Separate Effect Tests Research Articles

A new approach to investigate secondary hydriding phenomenon on M5Framatome clads under high–temperature LOCA conditions

The focus of this study is about a new experimental approach for a separate effects study of the secondary hydriding phenomenon under LOCA conditions. Many nuclear institutes perform semi–integrals tests to study the cladding behaviour during a LOCA transient. Those tests combined several phenomena and performing a detailed analysis of the secondary hydriding phenomenon using these tests can be challenging. A dedicated experimental protocol aiming at simulating secondary hydriding has been set up. Separate effects tests (SETs) were then carried out using this protocol to study the effects of both oxidation duration and temperature, on the hydrogen absorption during the oxidation stage of the LOCA transient on M5Framatome11M5 and M5Framatome are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries. cladding. The effects of gap size were also investigated. Metallographic analysis has been used to characterise the M5Framatome clad metallurgical transformation after the high–temperature (HT) oxidation.

Separate effect experimental test and theoretical investigate of pressure wave propagation and pressure rising process upon SGTR accident in advanced LFRs

Steam Generator Tube Rupture accident (SGTR) accident is one of the main safety events to the industrial application of Lead-cooled Fast Reactors (LFRs). This paper presents a separate effect test facility, called Lead-bismuth Eutectic Steam generator tube rupture Test facility (LEST), for testing high-pressure subcooled water jetting into a Lead Bismuth Eutectic (LBE) pool, and analyzes experimental data of maximum initial pressure peak, dynamic pressure attenuation together with the pressure rising process upon SGTR. The experimental analyses put forward a general equation form together with a dimensionless number of the initial pressure peak associated with a high-pressure Newtonian fluid injecting into another low-pressure Newtonian fluid, and the derivation of the general equation form is specifically applied to match a further semi-empirical equation for an initial maximum pressure peak with 5–10 MPa water injection in SGTR. The attenuation curve of dynamic pressure wave in each experiment is matched, and more weakening conditions can change the pressure peak from concentrated and non-period peak group to approximate period 1.4–3.7 s pulse group. The size relationship between pipe length L and nozzle diameter D, which L/D=12, is set to divide SGTR into small wall rupture and tube shear fracture, further three stages of pressure rising are distinguished in the tube shear fracture. This work provides experimental data to help determining the severity of SGTR and assessing the consequences of this event as part of the safety case for the industrial application of LFRs.

Global Sensitivity Analysis for Segmented Inverse Uncertainty Quantification in the Safety Analysis of Nuclear Power Plants

Within the Best Estimate Plus Uncertainty framework for the safety analysis of Nuclear Power Plants, the quantification of the uncertainties affecting the Thermal-Hydraulics (T-H) codes used is crucial. For this, Inverse Uncertainty Quantification (IUQ) methodologies are being developed for determining the probability density functions of relevant T-H codes input parameters, based on experimental data from Separate Effect Tests (SETs) experimental facilities. In practice, IUQ is challenged by the large range of variability of the experimental data in terms of Initial and Boundary Conditions (ICs & BCs), because the experimental campaigns are designed to cover the widest possible domain of conditions with the smallest number of experiments, so that same or similar ICs and BCs are seldomly repeated. To address this issue, we propose to use global sensitivity analysis, to tailor the IUQ on specific sub-regions described by segmented ICs & BCs domains. The methodology proposed is exemplified on two SETs, namely Sozzi-Sutherland and Super Moby Dick, whose experimental databases have been made available in the ATRIUM (Application Tests for Realization of Inverse Uncertainty quantification and validation Methodologies in thermal hydraulics) project promoted by the OECD/NEA/CSNI. The results obtained are superior to those of traditional IUQ methodologies for models highly sensitive to ICs & BCs.

Early-phase core degradation interpretation by interaction between fuel cladding and fuel components in Fukushima Daiichi nuclear power plant Unit 2

ABSTRACT The early phase of the core degradation at the Fukushima Daiichi Nuclear Power Station Unit 2 was investigated in terms of a fuel-component degradation by materials interaction by carrying out the accident analysis, evaluation of chemical conditions and experimental works. First, the accident scenario was evaluated to specify a target period (from 18:40 to 22:40 on NaN Invalid Date NaN), and then the temperature and H2/H2O pressure ratio in the core region during this period were calculated based on plant parameters. Subsequently, semi-integral experiments were performed using the mocked-up fuel components of a boiling water reactor (BWR). Furthermore, separate-effects tests using Ni-alloy/Zircaloy4 were also performed to understand the phenomenon observed in the semi-integral experiments, where the Ni-Zr reaction occurred even with the oxide layer on the Zr-based alloy. Consequently, it was estimated that the Zr-Ni eutectic melting reaction above 1100°C under steam-starved conditions resulted in a fuel failure at 19:08, melting of spacer locations at 19:37, and upper-tie-plate at 21:26. The first and second degradations could have caused the release of radioactive elements or particles leading to the neutron detection at 21:00 in front of the main gate, and the third degradation could have resulted in the separation of the upper-tie-plate.

