Pool temperature stratification analysis in CIRCE-ICE facility with RELAP5-3D© model and comparison with experimental tests

In the frame of heavy liquid metal (HLM) technology development, CIRCE pool facility at ENEA/Brasimone Research Center was updated by installing ICE (Integral Circulation Experiments) test section which simulates the thermal behavior of a primary system in a HLM cooled pool reactor. The experimental campaign led to the characterization of mixed convection and thermal stratification in a HLM pool in safety relevant conditions and to the distribution of experimental data for the validation of CFD and system codes. For this purpose, several thermocouples were installed into the pool using 4 vertical supports in different circumferential position for a total of 119 thermocouples [1][2]. The aim of this work is to investigate the capability of the system code RELAP5-3D (c) to simulate mixed convection and thermal stratification phenomena in a HLM pool in steady state conditions by comparing code results with experimental data. The pool has been simulated by a 3D component divided into 1728 volumes, 119 of which are centered in the exact position of the thermocouples. Three dimensional model of the pool is completed with a mono-dimensional nodalization of the primary main flow path. The results obtained by code simulations are compared with a steady state condition carried out in the experimental campaign. Results of axial, radial and azimuthal temperature profile into the pool are in agreement with the available experimental data Furthermore the code is able to well simulate operating conditions into the main flow path of the test section

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermofluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation

Post-test simulation of a PLOFA transient test in the CIRCE-HERO facility

CIRCE is a lead–bismuth eutectic alloy (LBE) pool facility aimed to simulate the primary system of a heavy liquid metal (HLM) cooled pool-type fast reactor. The experimental facility was implemented with a new test section, called HERO (Heavy liquid mEtal pRessurized water cOoled tubes), which consists of a steam generator composed of seven double-wall bayonet tubes (DWBT) with an active length of six meters. The experimental campaign aims to investigate HERO behavior, which is representative of the tubes that will compose ALFRED SG. In the framework of the Horizon 2020 SESAME project, a transient test was selected for the realization of a validation benchmark. The test consists of a protected loss of flow accident (PLOFA) simulating the shutdown of primary pumps, the reactor scram and the activation of the DHR system. A RELAP5-3D© nodalization scheme was developed in the pre-test phase at DIAEE of “Sapienza” University of Rome, providing useful information to the experimentalists. The model consisted to a mono-dimensional scheme of the primary flow path and the SG secondary side, and a multi-dimensional component simulating the large LBE pool. The analysis of experimental data, provided by ENEA, has suggested to improve the thermal–hydraulic model with a more detailed nodalization scheme of the secondary loop, looking to reproduce the asymmetries observed on the DWBTs operation. The paper summarizes the post-test activity performed in the frame of the H2020 SESAME project as a contribution of the benchmark activity, highlighting a global agreement between simulations and experiment for all the primary circuit physical quantities monitored. Then, the attention is focused on the secondary system operation, where uncertainties related to the boundary conditions affect the computational results

Investigation on RELAP5-3D© capability to predict thermal stratification in liquid metal pool-type system and comparison with experimental data

A numerical activity, aimed to evaluate the capability of RELAP5-3D© to reproduce the main thermal-hydraulic phenomena in an HLM pool-type facility, in different operative conditions, is presented. For this purpose, the experimental campaign performed in CIRCE-ICE test facility has been selected for the code assessment. Two experimental tests have been analyzed: TEST A consisting in a transition from no-power to a full power steady state conditions, and TEST I, consisting in a transition from gas-enhanced circulation to natural circulation, simulating a protected loss of heat sink plus a loss of flow accident. Three different pool modelling approaches are presented, consisting in a single vertical pipe, parallel pipes with cross junctions and multi-dimensional component. The comparison with experimental data has highlighted the need to divide the large pool in several sections to reproduce the natural convection, strictly correlated with the thermal stratification. The multi-dimensional component seems to be the best practice for the evaluation of this phenomenon even if the lack of specific correlation for heat transfer coefficient in quasi-stagnant conditions in large tanks is a limit for the accuracy of the results. In addition, the paper presents a detailed nodalization of the fuel pin bundle, highlighting quite good capabilities of RELAP5-3D as a subchannel analysis code

