Unprotected Loss Of Flow Accident Research Articles

Transient analysis of the safety characteristics on a super carbon-dioxide cooled micro modular reactor

The supercritical CO2 (S-CO2) demonstrates great potential as a coolant for nuclear reactors due to its low critical pressure and temperature. It offers improved thermal efficiency and decreased requirements for structural materials. By incorporating core physics models and updated thermal–hydraulic models, the SYST-SCO2 code developed in prior research has been enhanced and utilized for conducting transient safety analysis on the micro modular reactor (MMR) proposed by the Korea Advanced Institute of Science and Technology (KAIST). The analyzed transients include the Unprotected Loss of Flow Accident (ULOFA), Unprotected Reactivity Insertion Accident (URIA), Unprotected Loss of Heat Sink (ULOHS), Unprotected Over Cooling (UOC), and the Protected Loss of Flow Accident (PLOFA). It is found that the MMR experiences the exceeding of coolant outlet temperature limit of 676℃ in the ULOFA for over 155 s. With the reactor shutdown system put into operation, the reactor core can be maintained to operate with the fuel pellet, cladding and coolant temperatures below their limits respectively in the LOFA. Enough safety margin can be left for the MMR reactor core in URIA, ULOHS and UOC transients. Among the safety limits for the fuel and coolant, the core outlet temperature limit is the easiest to reach in the analyzed accidents.

Analysis of unprotected loss of flow accident in a sodium cooled fast reactor using coupled thermal hydraulics neutronics code ASTRA

The Unprotected loss of flow accident (ULOFA) is a severe accident that may lead to core disruption in a sodium cooled fast reactor (SFR). It is investigated as a part of the defense in depth concept in nuclear safety. A coupled thermal hydraulics neutronics model is required to model the severe accident scenario where reactivity is affected by the changes in the temperature of the core components, coolant boiling, and material distribution during the transient. The coupled code previously developed for the analysis of total instantaneous blockage (ASTRA) is improved for the study of ULOFA by adding various models, viz., (i) a one-dimensional two-fluid sodium boiling model, (ii) a simple primary hydraulics model, (iii) a simple clad motion model, and (iv) an improved point kinetics solver. The boiling model is validated with different loss of flow experiments. The improved code is used to analyze ULOFA in a 500 MWe medium size sodium cooled fast reactor up to the onset of fuel melting. At 21.5s, coolant boiling is initiated in the central channel. The temperature evolution of the core components, reactivity feedbacks, power evolutions, and coolant voiding propagation are calculated. The code predicts early fuel melting at 24.9s due to the power excursion by the early voiding in the fuel channels compared to the study without including the boiling model. Boiling is started in five out of ten representative fuel subassembly (SA) channels at the time of fuel melting.

Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Device Specifications

Abstract A new subassembly type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device passively provides high negative reactivity to the core. This study evaluated the nuclear and thermal properties of the device subassembly with metallic fuel to determine the device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWel-class mixed-oxide-fueled SFR core confirmed that a conventional homogeneous core maintains stable cooling of the core before coolant boiling in the driver fuel subassemblies. In contrast, the negative reactivity required to terminate the event by device operation was slightly higher in the low sodium void reactivity core than in the conventional homogeneous core.

A set of transient correlations for fast and unprotected loss of flow accident in VVER-1000 reactor using single-heated channel approach and Gene Expression Programming

Artificial intelligence methodologies along with human observations in the main control room of nuclear power plants can be applied for predictive analysis and accident detection in the early phase of accidents. This study set out with the aim of assessing the importance of accumulated information in the early phase of a fast and unprotected Loss of Flow Accident (LOFA) for exploring the behavior of significant operating parameters in the reactor. In this study, various fast and unprotected LOFA scenarios are numerically simulated using a single-hearted approach to generate a comprehensive dataset in a forward direction. The nondimensional transient dataset including inlet mass flux and outlet temperature of the core has been given to the Gene Expression Programming (GEP) algorithm for developing a set of correlations at different moments after initiating the accident. The Multi Time-Step Data (MTSD) and Single Time-Step Data (STSD) approaches have been used to extract the required correlations from the generated dataset. Mean Square Error (MSE) and coefficient of determination are used to measure the error of each approach. The most striking result to emerge from the computed results is that the MTSD approach based on trigonometric functions has exceptionally high prediction accuracy (MSE<10-5). Overall, the obtained results show that the GEP algorithm can be used as a powerful tool for generating a set of transient correlations with high accuracy to identify the progress in a fast and unprotected LOFA.

