Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

Gaëtan Girardin

During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. It is in this context that the Generation IV international forum (GIF) launched the initiative, in 2000, to collaborate on the research and development (R&D) efforts needed for the next generation, i.e. Generation IV, of nuclear reactors. These advanced systems will be ideally deployed beyond the year 2030, following the Generation III or III+ nuclear power plants, which are mainly based on light water technology and are currently entering deployment. A particular goal set for Generation IV systems is closure of the nuclear fuel cycle. Thus, apart from improvements in safety, they are expected to offer a better utilization of natural resources, as also a minimization of long-lived radioactive wastes. Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials (fuel pellets of mixed uranium-plutonium carbide within a plate-type, honeycomb structure made of SiC). It is in the context of the large, 2400 MWth reference GFR design that the present doctoral research has been conducted, the principal aim having been to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for this system. The work has been carried out in three successive and complementary phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behavior studies for hypothetical CA driven transients. The deterministic code system ERANOS and its associated nuclear data libraries for fast reactors were developed and validated for the previous generation of sodium-cooled reactors. The validation of ERANOS for GFR applications was, therefore, the first task to be realized in the present research. This has entailed a systematic reanalysis of the GFR-relevant, integral data generated at PSI during the GCFR-PROTEUS experimental program of the 1970's. Thus, during the first phase of the thesis, the reference PROTEUS test lattice from these experiments has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1, in order to derive a reference computational scheme to be used later for the GFR analysis. Additionally, benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. It has been found that, for the main reaction rate ratios, the new analysis of the GCFR-PROTEUS reference lattice generally yields good agreement – within 1σ measurement uncertainty – with experimental values and with the Monte Carlo simulations. As shown by the analysis, the predictions were in somewhat better agreement in the case of the adjusted ERALIB1 library. The applicability of ERANOS-2.0/ERALIB1 as the reference neutronics tool for the GFR analysis could thus be demonstrated. Furthermore, neutronics aspects related to the novel features of the GFR, for which new experimental investigations are needed, were highlighted. In the second phase of the research, the CA pattern was developed for the GFR, based on iterative neutronics and thermal-hydraulics calculations, 2D and 3D neutronics models for the reactor core having first been set up using the reference ERANOS-2.0/ERALIB1 computational scheme. For the thermal-hydraulics analysis, the CEA code COPERNIC was used. This design work was followed by the study of an appropriate CA implementation scheme (number of CAs and corresponding positions within the core). Detailed neutronics studies revealed the existence of large CA interaction effects, so-called shadowing/anti-shadowing effects leading to an amplification/reduction of the CA worth. The interactions between the absorber pins within a CA, and between the CAs themselves, were investigated in detail, with the goal to optimize the CA efficiency, in terms of the absorber fraction and minimization of the associated heterogeneity effects. The proposed CA pattern consists of 54 absorber pins placed in a triangular lattice. Each absorber pin is a stainless-steel tube filled with highly enriched 10B boron carbide pellets. As a result of the detailed investigations, the absorber pin diameter could be chosen such as to minimize the pin-to-pin influence within the assembly. In particular, a central part of the CA was designed without any absorber pins (zone filled with stagnant helium). A final reduction of the heterogeneity effect (difference between homogeneous and heterogeneous treatments) to 13% was achieved through this feature. Of special importance, the neutronics investigations performed for the reference GFR core ("2004-Core"), especially those related to the CA interactions, have directly contributed to a new core design ("2007-Core"), with the height-to-diameter ratio having been increased to 0.6, compared to 0.3 for the reference core. During the third phase, detailed coupled, 3D neutron-kinetics (NK) and 1D thermal-hydraulics (TH) models were developed for the GFR core, the aim being to arrive at an in-depth understanding of the 3D core behavior during CA driven transients, especially from the viewpoint of spatial effects. The coupled models were developed using the PARCS code for the 3D NK and the TRACE code for the 1D TH modeling. Particular attention was paid to have each individual fuel sub-assembly and CA represented, in order to allow the analysis of local deformations of the 3D distributions of power and safety related parameters, such as the coolant, cladding and fuel temperatures. The validation of the coupled full-core models was performed against reference ERANOS-VARIANT calculations. In particular, the CA worth and reactivity feedbacks were benchmarked, the discrepancies being shown to be relatively low (

EPFL2009

Experimental and numerical study of stochastic branching noise in nuclear reactors

