Nuclear Energy Agency

Abstract The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides $$^{235}\hbox {U}$$ 235U , $$^{238}\hbox {U}$$ 238U and $$^{239}\hbox {Pu}$$ 239Pu , on $$^{241}\hbox {Am}$$ 241Am and $$^{23}\hbox {Na}$$ 23Na , $$^{59}\hbox {Ni}$$ 59Ni , Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy.

The International Network of Nuclear Reaction Data Centres (NRDC) coordinated by the IAEA Nuclear Data Section (NDS) successfully collaborates in the maintenance and development of the EXFOR library. As the scope of published data expands (e.g. to higher energy, to heavier projectile) to meet the needs of research and applications, it has become a challenging task to maintain both the completeness and accuracy of the EXFOR library. Evolution of the library highlighting recent developments is described.

In the final objective of elaborating an optical sensor of dimensional change in a nuclear testing reactor, we present extreme high level irradiation tests, up to 1.3 1020nfast/cm2and 16 GGy, conducted on high purity silica and hollow core photonic bandgap (PBG) fibers. These tests are performed in the OSIRIS reactor at CEA Saclay. We measure continuously radiation induced absorption (RIA), at 3 optical wavelengths: 980 nm, 1064 nm and 1310 nm. The evolution with the time of the spectral dependency over larger wavelength range is also reported. As well as reporting RIA up to higher dose than presented before, we demonstrate here the excellent superiority of hollow core PBG fibers over their conventional counterparts in term of radiation resistance up to over 1020nfast/cm2and 10 GGy. We also report some optical time domain reflectometry (OTDR) measurements that probe the irradiated parts of the fibers and show a back scattering peak which appears, widens spatially then duplicates. We provide an interpretation based on the change of silica when it turns to metamict phase.

The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear technologies - 235,238U, 239Pu, 56Fe, 16O and 1H - with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform

We summarized temporal changes in air dose rates and radionuclide deposition densities over five years in the 80 km zone based on large-scale environmental monitoring data obtained continuously after the Fukushima Nuclear Power Plant (NPP) accident, including those already reported in the present and previous special issues. After the accident, multiple radionuclides deposited on the ground were detected over a wide area; radiocesium was found to be predominantly important from the viewpoint of long-term exposure. The relatively short physical half-life of 134 Cs (2.06 y) has led to considerable reductions in air dose rates. The reduction in air dose rates owing to the radioactive decay of radiocesium was more than 60% over five years. Furthermore, the air dose rates in environments associated with human lives decreased at a considerably faster rate than expected for radioactive decay. The average air dose rate originating from the radiocesium deposited in the 80 km zone was lower than that predicted from radioactive decay by a factor of 2–3 at five years after the accident. Vertical penetration of radiocesium into the ground contributed greatly to the reduction in air dose rate because of an increase in the shielding of gamma rays; the estimated average reduction in air dose rate was approximately 25% with penetration compared to that without penetration. The average air dose rate measured in undisturbed fields in the 80 km zone was estimated to be reduced owing to decontamination by approximately 20% compared to that without decontamination. The average deposition density of radiocesium in undisturbed fields has decreased owing to radioactive decay, indicating that the migration of radiocesium in the horizontal direction has generally been slow. Nevertheless, in human living environments, horizontal radiocesium movement is considered to contribute significantly to the reduction in air dose rate. The contribution of horizontal radiocesium movement to the decrease in air dose rate was estimated to vary by up to 30% on average. Massive amounts of environmental data were used in extended analyses, such as the development of a predictive model or integrated air dose rate maps according to different measurement results, which facilitated clearer characterization of the contamination conditions. Ecological half-lives were evaluated in several studies by using a bi-exponential model. Short-term ecological half-lives were shorter than one year in most cases, while long-term ecological half-lives were different across the studies. Even though the general tendency of decrease in air dose rates and deposition densities in the 80 km zone were elucidated as summarized above, their trend was found to vary significantly according to location. Therefore, site-specific analysis is an important task in the future.

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the U.S. Department of Energy. The ICSBEP became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Spain, and Israel are now participating. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled “International Handbook of Evaluated Criticality Safety Benchmark Experiments” (ICSBEP Handbook). The 2002 edition of the ICSBEP Handbook contains benchmark model specifications for 2881 critical or subcritical configurations that are intended for validating computer codes that calculate effective neutron multiplication and for testing basic nuclear data.

