Unprotected Transient Overpower 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.

Analysis of the stress field in the reactor vessel of the China Initiative Accelerator Driven System during postulated ULOF and UTOP transients

The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The subcritical reactor in CiADS is a liquid Lead Bismuth Eutectic (LBE) cooled fast reactor. When the reactor core is in operation, the LBE coolant will directly contact and corrode the inner surface of reactor vessel. Due to the high temperature, the corrosion will be more severe. If the stress on the reactor vessel exceeds the limit, the plastic deformation will occur, leading to the generation and expansion of defects and cracks, and the safety of the reactor will be affected. Therefore, evaluating the stress field of the reactor vessel under different operating conditions is a very important research project. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS, and the ASME Code was used as stress assessment standards. We can preliminarily prove that the stress assessments of the vessel during the postulated Unprotected Loss of Flow (ULOF) accidents satisfy the requirements of ASME Code. The limit reactivity insertion to protect the vessel from plastic deformation is 0.58$ in the postulated Unprotected Transient over Power (UTOP) accidents based on our current results. Therefore, we can preliminarily conclude that the current material selection and structural design of the reactor vessel in CiADS could survive most of the postulated transient accidents considering the stress effect.

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.

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

AbstractFollowing the Fukushima Nuclear Power Plant accident in 2011, it has become increasingly important for reactor safety designs to consider measures that can prevent the occurrence of severe accidents. This report proposes a novel subassembly-type passive reactor shutdown device that expands the diversity and robustness of core disruptive accident (CDA) prevention strategies for sodium-cooled fast reactors. The developed device contains pins with a fuel material that is in the solid state during normal operation but melts into a liquid when the temperature exceeds a certain value (i.e., during a potential accident). When an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident occurs, the device can passively provide significant negative reactivity by rapidly transferring liquefied device fuel into the lower plenum region of the pins via gravitation alone. The reactors containing some of the proposed devices in place of original fuel subassemblies become subcritical before the driver fuels are damaged, even if ULOF or UTOP transient events occur. The present study evaluates candidate materials for device fuels (e.g., metallic alloy, chloride), optimal device pin structures for liquefied fuel relocation, and nuclear and thermal-hydraulic characteristics of the device-loaded core under accident conditions to demonstrate the engineering applicability of the proposed device. This report discusses preliminary results regarding the nuclear requirements for inducing negative reactivity to achieve reactor shutdown under the expected device conditions during an accident.

Finite element analysis of the main reactor vessel in the China Initiative Accelerator Driven System

The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The reactor in CiADS is a subcritical fast neutron reactor cooled by a liquid lead–bismuth eutectic. The reactor operates at high temperature and bears high thermal stress. In addition to the heavy weight of the whole reactor, the vessel will bear large effective stress. If the effective stress exceeds the limit of the material, defects and cracks may occur on the main reactor vessel, which will affect the safety performances of the reactor. Therefore, it is very important to analyze the effective stress field of the reactor vessel. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS. We can preliminarily prove that the maximum effective stress that the vessel will bear during the postulated Unprotected Loss of Flow (ULOF) and Unprotected Transient over Power (UTOP) accidents is less than the yield strength of 316L stainless steel. Therefore, we can preliminarily conclude that the current material selection and structural design of the CiADS vessel could survive the postulated transient accidents considering the effective stress effect.

Transient analysis of the CiADS reactor core loaded with nitride fuel

The China initiative Accelerator Driven System (CiADS) is a nuclear energy system that can perform transmutation of long-lived radioactive nuclides in spent fuel. In this article, the CiADS reactor core loaded with nitride fuel was studied by using SAS4A/SASSYS-1 code. The preliminary simulation results of the unprotected transient over power (UTOP) at BOC (beginning of cycle) and EOC (end of cycle) showed that the maximum reactivity insertion to avoid cladding rupture was 500pcm. For unprotected loss of flow (ULOF), the fuel and cladding failures will not happen in all postulated ULOF cases according to our simulations.