Chemical interaction between uranium dioxide, boron carbide and stainless steel at 1900 °C — Application to a severe accident scenario in sodium cooled fast reactors

For the understanding of severe accidents in sodium cooled fast reactors (SFR), it is necessary to understand two prototypic accident scenarios such as ULOF (Unprotected Loss of Flow Accident) and UTOP (Unprotected Transient OverPower). As the base knowledge, it is also important to understand high temperature chemical interaction among major core materials such as MOx fuel (MOx: mixed oxide of uranium and plutonium), steel cladding and B4C neutron absorber have to be investigated. This study aims at providing experimental data on phase formation and phase-stability at various temperature and pressure conditions. A first series of samples containing a mixture of B4C and steel were prepared to obtain a homogenous metallic solid. In a second step, these metallic samples were mixed and melted with small UO2 pieces by arc melting. Then these samples underwent a heat treatment at 1900 °C for 1 hour. EDS, EBSD and EPMA analyses were performed to identify the phases formed during the solidification. In addition, thermodynamic calculations were performed for the interpretation of the results, revealing that a carbo-reduction reaction occurs: UO2 + 2 C = 2 CO + U. A significant amount of uranium from the fuel is dissolved in the metallic liquid phase, leading to the formation of mixed borides (UM3B2, UMB4, UM4B, M=Fe,Cr,Ni). In comparison with the UO2/steel interaction, the present results show that the presence of B and C in the melt improves the wetting behaviour of the metallic liquid towards UO2.

Performance test of conceptually designed hydraulically suspended passive shut down subassembly for CFR600

Hydraulically Suspended Passive Shutdown Subassembly (HS-PSS) of Sodium-cooled Fast Reactor (SFR), which can protect the reactor from beyond design basis accidents, is especially suitable for responding to the Unprotected Loss of Flow accident (ULOF). In this study, the performance of the conceptually designed HS-PSS for China Fast Reactor with 600 MW electric power (CFR600) that is being developed by the China Institute of Atomic Energy (CIAE) was experimentally investigated. Suspension tests, free drop performance tests, and hydraulic characteristic verification of the HS-PSS were carried out on a specially constructed vertical hydraulic test loop. Furthermore, the hydraulic characteristic data such as suspension critical flow rate, drop time, pressure drop, and hydraulic thrust on the Hydraulically Moving Body (HMB) inside the HS-PSS was obtained to verify whether they meet the design requirements. The HMB prototype has a drop time of 10.65s under the coast-down flow rate, and the initial falling critical flow rate is 1.80 kg/s, at 84 °C. Both agree well with the theoretical calculation results and meet the engineering design requirements.

Estimation of reactivity feedback and determination of safety criteria of inherent-safety fast reactors in unprotected transients based on the asymptotic approximation

The effect of feedback reactivity on the safety of a fast reactor has been studied based on the asymptotic approximation for unprotected transients. The asymptotic approximation adopts the steady state heat transfer equation for calculating temperatures of coolant and fuel with the use of the relevant core power and mass flow. The feedback reactivity can be approximately estimated from the temperatures. This approximation is applied to the UTOP (unprotected transient overpower) and ULOF (unprotected loss of flow) accidents of a sodium-cooled fast reactor with a sodium plenum and GEM (gas expansion module). The reactivity feedback has been estimated, and the constraints of the feedback reactivity for maintaining the safety of the core have been predicted. For the UTOP accident the constraint for the control rod reactivity and the Doppler (+fuel expansion) reactivity coefficient is presented to prevent the fuel melting. For the ULOF accident the constraint of the parameter R is introduced which is the normalized ratio of the power to mass flow rate. The parameter does not exceed the critical point corresponding to the onset of the coolant boiling, and the coolant boiling does not occur because of the large negative reactivity insertion by the GEM. Thus the core safety can be approximately estimated by the present method without detailed calculations, though the accurate analyses are required by the use of transient analyses.

