Idaho National Laboratory Research Articles

Coupled Thermal-Hydraulic Analysis and Species Mass Transport in a Versatile Experimental Salt Irradiation Loop (VESIL) Using CTF

With the resurgence of interest in molten salt reactors, there is a need for new experiments and modeling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho National Laboratory to be placed in the Advanced Test Reactor (ATR). One of the key phenomena this proposed experiment plans to elucidate is fission product speciation in the fuel-salt and the subsequent effects this has on the fuel-salt properties, source term generation, and corrosion control. Specifically, noble gases (Xe & Kr) will bubble out to a plenum or off-gas system, and noble metals (Mo, Tc, Te, etc.) will precipitate and deposit in specific zones in the loop. This work extends the mass transfer and species interaction models in CTF (Coolant-Boiling in Rod Arrays—Two Fluids) and applies these models to give a preliminary estimation of fission product behavior in the proposed VESIL design. A noble metal–helium bubble mass transfer model is coupled with the thermal-hydraulic results from CTF to determine the effectiveness of this insoluble fission product (IFP) extraction method for VESIL. Amounts of IFP species extracted to the off-gas system and species distributions in VESIL after a 60-day ATR cycle are reported.

Development and verification of the mathematical modeling for gas-pressurizer system

The accurate simulation of the dynamic characteristics of a pressure control system (Pressurizer) has a significant role in industrial applications, especially in the nuclear industry. In this paper, the pressurizer (PRZ) used in the water process engineering and nuclear reactors are classified into three categories: Steam PRZ, Steam-gas PRZ, and Gas PRZ. Nowadays, steam PRZs are typically used in pressurized nuclear reactors. In return, due to the proper characteristics of gas PRZ, these systems are expected to be a suitable alternative for use in Small Modular Reactors (SMRs) and even other pressurized research reactors. However, there is a lack of literature about the context of the gas PRZ. As one of the works about gas PRZ, this paper suggests a mathematical model to predict the gas PRZ’s dynamic behavior and implements this framework in MATLAB software. This proposed model's performance has been evaluated and verified for two pressurized water test facilities, including the University of Wisconsin High-Pressure Critical Heat Flux (WHPCHF) facility and the Idaho National Laboratory (INL) facility as the base loops. Throughout this analysis, the suggested model performance during the temperature and pressure transient conditions are compared with the results of the PRZ modeled in the Aspen HYSYS software for both base loops. Also, the diagrams of PRZ mass inventory changes, liquid level inside the PRZ and pressure in both different test loops with different non-condensable gases (nitrogen and argon) show that the proposed mathematical model is in good agreement with the results by Aspen HYSYS. Therefore, this proposed mathematical model can suitably model the dynamic behavior of a gas PRZ. The proposed model can also be used to simulate gas PRZs control systems and the extension in studying about the modeling of these PRZs.

Cost- and Risk-Based Seismic Design Optimization of Nuclear Power Plant Safety Systems

Seismic analysis, design, and qualification of systems, structures, and components (SSCs) is a significant contributor to the capital cost of a nuclear power plant. To reduce capital costs of advanced nuclear power plants and make commercial nuclear energy more competitive, innovations are needed in their structural design and construction, and not just in the reactor core and associated systems. Seismic isolation has been identified as an important cost-cutting technology that enables standardization of equipment across various sites. This paper develops and demonstrates a cost- and risk-based seismic design optimization of a representative safety system in a nuclear power plant with the dual goals of minimizing overnight capital cost and meeting safety goals. The design optimization can also include component seismic isolation, in which case, the optimized design includes a set of equipment that needs to be seismically isolated to minimize capital cost. The open-source codes MASTODON and Dakota are used for seismic probabilistic risk assessment and design optimization, respectively. A generic nuclear facility with a safety system comprising SSCs that are common to nuclear power plants is considered for the demonstration of the design optimization and is assumed to be located at the Idaho National Laboratory site. Generic costs and seismic design cost functions are assumed for the SSCs of the safety system. The sum of the costs of the SSCs is minimized in the optimization process, while the risk of failure of the safety system is provided as a constraint. Results show that the optimization process reduces capital costs significantly while automatically prioritizing the safety of SSCs that contribute most to the risk of the safety system.