Scaling methodologies and similarity analysis for thermal hydraulics test facility development for water-cooled small modular reactor

Small modular reactors (SMRs) represent a promising option for providing clean and sustainable energy due to their potential for enhanced safety, reduced capital costs, and increased siting flexibility. However, new reactor systems require the development and operation of representative scaled-down test facilities to support the verification and validation of system computer codes and models. This study reviews the research on scaling methodologies and similarity principles pivotal in developing non-nuclear integral effects test and separate effects test facilities for water-cooled SMRs. The study focuses on a review of the scaling methods, similarity approaches, and possible challenges posed by the unique and compact design features of integral-pressurized water reactor-type SMRs, and their representative test facilities. This study also reviews previous research related to scaling and similarity methodologies and provides insights into design considerations for achieving prototypic conditions in test facilities. The findings and recommendations emphasize the broader impact of appropriate scaling and similarity principles to ensure meaningful and transferable results from non-nuclear test facilities to accelerate the safe and efficient deployment of next-generation water-cooled SMRs.

Fractional scaling analysis of cladding temperature in large break loss of coolant accident

In integral test facilities, some phenomena relevant in the separate effect test are usually neglected due to scaling limitations and design compromises. The typical heat transfer phenomenon in reactor core in large break loss of coolant accident (LBLOCA) scenario is analyzed with fractional scaling analysis (FSA) method in this work. A conservation form of cladding temperature is derived on the basis of energy conservation. Then the scaling criteria and their importance are evaluated. It is found that the similarities of the criteria for convection heat transfer and heat accumulation should be guaranteed since their priorities are high. The contribution of initial pressure and distribution of coolant inventory to cladding temperature in LBLOCA scenario is too slight. This work is encouraging and may shed a light on the coupling application of hierarchical two-tiered scaling (H2TS) and FSA method in engineering application.

A Blowdown Critical Heat Flux Experimental Vehicle

In the most challenging nuclear power plant accidents, transient critical heat flux (CHF) is a primary phenomenon that drives peak cladding temperature, and ultimately, fuel failure. It has not yet been determined whether the use of steady-state CHF methods can accurately predict transient CHF under the conditions of a blowdown due to a loss-of-coolant accident. There are limited comprehensive experiments at prototypic conditions. To address this deficiency, a quality separate-effects test facility was built to simulate an electrically heated rod under blowdown conditions. Testing reached full pressurized water reactor thermal-hydraulic conditions. With scaled break sizes as large as a double-end cold leg break, CHF was repeatedly measured with depressurization rates ranging from 7 to 17 MPa s−1. These measurements at prototypic conditions acquired in a controlled methodology are novel to the body of knowledge. Several steady-state CHF methods and heater models were evaluated using RELAP5-3D simulations and the Dakota framework. The results showed that many steady-state CHF methods performed inadequately, but that recently developed wide-ranged, look-up table methods had the most acceptable results. Additionally, the results showed no significant correlation between prediction accuracy and the depressurization rates tested.

