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Nowadays, a great deal of attention is devoted to the development of best-estimate models able to produce more realistic outcomes. This is also the case for system codes, such as MELCOR, that are being mostly used in a conservative way especially when dealing with the licensing process. The above-mentioned need for more realistic results is at the core of this two-paper series related to the creation of a more accurate MELCOR model for the HI-STORM 100S dry cask. The findings obtained from the sensitivity studies carried out in the Part I are leveraged to set up an improved MELCOR model, the characteristics of which are consistent with the typical features of Spent Nuclear Fuel (SNF), and with geometrical and material properties of the cask itself. The addition of an axial power profile in the Fuel Assembly (FA), the better characterization of the flow losses in the air gap between internal metallic canister and external concrete-based overpack, and the choice of an appropriate value for the concrete thermal conductivity, are taken into account conjointly in this Part II. The outcomes from the improved MELCOR simulation are reported mainly in terms of the Peak Cladding Temperature (PCT), being the variable under regulatory surveillance. However, in addition to PCT, calculated temperature profiles are displayed and compared against the ones resulting from the previous model.

This paper presents the results of a severe accident parametric sensitivity study performed on a model of a generic integral pressurized water reactor (iPWR). The analyzed sequence is a loss of coolant accident (LOCA)-type scenario, postulating that safety systems are not available. In this work, a MELCOR 2.2 input deck of a generic iPWR was developed based on publicly available data. The iPWR used in this work is a generic iPWR with a thermal power of about 160 MWth, characterized by a compact steam generator and a submerged containment configuration. A hypothetical scenario considered in this work was an unmitigated small break LOCA leading to a severe accident with partial core degradation, due to the postulated assumptions. In the presented paper, 16 sensitivity cases were calculated and analyzed, focusing mainly on heat transfer, decay heat, and core degradation parameters. The selected parameters and their combination caused partial core degradation and showed significant differences in investigated variables such as core degradation and hydrogen generation, as well as examined CPU time consumption. This a preparatory work performed in the framework of the Horizon Euratom SASPAM-SA project, which aims to investigate the applicability and transfer of the operating large light-water reactor knowledge and know-how to the iPWR, taking into account European licensing analysis needs for the severe accident and emergency planning zone.

In this study, uncertainty and sensitivity analyses were performed with MELCOR 2.2.18 to study the hydrogen generation (figure-of-merit (FoM)) during the in-vessel phase of a severe accident in a light water reactor. The focus of this work was laid on a large generation-III pressurized water reactor (PWR) and a double-ended hot leg (HL) large break loss of coolant accident (LB-LOCA) without a safety injection (SI). The FPT-1 Phebus integral experiment emulating LOCA was studied, where the experiment outcomes were applied for the plant scale modelling. The best estimate calculations were supplemented with an uncertainty analysis (UA) based on 400 input-decks and Latin hypercube sampling (LHS). Additionally, the sensitivity analysis (SA) utilizing the linear regression and linear and rank correlation coefficients was performed. The study was prepared with a new open-source MELCOR sensitivity and uncertainty tool (MelSUA), which was supplemented with this work. The FPT-1 best-estimate model results were within the 10% experimental uncertainty band for the final FoM. It was shown that the hydrogen generation uncertainties in PWR were similar to the FPT-1, with the 95% percentile being covered inside a ~50% band and the 50% percentile inside a ~25% band around the FoM median. Two different power profiles for PWR were compared, indicating its impact on the uncertainty but also on the sensitivity results. Despite a similar setup, different uncertainty parameters impacted FoM, showing the difference between scales but also a significant impact of boundary conditions on the sensitivity analysis.

This work presents a thermohydraulic analysis of a postulated accident involving the rupture of the breeder primary cooling loop inside a heat exchanger (once through steam generator). After the detection of the loss of pressure inside the primary loop, a plasma shutdown is actuated with a consequent plasma disruption, isolation of the secondary loop, and shutoff of the pumps in the primary; no other safety counteractions are postulated. The objective of the work is to analyze the pressurization of the primary and secondary sides to show that the accidental overpressure in the two sides of the steam generators is safely accommodated. Furthermore, the effect of the plasma disruption on the FW, in terms of temperatures, should be analyzed. Lastly, the time transients of the pressures and temperatures in the HX and BB for a time span of up to 36 h should be obtained to assess the effect of the decay heat over a long period. A full nodalization of the OTSG was realized together with a simplified nodalization of the whole PHTS BB loop. The code utilized was MELCOR for fusion version 1.8.6. The accident was simulated by activating a flow path which directly connected one section of the primary with the parallel section of the secondary side. It is shown here that the pressures and the temperatures inside the whole PHTS system remain below the safety thresholds for the whole transient.