A study of the behavior of the floating absorber for safety at transient (FAST) in an innovative sodium-cooled fast reactor (iSFR)

Negative reactivity feedback is one of the most important inherent safety characteristics of reactors for suppressing changes in thermal power and component temperatures during accidents. The floating absorber for safety at transient (FAST) has been suggested to improve the coolant temperature coefficient (CTC) in a sodium-cooled fast reactor (SFR) for a better self-stabilization of the reactor system. Previous studies on the FAST application showed that the FAST causes power excursion and temperature oscillations under anticipated transient without scram (ATWS) conditions. This study explains the variables affecting the FAST behavior and oscillations with mathematical expressions. In addition, a damping FAST is proposed to passively remove oscillations without any mechanical system. The damping FAST increases the damping ratio by reducing the pin diameter at a height where the difference between the FAST bottom elevation and the equilibrium position is maximum. The performance evaluation for the damping FAST is conducted in an innovative sodium-cooled fast reactor (iSFR) under ATWS conditions. The 7-mm damping FAST unilaterally inserts the negative reactivity in unprotected transient overpower event, dramatically lowering the core power without oscillations. While the previous FAST shows oscillatory behavior repeatedly approaching the coolant boiling temperature, the damping FAST stably keeps the coolant temperature below 800 °C. Moreover, the damping FAST lowers the maximum coolant temperatures for other ATWS events. These findings show the potential of the FAST system to significantly improve the inherent safety of SFRs.

Improvement of a design of the molten salt fast reactor

This study provides an improved design of molten salt fast reactors to reduce the maximum magnitude velocity in the core. The results indicated that the improved design reduces the magnitude velocity with acceptable hot spot temperature where the maximum magnitude velocity is nearly equal to 4.59 m/s and the temperature of the hot spot is greater than the outlet temperature by 9°. The performed analyses are used to adjust the amount of fissile material started for keff = 1. Neutron flux and delayed Neutron distribution study where the position of the maximum Neutron flux is in the center and the position of the maximum thermal flux is in a fertile blanket. Simulations of unprotected loss of heat sink, unprotected loss of flow, unprotected transient overpower, and Unprotected over-cooling are performed and discussed with new geometry. Due to the negative reactive feedback, it is clear that MSFR has a high safety potential.

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)

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.

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.

Safety Analysis of Small Modular Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100 Under Unprotected Transient

SNCLFR-100 is a small modular natural circulation lead-cooled fast reactor, developed by University of Science and Technology of China, aiming at taking full advantage of the good economics and inherent safety of lead-cooled fast reactors to develop miniaturized, lightweight and multi-purpose special nuclear reactor technology. SNCLFR-100 is still in the conceptual design stage, in order to fully evaluate the safety features of the reactor and provide reference for the optimization design of the next stage, three typical transients are selected based on the analysis of the SNCLFR-100 initiating events by using the code Analysis of Thermal-hydraulics of Leaks and Transients (ATHLET), which are unprotected transient overpower (UTOP), unprotected loss of heat sink (ULOHS) and unprotected partial blockage in the hottest fuel assembly. For UTOP, the unexpected positive reactivity insertion of 0.7$ in 15s led to two large power peaks in the core quickly, and then the core power began to decrease and gradually stabilized under the action of various of negative feedbacks of the reactor, the peak temperatures of fuel and cladding rose rapidly with the increase of core power and eventually stabilized at a higher temperature. For ULOHS, as the reactor were driven by natural circulation, the coolant mass flow rate continued to decline after the transient, both core and cladding temperatures rose quickly and the temperature rise were smaller than that of UTOP transient, the reactor shutdown by itself and the peak temperatures of fuel and cladding were smaller than the safety limit. For unprotected partial blockage in the hottest fuel assembly, with the increase of the blockage rate of the hottest fuel assembly inlet, the coolant flow rate, the peak temperatures of coolant, fuel and cladding increased significantly, when the blockage rate increased to 0.9, the coolant flow rate of the hottest fuel assembly dropped to about 12.6% of the normal value, and the cladding peak temperature would exceed the cladding melting point, measures should be taken to avoid the happening of severe accident.

