Volume 90, Issue 1 - Special Issue: 50th anniversary on Structural Mechanics in Reactor Technology (SMiRT) conference series; Guest Editor: Hamid Sadegh-Azar

Unreinforced masonry walls are highly susceptible to out-of-plane loads during seismic events. The rocking behavior of masonry walls under seismic loads is a nonlinear, time-dependent process, best represented by nonlinear dynamic time-history simulations. However, discrete modeling of complete wall systems – including bricks, mortar joints, and boundary conditions – is typically impractical. To achieve simplicity and precision, an idealized single-degree-of-freedom (SDOF) model is often preferred. Several authors have developed methods to determine the force-displacement relationship of unreinforced masonry walls, but experimental investigations have shown that critical influencing factors are often neglected. A new model has been developed to address these gaps, incorporating real-world parameters such as vertical stiffness and dynamic effects, and has been experimentally verified through shaking table tests. In the present work, the model is extended to include the effects of cantilevered loads, which significantly impact the structural behavior of walls. Dynamic analyses using real earthquake records reveal that higher cantilevered loads lead to earlier failures and greater deformations, particularly in slender configurations. These findings highlight the critical need for precise modeling and consideration of cantilevered loads in seismic design, especially for safety-critical structures like nuclear facilities.

The present paper deals with the fragility analysis of the nuclear steam supply system (NSSS) at Gösgen nuclear power plant (KKG), a three-loop pressurized water reactor (PWR). The analysis covers the stability and integrity of the main NSSS components and piping: reactor pressure vessel (RPV), steam generators (SG), reactor coolant pumps (RCP), pressurizer, main coolant lines, surge line, main feedwater and main steam lines (inside the reactor building). Furthermore, the fragility of the reactor trip is analyzed with respect to failure modes associated with the RPV internals (RPVI). The fragilities are based on a probabilistic soil-structure interaction analysis with a coupled model of the reactor building and the above mentioned NSSS components/piping, see Rangelow et al. (2019. Probabilistic SSI analysis of a reactor building to facilitate fragility assessment. Transactions, SMiRT-25 , Charlotte, NC, USA). The ground motion is represented by time histories corresponding to the median uniform hazard spectra (UHS) with an annual probability of exceedance of 10 −4 . The fragility analysis is performed according to the separation-of-variables method, following guidance documents such as EPRI (2009. Seismic fragility applications guide update , Report 1019200. Palo Alto, CA) and EPRI (2018. Seismic fragility and seismic margin guidance for seismic probabilistic risk assessments , Report 3002012994. Palo Alto, CA). For the RPV internals, the fragility is based on two non-linear analysis models: a refined one for the evaluation of the fuel assembly spacer grids and a more global one for the remaining failure modes. A combined fragility has also been derived for the RPV internals, as the union of the individual failure modes. The paper discusses the method, the input data and selected results of the analysis.

The concrete containment building (CCB) is the third and last safety barrier to protect the environment from diffusion of radioactive products. In 2014, Électricité de France (EDF) launched the VERCORS project to improve the knowledge of leak-tightness in the CCB under air pressure loading and aging effects. VERCORS is in a 1:3 scaled CCB mock-up of a 1,300 MWe French pressurized water reactor (PWR). This paper presents a numerical methodology applied on VERCORS mock-up and numerical results focus on the gusset only, which is the lower part of the containment wall casted right after the base slab. Concrete is modeled through 3D elements while rebars and tendons are modeled by 1D elements. The simulation includes THM analysis of the gusset, prestressing phase and pressurization phase, and is conducted through a finite element code, Cast3M. The delayed strains including creep and shrinkage are especially considered. The instantaneous prestress losses, which result from friction; relaxation and pull-in at wedge, are taken into account. Regarding the long-term prestress loss due to the delayed strains, it is achieved by a kinematic relationship at interface of concrete-tendons. To describe cracks in concrete, an elastic damage model is used for the concrete elements, and the post-processing of leak-tightness is based on through crack leakage only. Those cracks appear usually at the early age phase given the restrained strains that the gusset undergoes during the hydration and hardening phase.

The first draft of the new developed ISO standard series ISO 4917 entitled “Design of Nuclear Power Plants against Seismic Events” comprises six parts where part 1 “Principles” represents the basis for the following five parts. Part 2 concentrates on “Geotechnical Aspects”, part 3 on “Civil Structures”, part 4 on “Components”, part 5 on “Seismic Instrumentation” and part 6 on “Post Seismic Measures”. All parts have been adapted to newly developed and existing IAEA guidelines. Many restrictive clauses have been opened to accommodate major relevant national guidelines and code of practices. Several new aspects from other newly developed seismic design guidelines and recent research results have been incorporated in the new drafts. In this paper the development and the corresponding scientific background of the new ISO standard series ISO 4917 is presented.

The aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO ® claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5 ® samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area.

Fitness-For-Service (FFS) and regulatory codes provide rules allowing determining whether multiple flaws shall be combined into a single flaw or evaluated separately. However, these rules set thresholds beyond which flaws are replaced by an envelope sometimes over conservative. The present study aims to develop a hybrid surrogate model accounting for crack interaction in an infinite isotropic homogeneous medium subjected to a uniform stress field normal to the reference plane. The paper presents closed form formulae for interaction factors relative to parallel equal through-wall cracks subjected to normal tension. Crack shielding is analysed in detail: the equation of the shielding limit is established and mode I and mode II variations of overlapped cracks are compared. Comparison with Finite Element calculations for all configurations validate the results.

The paper investigates the seismic SSI effects for a base-isolated reinforced concrete (RC) storage structure founded on firm soil subjected to coherent and incoherent motions. Three types of base-isolators are considered: (1) lead-rubber bearing (LRB) isolators, (2) triple pendulum friction (TPB) and (3) base control system (BCS) isolators including a combination of spring units and high-viscosity damper units. Two level of earthquake severity are considered: (1) 0.40 g for DBE and (2) 0.60 g for BDBE (1.5 DBE). Both deterministic and probabilistic simulations are considered. The paper also investigates the effects of motion incoherency on the SSI responses for base-isolated structure and shows that these effects could be? quite significant. The LRB and TPB isolators are modelled as hysteretic systems, while the BCS isolators are modelled using a combination of linear springs and frequency-dependent 3D high-viscosity damper (HVD) systems. Results highlight the significant additional benefits of the 3D-space BCS in comparison with the traditional 2D-space horizontal LRB and TPB systems, for reducing of the large structural amplifications due local vibration modes, reducing drastically the floor vertical vibrations, damping totally the detrimental amplifications due to the motion incoherency effects, and also reducing the structural moments more significantly than the other isolation systems.

The Fukushima accident prompted rapid development in the appreciation and codification of BDB methodology into regulatory guidance and this is still continuing, especially at IAEA, which has recently and is due to publish several new standards documents directly relevant to BDB analysis. The objective of this paper is to provide a state-of-art review on how BDB evaluations are undertaken in the UK nuclear industry. Then, using the work undertaken for the Generic Design Assessment (GDA) in the UK for UK HPR1000, this paper proposed a more systematic approach for “cliff-edge” evaluation based on a generic nuclear power plant design currently being developed in the UK.

Several types of dynamic tests were performed using a full-scale lead rubber bearing (LRB) that has a lead plug embedded at the center of laminated natural rubber. The horizontal and vertical characteristics of the LRB due to various shear strains, frequencies, cyclic effect and compressive stresses were obtained by the experiments. It was confirmed that the variations of the obtained horizontal and vertical characteristics satisfied allowable variations defined in European standard for Anti-seismic devices (EN15129-2009).