Engineering demonstration reactors: A stepping stone on the path to deployment of advanced nuclear energy in the United States

Nuclear energy is an important ultra-low carbon electricity source. Engineering demonstration reactors, such as Shippingport for Pressurized Water Reactor technology, have been used to advance nuclear reactor technology to deployment. It may be necessary to develop similar engineering demonstration reactors for other reactor concepts that have not yet been built. This paper reviews the historical role of engineering demonstration reactors and highlights the potential for future engineering demonstrations of Fluoride Salt-Cooled High Temperature Reactors (FHRs) and Molten Chloride Fast Reactors (MCFRs). Future engineering demonstration reactors could significantly accelerate deployment of larger-scale FHRs and MCFRs by the advanced nuclear industry via cost and risk reduction. Experience in the United States shows that sustained investment in a public-private partnership, such as the current Advanced Reactor Demonstration Program, is essential for deployment of advanced reactor technologies that have never been built before. The National Reactor Innovation Center at Idaho National Laboratory and Clinch River Site near Oak Ridge National Laboratory are examples of ideal locations for these demonstrations.

A Hybrid Energy System Workflow for Energy Portfolio Optimization

This manuscript develops a workflow, driven by data analytics algorithms, to support the optimization of the economic performance of an Integrated Energy System. The goal is to determine the optimum mix of capacities from a set of different energy producers (e.g., nuclear, gas, wind and solar). A stochastic-based optimizer is employed, based on Gaussian Process Modeling, which requires numerous samples for its training. Each sample represents a time series describing the demand, load, or other operational and economic profiles for various types of energy producers. These samples are synthetically generated using a reduced order modeling algorithm that reads a limited set of historical data, such as demand and load data from past years. Numerous data analysis methods are employed to construct the reduced order models, including, for example, the Auto Regressive Moving Average, Fourier series decomposition, and the peak detection algorithm. All these algorithms are designed to detrend the data and extract features that can be employed to generate synthetic time histories that preserve the statistical properties of the original limited historical data. The optimization cost function is based on an economic model that assesses the effective cost of energy based on two figures of merit: the specific cash flow stream for each energy producer and the total Net Present Value. An initial guess for the optimal capacities is obtained using the screening curve method. The results of the Gaussian Process model-based optimization are assessed using an exhaustive Monte Carlo search, with the results indicating reasonable optimization results. The workflow has been implemented inside the Idaho National Laboratory’s Risk Analysis and Virtual Environment (RAVEN) framework. The main contribution of this study addresses several challenges in the current optimization methods of the energy portfolios in IES: First, the feasibility of generating the synthetic time series of the periodic peak data; Second, the computational burden of the conventional stochastic optimization of the energy portfolio, associated with the need for repeated executions of system models; Third, the inadequacies of previous studies in terms of the comparisons of the impact of the economic parameters. The proposed workflow can provide a scientifically defendable strategy to support decision-making in the electricity market and to help energy distributors develop a better understanding of the performance of integrated energy systems.

Evaluation fuel slug eccentricity effects on fuel performance

Metallic fast reactor fuel is often formed via a casting process. This paper explores the potential role of mold casting tolerances in the formation of non-circular fuel slugs, and how such variability in shape may affect fuel swelling and the potential for excessive strain in the cladding at lower fuel burnups. The fuel performance simulation effort presented in this paper was executed using Idaho National Laboratory’s BISON fuel performance code. Three fuel configurations are included in this analysis: circular, elliptical, and maximum eccentric fuel all with the same 75% smeared density. For eccentric fuel cross sections, temperature gradients on the fuel surface along the long dimension of the oval are greater, leading to fuel anisotropic deformation due to creep. Stress emerging due to fuel-cladding contact forces fuel to creep into available cavities and creates very little cladding strain. The cladding experiences a hoop stress surge when contacted by the fuel. However, the magnitude of the hoop stress is inconsequential compared to the cladding yield stress. Thus, the dimensional variations possible for fuel slugs cast in common quartz tubing molds is not expected to impact overall fuel performance nor increase the probability of fuel cladding failure at early life.

