The degassing tower liquid cooler is an important equipment for boron recovery system of HPR1000 nuclear power plant. It is crucial to ensure the security function of degassing tower liquid cooler under earthquake. This report carried out a seismic analysis for the degassing tower liquid cooler. The finite element (FE) model for the Degassing Tower Liquid Cooler was established by using the ANSYS code. The response spectrum method was adopted for the equipment seismic analysis and evaluation. Through the seismic analysis, the design weakness of the degassing tower liquid cooler was identified. And the weakness was modified to meet the requirement of practical engineering according to the FE results and the relevant specification.

High confidence of low probability of failure (HCLPF) values of equipment, representing the seismic capacities of the equipment, are the fundamental ingredient in seismic probability safety assessment (SPSA) and seismic margin analyses (SMA). In this paper, two methods for calculating the HCLPF values of equipment were investigated, fragility analysis, and conservative deterministic failure margin (CDFM). These methods are linear methods. Based on these methods, HCLPF value of equipment can be computed conveniently by scaling the results of the existing seismic analysis. For a nonlinear systems, the HCLPF values based on these linear scaling methods are unrealistic. For a complicated nonlinear equipment or structure, a detail nonlinear model was used to derive the seismic capacity. The results by this method are realistic, but cost calculation time. In this paper, a nonlinear model of reactor coolant system coupled reactor building was built. This model includes the steam generator and considers the nonlinear factors of steam generator such as gap in the supports, plasticity of hot leg and cold leg. Forced motion was applied to the base of reactor building. And seismic response of the steam generator was calculated iteratively by scaling the ground motion level step by step. Based on these calculations, a curve of load on the supports VS peak ground acceleration (PGA) can be obtained. Then based on these curves and allowable load of supports of steam generator, which derived from stress analysis on support of steam generator, seismic capacity of the supports of steam generator was determined. Then the HCLPF Value of the supports of steam generator was obtained by this nonlinear time history analysis and was compared with the results based on the CDFM. The two results were different. Therefore, the HCLPF seismic capacity of equipment with nonlinearity, such as gap nonlinearity, should be calculated by nonlinear time history method.

Modeling water in passive containment cooling water storage tank (PCCWST) using fluid element will result in large amount of calculation when conducting seismic analysis of shield building or NI. Thus, it is necessary to simplify the modal of water so as to reduce the difficulty of seismic analysis under condition that the error is slight enough to be ignored. By formula deduction and analysis, on the one hand, this paper proofs that modeling “sloshing mass” as fixed mass on structure is unreasonable. On the other hand, this paper proposes that the reasonable simplified approach is to decouple “sloshing mass” totally from the structure system. Furthermore, conditions of utilizing decoupling method are illustrated.

Water sloshing of PCS water storage tank (PCSWST) can cause a significant effect on the dynamic response of Shield Building under seismic loads. It is complicated to perform the calculation of water sloshing especially for the tanks with irregular shapes. Consequently, it is important to establish an appropriate equivalent mechanical model for simulation [1], [2] . In this paper, the water sloshing is firstly investigated based on the potential flow theory, which including the seismic modal analysis. Based on the theoretical research, a highly efficient (simplified) calculation formula is derived, which mainly considering the impulse mass, convective mass, position function and spring stiffness etc., through this way the equivalent model for PCSWST is established by applying mass-spring element. The equivalent models based on Housner & Graham theory [3], [4] are also established. Additionally, the 3-D finite element model of water sloshing considering fluid-structure interaction is established by using the software of Ansys. Total of four models are built as shown in the paper, then modal analysis and dynamic response under earthquake excitation are performed using ANSYS. The results are compared to justify the equivalent model in this paper. The results indicate that Graham formula did not provide the correct location expressions for the convective masses. The expressions for the impulsive mass and its position given by Housner are not satisfactory. As a comparison, the results from the equivalent model, which is recommended in this paper, can best fit the data from finite model. From above results and comparisons, a more reasonable and refined equivalent model for PCSWST design is provided. Based on the equivalent model established, the influence on structure caused by the increase of water mass is analyzed. The results from the seismic analysis are compared, including member force, shear strain and shear force. Based on the research, the feasibility of the design is analyzed, which can provide important support for the structural design. Finally, the seismic reduction of water tank is studied using the finite element model established in this paper. The horizontal and vertical anti-sloshing baffles are designed. The maximum acceleration and displacement corresponding to different baffle length are compared to study the effect of the seismic reduction.

