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CORTEX is a EU H2020 project (2017-2021) devoted to the analysis of 'reactor neutron noise' in nuclear reactors, i.e. the small fluctuations occurring around the stationary state due to external or internal disturbances in the core. One important aspect of CORTEX is the development of neutron noise simulation codes capable of modeling the spatial variations of the noise distribution in a reactor. In this paper we illustrate the validation activities concerning the comparison of the simulation results obtained by several noise simulation codes with respect to experimental data produced at the zero-power reactors AKR-2 (operated at TUD, Germany) and CROCUS (operated at EPFL, Switzerland). Both research reactors are modeled in the time and frequency domains, using transport or diffusion theory. Overall, the noise simulators managed to capture the main features of the neutron noise behavior observed in the experimental campaigns carried out in CROCUS and AKR-2, even though computational biases exist close to the region where the noise-inducing mechanical vibration was located (the so-called "noise source"). In some of the experiments, it was possible to observe the spatial variation of the relative neutron noise, even relatively far from the noise source. This was achieved through reduced uncertainties using long measurements, the installation of numerous, robust and efficient detectors at a variety of positions in the near vicinity or inside the core, as well as new post-processing methods. For the numerical simulation tools, modeling the spatial variations of the neutron noise behavior in zero-power research reactors is an extremely challenging problem, because of the small magnitude of the noise field; and because deviations from a point-kinetics behavior are most visible in portions of the core that are especially difficult to be precisely represented by simulation codes, such as experimental channels. Nonetheless the limitations of the simulation tools reported in the paper were not an issue for the CORTEX project, as most of the computational biases are found close to the noise source. • Validation activities with dedicated experimental data from zero-power reactors. • Modeling in the time and frequency domain using transport or diffusion theory. • The noise simulators capture the main features of the neutron noise behavior. • Computational biases exist close to the noise source. [ABSTRACT FROM AUTHOR]

• The data analysis of the neutron noise experiment in the CROCUS reactor is performed. • The APOLLO3 MOC code for the 2D reactor configuration and an "open" leakage modeling are used. • A new "improved" point kinetics neutron noise simulator with varying shapes is presented. • The validation and interpretation of the experiments using the point kinetics model is carried out. In the framework of the European project CORTEX, included in the H2020 program, the experimental campaign carried out with the CROCUS reactor, at the École Polytechnique Fédérale de Lausanne (EPFL) in Switzerland, aims at setting up methodologies and tools for the analysis and the interpretation of neutron noise in view of their industrial applications. An important part of the CORTEX project deals with code development and validation. In this paper, a new methodology for the simulation of neutron noise has been formulated and implemented in the APOLLO3 code. In the first part of this paper, the experimental setup and the methodology for the experimental data analysis developed by CEA are presented. The experimental outcomes are reported in terms of Cross-Power Spectral Densities (CPSD) of detector responses. The results produced with CEA analysis are compared to those obtained by EPFL. The second part of the work presents the noise simulator based on an improved point-kinetics model employed in a fully heterogeneous 2D transport calculation including its theoretical derivation. The computed results, i.e. the temporal signals of the detector responses, are treated to recover the CPSDs to be compared to the experimental ones. In the third part of the work, the paper presents a series of interpretive exercises performed with the IPK noise model aiming at showing its simulation capabilities and at trying to address some of the discrepancies observed in the direct comparison. Beside the promising outcomes of the simulations, some work is still being done from the experimental and computational side for a better comprehension of the problem before the extension of the model to full 3D problems and industrial application. [ABSTRACT FROM AUTHOR]

