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The International Atomic Energy Agency (IAEA) has implemented a detailed E-learning on-line course in neutron activation analysis (NAA). Existing books and guidance documents on the concepts and execution of NAA are out of date and not sufficient anymore to ensure the transfer of current knowledge on its practice. The overall objective of the E-learning tool in NAA is to realize a ‘living book’, summarizing the basic concepts and providing practical information on the implementation of the methodologies which can be more easily updated than a common book, allowing also for visualization using contemporary media.

In this paper, the feasibility of high-level radioactive waste transmutation in accelerator driven system sub-critical reactor assembly, has been studied for two zone's model and with three different core configurations. The inner zone has a fast neutron spectrum and the outer one has a thermal neutron spectrum. The subcritical core is coupled with external neutron source of energy 14 MeV (D-T source). The effects of high level waste isotopes sample (238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm, and 245Cm) distribution on the neutron spectrum and burnup performance in the inner zone have been investigated and discussed, by proposed three core configurations non-uniform, uniform, and spiral. The burnup calculations have been performed for one-year operation cycle for all the all proposed models. This work shows that one can effectively transmute most of the actual minor actinides isotopes in the inner fast spectrum zone of the proposed system, with optimal distribution of these isotopes.

The Nuclear Science and Instrumentation Laboratory (NSIL) is currently establishing a Neutron Science Facility (NSF) based on two compact neutron generators: Deuterium-Deuterium (DD), resulting in 2.45 MeV neutrons, and Deuterium-Tritium (DT), resulting in 14.1 MeV neutrons, with maximum source intensities up to 5 × 106 n/s and 4 × 108 n/s over 4π, respectively. Neutron/X-ray radiography and tomography are two of the applications the NSF will be equipped with. In this paper, we report the state of the radiography and tomography system, and the neutron/X-ray radiography and tomography experiments we performed. Good radiographs and tomographs are obtained with X-ray. The spatial resolution of the X-ray radiography system is measured to be about 0.1 mm. The thermal neutron radiographs and fast neutron radiographs have proved that the DD neutron generator will be beneficial to demonstrate neutron radiography capabilities for educational purposes. Additionally, the detail of the collimator optimization of the DD neutron radiography was presented.

The International Atomic Energy Agency (IAEA) implemented an innovative project for assisting neutron activation analysis laboratories in improving the validity of their results by feedback workshops for discussion of results from participation in interlaboratory comparisons rounds in 2010. The participants learned during these meetings to identify the most probable sources of errors in their analytical procedures and how to implement corrective actions to prevent reoccurrence. The outcome of successive rounds between 2010 and 2018 is discussed and experiences during the feedback workshops are given. The quantitative evaluation of the results shows an overall improvement in satisfactory performance. Moreover, there is a clear indication that improvements are consolidated in most laboratories but also stimulate laboratories to develop to a higher level of excellence. Regional differences in performance are also analysed.

Samples of graphite from a RBMK-1500 reactor at the Ignalina Nuclear Power Plant from different construction elements (stack, sleeve, and bushing) were analyzed by the instrumental neutron activation analysis (INAA) method (LVR-15 experimental reactor of the Research Centre Řež, Ltd.) using the prompt gamma activation analysis (PGAA) method (Heinz Maier-Leibnitz Zentrum) and with an inductively coupled plasma mass spectrometer (ICP-MS) (CPST, Lithuania). These measurements were performed with the aim of obtaining the missing information on the impurity distribution in the RBMK-type nuclear graphite constructions as well as for intercomparison purposes, with the results measured in the graphite sleeve samples previously obtained by INAA & GDMS (Glow Discharge Mass Spectrometry) at CEA Saclay, France, and ICP-MS (CPST, Lithuania) methods. Validation of the ICP-MS method for the nuclear graphite impurity concentration determination was proven. The experimentally obtained RBMK-1500 graphite impurity values in different graphite constructions were compared with other measurements and new limits of the possible maximal concentrations of nuclear RBMK graphite impurity concentrations were obtained.

The methodology and the application range of neutron imaging techniques have been significantly improved at numerous facilities worldwide in the last decades. This progress has been achieved by new detector systems, the setup of dedicated, optimized and flexible beam lines and the much better understanding of the complete imaging process thanks to complementary simulations. Furthermore, new applications and research topics were found and implemented. However, since the quality and the number of neutron imaging facilities depend much on the access to suitable beam ports, there is still an enormous potential to implement state-of-the-art neutron imaging techniques at many more facilities. On the one hand, there are prominent and powerful sources which do not intend/accept the implementation of neutron imaging techniques due to the priorities set for neutron scattering and irradiation techniques exclusively. On the other hand, there are modern and useful devices which remain under-utilized and have either not the capacity or not the know-how to develop attractive user programs and/or industrial partnerships. In this overview of the international status of neutron imaging facilities, we will specify details about the current situation.

