nsc-doc97-1.html
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International Nuclear Data Measurement Activities - Newsletter No.2
English, published: 01/01/97
NEA#281
Available online at: http://www.oecd-nea.org/science/datameas/datameas2.html
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Modelling in Aquatic Chemistry
English, published: 01/01/97
NEA#291, ISBN: 92-64-15569-4
Ex-sale now free
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Neural Network Benchmark Report
English, published: 01/01/97
NEA#309
Available online at: http://www.oecd-nea.org/science/docs/1996/nsc-doc96-29.html
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Nucleon Nucleus Optical Model up to 200 MeV
Proceedings of a Specialist Meeting, Bruyères-le-Châtel, France, 13-15 November 1996
English, 235 pages, published: 01/01/97
NEA#528
Available online at: http://www.oecd-nea.org/science/om200/
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PWR Benchmark on Uncontrolled Rods Withdrawal at Zero Power
Final Report - September 1997
English, 77 pages, published: 01/01/97
NEA#406
Available online at: http://www.oecd-nea.org/science/docs/1996/nsc-doc96-20.pdf
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Update to JEF-PC Electronic Book
English, published: 01/01/97
NEA#260
Ex-sale now free

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Actinide Separation Chemistry in Nuclear Waste Streams and Materials
English, 115 pages, published: 01/01/96
NEA#163
Available online at: http://www.oecd-nea.org/science/docs/1997/nsc-doc97-19.pdf
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International Evaluation Co-operation
Comparison of Evaluated Data for Chromium-52, Iron-56 and Nickel-58 [Volume 1]
English, published: 01/01/96
NEA#172
Volume 1 of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume1/volume1.pdf
A Working Party on International Evaluation Co-operation was
established under the sponsorship of the OECD/NEA Nuclear Science Committee (NSC) to promote the exchange of information on nuclear data evaluations, validation, and related topics. Its aim is also to provide a framework for co-operative activities between members of the major nuclear data evaluation projects. This includes the possible exchange of scientists in order to encourage co-operation.

Requirements for experimental data resulting from this activity are
compiled. The Working Party determines common criteria for evaluated nuclear data files with a view to assessing and improving the quality and completeness of evaluated data.

The Parties to the project are: ENDF (United States), JEF/EFF (NEA Data Bank Member countries), and JENDL (Japan). Co-operation with evaluation projects of non-OECD countries are organised through the Nuclear Data Section of the International Atomic Energy Agency (IAEA).

* NEA/WPEC-1: Comparison of evaluated data for chromium-52, iron-56 and nickel-58 (Volume 1)

* NEA/WPEC-2: Generation of covariance files for iron-56 and natural iron (Volume 2)

* NEA/WPEC-3: Actinide data in the thermal range (Volume 3)

* NEA/WPEC-5: Plutonium-239 fission cross-section between 1 and 100 keV (Volume 5)

* NEA/WPEC-15: Cross-section fluctuations and self-shielding
effects in the unresolved resonance region (Volume 15)
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International Evaluation Co-operation
Cross-section Fluctuations and Self-shielding Effects in the Unresolved Resonance Region [Volume 15]
English, published: 01/01/96
NEA#176
Volume 15 of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume15/volume15.pdf
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International Evaluation Co-operation
Plutonium-239 Fission Cross-section Between 1 and 100 keV [Volume 5]
English, published: 01/01/96
NEA#175
Volume 5 of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume5/volume5.pdf
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International Evaluation Co-operation
Actinide Data in the Thermal Energy Range [Volume 3]
English, published: 01/01/96
NEA#174
Volume 3 of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume3/volume3.pdf
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International Evaluation Co-operation
Generation of Covariance Files for Iron-56 and Natural Iron [Volume 2]
English, published: 01/01/96
NEA#173
Volume 2 of the series: International Evaluation Co-operation
Available online at: http://www.oecd-nea.org/science/wpec/volume2/volume2.pdf
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SMORN-VII
Proceedings of the Seventh Symposium on Surveillance and Diagnostics in Nuclear Reactors
English, published: 01/01/96
NEA#217
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea0217-smorn7.html

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Scientific Issues in Fuel Behaviour
English, published: 01/01/95
NEA#213, ISBN: 92-64-14420-X
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea0213-fuel-2.pdf

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Blind Intercomparison of Nuclear Models for Predicting Charged Particle Emission
English, published: 01/01/94
NEA#166
Available online at: http://www.oecd-nea.org/science/docs/1993/nsc-doc93-4.pdf
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JEF Report 13: JEF 2.2 Radioactive Decay Data
English, published: 01/01/94
NEA#179
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-13.pdf
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JEF Report 14: Table of Simple Integral Neutron Cross-section Data from JEF-2.2, ENDF/B-VI, JENDL-3.2, BROND-2 and CENDL-2
English, published: 01/01/94
NEA#180
Available online at: http://www.oecd-nea.org/dbdata/nds_jefreports/jefreport-14.pdf
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NJOY and THEMIS Nuclear Data
Proceedings of a Seminar Saclay, France, 7-8 April 1992
English, published: 01/01/94
NEA#186
Available online at: http://www.oecd-nea.org/science/docs/pubs/nea0186-njoy.pdf
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Sensitivities of Calculated Cross-section of Fe56 to Model Parameters
English, published: 01/01/94
NEA#233
General Distribution Document Number:
OCDE/GD(94)21
Available online at: http://www.oecd-nea.org/science/docs/1993/nsc-doc93-05.pdf
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Summary of the Work of the NEANDC Task Force on U-238
English, published: 01/01/94
NEA#218
Available online at: http://www.oecd-nea.org/science/docs/1991/neandc1991-313-u.pdf

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Evaluation and Processing of Covariance Data
Proceedings of a Specialists' Meeting Oak Ridge National Laboratory, U.S.A., 7-9 October 1992
English, published: 01/01/93
NEA#167
Available online at: http://www.oecd-nea.org/science/docs/1993/nsc-doc93-3.pdf
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Results of LWR Core Transient Benchmarks
English, published: 01/01/93
NEA#196
Available online at: http://www.oecd-nea.org/science/docs/1993/nsc-doc93-25.pdf
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Strategic View on Nuclear Data Needs
English, published: 01/01/93
NEA#161
Available online at: http://www.oecd-nea.org/science/docs/pubs/data-strategy.html

Other language(s):
- Français: Evaluation de l'ensemble des besoins en matière de données nucléaires 

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Fission Product Nuclear Data
Proceedings of a Specialists' Meeting Tokai, Japan, May 1992
English, published: 01/01/92
NEA#169
Available online at: http://www.oecd-nea.org/science/docs/1992/nsc-doc92-9.pdf
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Nuclear Data Standards for Nuclear Measurements - 1991 Nuclear Standards File
English, published: 01/01/92
NEA#187
Available online at: http://www.oecd-nea.org/science/docs/1991/neandc1991-311-u.pdf

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Neutron Cross-section Standards for the Energy Region Above 20 MeV
Proceedings of a Specialists' Meeting, Uppsala, Sweden, May 1991
English, published: 01/01/91
NEA#182
Available online at: http://www.oecd-nea.org/science/docs/1991/neandc1991-305-u.pdf