Modelling and Uncertainty Quantification application to SA simulation codes in advanced SMR

In the framework of a global transition to a low-carbon energy mix, the interest in advanced nuclear Small Modular Reactors (SMRs) has been growing at the international level. Due to the high level of maturity reached by Severe Accident Codes for currently operating rectors, their applicability to advanced SMRs is starting to be studied. Within the present work of thesis and in the framework of a collaboration between ENEA, UNIBO and IRSN, an ASTEC code model of a generic IRIS reactor has been developed. The simulation of a DBA sequence involving the operation of all the passive safety systems of the generic IRIS has been carried out to investigate the code model capability in the prediction of the thermal-hydraulics characterizing an integral SMR adopting a passive mitigation strategy. The following simulation of 4 BDBAs sequences explores the applicability of Severe Accident Codes to advance SMRs in beyond-design and core-degradation conditions. The uncertainty affecting a code simulation can be estimated by using the method of Input Uncertainty Propagation, whose application has been realized through the RAVEN-ASTEC coupling and implementation on an HPC platform. This probabilistic methodology has been employed in a study of the uncertainty affecting the passive safety system operation in the DBA simulation of ASTEC, providing a further characterization of the thermal-hydraulics of this sequence. The application of the Uncertainty Quantification method to early core-melt phenomena has been investigated in the framework of a BEPU analysis of the ASTEC simulation of the QUENCH test-6 experiment. A possible solution to the encountered challenges has been proposed through the application of a Limit Surface search algorithm

Validation and uncertainty analysis of ASTEC in early degradation phase against QUENCH-06 experiment

Two pyrene-tetrazole conjugates were synthesized as photoreactive chromophores that allow for the first time the combination of metabolic labelling of DNA in cells and subsequent bioorthogonal “photoclick” modification triggered by visible light. Two strained alkenes and three alkene-modified nucleosides were used as reactive counterparts and revealed no major differences in their “photoclick” reactivity. This is a significant advantage because it allows 5-vinyl-2′-deoxyuridine to be applied as the smallest possible alkene-modified nucleoside for metabolic labelling of DNA in cells. Both pyrene-tetrazole conjugates show fluorogenicity during the “photoclick” reactions, which is a second advantage for cellular imaging. Living HeLa cells were incubated with 5-vinyl-2′-deoxyuridine for 48 h to ensure one cell division. After fixation, the newly synthesized genomic DNA was successfully labelled by irradiation with visible light at 405 nm and 450 nm. This method is an attractive tool for the visualization of genomic DNA in cells with full spatiotemporal control by the use of visible light as a reaction trigger

SBO analysis of a generic PWR-900 with ASTEC and MELCOR codes

Abstract After the Fukushima accident, the interest of the public to nuclear safety has growth and the international technical nuclear community has increased his attention in the investigation and the characterization of Severe Accident (SA) scenarios. In order to simulate the different, complex and multi-physical phenomena involved in a SA, computational tools, known as SA codes, have been developed in the last decades. In order to give some insights on the modelling capabilities of these tools and the differences in the calculation results, also related to the user-effect, an analysis of an unmitigated Station Black Out (SBO) occurring in a generic Western three-loops PWR 900 MWe has been carried out by the authors in the framework of the NUGENIA TA-2 ASCOM project. The simulation results of ASTEC code (study carried out with ASTEC V2, IRSN all rights reserved, [2019]), developed by IRSN, and MELCOR 2.2 code, developed by SANDIA for USNRC, have been compared and analyzed. The SBO scenario considered takes into account the intervention of the accumulators as only accident mitigation strategy. Several figures of merits related to the thermal-hydraulic (e.g. primary pressure, cladding temperature, etc.) and to the core degradation (e.g. hydrogen production, etc.) have been considered to describe the accident evolution until the vessel failure, for the two codes comparison

Association of kidney disease measures with risk of renal function worsening in patients with type 1 diabetes