As part of the International Atomic Energy Agency's (IAEA) Physics Section, the Nuclear Science and Instrumentation Laboratory (NSIL) helps Member States (MSs) to establish, operate and maintain various nuclear instrumentation and spectrometry techniques in support of a wide range of applications such as health care, food, agriculture, environment, forensics, cultural heritage, and materials science. NSIL is contributing to capacity building, transfer of knowledge and expertise sharing, including the development of instruments and validation of analytical methodologies. This paper describes NSIL's key activities and provides an overview of the recent developments and achievements, including the commissioning and utilization of an end-station at one of the beam lines in Elettra Sincrotrone Trieste (EST, Italy), the upgrade of an ultralight radiation monitoring system onboarded on a drone, as well as the development of a full field X-ray fluorescence analytical system for forensic applications and characterization of valuable art/archaeological objects.

The subject of this paper is to explore the possibility to obtain 99mTc from activation of 98Mo, using the TRIGA Mark II low flux research reactor (Vienna, Austria). Irradiation of both natural and enriched in 98Mo molybdenum oxides was compared. Aims of this work included the determination of neutron fluxes and 98Mo(n, γ)99Mo reaction effective cross section in the TRIGA Mark II reactor irradiation channels, calculation of 99Mo specific activities, determination of optimal irradiation conditions for the subsequent 99mTc separation from MoO3 targets using concentrating technologies. © 2012 Akademiai Kiado, Budapest, Hungary.

International audience; In order to reduce the volume and the radiotoxicityof the nuclear waste coming from the operation of existingpressurized water reactors, accelerator-driven systems(ADSs) have been envisioned.The Helium-cooled (He) European Facility forIndustrial-scale Transmutation (He-EFIT) concept is theEUROpean Research Programme for the TRANSmutationof High Level Nuclear Waste in Accelerator DrivenSystem (EUROTRANS) Integrated Project (IP) (EUROTRANSIP) backup option, whereas Pb-cooled EFITis the reference one. The plant has a power of ;400MW(thermal). Like all ADS plants, it consists of threemain components: the accelerator, the spallation targetmodule, and the subcritical core.This paper describes the He-EFIT design at the endof the EUROTRANS IP as well as the studies performedto support this design: spallation performances, trasmutationcapabilities, and plant safety analyses.No specific technology deadlock has been identified,and it might be possible to build such a plant given necessaryresearch and development in support.

Although the number of research reactors (RRs) is steadily decreasing, more than half of the opera- tional RRs are still heavily underutilized, and in most cases, underfunded. The decreasing and rather old fleet of RRs needs to ensure the provision of useful services to the community, in some cases with adequate revenue genera- tion for reliable, safe and secure facility management and operations. Enhancement of low and medium power research reactor (RR) utilization is often pursued by increasing the neutron activation analysis (NAA) activities. In this paper we will present the strategy and concrete actions how NAA as one of the most popular RR appli- cations can contribute to the above goals in particular through (a) RR coalitions and networks, (b) implementa- tion of automation in different stages of NAA, (c) QA/QC, including skills improvement of involved personnel, (d) dedicated proficiency tests performed by a number of targeted analytical laboratories. We also show that despite the IAEA's efforts, some of the NAA laboratories still perform badly in proficiency tests, do not have formal QA/ QC procedures implemented, have not implemented auto- mation to process large number of samples or lack of clear marketing strategies. Some concrete actions are proposed and outlined to address these issues in the near future.

