Andrea Alfonsi

New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors

The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential), specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead), were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

Comparison of the PHISICS/RELAP5-3D Ring and Block Model Results for Phase I of the OECD/NEA MHTGR-350 Benchmark

Abstract The Parallel and Highly Innovative Simulation for INL Code System (PHISICS) has been under development at Idaho National Laboratory since 2010. It consists of several modules providing improved coupled core simulation capability: INSTANT (Intelligent Nodal and Semi-structured Treatment for Advanced Neutron Transport) (three-dimensional nodal transport core calculations); MRTAU (Multi-Reactor Transmutation Analysis Utility) (depletion and decay heat generation); and modules performing criticality searches, fuel shuffling, and generalized perturbation. Coupling of the PHISICS code suite to the thermal-hydraulic system code RELAP5-3D was finalized in 2013, and as part of the verification and validation effort, the first phase of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MHTGR-350 benchmark has now been completed. The theoretical basis and latest development status of the coupled PHISICS/RELAP5-3D tool are described in more detail in a concurrent paper. This paper provides an overview of the OECD/NEA MHTGR-350 benchmark and presents the results of exercises 2 and 3 defined for phase I. Exercise 2 required the modeling of a stand-alone thermal fluids solution at the end of equilibrium cycle for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The RELAP5-3D results of four subcases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 required a coupled neutronics and thermal fluids solution, and the PHISICS/RELAP5-3D code suite was used to calculate the results of two subcases. The main focus of this paper is a comparison of results obtained with the traditional RELAP5-3D “ring” model approach against a much more detailed model that includes kinetics feedback on individual “block” level and thermal feedbacks on a triangular submesh. The higher fidelity that can be obtained by this block model is illustrated with comparison results on the temperature, power density, and flux distributions. It is shown that the ring model leads to significantly lower fuel temperatures (up to 10%) when compared with the higher-fidelity block model and that the additional model development and run-time efforts are worth the gains obtained in the improved spatial temperature and flux distributions.

Rethinking sensitivity analysis of nuclear simulations with topology

In nuclear engineering, understanding the safety margins of the nuclear reactor via simulations is arguably of paramount importance in predicting and preventing nuclear accidents. It is therefore crucial to perform sensitivity analysis to understand how changes in the model inputs affect the outputs. Modern nuclear simulation tools rely on numerical representations of the sensitivity information - inherently lacking in visual encodings - offering limited effectiveness in communicating and exploring the generated data. In this paper, we design a framework for sensitivity analysis and visualization of multidimensional nuclear simulation data using partition-based, topology-inspired regression models and report on its efficacy. We rely on the established Morse-Smale regression technique, which allows us to partition the domain into monotonic regions where easily interpretable linear models can be used to assess the influence of inputs on the output variability. The underlying computation is augmented with an intuitive and interactive visual design to effectively communicate sensitivity information to nuclear scientists. Our framework is being deployed into the multipurpose probabilistic risk assessment and uncertainty quantification framework RAVEN (Reactor Analysis and Virtual Control Environment). We evaluate our framework using a simulation dataset studying nuclear fuel performance.

BWR Station Blackout: A RISMC Analysis Using RAVEN and RELAP5-3D

Abstract The existing fleet of nuclear power plants is in the process of having its lifetime extended and having the power generated from these plants increased via power uprates and improved operations. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) pathway aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions and accident scenarios. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess the impact of power uprate of a boiling water reactor system during a station blackout accident scenario. We employ a system simulator code, RELAP5-3D, coupled with RAVEN, which performs the stochastic analysis. Our analysis is performed by (a) sampling values from a set of parameters from the uncertainty space of interest, (b) simulating the system behavior for that specific set of parameter values, and (c) analyzing the outcomes from the set of simulation runs.

Simulation of AER-DYN-002 and AER-DYN-003 Control Rod Ejection Benchmarks by RELAP5-3D/PHISICS Coupled Codes

Abstract ENEA “Casaccia” Research Center is collaborating with Idaho National Laboratory performing activities devoted to the validation of the Parallel and Highly Innovative Simulation for INL Code System (PHISICS) neutron simulation code. In such framework, the AER-DYN-002 and AER-DYN-003 control rod (CR) ejection benchmarks were used to validate the coupled codes RELAP5-3D/PHISICS. The AER-DYN-002 benchmark provides a test case of a CR ejection accident in a VVER-440 at hot-zero-power and end-of-cycle conditions assuming an adiabatic fuel and taking into account only the fuel temperature feedback. The AER-DYN-003 benchmark is based on the same problem; however, the moderator density feedback and the coolant heat removal are also considered. A RELAP5-3D core channel-by-channel, thermal-hydraulic nodalization was developed and coupled, first with the RELAP5-3D internal neutronic routine NESTLE and then with the PHISICS code. Analysis of the AER-DYN-002 results shows that the steady-state solutions are in good agreement with the other participants’ average solution, while some differences are shown in the transient simulations. In the AER-DYN-003 benchmark, however, both steady-state and transient results are in good agreement with the average solution.