Dynamic performance investigation on the hydraulic-suspended passive shutdown subassembly of SFR

Hydraulic suspended passive shutdown subassembly (HS-PSS) belongs to one of the diverse passive safety devices for further enhance safety for pool-type sodium-cooled fast reactor (SFR). The hydraulic moving body (HMB) in the PSS with neutron absorbers inside was designed to insert into the core automatically within a stipulated time when the mainstream below a critical flow rate. The dynamic displacement performance of the HMB is of particular importance to ensure the safety of SFR especially during the Unprotected loss of flow accident (ULOF). In this paper, the mathematical modeling related to the falling process, the hydraulic characteristics of HMB under different flows are detailed. A series result of the sensitivity analysis of the factors affecting HMB performance, such as the flow rate and the size of some key structures, are obtained through calculation. A PSS-mockup was tested at the hydraulic (water) test bench and the calculation program was validated by the experimental results.

VVUQ of a thermal-hydraulic multi-scale tool on unprotected loss of flow accident in SFR reactor

Within the framework of the French 4th-generation Sodium-cooled Fast Reactor safety assessment, methodology on VVUQ (Verification, Validation, Uncertainty Quantification) is conducted to demonstrate that the CEA's thermal-hydraulic Scientific Computation Tools (SCTs) are effective and operational for design and safety studies purposes on this type of reactor. This VVUQ-based qualification is a regulatory requirement from the French Nuclear Safety Authority (NSA). In this paper, the current practice of VVUQ approach application for a SFR accidental transient is described with regard to the NSA requirements. It constitutes the first practical, progressively improvable approach. As the SCT is qualified for a given version on a given scenario, the transient related to a total unprotected station blackout has been selected. As it is a very complex multi-scale transient, the SCT MATHYS (which is a coupling of the CATHARE2 tool at system scale, TrioMC tool at component scale and TrioCFD tool at local scale) is used. This paper presents the preliminary VVUQ application to the qualification of this tool on this selected transient. In addition, this work underlines some feedback on design and R&amp;D aspects that should be addressed in the future to improve the SCT.

2D and 3D numerical investigations of sodium boiling in sodium cooled fast reactor with MOX fuel and low sodium void reactivity effect during unprotected loss of flow accidents

• Behavior of SFR under ULOF simulated for Low Sodium Void Worth design. • No core melting simulated and stable boiling behavior predicted by both codes. • Different frequencies and amplitudes of boiling regimes with SIMMER and COREMELT. • 3D modeling performed with COREMELT confirms oscillating behavior and absence of core melting. • If confirmed by enough experimental evidence, could make safety demonstration of SFR more robust.

3D thermal hydraulic characteristics analysis of pool-type upper plenum for lead-cooled fast reactor with multi-scale coupling program

The large space plenum of the lead-cooled reactor has complex three-dimensional thermal hydraulic phenomena such as thermal stratification and coolant mixing, which affects the safety performance of the reactor. In this paper, based on the multi-scale coupling program ATHLET-OpenFOAM, three-dimensional simulation of the hot pool for lead-cooled fast reactor M2LFR-1000 is carried out. Considering the axis symmetry of the hot pool model and reducing the amount of calculation, 1/8 model of the hot pool is selected to build the model by CFD, and the rest is simulated by system program ATHLET. The steady-state operation condition is first calculated by coupling program. Based on the calculation of steady-state condition, the main thermal parameters of M2LFR-1000 primary coolant system and the complex three-dimensional phenomena of the upper plenum under unprotected loss of flow accident (ULOF) are studied, and the results of multi-scale coupling calculation are compared with those of ATHLET alone. The results show that there is an obvious thermal stratification in the inner region and the outer region of the upper plenum in the early period of the accident, which slows down the speed of establishing natural circulation. During ULOF accident, there are obvious vortex flow changes in the inlet outer side and the outlet area of the upper plenum.

