Study of the Ageing Effects on the Lower Head Failure in a PWR Reactor

The paper studies influence the ageing effects on the failure of a Reactor Pressure Vessel (RPV) during a severe accident with a core meltdown in a Nuclear Power Plant (NPP). The studied plant is a generic high-power Generation III Pressurized Water Reactor (PWR) developed in the frame of the EU NARSIS project. A Total Station Blackout (SBO) accident was simulated with MELCOR 2.2 severe accident integral computer code. Results of the analysis, temperatures in the lower head and pressures in the lower plenum were used as initial and boundary conditions for the Finite Element Method (FEM) simulations. Two FEM models were developed, a simple two-dimensional axis-symmetric model of the lower head to study fundamental phenomena and complex 3D model to include interactions with the RPV and reactor internals. Ageing effects of a lower head were incorporated into the FEM models to investigate its influence onto lower head response. The ageing phenomena are modelled in terms of degraded mechanical material properties as σ(T), E(T). The primary outcome of the study is the quantitative estimation of the influence of ageing process onto the timing of reactor vessel failure. Presented novel methodology and results can have an impact on future consideration about Long-Term Operation (LTO) of NPPs.

Development on Simulation Method for Two-Phase Flow in Large Diameter Pipes With 90 Degree Elbows

Two-phase flows are observed in various industrial plants and piping systems. Understanding two-phase flow behaviors such as flow patterns and unsteady void fraction in horizontal and vertical pipes are crucial in improving plant safety. Notably, the flow patterns observed in a large diameter pipe (approx. 4–6 in or larger) are significantly different from those observed in a medium diameter pipe. In a vertical large diameter pipe, no slug flow is observed due to the instantaneous slug bubble breakup caused by the surface instability. Besides, in a horizontal pipe, flow regime transition from stratification of liquid and gas to slug (plug) flow that induces unsteady flow should be taken into account. From this viewpoint, it is necessary to predict the flow regime in horizontal and vertical large diameter pipes with some elbows and to evaluate the unsteady flow regime. In this study, the simulation method based on the two-fluid model is developed. The two-fluid model is considered the most accurate model because the governing equations for mass, momentum, and energy transfer are formulated for each phase. When using the two-fluid model, some constitutive equations should be given in computing the momentum transfer between gas and liquid phases. In this study, several state-of-art constitutive equations of the bubble diameter, the interfacial drag force and non-drag forces such as the lift force and the bubble-bubble collision force, are implemented in the platform of ANSYS FLUENT. The developed simulation method is validated with visualization results and force from an air-water flow at the elbow of the piping system.

Development of Design Support System for Piping Route and Differential Pressure Flowmeter by Three-Dimensional Fluid Analysis

In power plants that becoming more compact, it will expend much time and effort to satisfy the requirement for the differential pressure flow measurement according to ISO’s standards. Therefore, it is difficult for engineers in the design phase to completely remove the potential for large errors in flow measurement. This paper presents the 3D fluid analysis system that is a lower cost than the conventional method to confirm the soundness of such measurement in the phase of piping route design. This system has the function to automatically generate the analysis models from general 3D piping CAD data. The analysis program is written by the open source code to reduce a license fee. Also, this system has the function of calculating the swirl strength along the pipe axis as one of the means for efficiently supporting the design change. In order to verify and validate the analysis system, we analyzed several flow paths, confirmed the response of the swirl strength and flow rate indication value of the differential pressure flowmeter model. The analysis result well simulated the increase or decrease swirl strength in the complex flow path, and fluctuation of the flow rate indication value. Also, the system supports to set the flowmeter in the appropriate position by providing visualization of the swirl strength along the pipe axis. In the flow path analysis in this validation, it took about one month to visualization of the swirl strength along the pipe axis from the generation of the analysis models. The 3D fluid analysis system collaborative with 3D piping CAD design system has been developed. This system enable to confirm the effects of swirl strength on flow measurement and the soundness of the differential pressure flow measurement at a lower cost in comparison with conventional method.