Optimization models for integrated biorefinery operations

Variations of physical and chemical characteristics of biomass lead to an uneven flow of biomass in a biorefinery, which reduces equipment utilization and increases operational costs. Uncertainty of biomass supply and high processing costs increase the risk of investing in the US’s cellulosic biofuel industry. We propose a stochastic programming model to streamline processes within a biorefinery. A chance constraint models system’s reliability requirement that the reactor is operating at a high utilization rate given uncertain biomass moisture content, particle size distribution, and equipment failure. The model identifies operating conditions of equipment and inventory level to maintain a continuous flow of biomass to the reactor. The sample average approximation method approximates the chance constraint and a bisection search-based heuristic solves this approximation. A case study is developed using real-life data collected at Idaho National Laboratory’s biomass processing facility. An extensive computational analysis indicates that sequencing of biomass bales based on moisture level, increasing storage capacity, and managing particle size distribution, increases utilization of the reactor and reduces operational costs.

MCNP modelling of radiation effects of the Dragonfly mission's RTG on Titan II: Atmospheric ionization effects

The Dragonfly mission to Titan represents the first use of a large-scale radioisotope power source in that environment; as such little is known of its potential effects on the surface or near-surface atmosphere. These effects represent backgrounds it will be necessary to subtract from the spacecraft's instrumentation data. Repurposing a Monte Carlo code model of the generator built for previous neutronic investigations with the addition of electron detectors allows for the determination of the degree to which the radiation fields from the generator may ionize the surrounding atmosphere and generate interference with the lander's conductivity measurements. Results showed good agreement with experimental data undertaken at the Idaho National Laboratory using the F-2 generator flown on the Perseverance Mars rover, and showed that the interference with readings taken in the daytime on Titan when atmospheric conductivity is expected to be low may be substantial.

Design of an experimental test loop for fast spectrum test conditions

This summary details the feasibility of placing a conceptual fast spectrum test loop within the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) that would allow for preliminary Versatile Test Reactor (VTR)-condition testing. This experimental loop - the ATR Fast Spectrum Test Loop (FSTL) - has been designed, based on benchmarking, against predicted in-reactor conditions that the VTR will produce, including the use of molten sodium as a reactor coolant. A discussion of the mechanical design of the in-pile vehicle, along with the material delineation of primary components, is included. Furthermore, a general physical arrangement of the hardware necessary to compliment the FSTL in-pile vehicle to provide a sodium-cooled environment within the ATR, passing through the Large-I position, is discussed in detail.

3D-2D coupling multi-dimension simulation for the heat pipe micro-reactor by MOOSE&SAM

In recent years, micro-reactor concepts have attracted increasing attention in the nuclear industry due to the market demand for flexible, reliable, and sustainable power and heat on-site for industrial or federal installations or remote communities. To help demonstrate and validate these innovative reactor concepts, the Micro-reactor AGile Non-nuclear Experimental Test-bed (MAGNET) is being constructed at Idaho National Laboratory (INL) with an initial focus on the thermal and structural performance of heat pipe cooled micro-reactors. At this time, the preliminary design parameters of the MAGNET facility have been specified. In this work, a simulation using the System Analysis Module (SAM) code is performed for the prototypical 37-heat-pipes test article to predict its experimental facility performance and associated uncertainties. We first carry out a benchmark demonstration of our modeling method with an example provided by Argonne National Laboratory (ANL). Then, we predict the thermal performance of the MAGNET facility under steady-state operation. Moreover, several sensitivity parameters are analyzed to investigate their impact on facility thermal performance. This MAGNET experiment simulation provides valuable information for researchers to validate the facility's initial design and steady-state operation.

Modeling of High-Temperature Corrosion of Zirconium Alloys Using the eXtended Finite Element Method (X-FEM)