Preliminary design and analysis of Liquid Fuel Molten Salt Reactor using multi-physics code GeN-Foam

In this study, a hypothetical fast spectrum Liquid Fuel Molten Salt Reactor (LFMSR) core is modeled using the multiphysics C++ code GeN-Foam (General Nuclear Foam). GeN-Foam is based on OpenFOAM, a C++ open-source library for solution of continuum mechanics problems. The code utilizes a unified fine/coarse mesh approach, modeling different physics such as neutron kinetics, thermal-hydraulics based on porous fluid equations, and structural thermal-mechanics. A steady state analysis of a simplified two-dimensional (2D) LFMSR model has been performed assuming rotational symmetry to cross verify the code with the commercial ANSYS Computational Fluid Dynamics (CFD) code Fluent. The calculations showed very good agreement between the two codes allowing progression to a three-dimensional (3D) model simulation. A coupled neutron kinetics and CFD steady state analysis of a right-cylindrical 3D LFMSR core has been performed modeling one quarter of the core while using symmetry boundaries to reduce the computational time. Mixed uranium and plutonium chloride fuel has been selected in this preliminary study. Both 2D and 3D simulations showed appearance of recirculation zones within the right-cylinder core. These zones can be a challenge for LFMSR control and materials. A new hyperboloid design is proposed to remove recirculation zones, which is based on eight symmetrical loops. An Unprotected Loss of Flow accident (ULOF), in which the pump head is instantaneously reduced to zero, has been selected to demonstrate the safety characteristics of the reactor in one of the most challenging possible situations for LFMSR. The obtained results (e.g., reduced total precursors concentration at the core inlet and reduction of the core nominal power following the transient) confirm that GeN-Foam is capable of performing coupled LFMSR transient analysis and can be used for design analysis and optimization. Although the current design needs further assessment and development, it shows encouraging performance under ULOF conditions paving the way to the next step in the optimization process.

Thermal-hydraulic analysis of pellet bed reactor for space nuclear electric propulsion

The pellet bed reactor (PeBR) featured with high-power density, small reactor size and full retention of fission products, has promising applications for space nuclear power system. According to the special features of the PeBR for nuclear electric propulsion (NEP), a transient analysis code is developed, where the six groups delay neutrons point kinetics model, coolant flow and heat transfer model as well as the fuel pellets heat conduction model are built and adopted. Then, the code is used to perform thermal-hydraulic analysis of PeBR, including the optimization of the frits, key parameters sensitivity analysis and the reactor transient characteristics. Results show that the optimized hot frit porosity (HFP) profile can eliminate the hot spot in the core at the steady full power operation. Conversely, changing the cold frit porosity (CFP) is inefficient on controlling the coolant flow through the PeBR core. Increasing the pellet bed porosity and decreasing the HFP from the base case have a negative effect on core flow distribution. Finally, when unprotected reactivity insertion accident (URIA) or unprotected loss of flow accident (ULOFA) occurs, the PeBR is capable of keeping the power and temperature stable and below the safety limits due to the reactor negative reactivity feedback mechanism. This study could provide suggestions and guidance for the design and safety analysis of the PeBR for NEP.

Design-oriented tool for Unprotected Loss Of Flow simulations in a Sodium Fast Reactor: Validation and application to stability analyses

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Commissariat à l’Energie Atomique et aux Energies Alternatives), the reactor behavior in case of severe accidents is assessed through experiments and simulations. Such accidents are usually simulated with mechanistic calculation tools (such as SAS-SFR and SIMMER-III). As a complement to these codes, which give reference accidental transient results, a new Best Estimated Plus Uncertainty approach has been developed by CEA; its final objective being to derive the variability of the main outputs of interest for the safety. This approach involves a fast-running description of extended accident sequences coupling physical models for the main phenomena with advanced statistical analysis techniques. It enables to perform a large number of simulations in a reasonable computational time and to describe all the possible bifurcations of the accident transient. In this context, this paper presents a physical tool (models and results assessment) dedicated to the primary phase of an Unprotected Loss Of Flow accident (ULOF) in a SFR, i.e. before the first hexagonal can failure. This paper focuses on the modeling of neutronic phenomena and of the pin degradation and on their validations, the two-phase sodium thermal–hydraulic behavior having largely been assessed in a previous paper. The physical tool called MACARENa is described before presenting the comparison of its results with experimental tests results for the pin degradation behavior modelling and with mechanistic SIMMER-III code results for the global core behavior evolution during an ULOF transient. This tool is demonstrated to be capable of reproducing the mass-flow rate, reactivity, power evolution, and global pin degradation behaviour during an ULOF with a low discrepancy regarding the same transient simulated with SIMMER-III. In particular, observed discrepancies concerning mass flow rates and power evolutions are under 3% before boiling occurs and, later during the transient, the power excursion starts only 2s earlier in the MACARENa simulation than in the SIMMER-III one. The fast-running tool has then been used for safety-informed design and flow stability analyses of fast reactor systems, more specifically dealing with the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) core concept. It allowed to emphasize the main dominant phenomena and the significant trends for safety assessment. This work has pointed out the strong impact of the modelling of reactivity feedback due to thermal expansion on the resulting core degradation during an ULOF. It has also highlighted the main scenario parameters and models influencing this transient evolution.

