• Energy Research

Abstract During the last decades, the safety analysis of nuclear power plants, have been shifting from conservative models and hypotheses to Best Estimate. Within this trend, the different safety analyses can be categorized depending on their associated conservatisms. The approach that includes Expanded Event Trees together with Best Estimate model and conditions plus uncertainty avoiding conservatisms in system availability can be referred as Extended BEPU (EBEPU). In this sense, an Extended BEPU safety analysis relies on Expanded Event Trees and Best Estimate models to study an accidental sequence. In the present paper, an EBEPU study is performed, using a loss of coolant accident in a pressurized water reactor modeled with the TRACE code. A 5% core power uprate is used as an example of safety margin EBEPU analysis. The paper presents parametric and non-parametric uncertainty approaches. Observing the results from all event tree sequences, it is seen that the branches that contribute the most to the Core Damage frequency are successful branches, with all safety systems available, and some Core Damage branches have a possibility of success.

Abstract AP1000® Generation III+ reactor bases its safety concept on passive systems, differently from the previous Generation II reactors. This fact has led the approximations and methodologies previously used for modeling active safety systems to be reviewed and adapted to simulate the physics of passive systems. Diverse studies about the AP1000 containment have demonstrated the difficulty to correctly model the occurring phenomenology. In this paper, an integral AP1000 3D containment GOTHIC model is presented, including the Passive Containment Cooling System (PCCS). The model includes the compartments inside and outside the metallic containment liner that influence the thermal–hydraulic behavior. The model is tested against a Large Break Loss of Coolant Accident (LBLOCA) to assess its thermal–hydraulic performance, assuming a PCS tank malfunction, what is a conservative hypothesis. The pressure and temperature evolution predicted by the 3D containment model is analyzed and compared with a single node Lumped Parameters model, allowing to evaluate some preliminary benefits of 3D modeling for containment safety analysis. The 3D containment model allows to predict the thermal evolution in each containment compartment capturing the heterogeneity of this phenomenon, with higher resolution than the lumped parameters models traditionally used in this kind of analyses. It allows to observe the thermohydraulic conditions locally at any time during the transient.

One of the main goals of severe accident management strategies is to maintain containment integrity to prevent radioactive release into the environment. As a result, containment internal loads need to be investigated during a postulated accident. In the present study, short-term containment thermal-hydraulic (TH) parameters of a VVER-1000/V446 reactor are analysed during a LB-LOCA. To achieve this goal, in the first step as-built 3D structure of VVER-1000/V446 containment was modelled in detail by using AutoCAD. The AutoCAD model has been processed to be prepared for GOTHIC 3D input. Meanwhile, an equivalent GOTHIC lumped parameter (LP) model is also prepared to validate the modelling procedure and results against the FSAR. Finally, LP profiles and 3D TH contour results of GOTHIC code were presented and discussed. LP results show a close agreement with FSAR reference and can approve the accuracy of the simulation procedure. 3D contours present all-coordinate detailed TH parameters versus time inside the containment. Spatial distribution of TH parameters and minor short-term effects of spray as Engineering Safety Features (ESFs) to tackle the containment pressurization can be evaluated by employing these contours. 3D simulation results can provide advantages for the precise locating and installation of ESFs in design and operation to achieve the highest efficiency in case of containment accident.

Hydrogen is a flammable gas that can generate thermal and mechanical loads which could jeopardise the containment integrity upon combustion inside nuclear power plants containment. Hydrogen can be generated from various sources and disperses into the containment atmosphere, mixing with steam and air following a loss of coolant accident and its progression. Therefore, the volumetric hydrogen concentration should be examined within the containment to determine whether a flammable mixture is formed or not. Codes with 3D capabilities could serve this examination by providing detailed contours/maps of the hydrogen distribution inside containment in view of the local stratification phenomenon. In this study, a 3D VVER-1000 as-built containment model was sketched in AutoCAD and then processed into GOTHIC nuclear containment analysis code for hydrogen evaluation. The model was modified to a great extent by installing 80 passive autocatalytic recombiners and locating hydrogen sources to evaluate the performance of the hydrogen removal system inside the containment on maintaining the hydrogen concentration below the flammability limit during a large break loss of coolant accident. 2D profiles and 3D contours of volumetric hydrogen concentration with and without PARs are presented as the simulation outcome of this study. The results were validated against the results of the Final Safety Analysis Report, which also demonstrates the effectiveness of the hydrogen removal system as an engineered safety feature to keep the containment within a safe margin. Detailed 3D contours of hydrogen distribution inside containment can be employed to evaluate the local hot spots of hydrogen, rearranging and optimising the number and location of PARs to avoid the hydrogen explosion inside containment.

