F. H. E. De Haan-De Wilde’s research while affiliated with NRG, Nuclear Research & consultancy Group and other places

Pressurized thermal shock (PTS) may cause a quick, catastrophic cleavage fracture in a reactor pressure vessel (RPV) of a pressurized water reactor (PWR). Low temperatures, thermal strains, and radiation embrittlement can all combine to create dangerous situations for structures, specifically thick-walled reactor pressure containers with fractures and welds as weak areas. A thorough picture of the temperature and stress intensity is required to determine the likelihood of the onset and spread of a cleavage crack. The ductile to brittle transition temperature affects the critical stress intensity for brittle cleavage fracture. This complicated combination of loads, absolute temperatures, and temperature gradients is combined with radiation damage to evaluate the likelihood that cleavage fracture will occur. In earlier works, simulations were carried out using combined computational fluid dynamics (CFD) and finite element method (FEM) simulations to get the most realistic picture of this issue. However, due to the complexity of the problem, the thermal mixing of the fluid and its effects on the RPV wall are simulated by models that are simplified in terms of geometric complexity and physics. This study investigates the effect of the interaction between multiple emergency core cooling (ECC) plumes on the thermal response of the RPV wall by considering a full (360 degree) RPV geometry with two loops for the ECC fluid injection. We first perform a transient conjugate heat transfer CFD simulation to compute the spatial and temporal evolutions of RPV wall temperature. The unsteady Reynolds-averaged Navier-Stokes equations are solved on the fluid side, and the unsteady heat transfer equation is solved on the solid side. Next, a static structural analysis using FEM is conducted using the temperature profile obtained from CFD analysis on a one-loop reactor as input. The goal of the FEM analysis is to investigate the link between the depth, length, and ratio of the crack and the probability of failure. A probabilistic approach is used to evaluate the possibility of failure. The ultimate goal of these studies is the generation of a code that can implement hydraulic models that replace the time and resource-demanding CFD and FEM analysis.

The Double Ended Guillotine Break (DEGB) is a pipe failure mode in nuclear power plant installations which is the postulated root cause of a Loss Of Coolant Accident (LOCA). Pipe whip restraints and jet impingement shield are safety systems that mitigate the effects of this postulated accident. The Leak Before Break (LBB) failure mode occurs when a small leak from a cracked pipe is detected prior to the DBEG failure. The application of this failure mode and the detection of cracked pipes would increase defense in depth and justify the removal from the plant of the pipe whip restraints and the jet impingement shield, decreasing costs and maintenance time. In this study a probabilistic assessment is performed with an in-house developed LBB software. The assessment models the complexity of the physical model and the diversity of the systems with stochastic input variables for the material properties and crack parameters. The developed deterministic code is based on the UK procedure for assessing the integrity of structures containing defects (R6). The Detectable Leak Before Break (DLBB) procedure is used, which is based on the Failure Assessment Diagram (FAD) Option 1 assessment procedure. The structural integrity assessment is then coupled with the Henry-Fauske two-phase critical flow model for the evaluation of the leakage rate. A sensitivity analysis is performed to reduce the number of probabilistic variables. The output from the assessment is the probability of defect detection prior to structural failure. The Second Order Reliability Method (SORM) is the probabilistic method used. The probabilistic results are then compared with the safety factors currently used for deterministic LBB assessments.

NPPs and research reactors built during the mid-20th century often have incomprehensive material characteristics of their concrete structures. This lack of quality records frequently leads to challenges when attempting to demonstrate safety during life extension projects or when specific modelling is necessary for portions of the plant when design regulations are updated with new or revised requirements. In the framework of LTO there is limited knowledge about ageing and the structural integrity of concrete structures. In order to increase the knowledge in the field of civil structures, this paper focuses on investigating the ageing mechanisms of civil structures at NRG, Petten, and determining an appropriate chloride ingress model together with the alteration of concrete material properties to account for possible degradation effects. Knowledge of the ageing mechanisms of civil structures, especially concrete, will improve concrete’s ageing management and assessment methods. The purpose of this research paper is to provide a methodology that can be used to account for possible concrete degradation. This will include a chlorine ingress model, calibrated to data obtained from the HFR chimney, that allows for the rebar reduction to be calculated. Furthermore, the design properties of concrete were determined in line with NEN-EN 1992-1-1:2004 [10] and NEN-EN 1990 [11] for specific concrete grades. The model generated will allow for user input if more detailed concrete properties are available for the area in question. This will allow the generation of area specific design properties for concrete that accounts for possible rebar reduction due to chloride ingress corrosion. The report illustrates a methodology to account for chloride ingress during modelling by reducing the rebar diameter. It further demonstrates the importance of the assumed concrete properties as starting inputs to the chloride model and the calculated rebar reduction and generated internal pressure. The determined concrete properties, together with calculated rebar diameter reduction, are used to update the previously completed FEA of the HFR [1] and provide a comparison of the previous results when considering reduced concrete strength and rebar diameter. The actual rebar corrosion due to chloride ingress within the HFR is expected to be less as surface corrosion conditions are expected to be less. In the future, more accurate material properties should be determined through physical measurement for the modelled slab. Further research and modelling can be completed to understand better steel corrosion failure in reinforced concrete with 3D FE analysis and measurements for chloride ingress in the HFR should be obtained.