Accident diagnosis of a PWR fuel pin during unprotected loss of flow accident with support vector machine

In this study, we conducted various flow rate change simulations for a fuel pin during unprotected LOFA using MARS. The obtained transient outlet temperature profiles were used to establish a relationship with peak fuel temperatures and flow rate changes, using Support Vector Machine (SVM). Unless the number of training data is scarce, the SVM trained with the core outlet temperature gives an accurate prediction (R2 > 0.9) for peak cladding surface temperature, and mass flow rate changes in the early phase of LOFA transience (~0.5 s). It illuminates that key accident characteristics are well reflected in the early response of reactor core behavior (i.e., core outlet temperature). This implies that the possibility of (1) realizing an accident diagnosis framework different from today's practice which relies on the accumulated response of reactor behavior over an extended accident progression, and (2) providing an effective guideline for accident mitigation strategies in the early phase of accident progression. The high predictability (i.e., R2 > 0.9) presented in the early phase of unprotected LOFA indicates core outlet temperature is strongly correlated to both flow rate change, and peak cladding surface temperature during the entire transience. With these strong correlations between different physical parameters, the traditional boundaries of physical locations and physical quantities in detecting accident response and progression may be reduced, allowing the possibility of interdependent detector systems.

Probabilistic evaluation of the energetics upper bound during the transition phase of an unprotected loss of flow accident for a sodium cooled fast reactor by using a Phenomenological Relationship Diagram

Abstract One of the main research goals of the GEN-IV systems is enhancing their safety compared with the former Sodium-Cooled Fast Reactor (SFR) designs. A key issue is the capability of accidents prevention as well as of demonstrating that their consequences do not violate the safety criteria. In order to fulfill such requirements, risk analyses of severe core disruptive accidents are performed. Since the beginning of the SFR development, Hypothetical Core Disruptive Accidents (HCDAs) have played an outstanding role. Numerous safety analyses have been performed for developing and licensing past SFR designs and nowadays a large database of results is available. In particular, a large amount of results of the mechanistic SIMMER-II and SIMMER-III/IV analyses for various core designs and different power classes is available at the Karlsruhe Institute of Technology (KIT). The current paper describes the probabilistic approach based on the Phenomenological Relationship Diagram (PRD), which is used to evaluate the Probability Distribution Function (PDF) of the thermal energy release during the transition phase of an unprotected loss of flow accident scenario for a SFR. The technique allows taking into account the mechanistic nature of the accident scenario. In fact, the available results of the mechanistic analyses of HCDAs in SFRs are used to assess the PDFs of the dominant phenomena affecting the thermal energy release, which are propagated in the PRD by employing a Monte Carlo method.

A two-phase model for simulation of MTR type research reactor during protected and unprotected LOFA

An optimized, accurate and fast model is investigated to simulate the transient behavior of a research reactor during fast and slow LOFA (loss of flow accident). Simulation is performed in two states when the reactor scram system is initiated at critical core power and when it doesn't work properly. The latter case leads to boiling condition and hence the reactor core will be encountered with two-phase flow condition. A lumped heat conduction model for fuel and clad accompanied with a two-phase flow model for coolant are coupled with point kinetic equations to simulate reactor behavior during the accident. A combination of Runge-Kutta method and up-wind finite difference technique are adopted for the solution to the set of differential equations. The calculation performed for IAEA 10 MW benchmark problem in HEU and LEU fuel. Some important parameters such as the peak temperatures of fuel and clad are obtained for both HEU and LEU conditions during transients. Results of numerical simulations, at the first, agree well with the general physical rules and also verified through comparison with the other computer code results. According to present simulation, under both protected and unprotected LOFA transient, the peak clad surface temperatures remain in safe mode.