Spent fuel pool and storage racks are important nuclear security structures and components. In order to prevent it from structural failure, which includes the loss of the structural integrity of the spent fuel pool and stability of the spent fuel storage racks, also includes the possibility of fallen down of storage racks under seismic loading. Besides the necessary static analysis of structures, the influence of seismic loading on the interaction between water and structure should be fully considered, Especially concerned the analysis of the shaking effect of water sloshing on the storage racks, the displacement and the possibility of fallen down of the storage racks. The present paper is concerned with the problem of modeling the fluid-structure interaction (FSI) in filled liquid and filled with spent fuel pool. The study focuses on the sloshing phenomena and on the coupling computational fluid dynamic (CFD) analysis with the finite element stress analysis (FEA) code LS-DYNA. By the results of the response of seismic, such as the displacement of the storage racks, pressure exerted on the plate of racks and the walls of the pool. This paper also evaluates the seismic performance of the structure and the safety margin. Various numerical methods can be used for analysis of liquid storage pools, among these we mention explicit finite element, implicit Lagrangian-Eulerian, hybrid finite element, Smoothed Particle Hydrodynamics volume of fluid. In this article the coupled sloshing dynamics in a rectangle pool were studied using a model developed in LS-DYNA environment. The main solution methodology is based on explicit time integration. In order to demonstrate the FSI results of the FEA models of the spent fuel pool on seismic analysis, a 3D FEA models were developed. The Finite element model composed of the spent fuel pool (steel plate concrete), spent fuel storage racks, cushion block, water and air. Solid element modeling is used in concrete, cushion block, water and air. Steel plate and storage racks employ the shell element. The constitutive model of solid element is linear elastic. And the constitutive model of fluid element is described by the Gruneisen equation. Arbitrary Lagrangian-Eulerian (ALE) formulation is thought of as algorithms that perform automatic rezoning. It realized the advection of water and air in the ALE multi-material group.

In 1990s the first generation of PWR fuel assembly FA300 was developed in China. With constantly design improvements over the following 20 years, three types of FA300 fuel assemblies have been developed with max fuel assembly average burnup reach 40000 MWd/tU and successfully used in 300MWe PWRs. By now over 1100 fuel assemblies of FA300 series have been successfully operated in three different 300MWe PWRs with satisfied operation reliability. The CAP1400 fuel assembly development program was launched in 2010 which includes new zirconium cladding material development, UO 2 pellet development, high performance fuel assembly mechanical structure development, fuel rod performance code and fuel assembly seismic analysis code development, out-of-pile fuel assembly test facility construction and in core irradiation program. The main purpose of CAP1400 fuel assembly development program is to meet the needs of self-sufficient fuel supply to CAP1400 reactor. Based on FA300 and CAP1400 fuel assembly development, a full series of PWR Fuel R&D technology and test system has been successfully established by SNERDI which will continuously support the fuel assembly improving and new type of fuel development. This paper will mainly describe the PWR fuel technology including cladding technology, pellet technology, fuel assembly mechanical design technology, fuel assembly test facility and test technology, fuel rod and fuel assembly code development progress. Furthermore, the development progress of CAP1400 fuel assembly will also be introduced. By now two types of new zirconium alloys have been selected as candidate alloys, all of the out-of pile performance tests have been finished. Both two alloys show good corrosion resistance. Through full series of fuel assembly components performance tests, the fuel assembly has finished design finalization. Most of the fuel assembly mechanical and hydraulic tests will be finished by the end of 2016. The test reactor irradiation program and commercial PWR irradiation program are also on the schedule. After the LTA program and commercial application licensing, the CAP1400 fuel assembly is anticipated to provide adequate burnup capability and operation reliability to CAP1400 reactor in China.