The current trends in development and installation of solar and wind energy poses certain challenges over the stability of the future electricity grids. New regulations or energy storage systems will be essential to enable this large renewable penetration. In the present paper, a zero-emissions scenario composed solely of nuclear energy and renewables is proposed. The article analyzes the potential for nuclear power plants to perform the demand regulation function by coupling them with a crypto-asset mining facility, which provides a fast adaptation to the load-following requirements. The plant, operated at full power, will divert part of its energy to the mining facility when the electricity grid is saturated with renewable energy. An integral economic assessment of the coupled system is performed. To account for the uncertainties involved, different scenarios are assessed, varying the cryptocurrency price, network hashrate, hardware capabilities and lifespan. A profitability analysis is made comparing the benefits of both mining and selling energy to the grid, to the cost of producing that energy to find out if new-built plants would be positively balanced. The results show that only under circumstances of very high cryptoasset price and low network hashrate the cryptomining project would be profitable.

Anticipated Transient Without Scram (ATWS) sequences belong to the beyond design basis events. Within ATWS sequences, the most limiting for PWRs is the Loss Of Normal Feedwater (LONF). This sequence produces a pressure peak in the reactor cooling system that can compromise its integrity. To cope with this sequence, systems like the AMSAC were installed in PWR Westinghouse reactors, helping to manage the transient. However, the consequences of this sequence are still strongly dependent on several variables such as the systems availability, reactor coolant system characteristics and the human actions performance. In order to study this sequence, a sensitivity analysis focused on the most influential parameters is performed. The parameter selection for these sensitivities is based on the literature found, and the analysis is made using a 3-Loop PWR-W TRACE model. Moreover, a proposal of a Phenomena Identification and Ranking Table (PIRT) for the ATWS-LONF sequence is developed. The sequence is analyzed for two different time phases (before and after the pressure peak), focusing on 6 different systems and assessing 20 phenomena. In order to rank the importance of the different phenomena, a review of the experiments, proceduralized human actions, PSA and DSA is performed, with the additional information obtained in the sensitivity analysis. The results show that the important phenomena are those related to reactivity feedback, the energy stored in the core, the AMSAC actuation, and the mass and energy release through pressurizer safety and relief valves.

The Best Estimate Plus Uncertainty (BEPU) approach is being used worldwide for nuclear power plants licensing. This method relies on the use of best estimate models to simulate the accident, evaluating the uncertainties involved. To assess these uncertainties, several methodologies have been developed, such as the non-parametric Wilks/Wald method, parametric methods that reconstruct a distribution from the data, or the binomial approach which interprets the results as two states. Additionally, sensitivity analyses can be performed to obtain the correlation of the output-inputs. Finally, a variability analysis of the most influential parameters made to find a combination of parameters that can lead to damage is also useful. In this paper, all previous techniques are described and applied, by performing a large Monte-Carlo set of simulations of a loss of coolant accident in a pressurized water reactor assessing two figures of merit. The comparison of the different methods show that the most conservative is the Wilks/Wald method; the least conservative is the parametric approach, and in between, the binomial one. The impact of the sample size is also studied for all methods, showing different behaviors for the different approaches

Containment safety analysis is a field in expansion given the complex phenomenology developed during an accident, and the recent code capabilities improvement. In this paper, a modeling strategy to create complex containments with porous CFDs in a single control volume is presented. This methodology involves using solely Computer Aided Design (CAD) tools, to manage the geometrical model, instead of using the GOTHIC code pre-processor interface. This methodology is applied to Trillo NPP (KWU 3 loops) where a LBLOCA is simulated to illustrate its capabilities. The results are compared against a Lumped Parameters model obtaining similar average values. Finally, a sensitivity analysis is made to quantify the influence of the modeling simplifications showing a small dependence. This implies that this modeling strategy is able to recreate models accurately, with an affordable computational cost, and with a time saving improvement, which nowadays is the main bottleneck for 3D containment models.