• Criticality safety evaluation of a final disposal repository for high-level waste. • Formulation of canister corrosion scenarios and investigation of system criticality. • Reported reactivity changes due to variability in porewater composition and density. • Investigation of system criticality considering canister manufacturing tolerances. Evaluating the criticality safety of a final disposal concept for high-level waste (HLW) in the repository post-closure phase is a regulatory requirement for many national radioactive waste management organisations. The Swiss National Cooperative for the Disposal of Radioactive Waste, Nagra, is presently elaborating the technical foundation for a criticality safety assessment for the Swiss HLW deep geological repository concept. The very long timescale (one million years in Switzerland) that must be considered in the assessment represents one of the main challenges. In this context, the evolution with time of the system defined by the final disposal canister (ELB) filled with spent nuclear fuel (SNF) and its potential impact on the system's neutron multiplication factor must be considered and evaluated. One of Nagra's key safety requirements for the ELB design is a canister lifetime of 10,000 years. Various interactions between the natural and engineered barriers can occur over the long timeframe, leading to alterations of the initial geometry and material composition of the canister and the surrounding backfill material. To study their potential effect on the system criticality change in terms of Δk eff (referred to in the following also as "reactivity change"), several conservative simplified ELB corrosion scenarios have been formulated by Nagra and investigated in collaboration with the Paul Scherrer Institute (PSI). The reactivity effects of the postulated ELB corrosion scenarios were studied at PSI by means of detailed criticality simulations carried out with the MCNP® code. This paper presents the conservative ELB corrosion scenarios considered in the analysis and their implementation. The observed relative impact on the system reactivity is then discussed. Additional preliminary considerations concerning the expected variability in porewater composition and canister manufacturing tolerances are also addressed. [ABSTRACT FROM AUTHOR]

Abstract: For the development of new acceptance criteria for the analysis of Rod-Ejection-Accidents (REA) in Pressurized Water Reactor UO2 cores, full-width-at-half-maximum power pulse widths in the range 25–40ms were employed in the analytical transient fuel behaviour studies. Due to the operation of MOX cores in Switzerland, a study using the CORETRAN 3D kinetics code was carried out at PSI to investigate and compare on the basis of real operated cycles MOX and UO2 cores, the pulse width magnitude as function of rod reactivity for postulated super-prompt critical REAs. This study is presented in a first part of this paper. Thereafter, to quantify the variation in pulse width when applying a different 3D kinetic solver, a benchmarking with the SIMULATE-3K code is carried out. This code-to-code comparison aims also to serve as an assessment of SIMULATE-3K for REA analyses of the Swiss cores since only CORETRAN had so far been benchmarked at PSI for this type of applications. Finally, to provide confidence in the results, particularly for pulse widths below 25ms, sensitivity studies are performed to assess the effects of specific modelling options and assumptions applied in the REA 3D kinetic analysis and to compare the sensitivities between UO2 and MOX cores. On the basis of all these studies, it is found that the MOX core pulse width is usually around 5 to 10ms lower than for UO2 cores. For rod worths close to 2.0$, corresponding to around 500 pcm above prompt criticality and which was obtained for many of the investigated MOX cycles particularly at end-of-cycle conditions, a representative MOX pulse width magnitude is found to be around 15ms. To shortly address the applicability of these results, i.e. obtained for very specific MOX cores, additional calculations were performed for hypothetical cases with increased MOX core fractions and with higher MOX assembly Pu enrichments. And it was found that even for these types of cores, the pulse width would, although being slightly reduced, remain in the range 10–15ms for rod reactivities up to 500 pcm above prompt criticality. [Copyright &y& Elsevier]

Abstract: This paper presents an integrated multicriteria analysis method for the quantitative evaluation of a state’s nuclear nonproliferation credibility level. Underscoring the implications of policy on the sources of political demand for nuclear weapons rather than focusing on efforts to restrict the supply of specific weapons technology from the “haves” to the “have-nots”, the proposed methodology considers the political, social, and cultural dimensions of nuclear proliferation. This methodology comprises three steps: (1) identifying the factors that influence credibility formation and employing them to construct a criteria tree that will illustrate the relationships among these factors; (2) defining the weight coefficients of each criterion through pairwise comparisons of the Analytical Hierarchy Process (AHP); and (3) assigning numerical scores to a state under each criterion and combining them with the weight coefficients in order to provide an overall assessment of the state. The functionality of this methodology is examined by assessing the current level of nuclear nonproliferation credibility of four countries: Japan, North Korea, South Korea, and Switzerland. [Copyright &y& Elsevier]