System code validation in periodic two-phase flow from low-pressure oscillations

Low-pressure two-phase flow instabilities can potentially challenge start-up transients of water-cooled nuclear reactors. Predicting oscillations caused by flow instability is therefore paramount to reactor safety. While many thermal hydraulics system codes are developed as general-purpose tools, their performance is not guaranteed outside their optimized application ranges. The current study therefore performs validations of two thermal hydraulics system analysis codes, the Adaptive SYStem Thermal-hydraulics Version 3 (ASYST VER3) and Reactor Excursion and Leak Analysis Program MOD3.3 (RELAP5/MOD3.3), in reproducing oscillatory two-phase flow induced by low-pressure flashing instability. Benchmark conditions are selected from a novel dataset resolving void fraction with radial resolution and quantifying periodic behavior with ensemble averaging. Focusing on the prediction of transient two-phase phenomena rather than the determination of stability, validations are conducted by simulating a single channel under prescribed periodic boundary conditions. ASYST exhibits a systematic underprediction of void fraction in an adiabatic chimney downstream of a heated section. The prominent causal discrepancy is identified as a lost travelling void wave due to overpredicted condensation. RELAP5 noticeably overpredicts subcooled boiling and underpredicts flashing, which is consistent with existing validations against steady-state separate-effect tests under low pressure. This degraded code performance also suggests that in low-pressure transients prone to flashing instability, the confidence in RELAP5 derived from existing validations against integral-effect tests shall be carefully limited to validated capabilities in integral systems. In general, the current study fills the previous gap of knowledge about the performance of ASYST and RELAP5 in predicting detailed transient two-phase phenomena in low-pressure flashing-induced oscillations. The identified code defects reveal future directions for code improvements eventually towards reliable application of ASYST and RELAP5 under low pressure beyond their original calibration.

Overview of separate effect and integral system tests on the passive containment cooling system of SMART100

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

Numerical Evaluation of Sodium Droplet Swarm Combustion and Its Modeling Optimization

Because of its superior thermal-hydraulic qualities, liquid sodium has been applied to a variety of industries, including energy storage, solar energy, sodium-cooled fast reactors, and aerospace. However, fires brought on by sodium leaks at high pressure can have major thermodynamic repercussions and put employees and equipment in use at risk directly or indirectly. As a result, a realistic and accurate forecast of the combustion behavior of sodium droplet swarm can offer technical backing for the use of liquid sodium in engineering as well as a way of sodium fire prevention and control. Spray dynamics (droplet settling, droplet particle size distribution, etc.), combustion kinetics (premixed combustion, gas phase combustion, etc.), sodium aerosol diffusion, and other specialized phenomena all contribute to the complex process of sodium droplet swarm combustion. The NACOM code created by Brookhaven National Laboratory for sodium droplet swarm combustion is utilized in this paper as a framework. The code is first validated using the benchmark of the sodium droplet swarm combustion tests carried out by prestigious institutions. The validation results demonstrate that the code’s drag model, droplet combustion model, and heat transfer model are to blame for the significantly overestimated thermodynamic effects of sodium droplet swarm combustion. NACOM is subsequently developed twice for the authors’ previously developed vapor-liquid two-layer-structure drag model, chemical kinetic combustion model, and suitable heat transfer coefficient. It is then thoroughly assessed for the separate-effects tests and integral-effects test. The evaluation results demonstrate that the optimized drag model accelerates the settling of sodium droplets due to the consideration of the sodium-vapor drag reduction effect, reducing the thermodynamic effects of liquid sodium combustion; the optimized premixed combustion model can accurately predict the low-temperature sodium droplet swarm combustion conditions, resolving the issue of serious misvaluation of the original version of NACOM. The associated research findings can serve as valuable resources and tools for deeper comprehension of the combustion effects and mechanisms of sodium droplet swarm under various operating settings (such as leakage rate and oxygen concentration).

Numerical analysis of reflood heat transfer and large-break LOCA including CRUD layer thermal effects

This study examined the effects of CRUD on reflood heat transfer behaviors of nuclear fuel rods during a loss-of-coolant-accident (LOCA) in a pressurized water reactor using a best-estimate thermal-hydraulic analysis code. Changes in thermal properties and boiling heat transfer characteristics of the CRUD layer were extensively reviewed, and a set of correction factors to reflect the changes was implemented into the code. A heat structure layer reflecting the effects of CRUDs on the properties was added to the outer surface of the fuel cladding. Numerical simulations were conducted to examine the effects of CRUDs on reflood cooling of overheated fuel rods for representative separate and integral effect tests, FLECHT-SEASET and LOFT. In LOFT analysis, the average cladding temperature was increased due to the low thermal conductivity of CRUD during steady-state operation; however, in both analyses, the peak cladding temperature decreased, and the quenching time was reduced. Obtained results revealed that when the porous CRUD layer is deposited on the fuel cladding, two opposite effects appear. Low thermal conductivity of the CRUD layer always increases fuel temperature during normal operation; however, its hydrophilic porous structures may contribute to accelerated reflood cooling of fuel rods during a LOCA.