Nuclear class equipment should be assessed for seismic safety before they are used in nuclear power plant. According to nuclear safety codes and regulations, all seismic category I equipments shall be designed enduring safety shutdown earthquake (SSE). That is, the stress evaluation needs to be accomplished for those structures. For some components, the deformation evaluation needs to be performed as well to assure the function integrity of the equipment. In this paper, the seismic analysis for an explosion-proof valve used in nuclear power plant, which exactly belongs to seismic category I equipment, has been conducted based on finite element method. The natural frequency, vibration mode and seismic response of the structure have been obtained through calculation, and the stress and deformation under the combined loadings of gravity, internal pressure, blast and seismic load have been evaluated according to ASME AG-1. The bolts of the structure have been qualified according to ASME III-NF as well. The results show that the design of the explosion-proof valve is in compliance with the requirement of corresponding nuclear safety standards.

Sodium-cooled fast reactor vessel is one of seismic grade I equipment, so the vessel seismic analysis considered the liquid sloshing is an important topic in fast reactor vessel design. In order to investigate the behavior of liquid sloshing under the condition of seismic, a fast reactor vessel model at 1:25.7 scale was designed. Using water simulate liquid sodium, we carried out modal experiment and seismic table test. By using hammering test measured natural frequency of model structure under the condition of water and without water; By utilizing seismic table frequency sweep test measured frequency of water sloshing in model; Then, input water sloshing sine wave of first order frequency and second order frequency respectively through seismic table.After that measure wave height and hydraulic pressure response of liquid sloshing. The experimental results show wave height and hydraulic pressure response of liquid sloshing in second order frequency is bigger than the first order frequency. This conclusion can provide favorable reference for the seismic design of the sodium-cooled fast reactor.

Safety enhancement is essential to spent fuel storage racks of nuclear power plant after Fukushima accident. Criticality safety, cooling safety and structural safety enhancement for CAP1400 (a larger advanced pressurized water reactor developed by China) spent fuel storage racks of are proposed and prominent results have been obtained. The criticality analysis results indicate that neutron absorber insert is efficient to increase criticality safety of storage rack. The CFD simulation and seismic analysis reveal that these storage racks can meet the cooling safety and structural integrity requirements. Both neutron absorber inserts and passive spent fuel pool cooling technology have good application prospect.

Following the accident at the Fukushima Daiichi Nuclear Power Plant resulting from the March 11, 2011 Great Tohoku Earthquake and subsequent tsunami, there was a general concern regarding the beyond design basis capability of existing nuclear power plants. The Condensate Storage Tanks (CST) in the nuclear power plant were originally designed to withstand an earthquake with a peak ground acceleration (PGA) of 0.3g. The government regulatory commission increased the required PGA to 0.4g, therefore, an upgrade of the design basis for the CSTs was required. Due to the vintage of the existing nuclear power plant, the United States Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46 methodology may be used for seismic upgrade work of the mechanical and electrical equipment. The Condensate Storage Tanks (CST) belong to the mechanical and electric equipment and therefore were required to have seismic upgrade work because of significant deficiencies that were found in the anchorage of the tanks to the concrete foundation. A seismic analysis and design upgrade of the Condensate Storage Tanks (CST) was performed to resolve exceedances for base shear, overturning moment and sloshing due to recently updated seismic loads. A detailed analysis of the CST showed that the as-built anchor chair detail did not provide sufficient margin for the Beyond Design Basis Earthquake Event (BDBEE). The as-built anchor chair detail did not provide the required strength to transfer the shear and pull-out loads from the tank shell to the concrete foundation, i.e. no clear load path was provided for the updated seismic loads. Therefore, as a result of the inadequate anchorage, the main issues to resolve in the CST were tank sliding, shell buckling and sloshing due to earthquake loading. The principal challenges encountered during the analysis, design and construction stages were (1) not allowing to loosen the double nut configuration attaching the anchor bolt to the existing anchor chair, and therefore allowing to (2) remove or replace only a few components of the as-built anchor chair, and (3) the retrofit design had to be implemented while tanks were operable, i.e. filled with fluid. An additional challenge faced during the design of the new anchor chair components was the limit in anchor chair height imposed by the numerous interferences in the CST such as nozzles, reinforcing plates and existing welds. A mitigation strategy is analyzed, designed and successfully implemented for retrofitting the ninety-six anchor chairs and allowing for full development of the anchor bolt shear and pull-out strength.