The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the current understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burn-up and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including as regards experimental approaches and interpretation and the use of experimental data relevant for safety. In 1986, a working group of the NEA Committee on the Safety of Nuclear Installations (CSNI) issued a state-of-the-art report on water reactor fuel behaviour in design-basis accident (DBA) conditions. The 1986 report was limited to the oxidation, embrittlement and deformation of pressurised water reactor (PWR) fuel in a loss-of-coolant accident (LOCA). Since then, considerable experimental and analytical work has been performed, which has led to a broader and deeper understanding of LOCA-related phenomena. Further, new cladding alloys have been produced, which might behave differently than the previously used Zircaloy-4, both under normal operating conditions and during transients. Compared with 20 years ago, fuel burn-up has been significantly increased, which requires extending the LOCA database in order to cover the high burnup range. There was also a clear need to address LOCA performance for reactor types other than PWRs. The present report has been prepared by the WGFS and covers the following technical aspects: - Description of different LOCA scenarios for major types of reactors: BWRs, PWRs, VVERs and to a lesser extent CANDUs. - LOCA phenomena: ballooning, burst, oxidation, fuel relocation and possible fracture at quench. - Details of high-temperature oxidation behaviour of various cladding materials. - Metallurgical phase change, effect of hydrogen and oxygen on residual cladding ductility. - Methods for LOCA testing, for example two-sided oxidation and ring compression for ductility, and integral quench test for strength. - Predictive computer codes.

The significant and complex effect of plastic deformation on corrosion behavior involves changes in not only dislocation density but also other metallurgical factors such as grain size, texture, chemical inhomogeneity, phase transformation and residual stress. With the advent of severe plastic deformation (SPD), the effect of plastic deformation on corrosion in the ultrahigh strain range is becoming an important issue. However, our understanding of corrosion properties of SPD materials lags far behind than that of their other properties, e.g. their mechanical properties. In this review, the role of dislocations and grain boundaries generated by SPD was highlighted in pure metals and single-phase materials, where plastic deformation and grain refinement proceed mainly by dislocation activity. Accordingly, the complicated effect of chemical inhomogeneity arising from impurity segregation and precipitation was excluded from discussion, while other implicit effects were included. It is essential to elucidate the effect of so-called ultrafine-grained (UFG) structures which develop progressively to a saturation over a very wide strain range. Unfortunately, the literature mainly compares the corrosion behavior of UFG and coarse-grained (CG) materials, and the degree of perfection of UFG formation and the resultant effects on corrosion vary between studies. The limited number of studies that examines corrosion behavior systematically over a wide strain range suggests that, in most cases, the effect of plastic deformation on corrosion extends into the SPD region gradually, with no anomalous change. That is, SPD improves the corrosion resistance to further degree in a passive environment, whereas it increases the dissolution rate in a non-passive environment. However, several works reported an abrupt change in corrosion resistance, which could be attributed to UFG formation. A marked improvement is observed in Fe–Cr alloys, where passivation becomes more protective owing to UFG formation induced by SPD. In severely deformed materials, structural alterations in dislocations and grain boundaries have a very high impact on the corrosion kinetics because of their closely spaced configuration.

Cross section measurements have been performed at the time-of-flight facility GELINA to determine the average capture cross section for 197 Au in the energy region between 3.5 keV and 84 keV. Prompt γ-rays, originating from neutron-induced capture events, were detected by two C 6 D 6 liquid scintillators. The sample was placed at about 13m distance from the neutron source. The total energy detection principle in combination with the pulse height weighting technique was applied. The energy dependence of the neutron flux was measured with a double Frisch-gridded ionization chamber based on the 10 B(n,α) reaction. The data have been normalized to the well-isolated and saturated 197 Au resonance at 4.9 eV. Special care was taken to reduce bias effects due to the weighting function, normalization, dead time and background corrections. The total uncertainty due to normalization, neutron flux and weighting function is 1.0%. An additional uncertainty of 0.5% results from the correction for self-shielding and multiple interaction events. Fluctuations due to resonance structures have been studied by complementary measurements at a 30m flight path station. The results reported in this work deviate systematically by more than 5% from the cross section that is recommended as a reference for astrophysical applications. They are about 2% lower compared to an evaluation of the 197 Au(n, γ) cross section, which was based on a least squares fit of experimental data available in the literature prior to this work. The average capture cross section as a function of neutron energy has been parameterized in terms of average resonance parameters. Maxwellian average cross sections at different temperatures have been calculated.

The TALYS Evaluated Nuclear Data Library (TENDL) has now 8 releases since 2008. Considerable experience has been acquired for the production of such general-purpose nuclear data library based on the feedback from users, evaluators and processing experts. The backbone of this achievement is simple and robust: completeness, quality and reproducibility. If TENDL is extensively used in many fields of applications, it is necessary to understand its strong points and remaining weaknesses. Alternatively, the essential knowledge is not the TENDL library itself, but rather the necessary method and tools, making the library a side product and focusing the efforts on the evaluation knowledge. The future of such approach will be discussed with the hope of nearby greater success.