Assessment study about the risk of Wilks’ formula for uncertainty quantification of design extension condition scenarios in prototype Gen-IV sodium fast reactor

ABSTRACT Sensitivity analysis and uncertainty quantification using Wilks’ formula and Monte Carlo for Unprotected Loss of Flow (ULOF) and Unprotected Transient OverPower (UTOP) accidents of prototype Gen-IV sodium-cooled fast reactor were performed. Multi-dimensional analysis for reactor safety for liquid metal reactors code calculations were conducted while simultaneously varying the values of all uncertain parameters according to their distribution using parallel computing platform integrated for uncertainty and sensitivity analysis to obtain uncertainty bands for Figures of Merit (FOM) – coolant, fuel centerline, and cladding temperature at the hottest fuel rod. To specify the uncertainty range of the parameters for each accident scenario, literature survey and expert judgments were consulted. By the sensitivity analysis, the importance ranking of 25 parameters in model identification and ranking table based on phenomena identification and ranking table was identified. Through Monte Carlo calculation, 95% upper limit and 95% confidence level were obtained, and about 2% and 5% under-prediction (risk) of FOM of ULOF and UTOP accidents using Wilks’ formula were confirmed, respectively.

System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions

This paper discusses system codes benchmarking activities on an ASTRID-like heterogeneous fast core under a representative design basis accident condition: the unprotected loss of flow accident (ULOF). The paper provides evidence that all the system codes used in this exercise are capable to simulate the transient behavior of heterogeneous SFR cores up to the initiation of sodium boiling. As a proof of this, a comparison of steady-state results and dynamic simulation results for a ULOF transient (simulated using system codes in combination with neutron point kinetics) are provided and discussed in this paper. The paper contains a brief description of the system codes (TRACE, CATHARE, SIM-SFR, SAS-SFR, ATHLET, SPECTRA, SAS4A) used by the participants (PSI, CEA, EDF, KIT, GRS, UPVLC, NRG, KTH), assumptions made during the simulations, as well as results obtained.

Physical tool for Unprotected Loss Of Flow transient simulations in a Sodium Fast Reactor

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Commissariat à l’Energie Atomique et aux Energies Alternatives), reactor behavior in case of severe accidents is assessed. Such accidents are usually simulated with mechanistic calculation tools (such as SAS and SIMMER). As a complement to these codes, which give reference accidental transient, a new physico-statistical approach is currently followed by CEA; its final objective being to derive the variability of the main results of interest for the safety. This approach involves a fast-running description of extended accident sequences coupling physical models for the main phenomena with advanced statistical analysis techniques. It enables to perform a large number of simulations in a reasonable computational time and to describe all the possible bifurcations of the accident transient.In this context, this paper presents a physical tool (models and results assessment) dedicated to the beginning of the primary phase of an Unprotected Loss Of Flow accident (i.e. before the first pin degradation). This study focuses on the sodium thermalhydraulic behavior during this phase, because according to some previous boiling tests, a stabilized boiling flow could be achieved at the top of some subassemblies, thus preventing pins complete dry-out and consequent core degradation. Experiments that demonstrated this possible boiling stabilization were carried out at constant power. However, in case of an unprotected accident, power will vary according to reactivity feed-back effects.This physical tool is described before presenting the comparison of its results with experimental tests results (static and dynamic aspects) for the thermalhydraulic behavior and with mechanistic SIMMER code results for the global core evolution (including neutronics) during an ULOF transient. Moreover, the flow evolutions obtained with this tool are demonstrated to be in good agreement with the Ledinegg’s quasi-static theory regarding boiling stabilization possibilities or flow excursion. This tool is demonstrated to be capable of reproducing the magnitude of mass-flow rate, reactivity, power evolution during an ULOF with a discrepancy of less than 3% regarding the same transient simulated with SIMMER. It will be used for safety-informed design and stability analyses of fast reactor systems, allowing to emphasize main dominant phenomena and significant trends for safety assessment. In the future, this physical tool, associated to statistical treatments of the effect of uncertainties, will enable large sensitivity analysis studies.