Development and verification of a methodology for neutron noise response to fuel assembly vibrations

Nuclear reactors are stochastic systems, in which several parameters fluctuate even during steady-state conditions. In addition, structural components like fuel assemblies, vibrate due to strong hydraulic forces. This stochasticity is the cause of the neutron population fluctuating behavior; a phenomenon referred as neutron noise. This paper systematically analyzes the capabilities of the Studsvik's codes to model the fuel assembly vibration and to assess the neutron noise response. To this aim, the CASMO-5 code is utilized for generating cross sections by parametrizing the water-gap thickness of fuel segments. The fuel assembly vibration modelling is approximated by the dynamic modification of the water-gap widths between adjacent assemblies using the transient code SIMULATE-3K. The cross sections dependency on the water-gap width is analyzed and successfully compared against Serpent-2 reference results. Last, the utilized codes capability is verified qualitatively or/and quantitatively through a series of cases, at both lattice and nodal level. (C) 2020 The Author(s). Published by Elsevier Ltd.

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

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

Gaëtan Girardin, Mathieu Hursin, Andreas Pautz, Adolfo Rais, Daniel Jerôme Siefman

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.

2015

Delayed gamma fraction determination in the zero power reactor CROCUS

Pavel Frajtag, Vincent Pierre Lamirand, Tom Mager, Oskari Ville Pakari, Andreas Pautz

Gamma rays are an inextricable part of a nuclear reactor's radiation field, and as such require characterization for dose rate estimations required for the radiation protection of personnel, material choices, and the design of nuclear facilities. Most commonplace radiation transport codes used for shielding calculations only included the prompt neutron induced component of the emitted gamma rays. The relative amount of gamma rays that are emitted from delayed processes - the delayed gamma fraction - amount to a significant contribution, e.g. in a typical zero power reactor at steady state is estimated to be roughly a third. Accurate predictions of gamma fields thus require an estimation of the delayed content in order to meaningfully contribute. As a consequence, recent code developments also include delayed gamma sources and require validation data. The CROCUS zero power research reactor at EPFL is part of the NEA IRPhE and has therefore been characterized for benchmark quality experiments. In order to provide the means for delayed gamma validation, a dedicated experimental campaign was conducted in the CROCUS reactor using its newly developed gamma detection capabilities based on scintillators. In this paper we present the experimental determination of the delayed gamma fraction in CROCUS using in-core neutron and gamma detectors in a benchmark reactor configuration. A consistent and flexibly applicable methodology on how to estimate the delayed gamma fraction in zero power reactors has hitherto not existed - we herein present a general experimental setup and analysis technique that can be applied to other facilities. We found that the build-up time of relevant short lived delayed gamma emitters is likely attributed to the activation of the aluminium cladding of the fuel. Using a CeBr3 scintillator in the control rod position of the CROCUS core, we determined a delayed gamma fraction of (30.6 +/- 0.6)%.