Seismic analysis of the primary system in nuclear plant design is very important. The change of reactor coolant pump has some impact on the support seismic loads of primary equipment. Using the numerical calculation method, preliminary analysis was completed and some valuable results are obtained. The results show the natural frequency of system should be far away from the spectral peak with optimizing the reactor coolant pump configuration to raise the safety and economy of the plant.

Japan Atomic Energy Agency (JAEA) has carried out research and development to establish the technical basis of High Temperature Gas-cooled Reactor (HTGR) by using High Temperature engineering Test Reactor (HTTR). On March 11th, 2011, the Great East Japan Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the HTTR had been stopped under the periodic inspection and maintenance of equipment and instrument. In the great earthquake, the maximum seismic acceleration observed at the HTTR exceeded the maximum value in seismic design. The visual inspection of HTTR facility was carried out for the seismic integrity conformation of HTTR. The seismic analysis was also carried out using the observed earthquake motion at HTTR site to confirm the integrity of HTTR. The concept of comprehensive integrity evaluation for the HTTR facility is divided into two parts. One is the “inspection of equipment and instrument”. The other is the “seismic response analysis” for the building structure, equipment and instrument using the observed earthquake. For the basic inspections of equipment and instrument were performed for all them related to the operation of reactor. The integrity of the facilities is confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the result of inspection of equipment and instrument and seismic response analysis, it was judged that there was no problem to operate the reactor, because there was no damage and performance deterioration, which affects the reactor operation. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015.

The aim of this study is to evaluate the influence of the damaging or failure effects induced by seismic loadings on the loading bearing capability of the overall nuclear structures, systems and components (SSCs). In particular it was analyzed the behaviour of the SSCs in terms of their required safety functions and capacity assuming the occurrence of an earthquake having magnitude beyond the design basis value (in agreement with the “stress tests” suggestions, presently foreseen by the European and International Associations). To the purpose, quite refined FEM models were set up and implemented considering suitable materials behaviour and constitutive laws for the reactor materials, particularly for the concrete material behaviour (which could suffer failure and damage mechanisms during the ground shaking). Several type of damaging mechanics were considered. The obtained results seemed to confirm the overall containment reliability even if with minor upgrading actions in operating procedures and design characteristics. Moreover they allowed to appropriately check the NPP containment strength reserve.

On March 11th, 2011, the 2011 Tohoku Earthquake which is one of the largest earthquakes in japan occurred and the maximum acceleration in observed seismic wave in the HTTR exceeded the design value in a part of input seismic motions. Therefore, a visual inspection, a seismic analysis and a performance confirmation test of facilities were carried out in order to confirm the integrity of facility after the earthquake. The seismic analysis was carried out for the reactor core structures by using the response magnification factor method. As the results of the evaluation, the generated stress in the graphite blocks in the reactor core at the earthquake were well below the allowable values of safety criteria, and thus the structural integrity of the reactor core was confirmed. The integrity of reactor core was also supported by the visual inspections of facilities and the operation without reactor power in cold conditions of HTTR.

This paper is aimed at analyzing the structural responses of a Generation IV heavy-liquid-metal-cooled reactor (lead-bismuth-cooled fast reactor) vessel in the event of earthquake. For a seismic design, the seismic time history response analyses are carried out for both a top support type and a bottom support type. It is found that the bottom support type exhibits the better performance. There is the gap between the reactor vessel and guard vessel being filled with [1] argon. It is indispensable to add bellows at vessel upper end for a bottom support type to achieve the connection and seal between the reactor vessel and guard vessel. This paper is the first attempt to evaluate the effects on structural seismic performance by adopting the equivalent springs to simulate this elastic connection. For a seismic investigation, there is a key issue that should be focused on, namely: the fluid-structure interaction due to seismic loading. The vessel is filled with a high-density fluid and might lead to severe hydrodynamic pressure significantly in the occurrence of earthquake. It will impair the structural integrity of reactor vessel. In order to study the structure effects of reactor vessel under seismic loading, an appropriate 3-D finite element analysis model has to be set up and the FEM code ANSYS has been implemented.