The purpose of this safety study, carried out within the EURISOL Design Study, was to characterize the radiation environment and design the appropriate shielding of the new-generation radioactive ion beam postaccelerator. Both variants oflinac layouts-without stripper (L#1) and with stripper (L#2)—were analyzed using the 132 Sn 25+ radioactive beam of unprecedented intensity, namely, up to ∼10 13 particles/s, as reference for simulations. In this work two scenarios were analyzed: (a) an accidental full beam loss during 1 s every day and (b) continuous beam loss of 10 -4 m -1 , representing normal operation conditions. Representative loss positions along the accelerator at variable energies of 21, 45.5, 76, 115, and 150 MeV u -1 were investigated. The lost ions were assumed to strike a stopping copper target. Dedicated simulations were performed by means of the PHITS code. The induced radioactivity in the accelerator components, concrete walls, and air inside the tunnel were estimated using the DCHAIN-SP-2001 code based on an external neutron source and spallation products derived from PHITS. Ambient dose equivalent rates due to the residual radiation were calculated with the MCNPX code using photon sources resulting from DCHAIN. The effect of implanted radioactive ions at low energies in the accelerator structure was also assessed.

Recently, new applications based on detection of delayed neutrons (DNs) or delayed photons produced during photofission have shown the need for new basic nuclear data. Indeed, available data for DNs are scarce, incomplete, and sometimes contradictory. An experimental campaign has been dedicated to DN measurements for some important actinides with a bremsstrahlung end-point energy spectrum up to 20 MeV. The experiments were carried out at the electron accelerator facility ELSA installed at the Commissariat a l'Energie Atomique (CEA) center of Bruyeres-le-Châtel, France. Delayed neutrons were measured by 12 3 He counters embedded in an annular polyethylene cylinder. Delayed neutron absolute yields were determined. Different irradiation-decay time combinations allowed the extraction of the averaged six-group parameters ofDNs. Similarly to the nuclear reactor physics, these groups are formed by lumping DN precursors according to their half-lives. The DN time dependence is then expressed by the sum of group contributions. In this paper, DN absolute yields and six-group parameters for 235 U at 15 and 18 MeV and 237 Np at 15 MeV are presented. The 235 U and 237 Np absolute DNyields are found slightly lower than those published in the literature. For 235 U, the six-group parameters show no dependence on incident electron energy from 10 to 18 MeV. In addition, the group parameters are very close to those of 237 Np. This work helped to resolve the discrepancies of the photofission DN parameters for 215 U while the six-group parameters for 237 Np are reported for the first time.

The He-EFIT studied under domain DESIGN of the EUROTRANS (FP6) integrated project is a prototype gas cooled accelerator driven subcritical reactor dedicated for transmutation of minor actinides. The paper contains neutronic analysis of the He-EFIT innovative spallation module and aims at the estimation of radiation damage and lifetime characteristics of the target in question.

An active and non-destructive method is being established for detection of 237Np in nuclear waste barrels. The unique high energy decay gamma signature produced after photo-fission is analysed to deduce decay time parameters. A high purity 237Np sample was irradiated with bremsstrahlung photons and high energy gamma decay spectra were measured with BGO detectors for various irradiation durations. The analysis of decay spectra resulted in the formation of five decay groups, which can reproduce the experimental decay spectra within 10% accuracy. These parameters together with neutron decay groups are absolutely essential to quantify the nuclear material in waste barrels by photo-fission. The delayed decay gamma groups for 237Np nucleus are being reported for the first time.

Modern Nuclear Physics requires access to higher yields of rare isotopes, that relies on further development of the In-flight and Isotope Separation On-Line (ISOL) production methods. The limits of the In-Flight method will be applied via the next generation facilities FAIR in Germany, RIKEN in Japan and RIBF in the USA. The ISOL method will be explored at facilities including ISAC-TRIUMF in Canada, SPIRAL-2 in France, SPES in Italy, ISOLDE at CERN and eventually at the very ambitious multi-MW EURISOL facility. ISOL and in-flight facilities are complementary entities. While in-flight facilities excel in the production of very short lived radioisotopes independently of their chemical nature, ISOL facilities provide high Radioisotope Beam (RIB) intensities and excellent beam quality for 70 elements. Both production schemes are opening vast and rich fields of nuclear physics research. In this article we will introduce the targets planned for the EURISOL facility and highlight some of the technical and safety challenges that are being addressed. The EURISOL Radioactive Ion Beam production relies on three 100 kW target stations and a 4 MW converter target station, and aims at producing orders of magnitude higher intensities of approximately one thousand different radioisotopes currently available, and to give access to new rare isotopes. As an illustrative example of its potential, beam intensities of the order of 1013 132Sn ions pe r second will be available from EURISOL, providing ideal primary beams for further fragmentation or fusion reactions studies.