Hydrodynamic design of the Separate Effect test facility for Flow-Accelerated Corrosion and Erosion (SEFACE) studies in liquid lead

Flow-accelerated corrosion and erosion (FACE) phenomena can be crucial for performance of structural elements in heavy liquid metal (HLM) cooled reactor systems. Existing experimental observations indicate that turbulent flow characteristic can affect FACE, but there is no quantitative data that can be used for model development and validation. Main recirculation pump impellers, which operate at high relative velocities and rotational flow conditions can be especially vulnerable to FACE. For comparison, the core internals operate at lower velocities and in axial flow conditions, but at higher temperatures and neutron fluence. Hence, systematic experimental data is needed to improve our knowledge on FACE phenomena. The Separate Effect Test Facility for Flow-Accelerated Corrosion and Erosion (SEFACE) is designed to obtain such experimental data including high relative velocities (up 20 ms−1) and high temperatures (400 to 550 °C) of liquid lead. This article focuses on the hydrodynamic design of SEFACE. The aim of the design is to achieve well defined flow conditions for experiments and ensure safe operation of the facility. First, we examine three design concepts (i.e., forced convection loop, rotating cylinder, and rotating disk) and motivate the choice of the rotating disk approach for SEFACE. Second, we discuss different design options, i.e., a confined rotor–stator test chamber and the unconfined rotating disk configuration. We used Reynolds-Averaged Navier Stokes (RANS) calculations to identify and solve the issues stemming from the high rotational speed. These include, for instance, lead free surface deformation, radial pressure buildup, and axial bending forces due to asymmetric test chamber. The CFD-derived torque and power predictions in rotor–stator and rotating disk systems are verified with selected empirical turbulent friction factor correlations or/and DNS calculations. We demonstrate that the developed hydrodynamic design of SEFACE solves identified issues and enables obtaining experimental data under well-defined flow conditions. The findings are deemed to also be applicable to the design of rotating disk-type FACE installations for other liquid mediums.

Analysis of forced convection in the HTTU experiment using numerical codes

The High Temperature Test Unit (HTTU) was an experimental set-up to conduct separate and integral effects tests of the Pebble Bed Modular Reactor (PBMR) core. The annular core consisted of a randomly packed bed of uniform spheres. Natural convection tests using both nitrogen and helium, and forced convection tests using nitrogen, were conducted. The maximum material temperature achieved during forced convection testing was 1200 °C. This paper presents the numerical analysis of the flow and temperature distribution for a forced convection test using 3D CFD as well as a 1D systems-CFD computer code. Several modelling approaches are possible, ranging from a fully explicit to a semi-implicit method that relies on correlations of their associated phenomena. For the comparison between codes, the analysis was performed using a porous media approach, where the conduction and radiative heat transfer were lumped together as an effective thermal conductivity and the convective heat transfer was correlated between the solid and gas phases. The results from both codes were validated against the experimental measurements. Favourable results were obtained, in particular by the systems-CFD code with minimal computational and time requirements.

OECD/NEA/CSNI state-of-the-art report on scaling in system thermal-hydraulics applications to nuclear reactor safety and design (The S-SOAR)

The present paper deals with scaling in nuclear-system thermal-hydraulics (SYS TH), including the connection with Nuclear Reactor Safety Technology (NST). The paper is entirely derived from the S-SOAR document issued by CSNI of NEA, NEA/CSNI/R(2016)14, 2016 (Bestion et al., 2016). Scaling has constituted ‘an issue’ since the beginning of the exploitation of nuclear energy for civil purposes, with main reference to the generation of electricity. A Nuclear Power Plant (NPP) constitutes a technologically complex industrial system and it is characterized by the impossibility of, or the large difficulty in, characterizing the system’s performance under the conditions of the design. So, models were designed, constructed, and operated under downscaled ranges of values for one or more of selected parameters (e.g. power, volume, height, pressure, etc). These features lay at the origin of the scaling issue, i.e. the difficulty in demonstrating that a model behaves like the prototype. Integrated definitions of the widely adopted terms, ‘scaling’, ‘scaling issue’, and ‘addressing the scaling issue’ are part of the present document. The related application domain includes the NST, and the licensing for water-cooled nuclear reactors under operation, under construction, or under an advanced design stage at the time of publication of the (Bestion et al., 2016). Scaling-related analyses are done in different areas of SYS TH and NST. These include the design of test facilities (both integral and separate-effect test facilities, IETF and SETF), the design of experiments (including Counterpart Test, CT), the demonstration of the capability of any computational tool, and the evaluation of uncertainty affecting the prediction of the same computational tools. A variety of approaches have been used to address the scaling issue, including non-dimensional analysis of mass, energy- and momentum-balance equations, derivation and application of scaling factors, including the hierarchy of relative importance, performing experiments at different scales, and running the SYS TH computer codes. This paper discusses the key areas and the key approach for scaling. It was found that the SYS TH computer codes, following their application to differently scaled experiments, demonstrate that the accuracy of their predictions may not depend upon the scale of the considered experiments. The TH codes also may constitute an additional valuable tool for addressing the issue of scaling.