To broaden the available rice genetic resources, we developed two populations of introgression lines of cultivated rice (Oryza sativa) carrying donor segments from the wild species Oryza glumaepatula and Oryza meridionalis. These lines contain overlapped introgressed donor segments that covered most parts of the genome of the two donors in the same of O. sativa genetic background (ssp. japonica, cv. Taichung 65). The introgression lines were developed through repeated backcrossing with Taichung 65 as a pollen parent and marker assisted selection. O. glumaepatula introgression lines consist of two sets of the lines (with O. glumaepatula and Taichung 65 cytoplasm, respectively): these comprise a total of 69 lines that cover 79.5 to 89.2% of the Oryza glumaepatula genome. The O. meridionalis introgression lines also consist of two sets of the lines (with O. meridionalis and Taichung 65 cytoplasm, respectively): these comprise a total of 78 lines covering the 81.5 to 98.0% of the Oryza meridionalis genome. These introgression lines significantly broaden rice genetic resources, and will facilitate analyses of the genetics of traits specific to the donor species.

Uncertainty analysis is a key element in nuclear power plant deterministic safety analysis using best-estimate thermal-hydraulic codes and best-estimate-plus-uncertainty methodologies. If forward uncertainty propagation methods have now become mature for industrial applications, the input uncertainty quantification (IUQ) on the physical models still requires further investigations. The Organisation for Economic Co-operation and Development/Nuclear Energy Agency PREMIUM project attempted to benchmark the available IUQ methods, but observed a strong user effect due to the lack of best practices guidance. The SAPIUM project has been proposed toward the construction of a clear and shared systematic approach for IUQ. The main outcome of the project is a first “good-practices” document that can be exploited for safety study in order to reach consensus among experts on recommended practices as well as to identify remaining open issues for further developments. This paper describes the systematic approach that consists of five elements in a step-by-step approach to perform a meaningful model IUQ and validation as well as some good-practice guideline recommendations for each step.

The SINBAD project started in the early 1990's as a collaboration between the Organization for Economic Cooperation and Development's Nuclear Energy Agency Data Bank (OECD/NEADB) and the Radiation Safety Information Computational Center (RSICC) at the Oak Ridge National Laboratory (ORNL) with the goal to preserve the information on the performed radiation shielding benchmarks and make them available in a standardised form to the international community.The SINBAD database comprises now 100 shielding benchmarks, covering fission reactor shielding, fusion blanket neutronics, and accelerator shielding.A thorough revision of the SINBAD benchmark experiments was undertaken recently in order to verify the completeness and consistency of the benchmark information, in particular concerning the evaluation of the experimental sources of uncertainty.New improved computer code models were prepared.This review process is expected to provide users with an easier choice and help them make better use of the experimental information.The OECD NEA Working Party on Scientific Issues of Reactor Systems (WPRS) Expert Group on Radiation Transport and Shielding (EGRTS) was started in 2011 with the mandate among others, to monitor, steer and support the continued development of the SINBAD database.SINBAD is available from RSICC and from the NEA Data Bank.

An uncertainty propagation methodology relying on marginalization techniques was recently developed to produce covariance matrices between existing model parameters involved in describing neutron-induced reactions. This work has been implemented in the nuclear data assimilation tool CONRAD. The performance of the code was demonstrated through simplified test cases based on a Reich-Moore description of the 155Gd(n,γ) reaction. Results are compared with those produced via Monte Carlo techniques.

The status of the Joint Evaluated Fission and Fusion file (JEFF) is described.JEFF-3.1 comprises a significant update of actinide evaluations, materials evaluations that have emerged from various European nuclear data projects, the activation library JEFF-3.1/A, the decay data and fission yield sub-libraries, and fusion-related data files from the EFF project.The revisions were motivated by the availability of new measurements, modelling capabilities and trends from integral experiments.Validations have been performed, mainly for criticality, reactivity temperature coefficients, fuel inventory and shielding of thermal and fast systems.Compared with earlier releases, JEFF-3.1 provides improved performance with respect to a variety of scientific and industrial applications.Following on from the public release of JEFF-3.1, the French nuclear power industry has selected this suite of nuclear applications libraries for inclusion in their production codes.

According to the OECD/NEA estimates, nuclear power plants (NPPs), whether with a large reactor or with small modular reactors (SMRs), are competitive with many other electricity generation technologies in a significant number of cases, one of the exceptions being natural gas in the USA with the current level of prices. However, SMRs have particular features and requirements setting conditions for their deployment. This paper presents the preliminary analysis by OECD/NEA of the economics, opportunities, and market for small nuclear reactors.