A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

Light Water Reactor Sustainability Program Status of Adaptive Surrogates within the RAVEN framework

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RAVEN Beta Release

This documents the release of the Risk Analysis Virtual Environment (RAVEN) code. A description of the RAVEN code is provided, and discussion of the release process for the M2LW-16IN0704045 milestone. The RAVEN code is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. RAVEN is capable of investigating the system response as well as the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. RAVEN has now increased in maturity enough for the Beta 1.0 release.

A Fuel Cycle and Core Design Analysis Method for New Cladding Acceptance Criteria Using PHISICS, RAVEN and RELAP5-3D

The Nuclear Regulatory Commission (NRC) has considered revising the 10 CFR 50.46C rule [1] for analyzing reactor accident scenarios to take the effects of burn-up rate into account. Both maximum temperature and oxidation of the cladding must be cast as functions of fuel exposure in order to find limiting conditions, making safety margins dynamic limits that evolve with the operation and reloading of the reactor. In order to perform such new analysis in a reasonable computational time with good accuracy, INL (Idaho National Laboratory) has developed new multi-physics tools by combining existing codes and adding new capabilities. The PHISICS (Parallel Highly Innovative Simulation INL Code System) toolkit [2,3] for neutronic and reactor physics is coupled with RELAP5-3D [4] (Reactor Excursion and Leak Analysis Program) for the LOCA (Loss of Coolant Accident) analysis and RAVEN [5] for the PRA (Probabilistic Risk Assessment) and margin characterization analysis. In order to perform this analysis, the sequence of RELAP53D input models had to get executed in a sequence of multiple input decks, each of them had to restart and slightly modify the previous model (in this case, on the neutronic side only) This new RELAP5-3D multi-deck processing capability has application to parameter studies and uncertainty quantification. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical PWR (Pressurized Water Reactor). INTRODUCTION The nuclear power industry is continually improving its designs, safety equipment, processes, and analysis methods. The NRC is considering a revision of the requirements in 10 CFR 50.46C rule, focused on the operation of the ECCS (Emergency Core Coolant System) in LOCA scenarios [1]. Novel analysis strategies will be required to account for the effects of fuel burn-up rate. It is necessary to cast the maximum temperature and oxidation of the cladding as functions of the fuel exposure in order to find the limiting conditions of the reactor, with its different design and different reloading patterns. This revision requires the development of new tools and capabilities to calculate the dynamic phenomena of the multiphysics system to the required accuracy in a reasonable amount of time. To perform such analysis, a rigorous Probabilistic Risk Assessment (PRA) strategy must be employed. The PHISICS code toolkit [2,3] is being developed at INL to provide state of the art analysis tools to nuclear engineers. It implements many choices of algorithms and meshing schemes for optimizing accuracy needs on available computational resources. Analysis tools currently available in the PHISICS package are a nodal and semi-structured transport core solver, INSTANT, a depletion module, MRTAU, a time-dependent solver, TimeIntegrator, a cross section interpolation and manipulation framework, MIXER, a criticality search module CRITICALITY, and a fuel management and shuffling tool SHUFFLE. The tools are developed as independent modules in a pluggable fashion in order to simplify maintenance and development. PHISICS can be run in parallel to takes advantage of multiple computer cores (workstations and highperformance computing systems). The package is directly coupled with the system safety analysis code RELAP5-3D [4] through a Fortran 95 interfacing module that contains communication subroutines that translate

DAKOTA reliability methods applied to RAVEN/RELAP-7.

This report summarizes the result of a NEAMS project focused on the use of reliability methods within the RAVEN and RELAP-7 software framework for assessing failure probabilities as part of probabilistic risk assessment for nuclear power plants. RAVEN is a software tool under development at the Idaho National Laboratory that acts as the control logic driver and post-processing tool for the newly developed Thermal-Hydraulic code RELAP-7. Dakota is a software tool developed at Sandia National Laboratories containing optimization, sensitivity analysis, and uncertainty quantification algorithms. Reliability methods are algorithms which transform the uncertainty problem to an optimization problem to solve for the failure probability, given uncertainty on problem inputs and a failure threshold on an output response. The goal of this work is to demonstrate the use of reliability methods in Dakota with RAVEN/RELAP-7. These capabilities are demonstrated on a demonstration of a Station Blackout analysis of a simplified Pressurized Water Reactor (PWR).