Unprotected Transient Overpower Accident Research Articles

Analysis of the thermal-mechanical behavior of SFR fuel pins during fast unprotected transient overpower accidents using the GERMINAL fuel performance code

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered.Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up.The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

UTOP accident analysis and mitigation measures of lead-cooled fast reactor considering coupling effect with SCO2 Brayton cycle

The lead-cooled fast reactor (LFR) coupled with SCO2 Brayton cycle is an advanced nuclear power generation system which has high efficiency and high compactness. The purpose of this study is to investigate the impact of coupling effect between LFR and SCO2 cycle on the safety analysis of unprotected transient overpower (UTOP) accident, and to propose mitigation measures to UTOP accident with the aid of SCO2 cycle. Firstly, the dynamic simulation model of a 100 MW small modular natural circulation LFR coupled with SCO2 Brayton cycle was developed using APROS software, and was verified through the steady-state and transient simulations. Secondly, the UTOP accident of the LFR coupled with SCO2 Brayton cycle was analyzed, including a comparative case which simplified SCO2 cycle as a fixed boundary in the analysis. Finally, two measures to mitigate the UTOP accident by charging more SCO2 to the circulation loop from the storage tank and by maintaining constant SCO2 temperature at the inlet of in-reactor heat exchangers (IRHXs) were proposed. The results indicate that the conventional method of simplifying power generation system as the fixed boundary underestimated the severity of accident. Specifically, the maximum fuel cladding temperature was about 40 °C higher when considering the coupling effect of reactor and SCO2 cycle than that of assuming SCO2 cycle as the fixed boundary, which greatly exceeded the safety limit. Moreover, within the scope of present UTOP accident conditions, the measures proposed in this study could mitigate the severity of accident. The maximum cladding temperature was reduced by 15–23 °C in the condition of maintaining constant SCO2 temperature at IRHX inlet, and was reduced by 5–7 °C in the condition of charging more SCO2 to the circulation loop, compared with the condition not applying any measures.

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.

Study on initiating phase of core disruptive accident (Validation study of SAS4A code for the unprotected transient overpower accident)

Study on initiating phase of core disruptive accident (Validation study of SAS4A code for the unprotected transient overpower accident)

A comparative study of reactivity feedbacks including in-pin motion and their impact on UTOPA in a 600 MW fast reactor

The design experience of PFBR (Prototype Fast Breeder Reactor) gives confidence in designing advanced Sodium cooled Fast Reactors (SFR) with improved economy and passive safety features towards enhanced safety. In this regard, Fast Breeder Reactor (FBR) is optimised with 600 MWe leading to better economy. To prevent severe core damage & large radioactivity release to the public and also to enhance the safety, improved inherent safety characteristics are adopted with passive shutdown features. Important safety measures taken to improve inherent safety are the optimization of sodium void reactivity beneath 1 $, which is in line with the current international thinking. To incorporate the enveloping beyond design basis events (BDBE) within the Design Extension Condition (DEC), additional passive safety mechanisms are envisaged to strengthen the reactor safety and practically eliminate the probability of Core Disruptive Accidents (CDA). One such BDBE is Unprotected Transient Over Power Accident (UTOPA), in which uncontrolled withdrawal of one control rod leads to rise in power and fuel temperature. Under UTOPA, it is possible for fuel to melt and undergo in-pin fuel motion through the available central hole. Upon initiation of melting, fission gases trapped within the fuel microstructure are released into the pellet cavity along with molten fuel. Consequently, a multi-phase flow occurs inside the pellet cavity, with several hydrodynamic effects influencing the fluid motion. The relocation of molten fuel from high fuel void worth to low fuel void worth region results in a negative reactivity feedback. This in-pin fuel motion or squirting feedback reduces the total reactor power as well as the hot-spot clad and coolant temperatures. With change in reactor power during UTOPA, the balance of the plant gets affected. This results in a change in the inlet coolant temperature, which affects the overall core temperature profile and its respective feedbacks. Considering the above said arguments, UTOPA study has been carried out with different initial conditions such as Constant Inlet Coolant Temperature (CICT), Time Dependent Inlet Coolant Temperature (TDICT), with and without In-Pin Fuel Motion (IPFM) feedback. Comparisons of the results are made to get better understanding of various feedbacks and their impact on UTOPA, highlighting the importance of in-pin fuel motion reactivity feedback in a medium sized fast reactor core. From the study, it is concluded that the case without considering IPFM feedback gives the most conservative results and if the transients converge to a safe state without considering IPFM feedback then, the transients are expected to settle down to a safe state with IPFM feedback.