Oxidation modeling in modern nuclear fuel performance codes is currently limited by the lack of coupling with mechanics, thus preventing proper description of how high-temperature oxidation impacts mechanical properties. This is mostly due to the fact that the finite difference formalism adopted in corrosion models is incompatible with the direct coupling with mechanics in the finite element modeling employed in modern nuclear fuel performance codes. In this study, a physically based zirconium alloy corrosion model called the Coupled-Current Charge Compensation (C4) model, which was initially developed for operating temperature conditions, has been updated to include high-temperature corrosion in order to provide additional critical information (e.g., oxygen concentration profile) under loss-of-coolant accident (LOCA) conditions—information lacking in existing empirical models. The C4 model was implemented in the MOOSE finite-element framework developed at Idaho National Laboratory, enabling it to be used in the BISON nuclear fuel performance code based on the MOOSE framework. To precisely track the different interfaces at a relatively low computational cost, the eXtended Finite Element Method (X-FEM) was applied in MOOSE. The model’s results were compared to those of existing empirical models as well as metallographic analysis of high-temperature oxidized Zircaloy-4 coupons. Oxygen diffusivities in the α and β phases resulting from this comparison closely agree with those found in the literature. The C4 model implemented with X-FEM in MOOSE now has the capability to accurately predict oxide, oxygen-stabilized α, and prior β phase layer growth kinetics under isothermal exposure at high temperature (1000–1500 °C). Furthermore, in contrast with the empirical models, the C4 model accounts for the finite thickness of the fuel cladding. It can predict the oxygen concentration profile evolution through the whole cladding, enabling evaluation of the remaining ductile thickness—a crucial variable for modeling the mechanical behavior of the fuel cladding under LOCA. This implementation allows direct coupling with mechanics, at a low computing cost, using finite-element-based nuclear fuel performance codes such as BISON.

Performance Testing of Dysprosium-Based Scintillation Screens and Demonstration of Digital Transfer Method Neutron Radiography of Highly Radioactive Samples

Scintillator screens consisting of a dysprosium neutron converter and various scintillator materials were tested in the Heinz Maier-Leibnitz Zentrum Forschungsreaktor München II (FRM II) ANTARES cold neutron beam with the goal of finding a suitable screen for digital transfer method neutron radiography. This work explores the cold neutron response of 16 scintillator screens, 7 of which were previously tested with thermal neutrons. Light yield, signal-to-noise ratio (SNR), and spatial resolution were measured to compare the scintillator screens and determine which were best suited for digital transfer method neutron radiography. Screens with a zinc sulfide (ZnS:Cu) scintillator were most suitable for digital transfer method radiography based on light output, spatial resolution, SNR, and gamma-ray insensitivity. Spatial resolutions between 65 and 220 μm were measured. The top-performing screens were then used to demonstrate the feasibility of a new digital transfer method neutron radiography to image highly radioactive (8.84 Sv/h at ≈1 cm) nuclear fuel at Idaho National Laboratory’s Neutron Radiography reactor (NRAD). These results suggest that digital transfer method neutron radiography can be used to indirectly image highly radioactive objects and/or use neutron beams with a large gamma-ray content on a timescale of ~10 min/image (~144 images/day), much faster than the >10 h required using the current transfer method with film (limited to ~14 radiographs/day at NRAD).

Research Advancement of Energy to Molecules and Materials (E2M2) at Idaho National Laboratory

Integrated Energy System (IES) is a core initiative at Idaho National Laboratory where energy to molecules and materials (E2M2) pillar plays a critical role, due to the widely available distributed feedstock, energy input (renewable and/or nuclear energy associated electricity and excess heat) as well as the resulting industrial electrification. These factors offer new opportunities to those technologies with smaller scale, modularized/distributed, and driven by an integrated system of renewable and other energy sources that can show promise for increased productivity with decreased energy, capital, and operating costs. Intermediate temperature electrochemical processing through the solid oxide cells is a promising strategy because it can efficiently utilize electrons and heat while overcoming and/or mitigating the challenges of their lower and higher temperature counterparts, including 1) the use of precious metal; 2) sluggish kinetics; 3) fast degradation; 4) costly interconnect materials. Despite of these promising features, this technology still faces many challenges that include but not limited to immature process, relatively low catalyst activity, and unproven operation durability. At Idaho National Laboratory, we have made significant achievements on electrosynthesis of intermediates, fuels and chemicals using natural gas, steam, carbon dioxide and nitrogen through protonic ceramic membrane reactors. The research outcomes have the potential to establish a globally competitive U.S. industry for converting the Nation’s abundant supply of feedstock into conversion-ready intermediates and/or high demand fuels and chemicals.

