New approaches to provide feedback from nuclear and covariance data adjustment for effective improvement of evaluated nuclear data files

A critical examination of the role of uncertainty assessment, target accuracies, role of integral experiment for validation and, consequently, of data adjustments methods is underway since several years at OECD-NEA, the objective being to provide criteria and practical approaches to use effectively the results of sensitivity analyses and cross section adjustments for feedback to evaluators and experimentalists in order to improve without ambiguities the knowledge of neutron cross sections, uncertainties, and correlations to be used in a wide range of applications and to meet new requirements and constraints for innovative reactor and fuel cycle system design. An approach will be described that expands as much as possible the use in the adjustment procedure of selected integral experiments that provide information on “elementary” phenomena, on separated individual physics effects related to specific isotopes or on specific energy ranges. An application to a large experimental data base has been performed and the results are discussed in the perspective of new evaluation projects like the CIELO initiative

Development of a new method to determine the axial void velocity profile in BWRs from measurements of the in-core neutron noise

Determination of the local void fraction in BWRs from in-core neutron noise measurements requires the knowledge of the axial velocity of the void. The purpose of this paper is to revisit the problem of determining the axial void velocity profile from the transit times of the void between axially placed detectors, determined from in-core neutron noise measurements. In order to determine a realistic velocity profile which shows an inflection point and hence has to be at least a third order polynomial, one needs four transit times and hence five in-core detectors at various axial elevations, whereas the standard instrumentation usually consists only of four in-core detectors. Attempts to determine a fourth transit time by adding a TIP detector to the existing four LPRMs and cross-correlate it with any of the LPRMs have been unsuccessful so far. In this paper we thus propose another approach, where the TIP detector is only used for the determination of the axial position of the onset of boiling. By this approach it is sufficient to use only three transit times. Moreover, with another parametrisation of the velocity profile, it is possible to reconstruct the velocity profile even without knowing the onset point of boiling, in which case the TIP is not needed, although at the expense of a less flexible modelling of the velocity profile. In the paper the principles are presented, and the strategy is demonstrated by concrete examples, with a comparison of the performance of the two different ways of modelling the velocity profile. The method is tested also on velocity profiles supplied by system codes, as well as on transit times from neutron noise measurements

Case Study of Data Assimilation Methods with the LWR-Proteus Phase II Experimental Campaign

This paper describes the application of data assimilation methods to CASMO-5 simulations of a Proteus research reactor experiment. Its focus is a comparison and evaluation of three prominent data assimilation methods: generalized linear least squares, MOCABA, and Bayesian Monte Carlo. These methods have not yet been extensively compared to date. The experiment is an interesting case study for this comparison because the measured reactivity worth response can be non-linear. This study investigates the effects that non-linearity has upon the agreement between the methods. The adjusted calculated values, calculation uncertainty, and nuclear data are all investigated to compare and evaluate the methods. The presented results provide evidence supporting the hypothesis that for linear responses, all of the data assimilation methods agree well. But when the responses become more non-linear, significant disagreements occur between generalized linear least squares results and those of MOCABA and Bayesian Monte Carlo

Uncertainty analyses of neutron noise simulations in a Zero-Power reactor

A comprehensive uncertainty analysis methodology has been established for the modeling of stationary\ua0neutron flux\ua0oscillations induced by fuel rods vibration in a zero-power reactor. The methodology includes uncertainty propagation and sensitivity analysis. The target event is based on an actual experimental campaign at the CROCUS zero-power reactor and corresponds to the simultaneous oscillation of 18 metallic uranium fuel rods in the periphery of the core. Both the uncertainty propagation and the sensitivity analysis commonly use a large part of the entire analysis process, from the selection of uncertain parameters to the actual code simulations. Applying a random sampling-based approach, the input parameters are sampled\ua0N\ua0times from their distribution information and used as inputs for\ua0N\ua0noise simulations using CORE SIM\ua0+. The quantity of interest (QoI) is the amplitude of the Auto-Power Spectral Density at various detector locations, which is normalized by the amplitude of the Cross-Power Spectral Density of the reference detector. Their uncertainties are determined following the 4th order Wilks’ formula for two-sided limits. Through the determination of correlations among QoI at the installed detector locations, it is demonstrated that the neutron noise near the area of oscillating fuel rods (noise source) have different behavior compared to the neutron noise further away from the noise source. The following sensitivity analyses are carried out using multiple\ua0correlation coefficients\ua0within grouped parameters. As expected from the QoI correlations, the QoIs at two different locations (near and far from the noise source) are influenced by different input parameters. Near the noise source, the QoI uncertainty is driven by the uncertainties in the position of the noise source, while the uncertainties in the nuclear data for U-235 and U-238 are the leading contributors further away from the source. This paper provides general information on how to perform the uncertainty analyses for neutron noise simulations, as well as quantitative estimates of the computational uncertainty required for the validation of the computer programs under development for the simulation of neutron noise

VERIFICATION OF A REACTOR PHYSICS CALCULATION SCHEME FOR THE CROCUS REACTOR

CROCUS is a zero power (100 W) reactor of the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology, Lausanne (EPFL). It is used for teaching and research purposes. Its modeling has relied so far on diffusion theory and point kinetics for the neutronic analysis and simplified thermal hydraulics models for accident analysis. Recently, an effort has started within the LRS to improve its modeling capabilities, the long term goal being to update the CROCUS Safety Analysis Report (SAR) for improved operational flexibility. The present work is focused on the static neutron analysis of CROCUS through the development and preliminary verification of a 3D nodal simulator (e.g. PARCS) model of the reactor, a methodology typically used in the industry for modeling of Light Water Reactors (LWR). The set of homogenized macroscopic cross-sections needed by the core simulator, referred in this work as nuclear data library, is generated by a Monte Carlo based code (e.g. Serpent). The quantities of interest for the verification of the model are the keff, and the control rod worths. An innovative homogenization approach to generate the nuclear data library is considered due to the irregular radial geometry of the CROCUS reactor. The reference solution is provided by another Monte Carlo code, MCNP5. The uncertainty due to the nuclear data in the keff prediction of Serpent is also investigated and amounts to about 500pcm which covers the deviation from unity of keff prediction by MCNP5 and Serpent for a critical CROCUS configuration. PARCS keff predictions are within 400 pcm of the Serpent results