Abstract: This paper outlines the approach to the evaluation of sustainability of current and future electricity supply options of interest for a major Swiss utility Axpo Holding AG. The motivation behind this effort has been to provide a solid basis for a state-of-the-art interdisciplinary assessment and use this framework within a dialog with a wide spectrum of stakeholders. The development and implementation of the methodology was coordinated by Axpo in co-operation with the Paul Scherrer Institut (PSI) and other scientific institutions. The evaluation covers environmental, social and economic dimensions of sustainability. Methods used include among others life cycle assessment (LCA), impact pathway approach (IPA) and probabilistic safety assessment (PSA). The associated databases developed by PSI have been extensively used, subject to major extensions necessary for analyzing the future technologies. Learning curves were employed for future cost estimates. Furthermore, particularly in the social area expert surveys were used. The results were aggregated using total (internal plus external) costs approach and multi-criteria decision analysis (MCDA). For MCDA a set of criteria and the associated indicators was established. In total 75 indicators were quantified, including 11 environmental, 33 social and 31 economic. Eighteen current and 18 future technologies have been analysed including nuclear as well as fossil and renewable technologies. Total costs were estimated for these technologies providing a clear ranking with nuclear having the lowest costs and some of the renewables showing remarkable cost reductions until 2030. This ranking is partially controversial mainly due to the limited representation of social aspects in the total costs. The results of MCDA-applications involving elicitation of preferences from a relatively homogeneous stakeholder group, i.e. 85 employees of the Axpo Group (including also NOK, EGL, CKW and Axpo IT), are summarized. In addition, sensitivity of technology ranking to preference profiles is demonstrated. Broader consideration of social factors favours renewables and depending on the specifics of preference profiles may lower the ranking of nuclear. Further applications of the MCDA-approach with various stakeholder groups are planned. [Copyright &y& Elsevier]

Abstract: This paper presents the development and application of a methodology to estimate, using conventional deterministic lattice/core analysis methods, the fast neutron fluence at the tips of control rods inserted during operation in PWR reactors. The developed methodology is based on a 3-D Nodal Diffusion/2-D Lattice Transport multi-step calculation scheme, in which the operating history of the nuclear environment around the tip is tracked in 3-D core follow analyses and used thereafter to provide boundary conditions for 2-D transport calculations to compute the fast neutron flux. In subsequent steps, radial and axial correction factors, both based on fast flux results from the 3-D core simulator, are applied to the calculated 2-D transport flux in order to take into account radial leakage effects as well as axial flux gradients around the tip. The fluence is finally estimated through a time-integration of the corrected 2-D transport fast flux. The developed methodology has been applied to estimate the fluence for a total of 15 control rods, over 21 operating cycles of a Swiss nuclear power plant. [Copyright &y& Elsevier]

This paper gives a brief description of a filter back projection (FBP) and Algebraic Reconstruction techniques (ART), which are used for reconstruction of projected data obtained from transmission measurements. Both techniques are tested by reconstructing a simulated object, showing that the FBP technique yields images with better resolution and higher contrast, thus resulting in more accurate information than obtained by ART.Experimental data has been obtained for unheated and heated ordinary and heavy concrete samples by transmission measurements using a thermal neutron beam emitted from the spallation neutron source SINQ at PSI, Switzerland. The data are used to reconstruct the images of the examined samples using the FBP technique with different filter functions, including the Shepp Logan filter function produces images of higher contract, more smoothing and slightly better resolution than those obtained using the absolute filter function. Based on these images the samples were investigated for water content, cracks, homogeneity, etc. [Copyright &y& Elsevier]