Oskari Ville Pakari

Noise analysis applied to nuclear reactor physics is a powerful tool to investigate a reactor's kinetic parameters, and more generally underlying physical processes determining core behavior. The kinetic parameters are the coefficients of simplified time-dependent neutron population equations, the so-called point kinetics equations. Experimentally determined kinetic parameters aid to validate codes and to potentially evaluate nuclear data. This thesis focuses on improvements to the kinetic parameter and uncertainty quantification, the experimental techniques, and the direct simulation of noise experiments. In particular, the notion of spatial effects, i.e. effects that render point kinetics assumptions inaccurate for noise measurements, was investigated. This was achieved by drawing on experiments conducted in the zero power reactor CROCUS. New detection instrumentation for noise experiments was developed for CROCUS, namely current mode amplifiers for neutron detectors to allow for higher detection efficiency and thus precision; and a scintillation-based gamma detection array, called LEAF, to enable the study of gamma noise. A set of reference neutron and gamma noise experiments were conducted to exemplify an improved method of parameter and uncertainty estimation based on bootstrapping. It lead to a full uncertainty budget showing a relative uncertainty of 3.6% and 1.3% on the prompt decay constant with neutron and gamma noise, respectively. Moreover, reporting of full distributions of kinetic parameters was shown to be required in order to provide an accurate representation of the experimental results. Gamma noise was shown to be superior in terms of precision by a factor of at least two compared to neutron noise when determining the prompt decay constant. To study spatial effects and the spatial extent of the noise field of CROCUS, a set of experiments with varying detector locations, reactivity, and particle type were conducted. The neutron noise field study showed that measurements are limited to the immediate proximity of the core, and that CROCUS is reliably modeled as a point kinetic reactor for static configurations. Nonetheless, a systematic trend was shown when comparing to code predictions, pointing to a weak spatial effect biasing neutron noise measurements at a distance. The observable gamma noise field using high efficiency detectors was shown to extend beyond neutron noise limits, enabling ex-vessel measurements. The prompt decay constant was determined at about 0.9 meters to the core center with comparable accuracy to that of gamma in-core reference experiments using less-efficient detectors. Gamma correlations were be observed outside of the reactor cavity in front of an experimental channel, enabling a prompt decay constant determination at 7 meters to the core center. To complement the experimental investigation and to study the underlying physics, a simulation methodology to estimate noise responses was developed. Specifically, the explicit simulation of noise experiments using analog Monte Carlo transport coupled to a fission model was studied. An established code with an existing fission library coupling, TRIPOLI-4, was used as reference. In addition, a newly developed coupling of a fission library to Serpent 2 allowed for code-to-code comparison. The full methodology, verification, and validation to the aforementioned CROCUS experiments is presented.

EPFL2020

Development of Advanced Methodologies for Monitoring and Modelling of Neutron Noise in Modern LWR Cores

Dionysios Chionis

Nuclear reactors are inherently stochastic systems, in which neutronic and thermal-hydraulic parameters fluctuate continuously even during steady-state conditions. In addition, structural components vibrate due to the coolant hydraulic forces. This stochasticity is the cause of the neutron population fluctuating behavior, a phenomenon referred as neutron noise. The neutron noise is monitored over the reactor lifetime, providing valuable knowledge of the core behavior. More importantly, the neutron noise monitoring is used for the early detection of reactor anomalies. The neutron noise phenomenon is an intensively-studied research topic, leading the development of noise surveillance methods, signal processing techniques and analytical solvers. Nevertheless, an unexpected neutron noise level increase trend, observed in the last decade in several reactors causing undesirable costly operational consequences, triggered an increasing interest. The new observations revealed the need for an improved neutron noise behavior understanding. In this context, the current doctoral research intents to systematically analyze the neutron noise phenomenon by developing innovative noise modelling methodologies. First, main attention is given to the neutron noise characteristics identification of the Swiss Gösgen nuclear reactor (KKG). KKG plant data are analyzed in the time and frequency domains using traditional signal processing techniques, and key aspects of the KKG noise phenomenology are revealed, allowing the better characterization of the noise sources affecting its dynamic behavior. Then, a neutron noise modelling methodology is developed, utilizing advanced neutronic solvers. These codes are used to systematically model key noise sources (i.e. fuel assembly vibration, and inlet coolant temperature and flow fluctuations), and to study their impact on the neutron noise. For the first time, the fuel assembly vibration model in the utilized codes is systematically studied and is qualified at a 3D full core level. This detailed work, demonstrate the capabilities and the robustness of the newly developed PSI neutron noise methodology to successfully model key noise sources and to reproduce neutron noise phenomena. In addition, a comparative study between the simulated results and the KKG measured data revealed that, the stochastic fluctuation of the inlet coolant temperature in combination with the fuel assembly vibration have a primary role on the KKG neutron noise behavior. Most importantly, it is observed that, the neutron noise increase trend observed in KKG could be explained, at some extent, by the introduction in the core of a newer fuel design which is more susceptible in lateral vibrations. Last, a new in-house methodology is established serving as a supportive diagnostic tool for the detailed identification of signals connectivity patterns. To this aim, the PSI connectivity analysis methodology is based on the causality analysis principles to indicate the cause-and-effect interactions between reactor signals. The nuclear core is analyzed utilizing the most prominent causality analysis techniques in the frequency domain (i.e. rPDC and DTF). The current research is the first application of causality analysis techniques on nuclear reactor data. The application of the developed methodology in measured and simulated datasets showed that, it can successfully indicate the causal interconnections and the perturbation root-cause.

EPFL2020

Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

Gaëtan Girardin

EPFL2009

Experimental and numerical study of stochastic branching noise in nuclear reactors

Oskari Ville Pakari

EPFL2020

Development of Advanced Methodologies for Monitoring and Modelling of Neutron Noise in Modern LWR Cores

Dionysios Chionis

EPFL2020