Sensitivity analysis in core diagnostics

In the CORTEX project, methods to simulate neutron flux oscillations were enhanced and machine-learning based tools to determine the causes of measured neutron flux oscillations were developed, using the results of simulations as training and validation data. For a selected combination of those methods and tools, several sensitivity analyses were performed to assess their robustness and trustworthiness. The neutron flux oscillations were simulated using the tool CORE SIM+. It calculates the three-dimensional field of the neutron flux oscillations, which can be used to determine the response of neutron detectors at given locations. For the sensitivity analysis, the neutron flux oscillations were assumed to be caused by the vibration of one fuel element. It was investigated how selected input parameters like the core loading pattern, the burn up of the fuel elements, the neutronic core data, the geometry details of the vibrating fuel element, the chosen detectors, and other noise source parameters like the amplitude of the fuel element vibrations, affect the simulated neutron flux oscillations. A three dimensional fully convolutional neural network had been developed and trained during the CORTEX project to determine the cause and location of perturbations causing given measurements of in-core detectors in pressurized water reactors. The robustness of this network was tested by applying it to the simulated detector readings created during the sensitivity analysis

Convergence Analysis and Criterion for Data Assimilation with Sensitivities from Monte Carlo Neutron Transport Codes

Sensitivity coefficients calculated with Monte Carlo neutron transport codes are subject to statistical fluctuations. The fluctuations affect parameters that are calculated with the sensitivity coefficients. The convergence study presented here describes the effects that statistically uncertain sensitivities have on first-order perturbation theory, uncertainty quan-tification, and data assimilation. The results show that for data assimilation, posterior nuclear data were remarkably uninfluenced by fluctuations in sensitivity mean values and by sensitivity uncertainties. Posterior calculated values computed with first-order perturbation theory showed larger dependence on sensitivity mean-value convergence and small uncertainty arising from the sensitivities' uncertainties. A convergence criterion is proposed for stopping simulations once the sensitivity means are sufficiently converged and their uncertainties are sufficiently small. Employing this criterion economizes computational resources by preventing an excess of particle histories from being used once convergence is achieved. The criterion's advantage is that it circumvents the need to set up the full data assimilation procedure, but is still applicable to data assimilation results

Future experimental programmes in the CROCUS reactor

CROCUS is a teaching and research zero-power reactor operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology (EPFL). Three new experimental programmes are scheduled for the forthcoming years. The first programme consists in an experimental investigation of mechanical noise induced by fuel rods vibrations. An in-core device has been designed for allowing the displacement of up to 18 uranium metal fuel rods in the core periphery. The vibration amplitude will be 6 mm in the radial direction (±3 mm around the central position), while the frequency can be tuned between 0.1 and 5 Hz. The experiments will be used to validate computational dynamic tools currently under development, which are based on DORT-TD and CASMO/S3K code systems. The second programme concerns the measurement of in-core neutron noise for axial void profile reconstruction. Simulations performed at Chalmers University have shown how the void fraction and velocity profiles can be reconstructed from noise measurements. The motivation of these experiments is to develop an experimental setup to validate in-core the method in partnership with Chalmers University. The third experimental programme aims at continuing the validation effort on the nuclear data required in the calculation of GEN-III PWR reactors with heavy steel reflectors. This is a collaboration with CEA Cadarache that extends the results of the PERLE experiments carried out in the E reactor at CEA. Scattering cross sections at around 1 MeV will be studied separately by replacing successively the water reflector by sheets of stainless steel alloy and pure metals – iron, nickel, and chromium. Data will be extracted from the measured flux attenuation using foils in the metal reflector and from the criticality effects of these reflectors. In parallel to the three reactor experiments, we develop in-core detectors and measurement systems. Following the last development of a neutron noise measurement station in pulse mode, a second neutron noise station in current mode is being designed. In current mode the reactor can be used at higher power without dead time effects. It allows faster measurement time or lower results uncertainties. Finally, a joint development of a full new detection system based on chemical vapour deposited (sCVD) diamond has been started with the CIVIDEC instrumentation start-up company. A first prototype has been tested in November 2015 in CROCUS. One of the main purposes is to work on the discrimination of gammas, thermal and fast neutrons for demonstrating the interest of this detector type in a mixed neutron-gamma field

Modeling noise experiments performed at AKR-2 and CROCUS zero-power reactors

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

New approaches to provide feedback from nuclear and covariance data adjustment for effective improvement of evaluated nuclear data files

A critical examination of the role of uncertainty assessment, target accuracies, role of integral experiment for validation and, consequently, of data adjustments methods is underway since several years at OECD-NEA, the objective being to provide criteria and practical approaches to use effectively the results of sensitivity analyses and cross section adjustments for feedback to evaluators and experimentalists in order to improve without ambiguities the knowledge of neutron cross sections, uncertainties, and correlations to be used in a wide range of applications and to meet new requirements and constraints for innovative reactor and fuel cycle system design. An approach will be described that expands as much as possible the use in the adjustment procedure of selected integral experiments that provide information on “elementary” phenomena, on separated individual physics effects related to specific isotopes or on specific energy ranges. An application to a large experimental data base has been performed and the results are discussed in the perspective of new evaluation projects like the CIELO initiative