Investigation of system optimization and control logic on a solar geothermal hybrid heat pump system based on integral effect test data

The solar geothermal hybrid heat pump systems hold great promise for sustainable heating and cooling applications. However, the practical challenges and limitations have resulted in a lack of extensive field trials due to the time and financial constraints associated with conducting field trials. As a result, previous studies have primarily relied on dynamic simulation tools to estimate the performance of prototype systems. The annual performance and efficiency of these hybrid systems are heavily influenced by the control logic of their individual components such as PVT and GSHP unit and the overall operation of the system. Consequently, there is a pressing need to conduct a comprehensive study on the sensitivity of system performance to variations in the operation control, particularly when employing advanced logic, such as fuzzy logic control. When it comes to real-world implementation, there are concerns about the feasibility of integrating additional components like PVT and GSHP into the system. Advanced control strategies like fuzzy logic may also raise questions about the initial capital costs and whether the benefits outweigh the investments. Thus, the system optimization study needs regarding to energy consumption and economic feasibility analysis based on precisely predictable simulation model and practical initial capital cost. In the present study, a solar geothermal hybrid heat pump system has been modeled with simulation and designed IET (Integral Effect Test) facility in an industrial building based on component SET (Separate Effect Test) data, conducted a comparison study for validation of the simulation model based on real field heating and cooling weekly IET data. To optimize the system's performance, annual Operation and Maintenance (O&M) costs sensitivity study was conducted for four different Cases, considering the effects of PVT, GSHP components, and fuzzy logic control. Furthermore, an Net Present Value (NPV) method economic analysis was conducted to evaluate the economic feasibility analysis, considering the initial capital cost, inflation rate, and annual O&M costs savings. The results demonstrate that the simulation model can precisely estimate annual energy consumption within 7.4 % compared to heating and cooling IET operation data. The sensitivity study revealed that Case 4, incorporating fuzzy logic control, exhibited the lowest energy consumption of 9450 kWh/year and the highest annual O&M savings of $1869 among the four Cases studied. However, the advanced system control logic increased the initial capital cost by 34 % ($12,490) compared to Case 1's single-stage GSHP. The NPV economic feasibility analysis, considering the initial capital cost, inflation rate, and annual O&M savings in comparison to Case 1, indicated that Case 4 with fuzzy logic control system showed the shortest payback period of 8.5 years.

The deformation and burst behavior of Zircaloy-4 cladding tubes with hydride rim features subject to internal pressure loads

This study is motivated by the desire to conduct a separate effects test replicating specific aspects of an in-pile pulse reactor test on high exposure light water reactor (LWR) fuel rods. The principal purpose of the tests was to determine whether temperature and or hydrogen concentration thresholds exist in assessing the vulnerability of a stress relieved zirconium alloy cladding tubes to low strain ruptures. A secondary objective was to determine whether observations of the cladding stress during deformation could aid in the understanding of the crack formation and propagation in cladding tubes with hydride rim features. The study utilized artificially hydrided Zircaloy-4 cladding tubes and an isothermal, internal pressure test that was controlled with an in-situ strain measurement. The test results show a clear transition in failure vulnerability between room temperature and 150 °C. Below 150 °C, cladding tubes rupture at very low uniform plastic strain—even at relatively modest cladding hydrogen levels. At and above 150 °C, cladding tubes only rupture at low uniform plastic strain if the cladding hydrogen content is greater than 500 ppm and usually experience at least some modest amount of plastic deformation prior to rupture. Post-test metallographs from non-ruptured cladding tubes show numerous primary cracks forming in the hydride rim and arresting at the metal/hydride interface. This observation combined with the decrease in bulk yielding of hydrided cladding tubes when compared to fresh non-hydrided tubes leads to the conclusion that the fracture stress of the hydride rims is related to and bounded by the yield stress of the zirconium metal. Some evidence is presented for a change in crack propagation mechanism from one of void formation on hydride plates and linkage at temperatures below 150 °C to more classical ductile shearing beneath the primary crack tip at temperatures above 150 °C. This apparent change in crack propagation mechanism may help explain the observed temperature and hydrogen concentration thresholds.