In this paper we present for the first time the measured delayed neutron (DN) yields and time spectra from high-energy protons interacting with thick natPb targets. The 1GeV protons from the accelerator impinged on targets of different thicknesses producing a huge number of spallation-fission products, some of which can be DN precursors. After the beam is switched off, the DNs were detected with optimized 3He counter. The production yields of light DN precursors as 17N and “usual” fission products as 87Br and 88Br, which dominate the total DN activity, are obtained both for thin and thick targets. These new data are of great interest for the new generation high-power spallation targets based on liquid-metal technologies. Our findings also should help to constrain the physics models within the simulations codes.

Recently a renewed interest in photonuclear reactions has appeared. It is motivated by a number of different applications where progress in reliable and, in some cases, very high-intensity electron accelerators was awaited. In particular, today's interest is linked to the nuclear material interrogation and non-destructive nuclear waste characterization, both based either on prompt neutron, or delayed neutron, or delayed gamma detection following photofission. The knowledge of photonuclear reactions is also extremely important in the design of electron accelerators used for medical applications, nuclear physics, photoneutron sources, radioactive ion beam production, etc. In this paper we present the photonuclear activation file (PAF) under development to be included into any material depletion code. By now the PAF contains the photonuclear cross-sections for more than 600 isotopes, fission fragment distributions and delayed neutron evaluations for most of the actinides in the photon energy range from 0 up to 25 MeV. Theoretical predictions, evaluated data files and available experimental data were used to construct the PAF, and some benchmarks are in progress to ensure its quality. The release of the first version of PAF is planned in 2006.

Experimental results for the photofission of 238U with an endpoint Bremsstrahlung emission of 15 MeV are presented. Absolute yield and time characteristics of the delayed neutrons are extracted. In parallel, calculations for fission fragment distributions and corresponding delayed neutron parameters are given and compared to data.

Generally, a very intensive neutron source is needed for the efficient transmutation process of nuclear waste and minor actinides in particular. A plasma based fusion device is naturally a candidate neutron source. In this paper, we analyze a molten salt (flibe) blanket, driven by a plasma based fusion device, dedicated to burn actinides extracted from the LWR spent fuel. Performance parameters such as kscr, burn-up, fluxes and equilibrium conditions were calculated to assess the characteristics of a flibe based transmutation plant. It seems that nearly 1.1metric ton of spent fuel TRU could be burned annually with an output of 3 GWth fission power. Detailed burn-up calculations show that equilibrium conditions of the fuel concentration in the flibe transmutation blanket could be achieved, provided that fission products were removed at least in part and fresh spent fuel TRU was added during burn-up on-line.

A possibility to use a fusion–fission molten salt (flibe) hybrid system to incinerate nuclear waste minor actinides (Np, Am, Cm) is considered. In this work an optimization procedure was developed in order to determine the fission-blanket composition corresponding to fast incineration rate of minor actinides. Performance parameters such as k src , burn-up, neutron fluxes and equilibrium conditions were obtained. Monteburns code system (MCNP+ORIGEN) was used for system modeling. Approximately 1.1 ton of minor actinides, originating from the spent nuclear fuel, could be incinerated annually with an output of 3 GWth fission power. Detailed burn-up calculations show that the equilibrium conditions of the fuel concentration in the flibe transmutation blanket could be achieved, if the originating tritium and fission products were removed at least in part and fresh actinides were added during the burn-up on line. The optimal transmutation blanket in equilibrium is dominated by minor actinides for which the cross sections may vary depending on the nuclear-data libraries used. Therefore, the system performance was tested comparing different sets of data libraries. A number of major differences among ENDF, JENDL and JEF data files were identified and quantified in terms of the averaged one-group cross sections. In this context, our study could be directly generalized for other transmutation systems with similar neutron-energy spectra.

A measurement of the polonium production in a solid Bi target, placed in a high thermal neutron flux of 2.3 10 13 n/cm 2 /s, was performed. Two different activation methods, based on α and γ spectroscopy, were used to cross check the results. Values of (16.08 ± 1.8) mb and (18.4 ± 0.9) mb for the 210 P o (i.e. 210gs Bi) formation cross section were obtained and a recommended value of (17.9 ± 0.8) mb was proposed. Our results are in disagreement with adopted cross section values in the main nuclear data libraries, namely smaller by 25% for 210gs Bi. These new values will account for more precise evaluations of 210 P o formation in moderated spallation targets containing 209 Bi.