Background: Albuminuria has been classically considered a marker of kidney damage progression in diabetic patients and it is routinely assessed to monitor kidney function. However, the role of a mild GFR reduction on the development of stage 653 CKD has been less explored in type 1 diabetes mellitus (T1DM) patients. Aim of the present study was to evaluate the prognostic role of kidney disease measures, namely albuminuria and reduced GFR, on the development of stage 653 CKD in a large cohort of patients affected by T1DM. Methods: A total of 4284 patients affected by T1DM followed-up at 76 diabetes centers participating to the Italian Association of Clinical Diabetologists (Associazione Medici Diabetologi, AMD) initiative constitutes the study population. Urinary albumin excretion (ACR) and estimated GFR (eGFR) were retrieved and analyzed. The incidence of stage 653 CKD (eGFR < 60 mL/min/1.73 m2) or eGFR reduction > 30% from baseline was evaluated. Results: The mean estimated GFR was 98 \ub1 17 mL/min/1.73m2 and the proportion of patients with albuminuria was 15.3% (n = 654) at baseline. About 8% (n = 337) of patients developed one of the two renal endpoints during the 4-year follow-up period. Age, albuminuria (micro or macro) and baseline eGFR < 90 ml/min/m2 were independent risk factors for stage 653 CKD and renal function worsening. When compared to patients with eGFR > 90 ml/min/1.73m2 and normoalbuminuria, those with albuminuria at baseline had a 1.69 greater risk of reaching stage 3 CKD, while patients with mild eGFR reduction (i.e. eGFR between 90 and 60 mL/min/1.73 m2) show a 3.81 greater risk that rose to 8.24 for those patients with albuminuria and mild eGFR reduction at baseline. Conclusions: Albuminuria and eGFR reduction represent independent risk factors for incident stage 653 CKD in T1DM patients. The simultaneous occurrence of reduced eGFR and albuminuria have a synergistic effect on renal function worsening

BEPU analysis of a 2-in DVI break in a generic IRIS SMR by ASTEC -RAVEN coupling

International audienceAdvanced light-water Small Modular Reactors (SMRs) have acquired great interest in the international framework due to recognized advantages in terms of flexibility, capital cost and especially safety features, achieved by their designs. The "inherent safety" of SMRs is guaranteed by lower decay-heat, allowing the use of integral configuration for the primary coolant system and of passive safety systems. In the framework of NUGENIA TA-2 ASCOM collaborative project, coordinated by IRSN, an ASTEC code generic input-deck, based on an IRIS-like reactor, has been recently developed with the aim of studying the code capability to simulate SMR designs in challenging conditions. In the present paper, a Best-Estimate Plus Uncertainty (BEPU) method has been applied to study the safety criteria in a Design Basis Accident (DBA) due to a 2-in guillotine break of a Direct Vessel Injection (DVI) line and by assuming the operability of the passive safety systems of the generic IRIS. The uncertainty quantification study is performed through the propagation of the input uncertainty methodology, by implementing the RAVEN-ASTEC coupling on a multi-core cluster. The input uncertainty parameters perturbing the system are selected among the main reactor's initial and boundary conditions as well as related to passive safety systems. The statistical study of the reactor response in terms of output variation of the main safety Figures Of Merit (FOMs) is carried out by analyzing the sensitivity of the FOMs, with respect to the variation of the input uncertainties. The study is aimed to provide information regarding the role played by passive safety systems in the mitigation strategy; to characterize the thermal-hydraulic response of the code model and its capability to simulate the main natural-driven phenomena of passive advanced SMRs; and to develop a first uncertainty analysis regarding ASTEC application to SMR

Design of a Prototypical Mock-Up for the Experimental Investigation of WCLL First-Wall Performances