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.

Uncertainty analysis in coupled neutronic/thermal-hydraulic calculations based on computational fluid dynamics

There is an inevitable uncertainty in the coupled neutronic/thermal–hydraulic calculations, which are used for simulating the performance of nuclear reactors, due to the influence of design parameters and initial conditions. The application of uncertainty analysis methods in such calculations can reduce the “uncertainty zone” of calculation results as well as the unnecessary safety margin of the reactor. In this study, based on the combination of user-defined function (UDF) of FLUENT, a computational fluid dynamics (CFD) code, with the neutron dynamics model, fuel pin heat transfer model, and SIMLAB software, an uncertainty analysis platform called CFD/PFS is developed. The CFD/PFS platform is used to simulate the unprotected transient overpower (UTOP) accident in the small modular natural circulation lead–bismuth alloy cooled fast reactor-10MWth (SNCLFR-10) reactor, which is followed by uncertainty quantification. The results of uncertainty and sensitivity analysis justify the good reliability and accuracy of coupled neutronic/thermal–hydraulic calculations. The titular values of transient parameters, such as peak reactivity and peak fuel temperature, are within the bilateral tolerance limits, and the relative deviation between these values and the limit values is less than 3.95%. As far as UTOP accident, Doppler coefficient is found to be the main source of uncertainty, which has the most significant effect on reactor safety.

Modelling, verifications and safety feedback assessment of annular fast reactor fuel pins with severe accident code MITRA

Fuel of fast reactors is designed in the form of cylindrical pellets of solid or annular geometry. The annular configuration provides advantages in terms of higher achievable energy extraction and lower fuel-clad mechanical interaction. Another advantage of this configuration is the hydrodynamic flow of molten fuel inside the annular pellets during accidental meltdown, known as in-pin fuel motion. This motion can provide safety benefits during an unprotected transient overpower accident (UTOPA), if the molten fuel is dispersed significantly away from the core mid-plane. The physics behind the flow is complex. Intricate theoretical modelling and detailed experimental validation are pre-requisites for the reliable estimation of safety benefits. To meet this requirement, a Multi-phase In-pin Thermal hydraulic Relocation Algorithm (MITRA) is developed under the purview of the pre-disassembly analysis code, PREDIS. This paper presents a theoretical foundation of the algorithm followed by experimental verifications, burnup and top blanket design sensitivity analyses and whole core UTOPA simulations. Results show that regardless of the fuel burnup level, the melt tends to agglomerate into a column, slightly below the core mid-plane. Upon further melting, this column grows slowly, devoid of the rapid dispersion associated with fuel squirting. Minor deviations in the melt position arise due to variations in thermal parameters with burnup. UTOPA analysis shows a reduction in the safety feedback and increased melting. Modifying the conventionally solid top axial blanket to annular geometry enhances the safety feedback only after 34.2% melting.