Feasibility study of a gas cartridge loop for the versatile test reactor

A feasibility study of a Gas Cartridge Loop (GCL) for the Versatile Test Reactor (VTR) was carried out to determine the performance requirements of major components and evaluate the performance during normal operation and transient conditions. The major components consist of a circulator/motor, a heater, a test article, a flow guide and a pressure vessel. The pressure vessel incorporates a pressurized, flowing helium loop with capability to test heat-producing test articles up to a 200 mm length of Energy Multiplier Module (EM2) fuel rod at a linear heat rate of 55 kW/m. The heat is rejected to VTR primary coolant sodium flowing along the outside of the GCL pressure vessel. For the baseline design, the helium temperatures at the test article inlet and outlet are 650 °C and 850 °C, respectively, when the mass flow is ~0.01 kg/s. The circulator inlet temperature is 362 °C and the motor power is 11.9 W. An electric heater power is 15.8 kW which is required to maintain the inlet temperature of the test article. Under the normal operating condition, the estimated fuel and cladding temperatures are 1233–1610 °C and 917–1219 °C, respectively.The controllability of the GCL was evaluated for the operational transients during a single fuel cycle by maneuvering either the motor or heater power to keep the test article outlet temperature at 850 °C. For an estimated perturbation of the test article power by ±3.5%, the required motor or heater power variations are ± 4.1% or ± 2.4%, respectively. The safety feature of the GCL was evaluated for the postulated accidents initiated by the VTR and GCL itself under the assumption that the helium flow is pseudo-steady. For the reactivity-initiated accident (RIA) like transient overpower (TOP) of the VTR, the simulation predicted that the margin to fuel melting is 24% and 5% for the protected and unprotected TOP, respectively. For GCL-initiating accidents, the simulation predicted the fuel reaches its melting point when the system loses mass flow by 52% with the nominal heater power or when the system pressure drops by 56% with the nominal motor and heater power level. For both the loss-of-flow and loss-of-pressure cases, the system shall terminate the heater and fission reaction at pre-determined set points. For the loss-of-power case, the GCL requires coastdown capability of the motor for at least several seconds to avoid the test article failure due to fuel melting.

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.

IMPLEMENTATION OF AN AUTONOMOUS REACTIVITY CONTROL SYSTEM IN A SMALL LEAD-COOLED FAST REACTOR

This paper describes the design, implementation and characterisation of an Autonomous Reactivity Control (ARC) system in a small modular lead-cooled fast reactor. The aim of this work was to demonstrate the applicability of the ARC system and to study its dynamic behaviour during an anticipated transient without scram. A simplified one-dimensional model was developed to calculate the heat transfer within the ARC system, and the reactivity worth as a function of the neutron poison’s insertion into the active core was obtained via static neutronic calculations. By coupling the aforementioned models, the ARC’s time-dependent reactivity was derived as a function of the coolant outlet temperature variation. This model was implemented into the BELLA multi-point dynamics code and transient simulations were run. A control rod ejection accident was studied leading to an unprotected transient overpower scenario, in which 350 pcm reactivity was inserted during one second. It was shown that the ARC system provides a forceful negative reactivity feedback and that steady-state temperatures after the transient were reduced by almost 300 K compared to an identical transient without its action. In this scenario, the ARC system managed to stabilise the coolant outlet temperature at a value 100 K above nominal conditions. The implementation of an ARC system provided the reactor with a passively actuated self-regulating reactivity control system able to insert large amounts of negative reactivity in a short amount of time.

Mitigation of severe accidents for SFR and associated event sequence assessment

This article deals with the assessment of severe accident event sequence and mitigation strategy elaboration performed in the frame of ASTRID SFR Gen IV reactor project (Le Coz et al., 2013), including the Japanese/French collaboration involving JAEA, MFBR, Framatome and CEA. In the first part of the paper, the mitigation strategy and the approach defined for the design of mitigation features, which are transverse tubes called DCS-M-TT are presented. By considering 21 DCS-M-TTs into the core as defined at the end of the conceptual design, which enable to extract the core materials and to significantly reduce the core reactivity within the order of 10 s in some typical accidental conditions, the accidental power excursions in ASTRID should not lead to a significant mechanical energy release, thus ensuring the integrity of the main vessel. Nevertheless, some loading assumptions independent from event evolutions have been defined in order to design the main vessel and the core catcher in order to cover extreme accident consequences and to get rid of event sequence dependency when designing the reactor. Then, SIMMER-IV (3D) the main calculation results of ULOF simulations whose goal was to verify the DCS-M-TT efficiency and to support a PIRT are highlighted in the paper. They confirmed that fuel discharge through discharge tubes has a large mitigating impact on power excursions. The last part of this paper deals with an approach dedicated to the description of severe accident event sequences through generic event trees (GETs), which logically summarize the representative courses of accident progression showing relationship between critical causal events and major outcomes of the accidents. A set of trees has been elaborated for the description of each event sequence able to cause severe accidents (ULOF: unprotected loss of flow, UTOP: unprotected transient overpower and USAF: unprotected sub-assembly fault). The event-tree bifurcation points towards one branch or another of the possible event evolution could be examined alternatively with a help of phenomenological event chart (PECs). Afterwards, in association with the GET analysis, a PIRT (Phenomenon identification ranking table) on ULOF has been developed conjointly by Japanese and French sides in order to prioritize R&D needs. About 100 phenomena have been retained and classified depending on the various time periods of the accident event sequences; the primary phase, the short-term relocation phase, the expansion phase and the long-term relocation and cooling phase.