Core subcriticality can play an important role if the safety enhancement of a nuclear system is necessary, in particular, when minor actinides submitted for transmutation cause essential degradation of the reactivity feedback effects or/and significant reduction of the delayed neutron fraction. The present work shows that core subcriticality together with thermohydraulics optimization can compensate for the possible degradation of the Doppler effect and the reduction of the delayed neutron fraction. The particular dependence of the spallation neutron yield allows the creation of a supplementary negative feedback effect in case of accelerator coupled hybrid systems. A number of quantitative examples are provided in this context.

After decommissioning of Unit 1 of the Ignalina Nuclear Power Plant, the problem of the radioactive waste management emerged. Among radioactive waste there is an inventory of about 1700 tons of the graphite containing 14 C radioisotope as an activation product. The estimates show that the maximal total inventory of 14 C in graphite from Unit 1 is around 7 � 10 14 Bq. One of the possible ways for utilization of the graphite is its incineration in the radioactive waste processing plant. Unfortunately, in this case a significant amount of the radionuclide would be released into the atmosphere in the form of CO2 and the released radiocarbon would cause additional exposure of the population. Possible radiological consequences for the Lithuanian inhabitants are evaluated using the model of radiocarbon dispersion in the environment and considering several scenarios of the graphite incineration. Dispersion of the incineration gas is modelled using the Gaussian dispersion model. Assimilation of CO2 by the vegetation due to photosynthesis as well as washout of CO2 from the atmosphere by rain, uptake of the deposited 14 C by the plants from soil, and the eventual contamination of food products are considered. An estimated additional exposure effective dose to the critical group of the local population due to continuous releases of the total inventory of 14 C from the incinerator is of the order of 2.7 mSv. The consumption of the contaminated locally produced food products is the main contribution to the dose. Such continuous incineration of graphite would be acceptable if it were extended

at different fuel burn-up. The comparison of the experimentally measured and calculated isotopic composition for the main actinides and fission products allowed validation of MCNP and ORIGEN codes for RBMK spent nuclear fuel calculations. Variations of RBMK spent nuclear fuel burn-up and plutonium isotopic ratios depending on fuel location in the assembly were determined.

The techniques used to produce a 177 m Lu (J π =23/2 − ,T 1/2 =160.4 days) target are described in this paper. Firstly, an isotopic separation of an enriched lutetium sample was used to reach a purity of 176 Lu close to 99.993%. Afterwards, the high neutron flux of the Grenoble Institut Laue-Langevin reactor was used to produce the 177 m Lu isomer by the 176 Lu(n,γ) reaction. Finally, a chemical separation was performed to extract 1013 nuclei of 177 m Lu . Thanks to this experiment, we have been able to estimate the destruction cross-section of the 177 m Lu .

Recently a worldwide interest in HTR technology has been experienced. In this context the gas turbine modular helium-cooled reactor (GT-MHR) is a potential candidate for the maximum 239Pu destruction in a once-through cycle. A particular feature of GT-MHR is that its coated fuel (TRISO particles) is supposed to provide an impermeable barrier to the release of fission products and, at the same time, to resist very deep burn-up rates (more than 90% for 239Pu). In this paper a Monte Carlo approach is employed to characterise the neutron fluxes and the fuel evolution inside the tiny 200-μm diameter fuel kernels with an exact and finite geometry description. Our major goal is to obtain a quantitative comparison of different geometry sets, namely homogeneous versus single-heterogeneous and double-heterogeneous, in terms of keff eigenvalues, the length of the fuel cycle, neutron characteristics and the evolution of fuel composition in particular. In all cases the same Monteburns (MCNP+ORIGEN) code system is used. We show that the performance of GT-MHR is considerably influenced by the way its geometry is modelled within the Monte Carlo approach. The spatial and energy shielding of the neutron flux even in such small particles cannot be neglected for important isotopes which have high resonance cross sections as 240Pu, 241Pu and 167Er. Namely, the formation of 241Pu and burn-up of 167Er are responsible for the different length of the fuel cycle, being the shortest for a double-heterogeneous geometry. On the other hand, the evolution of 239Pu at a constant reactor power and comparable neutron fluence is very similar for all three geometry configurations.

A brief summary of Po formation processes and release mechanisms from polonium-contained media is provided. Quantitative estimations of 210 Po losses from the solid 209 Bi target irradiated in the thermal neutron flux is given as an example. Other potential applications where this analysis could be directly applicable are mentioned.