Advanced Studies and Statistical Treatment for Sodium-Cooled Fast Reactor Pin Failures During Unprotected Transient Overpower Accident

Usually, simulation tools are validated based on experimental data considering a best estimate simulation case; however, there is no quantification of this validation, which remains based on rough expert judgment. This technical note presents advanced validation treatment of the simulation tool OCARINa devoted to unprotected transient overpower (UTOP) accidents on two CABRI tests considering this time a Best Estimate Plus Uncertainty (BEPU) approach. The output results of interest are both scalar physical data such as the time and location of the pin failure and associated molten mass and vector data such as temperature axial distribution or temperature evolution versus time. This approach is a first step in quantifying the degree of agreement between the calculation results and the experimental results. It is of great interest for the verification, validation, and uncertainty quantification approach, which leads to the qualification of scientific calculation tools. Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) research and development project in which the CEA is involved, OCARINa is a physical tool, relevant for performing preconceptual design studies and devoted to simulation of UTOP accidents on heterogeneous cores. Such accidents could not be simulated with mechanistic calculation tools such as SAS4A or SIMMER with their current capabilities; the thermomechanical models are not finalized in the SIMMER tool, and the SAS4A tool is validated only for homogeneous cores. The final objective aims at deriving the variability of the main results of interest to quantify the safety margins. The final use of the OCARINa tool being to perform sensitivity studies on the various possible sodium fast nuclear preconceptual core designs, the validation of this tool is first discussed at the pin scale (where separate-effects test measurements are available) based on statistical treatment. This enables one to determine the lacks and uncertainties of this tool. The modeling is then extended from local pin behavior to global core behavior adding a point-kinetics neutronic model. Final simulations of UTOP accidents caused by a uniform space reactivity ramp on an SFR core are realized taking into account the specificities of the pins of the various assemblies. The orders of magnitude of mechanical energy released are derived.

In-pin fuel motion dynamics for beginning-of-life core in fast breeder reactors

Melting of nuclear fuel can occur during any unexpected accident where there is rise in power. Specifically, in the unprotected transient overpower accident (hereafter UTOPA), fuel melts and flows axially inside the fuel pin. A previous study of this phenomenon, known as in-pin fuel motion, in a medium-sized fast reactor with an equilibrium core concluded that it basically enhances the inherent safety features. The objective of the present study is to confirm whether such an enhancement also occurs for the beginning of life core (hereafter BOL) and to quantify the differences. Therefore, this article reports the best estimate and conservative UTOPA analyses for both BOL and equilibrium cores. A fuel mass relocation analysis of both cores shows that due to the absence of fission gas perturbations upon melting, the axial relocation of molten fuel is least in the BOL core. This leads to a smaller relocation feedback in the BOL core for the same amount of melting. Results show that while the molten fuel relocation feedback does act as an inherent feedback for the BOL core, it is unable to prevent extensive fuel melting, especially under the conservative UTOPA analyses. On the contrary, such an extensive amount of melting is altogether avoided in the equilibrium core due to greater relocation feedback for lesser amount of melting. A thermodynamic analysis of molten fuel during UTOPA is carried out and it is found that there is a greater possibility of vapourization of molten fuel in the BOL core as compared to the equilibrium core.

Sodium-cooled fast reactor pin model for predicting pin failure during a power excursion