MASTODON: An Open-Source Software for Seismic Analysis and Risk Assessment of Critical Infrastructure

Seismic analysis and risk assessment of safety-critical infrastructure like hospitals, nuclear power plants, dams, and facilities handling radioactive materials involve computationally intensive numerical models and coupled multiphysics scenarios. They are also performed in a strict regulatory environment that requires high software quality assurance standards, and in the case of safety-related nuclear facilities, a conformance to the American Society of Mechanical Engineers Nuclear Quality Assurance (NQA-1) standard. This paper introduces the open-source finite-element software, MASTODON (Multi-hazard Analysis of Stochastic Time-Domain Phenomena), which implements state-of-the-art seismic analysis and risk assessment tools in a quality-controlled environment. MASTODON is built on MOOSE (Multi-physics Object-Oriented Simulation Environment), which is a highly parallelizable, NQA-1 conforming, coupled multiphysics, finite-element framework developed at Idaho National Laboratory. MASTODON is capable of fault rupture and source-to-site wave propagation using the domain reduction method, nonlinear site response, and soil-structure interaction analysis, implicit and explicit time integration, automated stochastic simulations, and seismic probabilistic risk assessment. When coupled with other MOOSE applications, MASTODON can also solve strongly and weakly coupled multiphysics problems. This paper presents a summary of the capabilities of MASTODON and some demonstrative examples.

Computational fluid dynamics assessment of the Helium-3 Enhanced Negative Reactivity Insertion (HENRI) system for the Transient REActor Test (TREAT) facility

The Transient REActor Test facility (TREAT) is an air-cooled, graphite moderated nuclear reactor that is used for irradiation testing of nuclear fuels and materials. The shutdown of the Halden reactor in Norway lead to an increased interest of an indigenous nuclear facility that can perform Reactivity Initiated Accident (RIA) tests for Light Water Reactors (LWRs). Idaho National Laboratory (INL) has been leading the effort of making minor modifications to the TREAT facility to enhance its testing capabilities so it can mimic the neutronic boundary conditions representative of a RIA for a LWR. The power burst of the TREAT reactor needs to be shortened from 89 to 40 msec. to properly irradiate the fuel sample under RIA conditions. INL has proposed to use helium-3, a strong neutron absorber, as a form of negative reactivity to shorten the pulse of the reactor. In support of the development of a transient control rod for the TREAT, a Computational Fluid Dynamics (CFD) model is needed to confidently simulate the density evolution of helium-3 in this system.

Transient testing of uranium silicide fuel in zircaloy and silicon carbide cladding

Using the newly restarted Transient Reactor Test (TREAT) facility at Idaho National Laboratories, fresh (unirradiated) U3Si2 fuel pellets were tested under reactivity-initiated accident conditions in dry helium capsules. These tests were performed with total energy depositions sufficient to induce melting of the U3Si2 fuel pellets based on available thermophysical property data. The instrumented experiments utilized in-pile thermocouples and infrared pyrometers to measure the cladding temperature during the transient, which are in reasonable agreement with predicted peak cladding temperatures using a simple BISON fuel performance model. This same model is then used to predict expected cladding stresses during the transients in areas where pellet cladding interaction is expected. It was shown that for transient energy depositions up to 528 J/g rod like geometry of the test articles was maintained. While it is acknowledged that more confirmatory testing is required in more prototypic environments, these first of their kind experiments have demonstrated that unirradiated U3Si2 fuel and unirradiated SiC-SiC claddings can likely retain a “rod like” geometry for transient energy depositions up to 528 J/g.

In-Canal Assay of High Specific Activity 60Co at the Advanced Test Reactor

Production of high specific activity (HSA) 60Co has recently resumed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The technical steps of performing in-canal assay of HSA 60Co targets at the ATR are described herein. The HSA targets are assayed on a regular basis, between cycles at the ATR, in order to assess the progress of activation. The targets are also assayed at the conclusion of the irradiation in order to provide activity estimates for the distributor and to be used for safety and shipping evaluations. These target assay activities must take place in the ATR canal to provide sufficient radiological shielding. A specialized assay fixture is used in conjunction with custom 60Co radiation standards to assess the irradiated target. The specific activity of each irradiated target was determined by first measuring the 60Co standards to determine a sensitivity factor. Then, each irradiated target was measured, and the integrated measurement was multiplied by the sensitivity factor to determine the total activity of the irradiated target. Finally, a correction factor was determined to adjust the reported activity, accounting for differences in the physical geometries of the standards and targets. These methods were used to accurately assay the gross activity of nine irradiated 60Co targets, two of which were delivered to the distributor where confirmatory hot-cell assays were performed verifying the accuracy of the in-canal assay method.