Development of a Multiphase Particle Method for Melt-Jet Breakup Behavior of Molten Core in Severe Accident

In a hypothetical severe accident in a light water reactor (LWR) nuclear power plant, there is a possibility that molten core released from the reactor vessel gets in contact with water in the containment vessel. In this so-called fuel-coolant interactions (FCIs) process, the melt jet will breakup into fragments, which is one of the important factors for a steam explosion, as a potential threat to the integrity of the containment vessel. The particle method could directly and easily capture the large deformed interfaces by particle motions, benefiting from its Lagrangian description and meshless framework. In order to investigate the melt-jet breakup with solidification processes, a multiphase particle method with arbitrary high order scheme is presented in this study. In addition, an interfacial particle shifting scheme is developed to suppress the unnatural particle penetration between different phases. The convergence rate with different order is firstly confirmed by a verification test in terms of both explicit and implicit calculations. Then, a transient heat conduction between two materials is carried out and quite good results are obtained. After that, a rising bubble benchmark is performed to show the feasibility of modelling for deformation and collapse. Improvements of clear interface are indicated compared with previous reported results. Two important multiphase instabilities, namely the Rayleigh-Taylor instability and the Kelvin-Helmholtz instability, are studied since they play important roles during the melt-jet breakup. The results achieved so far indicate that the developed particle method is capable to analyze the melt-jet breakup with heat transfer.

Determination of Gross Alpha and Gross Beta Activity Concentration in Bahe River Water Samples Within Xi'an Chanba Area

As a screening procedure, gross alpha and gross beta activity have been developed to determine whether further analysis of water samples related to specific radionuclide is necessary. In China, the determination of gross alpha and gross beta in drinking water was generally based on the HJ standard method: HJ 898-2017, water quality — determination of gross alpha activity — thick source method, and HJ 899-2017, water quality — determination of gross beta activity — thick source method. In this study, 15 water samples from Bahe river in Chan Ba region of Xi’an in China, were pretreated and analyzed by BH1216-III low background alpha and beta scintillation counter. The water samples were collected nearby residential area, wetland park, water conservation district and urban sewage treatment plant as well as other important locations which probably influence on the radioactivity level. The values of the activity concentrations of the gross alpha and beta measured in the water samples ranged from less than LD to 0.183Bq/L with a mean of 0.077Bq/L and 0.073–0.151Bq/L with a mean 0.102Bq/L respectively. All values of samples were lower than the limit level of 0.500Bq/L for gross alpha and 1.000Bq/L for gross beta, indicating that the radioactivity level in Bahe water between Chan Ba region of Xi’an is basically within the normal environmental background.

FUELPOOL: A Computer Program to Model CANDU Spent Fuel Pool Severe Accident Progression and Consequences

CANDU PHWR spent fuel pools (SFPs), smaller than a tennis court, contain up to 38,000 or more (49,000 in Wolsong)fuel bundles in geometries not replicated in any other power reactor. Therefore, the phenomenological issues, accident progression pathways and effectiveness of mitigative actions are somewhat different. This requires a dedicated approach in progression and consequence assessments of potential accidents and development of mitigation measures. The SFPs house densely packed fuel bundles stacked in about a hundred vertical stainless steel tray towers, each containing 24 spent fuel bundles in each of the 16 or more (19 in Wolsong) horizontal fish basket style steel trays. Some of theupto 10 year worth of the on-line refuelled bundles in the SFP are at relatively high decay powers as fuel trays are prepped for the towers in near daily basis. In addition, there is a provision (see Figure 1) that a full core of bundles 20 days after being at full power can be transferred to the spent fuel bay at any time. About 4.5m of additional water layer on top of the tray towers provide radiation protection and a healthy margin to small rate of fluid loss.

Thermo-Mechanical Analysis of Instrumentation Guide Tube Failure During a Severe Accident in a Nordic Boiling Water Reactor