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) in which the CEA (French Commissariat à l’Energie Atomique et aux Energies Alternatives) is involved, the French innovative reactor design behavior under severe accidents conditions has to be assessed. Such accidents have mainly been simulated with mechanistic calculation tools (such as SAS-SFR and SIMMER-III). As a complement to these codes, which provide reference accidental transient calculations, a new physico-statistical approach is being developed at CEA; its final objective being to derive the variability of the main results of interest to quantify the safety margins. This approach requires fast-running tools to simulate extended accident sequences, by coupling models of the main physical phenomena with advanced statistical analysis techniques. The tool enables to perform a large number of simulations in a reasonable computational time and to describe all the possible scenario progressions of the hypothetical accidents. This general approach, combining mechanistic codes and evaluation tools, has already been conducted for some accidental initiator families (USAF – Unprotected SubAssembly Fault (Marie et al., 2016) and ULOF – Unprotected Loss Of Flow (Droin et al., 2017). In this context, this paper presents a physical tool (numerical models and result’s assessment) dedicated to the simulation of the beginning of the primary phase of the Unprotected Transient OverPower accidents (i.e. before failure of sub-assembly wrapper).At the beginning of this primary phase, the fast increase of nuclear power induces a strong temperature rise in the fuel pellets leading to strong mechanical and thermal loads on the cladding which could lead to clad failure or/and fuel meltdown. These phenomena are described and modelled analytically in single pin geometry in accordance to the level of details required to catch all the decisive phenomena.Slow power increase transients, such as control rod withdrawal, and fast power increase transients have been investigated in the past. Experimental validation on CABRI (experimental reactor dedicated to safety studies) and CESAR (Circuit d’Etude de l’ébullition du Sodium lors d’un Accident de Réactivité) experiments were carried out focusing on the amount of molten fuel formed during the transient, on the propagation of the void front in the channel in case of sodium boiling and on the pin failure mechanisms. Furthermore, a comparison of the physical tool calculation results was performed against reference accident SIMMER-III calculations.The tool is demonstrated to be able to predict the radial propagation of the molten zone in the fuel pin, the pin failure mechanism and void front propagation with a discrepancy of less than 10%. In the future, this physical tool, associated with a point kinetic neutronic model, will be used to simulate the global core behavior under an UTOP transient following a ramp excursion.

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.

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a Fluoride-Salt-Cooled High-Temperature Test Reactor

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses high-temperature tristructural isotropic (TRISO) fuel with a low-pressure liquid salt coolant. Design of the fluoride-salt-cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble-bed core design with a coolant temperature of 600–700°C is being planned for construction by the Chinese Academy of Sciences’ (CAS) Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal-hydraulic transient analyses of an FHTR using SINAP’s pebble-bed core design as a reference case. A point kinetic model is implemented using computer code by coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating several transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that SINAP’s pebble-bed core is a very safe reactor design.

Perturbation analysis of prototype fast breeder reactor equilibrium core using IGCAR and ERANOS code systems

First order approximation in standard perturbation theory has been used for the safety analysis of Indian prototype fast breeder reactor (PFBR). Material void and Doppler worth distribution obtained with first order approximation has been used to find the response of PFBR core during reactivity transients like unprotected loss of flow accident (ULOFA) and unprotected transient over power accident (UTOPA) up to sodium voiding. The validity of first order approximation applied to larger perturbations during such accidents has been tested by developing an exact perturbation code PERTX. 2D diffusion code ALCIALMI and ABBN-93 evaluated nuclear data has been used for the analysis.European Reactor Analysis Optimized Calculation System (ERANOS) is a code system used for reliable neutronic calculations of current and advanced fast reactor cores. The results of perturbation analysis of PFBR done with IGCAR code systems have been compared with those obtained with ERANOS 2.1 and JEFF 3.1 nuclear data. The present study reveals the adequacy of the IGCAR perturbation methods in 2D for the safety analysis of PFBR core involving transients with fuel, structure and coolant temperatures change.

Preliminary assessments of carbide fuel pins during mild overpower transients in LMFBRs

Mixed carbide fuel has been proposed for advanced liquid metal fast breeder reactors (LMFBRs). In this paper the general characteristics of carbide fuel are reviewed and a survey of available data for use in safety studies is presented. A preliminary investigation of the unprotected transient overpower accident sequence postulated for a large carbide fueled reactor is also presented. The HOPE computer code is the computational tool used to analyse a conceptual core design fueled with mixed carbides of uranium and plutonium. The details of this design are presented and compared with those of the Clinch River Breeder Reactor Plant. Results of HOPE calculations for both helium and sodium bonded carbide fueled cores are presented and compared with a similar study performed on an oxide core. Conclusions are drawn based on the findings of the preliminary transient overpower studies conducted.