As a result of an international cooperation a recommended file of photonuclear reactions of 164 isotopes became available. The formal and practical quality of these data are examined by using them for calculation of photoneutron production in thick targets bombarded by photons and electrons. The Monte Carlo calculations are accomplished with a recent version of the MCNP code enhanced with a photonuclear capability. The energy range of our investigations is from the photonuclear threshold up to 150 MeV. In some cases, a comparison between our calculations and a new version of the MCNPX code, recently updated with the LANL photonuclear data files (12 isotopes) is established. Our results are also compared with the experimental data when available. These data include early measurements of the total neutron yields from thick targets bombarded by electrons as a function of energy, target thickness and atomic number, also, recent measurements of 238 U photofission yields. We show that in general the agreement b...

The gas turbine modular helium-cooled reactor (GT-MHR)1) is known probably as the best option for the maximum plutonium destruction in once-through cycle, even though the industrial fabrication of coated particle fuel still has to be proved. We perform detailed simulations along these lines by comparing different sets of data libraries in terms of keff eigenvalues, the length of the fuel cycle, neutronic characteristics and the evolution of fuel composition in particular. In all cases the same Monteburns code system2) is used making our results dependent only on the evaluated data tables. We show that in general the performance of GT-MHR is not considerably influenced by the choice of the data libraries employed. Nevertheless, a number of major differences among ENDF, 3) JENDL4) and JEF5) data files are identified and quantified in terms of the averaged one-group cross sections both for military (MPu) and civil (CPu) plutonium based fuel cycles.

A new experimental installation has been recently commissioned at the High Flux Reactor of Institut Laue-Langevin (ILL) in Grenoble (France). It gives access to high intensity neutron spectra from pure thermal (5.6 1014 n/s/cm2) to epithermal (2 1015 n/s/cm2). Several of low mass (10 µg) mono-isotopic targets of actinide elements are in the process of being irradiated and analyzed by a number of techniques, from nuclear spectroscopy to off-line mass spectrometry and innovative double-deposit fission micro-chambers.In the present paper we will present the first experiments carried out at the thermal neutron spectrum installation with 242Pu, 241-243Am samples.

The Incineration by Accelerator (INCA) project of the Directorate for Science of Matter of the French Atomic Energy Authority (CEA/DSM) aims to outline the ideal physical conditions to transmute minor actinides in a high intensity neutron flux obtained either by hybrid systems or innovative critical reactors. To measure on-line the incineration rates of minor actinides, we are developing an innovative Double Deposit Fission Chamber (DDFC) working in current mode. Our method is based on a comparison between the isotope under study and a reference material whose nuclear parameters are well known, as 235U and 239Pu. This new fission chamber will be used in the High Flux Reactor in Grenoble/France in a neutron flux of 1.2×1015 n cm−2 s−1 for 50 days, the operating cycle of the reactor. These specific experimental conditions require substantial modifications of the existing chambers. The first experiment will be carried out in fall 2000.

The Neutrons For Science (NFS) facility is a component of SPIRAL-2 laboratory under construction at Caen (France). SPIRAL-2 is dedicated to the production of high intensity Radioactive Ions Beams (RIB). It is based on a high-power linear accelerator (LINAG) to accelerate deuterons beams in order to produce neutrons by breakup reactions on a C converter. These neutrons will induce fission in 238U for production of radioactive isotopes. Additionally to the RIB production, the proton and deuteron beams delivered by the accelerator will be used in the NFS facility. NFS is composed of a pulsed neutron beam and irradiation stations for cross-section measurements and material studies. The beams delivered by the LINAG will allow producing intense neutron beams in the 100 keV-40 MeV energy range with either a continuous or quasi-mono-energetic spectrum. At NFS available average fluxes will be up to 2 orders of magnitude higher than those of other existing time-of-flight facilities in the 1 MeV - 40 MeV range. NFS will be a very powerful tool for fundamental physics and application related research in support of the transmutation of nuclear waste, design of future fission and fusion reactors, nuclear medicine or test and development of new detectors. The facility and its characteristics are described, and several examples of the first potential experiments are presented., JRC.D.4-Standards for Nuclear Safety, Security and Safeguards

We calculate in-target production rates of radioactive nuclei in the case of single-stage spallation targets designed for EURISOL (European Isotope Separation On-Line Radioactive Ion Beam Facility). Using particle generation-transport codes, we estimate the isotopic distributions of the reaction products, in terms of their charge and mass numbers, for 320 configurations of cylindrical targets. For 11 elements of the highest priority (Li, Be, Ne, Mg, Ar, Ni, Ga, Kr, Sn, Hg, Fr) and 23 of their isotopes far from stability, this study helps in deciding which target configuration can be retained or discarded to reach the highest production rates.