Computational model development and validation of fuel dispersal phenomena

As the US nuclear industry is renewing efforts to extend the rod average burnup limits in existing pressurized water reactors, it becomes imperative to investigate the safety implications associated with high burnup fuel fragmentation and the potential dispersal of fragmented fuel into the reactor coolant system. Additionally, as there is a growing interest in transitioning to new nuclear fuel designs to accommodate higher fuel burnup levels, it becomes essential to conduct studies on fuel behavior during design basis accident scenarios, particularly during a postulated loss-of-coolant accident, in order to ensure the safe operation of light-water reactors. The present study develops a computational model to simulate the complex three-phase flow of the fuel dispersal phenomena that occur following a breach in the cladding, when the stream of solid fuel particles and carrier fission gas stream may interact with the surrounding liquid medium. The proposed framework considers the solid phase as discrete Lagrangian particles while the interpenetrating gas–liquid continuum is modeled using a Eulerian framework. The developed simulation framework is validated with experimental results performed in a separate-effect test facility by comparing predicted particle settlement locations with the experimental observations reported in the literature. Key highlights of this study include the interpretation of the high-pressure boundary conditions and the three-phase flow coupling strategy. Also discussed are the gas–liquid hydrodynamics and the behavior of particles as they are transported through the gas and liquid phases.

The LIFUS5 separate effect test facility experimental programme

The LIFUS5 facility is a separate effect test facility aimed at investigating the heavy liquid metal-water interaction. It has been designed and constructed to withstand high pressure and temperature (i.e. up to 200 bar and up to 500 °C) and to record the fast pressure transients in heavy liquid metal melt. This type of transient is typical of a Steam Generator Tube Rupture event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system, and should be investigated from the safety point of view because it could potentially induce, beyond the damaging of the internal structures (HX tube bundle, above core structures, Fuel Assembly, Control Rods, etc.) several negative effects on the operation of the reactor. These include an insertion of positive reactivity into the system or reduced cooling efficiency due to steam dragging into the core. It will also have an effect on the chemistry control of the cooling. All these effects compromise the safety and the reliability of the system. The twenty years experimental programme and the different LIFUS5 configurations are presented in the paper, as well as the experimental campaigns and test matrix characterizing the heavy liquid metal-water interaction phenomena and data for codes validation.

Verification and validation (V&V) for fast reactor system analysis code FASYS

The fast reactor system analysis code FASYS is developed by CIAE for transient accident analysis in CEFR. By referring to the application scope of FASYS, the basic requirements of FASYS and PIRT in CEFR are established. A comprehensive verification and validation for the accuracy of FASYS is performed, which consist of bottom-up assessment, top-down assessment and uncertainty assessment. The bottom-up assessment applies the static analysis and dynamic analysis to verify the basic function of FASYS. The separate effect tests and integral effect tests are employed in the top-down assessment to validate the availability to simulate model and system. The calculated results confirms that the basic function and model in FASYS are reasonable and credible. The uncertainty assessment results show that FASYS can still calculate IET under the reasonable uncertainty of input parameters, indicating that the FASYS can simulate and analyze the transient accident processes of sodium-cooled fast reactor.

Progress towards the validation of SIMMER-III code model for lead-lithium water chemical interaction

• SIMMER-III code validation for WCLL in-box LOCA assessment • Lead-lithium eutectic and water chemical interaction • Numerical Analysis of Experimental Test E5.2 • Execution of the Series E experimental campaign employing the facility, LIFUS5/Mod3 Research activities are continuing between the University of Pisa and ENEA Brasimone Research Center to better understand the phenomena and processes that occur during a postulated in-box LOCA in the Water-Cooled lead-lithium Breeding Blanket (WCLL BB) and during the system's safety response. Various activities are also going on to strengthen the reliability of numerical tools and validate computer models, codes, and procedures for their implementations. The study presented here aims to assist in the validation process of the SIMMER-III code model, to simulate the lead-lithium and water interaction under conditions like those expected for the WCLL BB during accident conditions. In particular, the current work has been performed by numerically reproducing the experimental Test E5.2, executed in the separate effect test facility LIFUS5/Mod3 (installed at ENEA Brasimone Research Center).The code used for the simulation was SIMMER-III computational code, in a version modified by the University of Pisa in order to take into account also the chemical interaction between the two fluids. A comparison of significant parameters computed by the SIMMER code with those obtained in the experimental test during the transient has been carried out and it is presented and described in this paper.