Spreadsheets are important business tools. Increasingly they are being used for simulation e.g. to perform risk analysis. Such tasks have far greater computational demands than traditional spreadsheet bookkeeping applications. In this paper we show how spreadsheets can support supercomputing. This is achieved without requiring the spreadsheet user to have specialist tools or knowledge. The key technical innovation is a mechanism enabling the concurrent evaluation of spreadsheet functions. Furthermore the mechanism does not require modification of the standard spreadsheet evaluation engine. INTRODUCTION Spreadsheets are popular because they are easy to use and modify, and they support numerical data analysis without programming. A feature of modern spreadsheets is their capability to be extended; in particular custom functions may be incorporated into them, for example complex simulation functions and computational experiments. For these reasons spreadsheets make an ideal front-end f...

Modification of silica coated on magnetic material iron sand with chitosan for adsorption of Au(III) has been carried out. Magnetic material (MM) from iron sand was separated using an external magnet, washed with water and HF solution 10%. MM–silica–chitosan material (MMSC) was synthesized via sol gel process and the product was characterized with Fourier Transform Infrared (FT-IR) spectrophotometer, X–ray diffractometer, scanning electron microscopy and energy dispersive X–ray, thermogravimetric analysis, and vibration sample magnetometer. Additionally, the effect of pH on the stability of MMSC has also been tested. The adsorption of Au(III) on MMSC was proceeded in a batch system with variation of pH, contact time, and concentration of absorbate. Adsorbent was separated using external magnet and concentration of Au(III) not adsorbed was analyzed using Atomic Absorption Spectrometer. Characterization result indicated that MMSC was successfully synthesized. Adsorption of Au(III) on MMSC followed pseudo second-order kinetic model with the value of adsorption rate constant (k) of 4.10 x 10-3 g mg-1 min-1 and adsorption isotherm fixed with Langmuir model with the adsorption capacity (qmax) of 149.25 mg g-1.

BACKGROUND: The purpose of toxicology is to protect human health and the environment. To support this, the Organisation for Economic Co-operation and Development (OECD), operating via its Extended Advisory Group for Molecular Screening and Toxicogenomics (EAGMST), has been developing the Adverse Outcome Pathway (AOP) approach to consolidate evidence for chemical toxicity spanning multiple levels of biological organization. The knowledge transcribed into AOPs provides a structured framework to transparently organize data, examine the weight of evidence of the AOP, and identify causal relationships between exposure to stressors and adverse effects of regulatory relevance. The AOP framework has undergone substantial maturation in the field of hazard characterization of chemicals over the last decade, and has also recently gained attention from the radiation community as a means to advance the mechanistic understanding of human and ecological health effects from exposure to ionizing radiation at low dose and low dose-rates. To fully exploit the value of such approaches for facilitating risk assessment and management in the field of radiation protection, solicitation of experiences and active cooperation between chemical and radiation communities are needed. As a result, the Radiation and Chemical (Rad/Chem) AOP joint topical group was formed on June 1, 2021 as part of the initiative from the High Level Group on Low Dose Research (HLG-LDR). HLG-LDR is overseen by the OECD Nuclear Energy Agency (NEA) Committee on Radiation Protection and Public Health (CRPPH). The main aims of the joint AOP topical group are to advance the use of AOPs in radiation research and foster broader implementation of AOPs into hazard and risk assessment. With global representation, it serves as a forum to discuss, identify and develop joint initiatives that support research and take on regulatory challenges. CONCLUSION: The Rad/Chem AOP joint topical group will specifically engage, promote, and implement the use of the AOP framework to: (a) organize and evaluate mechanistic knowledge relevant to the protection of human and ecosystem health from radiation; (b) identify data gaps and research needs pertinent to expanding knowledge of low dose and low dose-rate radiation effects; and (c) demonstrate utility to support risk assessment by developing radiation-relevant case studies. It is envisioned that the Rad/Chem AOP joint topical group will actively liaise with the OECD EAGMST AOP developmental program to collectively advance areas of common interest and, specifically, provide recommendations for harmonization of the AOP framework to accommodate non-chemical stressors, such as radiation.

The OECD-NEA High Priority Request List (HPRL) is a point of reference to guide and stimulate the improvement of nuclear data for nuclear energy and other applications, and a tool to bridge the gap between data users and producers. The HPRL is application-driven and the requests are submitted by nuclear data users or representatives of the user’s communities. A panel of international experts reviews and monitors the requests in the framework of an Expert Group mandated by the NEA Nuclear Science Committee Working Party on International Nuclear Data Evaluation Cooperation (WPEC). After approval, individual requests are classified to three categories: high priority requests, general requests, and special purpose requests (e.g., dosimetry, standards). The HPRL is hosted by the NEA in the form of a relational database publicly available on the web. This paper provides an overview of HPRL entries, status and outlook. Examples of requests successfully completed are given and new requests are described with emphasis on updated nuclear data needs in the fields of nuclear energy, neutron standards and dosimetry.