FAST (floating absorber for safety at transient) for the improved safety of sodium-cooled burner fast reactors

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

Nuclear Data Uncertainty Quantification and Propagation for Safety Analysis of Lead-Cooled Fast Reactors

In this study, the Best Estimate Plus Uncertainty (BEPU) approach is developed for the systematic quantification and propagation of uncertainties in the modelling and simulation of lead-cooled fast reactors (LFRs) and applied to the demonstration LFR (DLFR) initially investigated by Westinghouse. The impact of nuclear data uncertainties based on ENDF/B-VII.0 covariances is quantified on lattice level using the generalized perturbation theory implemented with the Monte Carlo code Serpent and the deterministic code PERSENT of the Argonne Reactor Computational (ARC) suite. The quantities of interest are the main eigenvalue and selected reactivity coefficients such as Doppler, radial expansion, and fuel/clad/coolant density coefficients. These uncertainties are then propagated through safety analysis, carried out using the MiniSAS code, following the stochastic sampling approach in DAKOTA. An unprotected transient overpower (UTOP) scenario is considered to assess the effect of input uncertainties on safety parameters such as peak fuel and clad temperatures. It is found that in steady state, the multiplication factor shows the most sensitivity to perturbations in 235U fission, 235U ν, and 238U capture cross sections. The uncertainties of 239Pu and 238U capture cross sections become more significant as the fuel is irradiated. The covariance of various reactivity feedback coefficients is constructed by tracing back to common uncertainty contributors (i.e., nuclide-reaction pairs), including 238U inelastic, 238U capture, and 239Pu capture cross sections. It is also observed that nuclear data uncertainty propagates to uncertainty on peak clad and fuel temperatures of 28.5 K and 70.0 K, respectively. Such uncertainties do not impose per se threat to the integrity of the fuel rod; however, they sum to other sources of uncertainties in verifying the compliance of the assumed safety margins, suggesting the developed BEPU method necessary to provide one of the required insights on the impact of uncertainties on core safety characteristics.

Preliminary neutron kinetic – thermal hydraulic coupled analysis of the ALFRED reactor using PHISICS/RELAP5-3D

The development of a lead-cooled fast reactor (LFR) demonstrator was proposed, mainly in EU, to investigate the feasibility of an industrial size ELFR (European Lead-cooled Fast Reactor). The demonstrator, called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator), consists of a pool-type lead-cooled fast reactor, with a nominal thermal power of 300 MWt. This paper aims to verify the capability of the PHISICS/RELAP5-3D coupled approach to simulate transients of such reactor and to evaluate the effects of accidental scenarios relevant for the safety analysis on the system thermal-hydraulics and on the core power spatial evolution. RELAP5-3D©, developed at Idaho National Laboratory (INL), is a thermal-hydraulic system code, validated for a wide range of transient simulations. The code provides the possibility to simulate innovative working fluids (such as lead and lead alloys) and to use a fully integrated multi-dimensional nodalization. In addition, the need to study complex neutronic systems recommended the development of a new kinetic model allowing the calculation with any number of energy groups and also considering the transport for the angular variations. At this purpose, INL developed PHISICS (Parallel and Highly Innovative Simulation for INL Code System) and its coupling methodology with RELAP5-3D. The simulation activity described in this paper has been focused on the safety analysis of ALFRED reactor assuming the occurrence of two unprotected transient scenarios: unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP). At this purpose, a thermal-hydraulic geometrical scheme of the whole reactor has been developed. The models and the outcomes of the calculations are described and discussed in the paper, highlighting the capability of the coupling approach to obtain results consistent with the ones available in the literature.