Proc on line; International audience; 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.

Design and decommissioning of large-scale nuclear facilities is nowadays a high-interest activity for the nuclear community. Nuclear data have an important role in R&D studies for accelerator-based applications and decommissioning of nuclear facilities including nuclear reactors. Most of these studies use simulations to predict activity levels and dose rates, to design shielding components, etc. The different data libraries and/or reaction models included in the transport codes are essential for these calculations. In this contribution we present several examples of the nuclear data usage for this type of applications. We show the importance of a correct use of nuclear data and reaction models for the accuracy of the results obtained by simulations. Installations able to provide high intensity radioactive beams (SPIRAL2, EURISOL...) are subject of sustained R&D ef- forts. There are also an important number of small cyclotrons, most of them dedicated to medical applications, with specific needs in terms of R&D studies. Decommissioning of nuclear facilities also needs dedicated activity studies. In this work we present two different examples: R&D studies for accelerator- based applications and decommissioning of nuclear facilities. Both examples clearly show the role played by nuclear data and their employment in reaction codes for the accuracy of the results obtained by simulations.

In the framework of the Neutronic and Nuclear Assessment Task Group of the MEGAPIE experiment we measured the delayed neutron (DN) flux at the top of the target. The measurement was proposed mainly for radioprotection purposes since the DN flux at the top of the target has been estimated to be of the same order of magnitude as the prompt neutron flux. Given the strong model-dependence of DN predictions, the measurement of DN contribution to the total neutron activity at the top of the target was thus desired. Moreover, this measurement is complementary to the DN experiments performed at PNPI (Gatchina) on solid lead and bismuth targets. The DN measurement at MEGAPIE was performed during the start-up phase of the target. In this paper we present a detailed description of the experimental setup and some preliminary results on decay spectra., Comment: 4 pages, 4 figures, Proceedings of International Conference on Nuclear Data for Science and Technology (ND 2007), Nice, France, 22-27 Apr 2007

The total delayed neutron (DN) yields from neutron induced fission on 232 Th were measured for neutron energies of 2.0, 2.4 and 16.0 MeV. In this paper, the experimental results are presented and compared to existing experimental data and evaluated data libraries. The future measurements are outlined.

Photons with energies above 6 MeV can be used to detect small amounts of nuclear material inside large cargo containers. The method consists in using an intense beam of high-energy photons (bremsstrahlung radiation) in order to induce reactions of photofission on actinides. The measurement of delayed neutrons and delayed gammas emitted by fission products brings specific information on localization and quantification of the nuclear material. A simultaneous measurement of both of these delayed signals can overcome some important limitations due to matrix effects like heavy shielding and/or the presence of light elements as hydrogen. We have a long experience in the field of nuclear waste package characterization by photon interrogation and we have demonstrated that presently the detection limit can be less than one gram of actinide per ton of package. Recently we tried to extend our knowledge to assess the performance of this method for the detection of special nuclear materials in sea and air freights. This paper presents our first results based on experimental measurements carried out in the SAPHIR facility, which houses a linear electron accelerator with the energy range from 15 MeV to 30 MeV. Our experiments were also modeled using the full scale Monte Carlo techniques. In addition, and in a more general frame, due to the lack of consistent data on photonuclear reactions, we have been working on the development of a new photonuclear activation file (PAF), which includes cross sections for more than 600 isotopes including photofission fragment distributions and delayed neutron tables for actinides. Therefore, this work includes also some experimental results obtained at the ELSA electron accelerator, which is more adapted for precise basic nuclear data measurements.

In this work a new approach to the realisation of an accelerator-coupled hybrid system (ACS) is proposed. A significant improvement of the feedback effect due to the particularities of the neutron production in a spallation target is expected. In the present study, we explain the principles of system functioning as well as the advantages and disadvantages of the proposed concept. The quantitative analysis of the innovative ACS operation is based on a generalised point kinetics approach. In the framework of this simplified model, we show that the particular dependence of the spallation neutron yield allows for the creation of a supplementary negative feedback effect (Doppler-like). Implementation of this concept should compensate, to some extent, the eventual feedback degradation in the cores dedicated to the transmutation of nuclear waste.