Vessel penetrations are important features of both pressurized water reactors and boiling water reactors. The thermal and structural behaviour of instrumentation guide tubes (IGTs) and control rod guide tubes (CRGTs) during a severe accident is vital in the assessment of the structure integrity of the reactor pressure vessel. Penetrations may fail due to welding failure, nozzle rupture, melt-through, etc. It is thus important to assess the failure mechanisms of penetrations with sufficient details. The objective of this paper is to assess the timing and failure modes of IGTs at the lower head during a severe accident in a Nordic boiling water reactor. In this study, a three-dimensional local finite element model was established using Ansys Mechanical that includes the vessel wall, the nozzle, and the weld joint. The thermo-mechanical loads of the finite element model were based on MELCOR results of a station blackout accident (SBO) combined with a large-break loss-of-coolant accident (LBLOCA) including an external vessel cooling by water as a severe accident management strategy. Given the temperature, creep strain, elastic strain, plastic strain, stress and displacement from the ANSYS simulations, the results showed the timing and failure modes of IGTs. Failure of the IGT penetration by nozzle creep is found to be the dominant failure mode of the vessel. However, it was also found that the IGT is clamped by the flow limiter before the nozzle creep, which means that IGT ejection is unlikely.

Numerical Simulation of Liquid Jet Behavior in Shallow Pool by Interface Tracking Method

In the severe accident (SA) of nuclear reactors, fuel and components melt, and melted materials fall to a lower part of a reactor vessel. In the lower part of a reactor vessel, in some sections of the SAs, it is considered that there is a water pool. Then, the melted core materials fall into a water pool in the lower plenum as a jet. The molten material jet is broken up, and heat transfer between molten material and coolant may occur. This process is called a fuel-coolant interaction (FCI). FCI is one of the important phenomena to consider the coolability and distribution of core materials. In this study, the numerical simulation of jet breakup phenomena with a shallow pool was performed by using the developed method (TPFIT). We try to understand the hydrodynamic interaction under various, such as penetration, reach to the bottom, spread, accumulation of the molten material jet. Also, we evaluated a detailed jet spread behavior and examined the influence of lattice resolution and the contact angle. Furthermore, the diameters of atomized droplets were evaluated by using numerical simulation data.

Unmet Challenges to Successfully Mitigating Severe Accidents in Multi Unit CANDU Reactors

One sees eerie similarities here in Canada to the cozy relationship between regulator and utilities in ‘pre-Fukushima’ Japan. Such ties are hardly limited to Canada though. The chronic degradation of real commitments to continued improvements in reactor safety systems and a decline in overall safety culture that discourages critical design reviews and willfully ignores well justified, safety critical hardware upgrades, has created alarming conditions that are likely inching us towards another nuclear disaster. Operating CANDU reactors are now close to being obsolete but have barely seen any substantive severe accident related risk reduction upgrades nine years after Fukushima, hoopla in Canada around some minor improvements and premature closure of even otherwise sparse and what were really weak regulatory ‘Fukushima Action Items’ notwithstanding.

Development and Assessment of a Nearly Autonomous Management and Control System During a Single Loss of Flow Accident

In this work, a Nearly Autonomous Management and Control (NAMAC) system is designed to diagnose the reactor state and provide recommendations to the operator for maintaining the safety and performance of the reactor. A three layer-hierarchical workflow is suggested to guide the design and development of the NAMAC system. The three layers in this workflow corresponds to knowledge base, digital twin developmental layer (for different NAMAC functions), and NAMAC operational layer. Digital twin in NAMAC is described as knowledge acquisition system to support different autonomous control functions. Therefore, based on the knowledge base, a set of digital twin models is trained to determine the plant state, predict behavior of physical components or systems, and rank available control options. The trained digital twin models are assembled according to NAMAC operational workflow to support decision-making process in selecting the optimal control actions during an accident scenario. To demonstrate the capability of the NAMAC system, a case study is designed, where a baseline NAMAC is implemented for operating a simulator of the Experimental Breeder Reactor II (EBR-II) during a single loss of flow accident. Training database for development of digital twin models is obtained by sampling the control parameters in the GOTHIC data generation engine. After the training and testing, the digital twins are assembled into a NAMAC system according to the operational workflow. This NAMAC system is coupled with the GOTHIC plant simulator, and a confusion matrix is generated to illustrate the accuracy and robustness of implemented NAMAC system. It is found that within the training databases, NAMAC can make reasonable recommendations with zero confusion rate. However, when the scenario is beyond the training cases, the confusion rate increases, especially when the scenarios are more severe. Therefore, a discrepancy checker is added to detect unexpected reactor states and alert operators for safety-minded actions.