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

System thermal-hydraulic modelling of the phénix dissymmetric test benchmark

Phénix is a French pool-type sodium-cooled prototype reactor; before the definitive shutdown, occurred in 2009, a final set of experimental tests are carried out in order to increase the knowledge on the operation and the safety aspect of the pool-type liquid metal-cooled reactors. One of the experiments was the Dissymmetric End-of-Life Test which was selected for the validation benchmark activity in the frame of SESAME project. The computer code validation plays a key role in the safety assessment of the innovative nuclear reactors and the Phénix dissymmetric test provides useful experimental data to verify the computer codes capability in the asymmetric thermal-hydraulic behaviour into a pool-type liquid metal-cooled reactor. This paper shows the comparison of the outcomes obtained with six different System Thermal-Hydraulic (STH) codes: RELAP5-3D©, SPECTRA, ATHLET, SAS4A/SASSYS-1, ASTEC-Na and CATHARE. The nodalization scheme of the reactor was individually achieved by the participants; during the development of the thermal-hydraulic model, the pool nodalization methodology had a special attention in order to investigate the capability of the STH codes to reproduce the dissymmetric effects which occur in each loop and into pools, caused by the azimuthal asymmetry of the boundary conditions. The modelling methodology of the participants is discussed and the main results are compared in this paper to obtain useful guide lines for the future modelling of innovative liquid metal pool-type reactors

Real-life effectiveness and safety of guselkumab in patients with psoriasis who have an inadequate response to ustekinumab: A 104-week multicenter retrospective study – IL PSO (ITALIAN LANDSCAPE PSORIASIS)

BackgroundGuselkumab is a humanized monoclonal antibody that binds selectively to the p19 subunit of interleukin-23, which has shown efficacy in patients with previous incomplete response to ustekinumab in the NAVIGATE clinical trial. ObjectivesWe conducted a 104-week multicenter retrospective study to assess the effectiveness and safety of guselkumab in patients affected by plaque psoriasis with an inadequate response to ustekinumab in a real-life setting. MethodsOur retrospective study included 233 adults affected by moderate-to-severe plaque psoriasis, enrolled in 14 different Italian centres, and treated with guselkumab after failing therapy with ustekinumab. Patient characteristics and PASI (Psoriasis Area and Severity Index) score at each visit (baseline, weeks 16, 52 and 104) were recorded. The percentages of patients achieving 75%, 90% and 100% (PASI 75, PASI 90 and PASI 100) improvement in PASI, compared with baseline, were registered. ResultsAt week 52, PASI 75 was reached by 89.88% of patients, PASI 90 by 71.43%, PASI 100 by 58.83% and absolute PASI <= 2 by 90.48%. At week 104, similar effectiveness results were observed. Compared to the NAVIGATE trial, we observed higher rates of PASI 75/90/100. Patients with the involvement of difficult-to-treat areas were significantly less likely to achieve PASI90 and PASI100 at week 16. Obese patients had significantly lower rates of PASI75 and PASI <= 2 at week 52. At week 104, comparable responses were observed among all patients' subgroups, regardless of BMI status, involvement of difficult-to-treat areas, presence of cardiometabolic comorbidities and concomitant psoriatic arthritis. No significant safety findings were reported throughout the study. ConclusionOur data suggest that the efficacy of guselkumab in patients with inadequate response to ustekinumab for plaque psoriasis in 'real-life' clinical practice is comparable with NAVIGATE study with higher percentages of patients achieving PASI90 and PASI100 at weeks 16, 52 and 104

Tokamak cooling systems and power conversion system options

DEMO will be a fusion power plant demonstrating the integration into the grid architecture of an electric utility grid. The design of the power conversion chain is of particular importance, as it must adequately account for the specifics of nuclear fusion on the generation side and ensure compatibility with the electric utility grid at all times. One of the special challenges the foreseen pulsed operation, which affects the operation of the entire heat transport chain. This requires a time-dependant analysis of different concept design approaches to ensure proof of reliable operation and efficiency to obtain nuclear licensing. Several architectures of Balance of Plant were conceived and developed during the DEMO Pre-Concept Design Phase in order to suit needs and constraints of the in-vessel systems, with particular regard to the different blanket concepts. At this early design stage, emphasis was given to the achievement of robust solutions for all essential Balance of Plant systems, which have chiefly to ensure feasible and flexible operation modes during the main DEMO operating phases – Pulse, Dwell and ramp-up/down – and to adsorb and compensate for potential fusion power fluctuations during plasma flat-top. Although some criticalities, requiring further design improvements were identified, these preliminary assessments showed that the investigated cooling system architectures have the capability to restore nominal conditions after any of the abovementioned cases and that the overall availability could meet the DEMO top-level requirements. This paper describes the results of the studies on the tokamak coolant and Power Conversion System (PCS) options and critically highlights the aspects that require further work