A large research effort is currently ongoing within the framework of the EUROfusion consortium for the study and design of a water-cooled lithium–lead (WCLL) breeding blanket (BB). This concept will be tested in ITER through the installation of a test blanket module (TBM) and it is one of the two candidates adopted as driver BBs in DEMO. In this framework, at the ENEA research centre of Brasimone, the realization of the experimental platform, W-HYDRA, is envisaged. The platform is dedicated to the support of the development of WCLL BB and ITER TBM and the investigation of the DEMO balance of plants. One of the most important experimental infrastructures is the water-loop facility, the aim of which is to provide water at a high pressure and temperature (PWR conditions), with a sufficient mass-flow rate and power for the experimental testing of BB and TBM components. The facility will be equipped with a vacuum chamber and an electron beam gun for the reproduction of high surface heat flux on plasma-facing components. In the present work, the design of a prototypical mock-up (MU) of the WCLL BB first wall is described. The MU is used to investigate the thermal, hydraulic and structural behavior of the current first-wall design under relevant heat loads at the expected operational conditions. The delineation of the main experimental test’s features and the instrumentation needed is assessed in the paper. A preliminary CFD calculation on the prototypical MU and the computational results are also presented

Design and Integration of the EU-DEMO Water-Cooled Lead Lithium Breeding Blanket

The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have allowed a quite advanced reactor architecture to be achieved. Moreover, significant efforts have been made in order to develop the WCLL BB pre-conceptual design following a holistic approach, identifying interfaces between components and systems while respecting a system engineering approach. This paper reports a description of the current WCLL BB architecture, focusing on the latest modifications in the BB reference layout aimed at evolving the design from its pre-conceptual version into a robust conceptual layout. In particular, the main rationale behind design choices and the BB’s overall performances are highlighted. The present paper also gives an overview of the integration between the BB and the different in-vessel systems interacting with it. In particular, interfaces with the tritium extraction and removal (TER) system and the primary heat transfer system (PHTS) are described. Attention is also paid to auxiliary systems devoted to heat the plasma, such as electron cyclotron heating (ECH). Indeed, the integration of this system in the BB will strongly impact the segment design since it envisages the introduction of significant cut-outs in the BB layout. A preliminary CAD model of the central outboard blanket (COB) segment housing the ECH cut-out has been set up and is reported in this paper. The chosen modeling strategy, adopted loads and boundary conditions, as well as obtained results, are reported in the paper and critically discussed

Uncertainty quantification for a severe accident sequence in a SFP in the frame of the H-2020 project MUSA: First outcomes

International audienceThe Management and Uncertainties of Severe Accidents (MUSA) project, funded in HORIZON 2020 and coordinated by CIEMAT (Spain), aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) focusing on Source Term (ST). In this framework, the objectives of the Innovative Management of Spent Fuel Pool Accidents (IMSFP – WP6), led by IRSN (France), are to quantify and rank the uncertainties affecting accident analyses in a Spent Fuel Pool (SFP), to review existing and contemplated SA management measures and systems and to assess their possible benefits in terms of reduction of radiological onsequences.To quantify the propagation of the uncertainties of the input parameters to the output uncertainties of severe accident codes (ASTEC, MELCOR, RELAP/SCDAP), a diverse set of uncertainty quantification (UQ) tools (DAKOTA, RAVEN, SUNSET, SUSA) are used. The statistical framework used by the different UQ-tools is similar e.g. pure random (Monte Carlo) and Latin hypercube sampling (LHS).Fourteen partners from three different world regions are involved in the WP6 activities. The target of this paper is to describe the achievements during the first three years ofthe project. In a first part, a description is given of the SFP accidental scenario, of the key target variables and radionuclides chosen as ST Figure of Merits (FoM) and of theidentified uncertainty sources in models and input parameters. A key element when defining the SFP scenario has been the consideration (or not) of the reactor building,as it is expected to significantly affect analyses. In a second part, the first insights coming out from the calculation phase of the project are presented. The review of existing SA management measures is also exposed, as well as systems whose benefits will be assessed in the second phase of the project. Finally, challenges that arise from such an exercise are discussed, as well as major difficulties found when applying UQ methodologies to SFP scenarios and solutions adopted