Development of fuel performance analysis code, BISON for MOX, named Okami: Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

• 2-D analyses with modelling for MOX fuel restructuring in use of fast reactor. • Effects of O/M ratio and pellet-cladding gap width on pore migration behavior. • Irradiation behavior evaluation for the MA-bearing MOX irradiation test in Joyo. • Influence of pellet eccentricity on central void formation in 2-D analyses. Fuel composition, i.e. Pu content, minor actinide (MA) content, oxygen-to-metal (O/M) ratio, etc., affects the irradiation behavior of MA-bearing mixed oxide (MOX) for fast reactor. To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with Idaho National Laboratory (INL) and have installed a pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation in the experimental fast reactor Joyo was evaluated by a two-dimensional (2-D) analysis. To solve partial differential equations for energy conservation and advection diffusion equations to calculate temperature and pore migration, respectively, a correlation of thermal conductivity as a function of porosity, temperature, deviation from stoichiometry, Am content and Np content was used. In addition, the pore migration velocity was calculated by using a correlation depending on vapor pressure applying Rand-Markin's model. Four MA-bearing MOX test pins with different O/M ratio and pellet/cladding gap size were irradiated in B14 test in Joyo to study the effects of O/M ratio and gap size on irradiation behavior at the beginning of irradiation. In post-irradiation examinations (PIE), a remarkable restructuring was observed in stoichiometric fuels, and a discrepancy in central void diameter among various O/M ratios. This restructuring behavior could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void is assumed to offset from the geometric center to the larger-gap side due to the pellet eccentricity, but the PIE results were inconsistent in this regard. It is expected that the irradiated pellet might displace to another side during sample preparation in PIE from a reverse offset location during irradiation by comparing the result of 2-D analysis and radial profile of fuel density.

Neutron Beam Characterization at Neutron Radiography (NRAD) Reactor East Beam Following Reactor Modifications

The Neutron Radiography Reactor at Idaho National Laboratory (INL) has two beamlines extending radially outward from the east and north faces of the reactor core. The control rod withdrawal procedure has recently been altered, potentially changing power distribution of the reactor and thus the properties of the neutron beams, calling for characterization of the neutron beams. The characterization of the East Radiography Station involved experiments used to measure the following characteristics: Neutron flux, neutron beam uniformity, cadmium ratio, image quality, and the neutron energy spectrum. The ERS is a Category-I neutron radiography facility signifying it has the highest possible rank a radiography station can achieve. The thermal equivalent neutron flux was measured using gold foil activation and determined to be 9.61 × 106 ± 2.47 × 105 n/cm2-s with a relatively uniform profile across the image plane. The cadmium ratio measurement was performed using bare and cadmium-covered gold foils and measured to be 2.05 ± 2.9%, indicating large epithermal and fast neutron content in the beam. The neutron energy spectrum was measured using foil activation coupled with unfolding algorithms provided by the software package Unfolding with MAXED and GRAVEL (UMG). The Monte-Carlo N-Particle (MCNP6) transport code was used to assist with the unfolding process. UMG, MCNP6, and measured foil activities were used to determine a neutron energy spectrum which was implemented into the MCNP6 model of the east neutron beam to contribute to future studies.

BEPU analysis of a passive decay heat removal system with RELAP5/MOD3.3 and RELAP5-3D

Passive safety systems are currently implemented or under consideration in several advanced nuclear plants. In the Best-Estimate Plus Uncertainty (BEPU) approach it is fundamental to qualify Best-Estimate thermal-hydraulic system codes and models for the phenomena typical of passive systems and to quantify the uncertainty of calculations. In the present paper, through RELAP5/MOD3.3 and RELAP5-3D codes, thermal-hydraulic analyses were carried out and the obtained results compared against the experimental data collected in the experimental test facility PERSEO. For the present analysis, Test 7 Part 2, conducted at the nominal primary pressure of 7 MPa, has been selected. An accuracy evaluation has been performed on the reference calculation carried out with the two codes. For the accuracy evaluation both a qualitative and quantitative approach have been considered. The uncertainty of both calculations has been quantified by means of the probabilistic method of propagation of input uncertainties. Uncertain input parameters have been selected considering initial and boundary conditions, physical parameters and model uncertainties. The Uncertainty Analysis has been carried out through the RAVEN framework developed by Idaho National Laboratory. Additionally, regression and correlation indices have been used to characterize the importance of the selected uncertain input parameters for the selected Figure Of Merit. The effect of sampling methodologies on the results has also been evaluated.