The United States (US) Department Of Energy (DOE) has addressed the thermal analysis of the Spent Nuclear Fuel (SNF) stored within a dry cask system as a matter of high priority. In this regard, it is of utmost importance that simulation tools effectively reproduce the general thermal behavior of the modelled cask, including heat exchange and removal. Temperature distribution in the different components of the system is usually the focus of performed thermal analyses. In particular, attention is paid to the maximum temperature reached in the fuel cladding, namely the Peak Cladding Temperature (PCT). Within this framework, the present paper is the first of a two-paper series aimed at developing a more accurate model for the HI-STORM 100S cask. The dry cask in question is modelled and its behavior is simulated by means of the MELCOR code (version 2.2.18019). Stressing the need for a more realistic model rather than a conservative one, this paper reports the efforts undertaken to evaluate the influence of some specific modelling choices on the PCT. The study of the cask performance is therefore conducted taking into consideration three main factors: the axial power distribution in the Fuel Assembly (FA), the flow losses in the air gap between the internal canister and the external overpack, and the conductivity of the overpack concrete.

As part of the pre-conceptual design activities for the European DEMOnstration plant, a carefully selected set of safety analyses have been performed to assess plant integrated performance and the capability to achieve expected targets while keeping it in a safe operation domain. The DEMO divertor is the in-vessel component in charge of exhausting the major part of the plasma ions’ thermal power in a region far from the plasma core to control plasma pollution. The divertor system accomplishes this goal by means of assemblies of cassette and target plasma facing units modules, respectively cooled with two independentheat-transfer systems. A deterministic assessment of a divertor in-vessel Loss-of-Coolant Accident is here considered. Both Design Basis Accident case simulating the rupture of an in-vessel pipe for the divertor cassette cooling loop, and a Design Extension Conditions accident case considering the additional rupture of an independent divertor target cooling loop are assessed. The plant response to such accidents is investigated, a comparison of the transient evolution in the two cases is provided, and design robustness with respect to safety objectives is discussed.

In the WCLL-BB liquid blanket concept, among the postulated accidental scenarios to be investigated regarding the PbLi system, a loss of liquid metal is one of the most crucial for safety purposes. As a consequence of this loss, chemical reactions might occur between lithium, air, and water. These reactions may lead to significant increases in temperature and pressure and to the formation of hydrogen inside the building; the evaluation of the impact of these events is essential to define future safety guidelines. Using MELCOR for fusion (v.1.8.6.), a first approach for investigating an out-vessel loss of PbLi was presented in this work. The temperature and pressure trends were investigated through the default MELCOR package for lithium air chemical reactions; however, this package works only with pure lithium. Therefore, a new model of the reaction was developed in this work using MELCOR Control Functions, which allows the simulation of the chemical reactions using PbLi as working fluid, in order to investigate a more realistic scenario. The results from different approaches were compared, and the limits of the code were outlined. It was found that the final pressure and temperature in the tokamak building might reach critical values at the end of the transient, if the energy released by the chemical reaction was entirely considered; however, it is important to note that, due to the assumptions and simplifications adopted, the results were very conservative in terms of temperature and pressure reached in the system, and further investigations were suggested in this work.

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO.

A Demo-Oriented early NEutron Source (DONES) facility for material irradiation with nuclear is currently being designed. DONES aims to produce neutrons with fusion-relevant spectrum and fluence by means of D–Li stripping reactions occurring between a deuteron beam impacting a stable liquid lithium flowing film implementing the target. Given the hazard constituted by the liquid lithium inventory and the potential risk of reactions with water, air, and concrete eventually resulting in fire events, the Target Test Cell (TTC) is filled with helium and the reinforced concrete walls forming the bio-shield are covered with steel liners. A loss of Li in TTC, due to a large break in the Quench Tank, is postulated, and consequences are deterministically studied. With the TTC liner being water-cooled, the impact of the liner temperature rise following a leakage event is evaluated. Two separate MELCOR code models have been defined for the liquid lithium loop and water-cooled loop and are numerically coupled. The amount of leaked inventory dependent on the implemented safety logic and impact on TTC containment is evaluated. The water pressurization pattern within the liner cooling loop is studied to highlight possible risks of lithium–water/concrete reactions.

LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by forced circulation. In 2011, the fuel was replaced, going from high-enriched uranium (HEU) to low-enriched uranium (LEU). After 2017, the State Office for Nuclear Safety (SUJB) asked CVR to evaluate the LVR-15 under Design Extended Conditions B (DEC-B). For this reason, a new model was developed in the MELCOR code, which allows for modelling the progression of a severe accident (SA) in light-water nuclear power plants and estimating the behaviour of the reactor under SA conditions. The model was built by collecting information about the LVR-15. Since the research reactor can have different core configurations according to the location of the core components, the core configuration with the most fuel (hottest campaign K221) was selected. Then, to create the radial nodalisation, the details of the core components were obtained and grouped in five radial rings and 27 axial levels. The simulation was run with the boundary conditions collected from campaign K221, and the results were compared with the reference values of the campaign with a negligible percentage of error. For the coolant inlet and outlet temperature, the reference values were 318.18 K and 323.5 K, respectively, while for the simulation, the steady state reached 319 K for the inlet temperature and 324 K for the outlet temperature. Additionally, the cladding temperature of the hottest assembly was compared with the reference value (353.72 K) and the steady-state simulation results (362 K). In future work, different transients leading to severe accidents will be simulated. When simulating the LVR-15 reactor with MELCOR, specific attention is required for the aluminium-cladded fuel assemblies, as the model requires some assumptions to cope with the phenomenological limitations.