The decommissioning and dismantling of nuclear installations after their service life involves the necessary disassembling, handling and disposing of a large amount of radioactive equipment and structures. In particular, the concrete that has been used as a biological reactor shield and graphite that has been used as a moderator-reflector represent the majority of waste, requiring geological disposal. To reduce this undesirable volume to the minimum and to successfully plan the dismantling and disposal of radioactive materials to storage facilities, the activations of the structures should be accurately evaluated. In the framework of the decommissioning and the dismantling of the experimental reactor of the University of Strasbourg, detailed activation estimates have been conducted to characterise the graphite and the structural materials present in the reactor environment. For this purpose, the chemical compositions of fresh graphite samples and different types of concrete have been determined by activation analysis in the research reactors OSIRIS and ORPHEE of CEA Saclay (France). Then, the activations of graphite, concrete and other materials have been calculated in the whole reactor, as a function of the three main nuclear data libraries, i.e. ENDF, JEF and JENDL. In parallel, the activations of representative graphite and concrete samples have been measured experimentally. The comparison of theoretical predictions with experimental values validates the approach and the methodology used in the present study and tests the consistency and the reliability of the nuclear data used for activation analysis. We believe that a similar approach could also be used for the decommissioning of industrial nuclear reactors.

In the framework of nuclear waste transmutation studies, the Mini‐INCA project has been initiated at CEA/DSM with objectives to determine optimal conditions for transmutation and incineration of Minor Actinides (MA) in high intensity neutron fluxes. Our experimental tools based on alpha‐ and gamma‐spectroscopy of irradiated samples and the development of fission micro‐chambers could gather both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials. Cross sections of selected actinides (241Am, 242Am, 242Pu, 237Np, 238Np) have already been measured at ILL, showing some discrepancies when compared to evaluated data libraries but in overall good agreement with recent experimental data.

The next generation spallation neutron sources, neutrino factories or RIB production facilities currently being designed and constructed world wide will increase the average proton beam power on target by a few orders of magnitude. Increased proton beam power results in target thermal hydraulic issues leading to new target designs, very often based on liquid metal technologies such as Hg, Pb, or PbBi. Radioactive nuclides produced in liquid metal targets are transported into hot cells, into pumps or close to electronics with radiation sensitive components. Besides the considerable amount of photon activity in the irradiated liquid metal, a significant amount of the Delayed Neutron (DN) precursor activity can be accumulated in the target fluid. The transit time from the front of a liquid metal target into areas, where DNs may be important, can be as short as a few seconds, i.e. well within one half‐life of many DN precursors. Therefore, it seems very important to evaluate the DN flux as a function of position and determine if DNs may contribute significantly to the activation and dose rates. The multi‐particle transport code MCNPX combined with the material evolution program CINDER’90 is used to predict the DN precursors and construct the DN tables. These DN tables are employed within the generalized geometrical model of the MegaPie spallation target at PSI (Switzerland). We show that the contribution of DNs and prompt spallation neutrons to the total neutron flux is comparable at the very top of the liquid PbBi loop. We also demonstrate that these estimates of DNs within MCNPX are very much model‐dependent. No experimental data are available for DN yields and time spectra from high energy fission‐spallation reactions. An experiment to perform these measurements is proposed.

This article presents calculations of photonuclear reaction cross sections for actinides by using different reaction codes, namely an improved version of HMS‐ALICE and GNASH. The GDR parameters from the RIPL2 data library are used to describe the total photo‐absorption cross section. In general, our results show satisfactory agreement with experimental data and IAEA evaluations. We plan to include our GNASH evaluations for 235U, 238U, 237Np, 239Pu, 240Pu and 241Am into a new release of the ENDF‐B/VII data files.

We propose to measure the neutron flux in the liquid Pb‐Bi 1 MW MEGAPIE target at PSI. Micro fission‐chambers will be placed inside the central rod of the target to measure both thermal and non‐thermal components of the neutron flux. In principle, both time‐ and space‐dependent variations of the flux will be on line monitored with a precision better than 5%.

In this article we present the development of a new photonuclear activation library for the CINDER’90 evolution code (LANL). The IAEA evaluations for 164 isotopes are explicitly included and more than 600 isotopes are added using the HMS‐ALICE predictions. GNASH evaluations are used in the case of actinides.

Fission fragment distributions and delayed neutron yields for 235U and 238U are provided by a complete modelization of the photofission process below 25 MeV. The absorption cross‐section parameterization and the fission fragment distributions are given and compared to experimental data. The delayed neutron yields and the half‐lives in terms of six groups are presented and compared to data obtained with a bremsstrahlung spectrum of 15 MeV.