Activated Corrosion Products (ACPs) may represent a significant source of radiological hazard in nuclear fusion reactors. Corrosion and erosion phenomena mobilize activated materials which are transported by the working fluid in regions of the cooling system accessed by worker personnel. Predicting contaminant transport in tokamak cooling circuits may benefit radiation exposure assessment, design optimization, waste management, maintenance plan definition, and source terms identification. Several calculation tools have been made available for ACPs determination in nuclear systems. Among these, the OSCAR-Fusion code, developed by the CEA (France), allows to predict ACPs generation and transport in closed water-cooled loops for fusion applications. This work aims to show a straightforward sensitivity analysis methodology and uncertainty quantification. A single ACP assessment involves several neutronics, thermal-hydraulics, geometrical, and water chemistry parameters. Coupling OSCAR-Fusion to RAVEN, a multi-purpose framework developed by INL (USA), allows sensitivity and uncertainty quantification analyses that might provide useful indications to the designers and safety analysts. This work presents a general methodology, showing preliminary results obtained for the EU-DEMO divertor cassette primary heat transfer systems.

The present paper deals with the assessment of the original and a modified version of RELAP5/MOD3.3 against the OSU Multi Application Small Light Water Reactor (OSU-MASLWR). The new implemented features regard suitable correlations for the heat transfer coefficient evaluation in helical geometry. Furthermore, two different modelling of the Helical Coil Steam Generator (HCSG) are assessed. In the first approach, HCSG’s primary and secondary sides are collapsed in a single pipe component. In the second model, three equivalent pipes are conceived for the simulation of the three ranks composing the HCSG. Two different power manoeuvring experiments are reproduced. The simulations highlight a satisfactory agreement in both the transients. Nevertheless, the modified code shows enhanced capabilities in the prediction of the HCSG operation. This is due to the improvements adopted in the modified version of RELAP5/MOD3.3 that allows a better modelling of the dryout phenomena occurring within helical tubes, as well as a better estimation of the primary side heat transfer coefficient. The better agreement of the heat exchange is propagated to the primary system, resulting in a more accurate prediction of the inlet and outlet core temperatures, and primary flow rate.

Activated Corrosion Products (ACPs) formation and deposition pose a critical safety issue for nuclear fusion reactors. The working fluid transports the ACPs towards regions accessible to worker personnel. Predicting ACPs formation deposition and transport is fundamental for source term identification, reduction of radiation exposure assessment, maintenance plan definition, design optimization, and waste management. The code OSCAR-Fusion has been developed by the CEA (France) to evaluate the ACPs generation and transport in closed water-cooled loops for fusion application. This work aims at assessing the impact of water chemistry on the transport, precipitation, and deposition of corrosion products for the EU-DEMO divertor Plasma Facing Unit Primary Heat Transfer System. Sensitivity analyses and uncertainty quantification are needed due to the multi-physics phenomena involved in ACPs formation and transport. The OSCAR-Fusion/RAVEN code coupling developed by the Sapienza University of Rome and ENEA has been used. This work presents the perturbation results of different parameters chosen for a closed water-cooled loop considering a continuous scenario of 1888 days. The aim of this work is to assess the variation of build-up of ACPs, perturbing the alkalizing agent concentration into the coolant and the corrosion and release rates of different materials.

The main outcome of the present paper is the feasibility analysis of SIRIO (Sistema di rimozione della Potenza di decadimento per Reattori InnOvativi) facility with conditions based on those of its reference facility. The aim of SIRIO project is to study an innovative Decay Heat Removal System (DHRS) for liquid metal reactor and advanced Light Water Reactor (LWR). Such system must ensure passive control of the power removed from the primary system in abnormal condition, and must ensure reactor cooling in both short and long term. This study present numerical simulations developed with RELAP5/MOD3.3, of two operational procedures: the first one is a steady-state and the second one is a transient phase with decay heat generation. The thermal-hydraulic model, developed with RELAP5/MOD3.3, simulates the whole facility including lines, valves, water and gas tanks, and the Molten Salts (MS) gap. Since there is not experimental data, the present paper is a pre-test study based on SIRO facility design.