Graphite plays an important role in the pebble-bed high temperature gas-cooled reactors (HTR) as moderator, reflector as well as internal structural material. The HTR core consists of a large number of graphite bricks interconnected with keys. It is required that the structural integrity of the HTR core be maintained when subjected to the seismic load. Hence it is important from the viewpoint of seismic design to investigate the seismic responses of the graphite bricks. Considering the pebble-bed HTR has various graphite shapes, a generalized three-dimensional model with the associated computer code is developed to treat these interconnected graphite bricks with arbitrary shapes. In this model, each brick is treated as a rigid body with six degrees-of-freedom: three translational displacements and three rotations around the brick center of gravity. A nonlinear spring dashpot model is applied to present the collision between adjacent bricks and the interaction forces through the key systems. In the numerical tests, the code is verified by comparing predicted responses with exact solutions for two cases and good agreement is observed. The model is then used for the dynamic analysis of the side reflectors of the pebble-bed HTR core under a given seismic load. The calculated response behaviour of the side reflector column is summarized and discussed.

For high earthquake resistance and ease of installation, free standing racks which are not anchored to the pool floor or walls has been adopted in many countries. Under the earthquake, the response of the free standing rack is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning) and impacts between the fuel assemblies and the fuel cell walls, rack-to-rack, and the pit floor and rack pedestals. To obtain an accurate simulation of the free standing rack, the seismic analysis requires careful considerations of these complex phenomena (sliding, rocking, twisting, and turning), fluid coupling effects and frictional effects. We carried out seismic experiments on the full-scale rack model in water and dry conditions to obtain the fundamental data about free standing rack (sliding, rocking and turning motions). We have developed the nonlinear dynamic analysis method to predict seismic response for the free standing rack utilizing the full-scale test result and verified the analysis evaluation method of the rack by comparison of test result.

The Atucha II nuclear power plant is a pressurized heavy water reactor being constructed in Argentina. Nuclear power plants must be designed to maintain their integrity and performance of safety functions for a bounding set of normal operational events as well as abnormal events that might occur during the lifetime of the plant. Seismic fracture mechanics evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to an event with a probability of 10 −6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size for this case is 94-percent of the circumference. These analyses are performed by placing cracked-pipe-elements into a complete model of the primary cooling system including the reactor pressure vessel, pumps, and steam generators as summarized in the paper. This paper summarizes these results and further shows how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent (about 120°) around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 10 −6 seismic event accelerations, or 240 times higher than the SSE accelerations. This suggests there is a huge safety margin for beyond design basis seismic events and Atucha II plant rupture is pragmatically impossible. These surprising results are discussed and could be potentially applicable to other nuclear power plants as well.

Graphite blocks are important core components of the high temperature gas-cooled reactor. As these blocks are simply stacked in array, collisions among neighboring components may occur during earthquakes or accidents. Thus, it is important to develop a reliable seismic model of the stacked graphite blocks and have them designed to maintain their structural integrity during the anticipated occurrences. Various aspects involved in modeling and calculating impact-contact dynamics can affect the resulting behavior of the graphite block. These include mesh size, time step, contact behavior, mechanical constraint formulation of impact-contact analysis, etc. This work is dedicated to perform comparative studies and the effects of these parameters will be identified. The insights obtained through these studies will help build a realistic impact-contact model of the graphite block, from which a lumped or reduced dynamics model will be developed for the seismic analysis of the reactor including these graphite components.

The seismic analysis on reactor structure was performed with a new generation of finite element software. The amount of freedom degree of the model was more than twenty millions. The typical responses to operational basis earthquake excitation were given. They are larger than those with two-dimensional simplified finite element method, and the reasons of this phenomenon were analyzed. The feasibility of seismic analysis on large-scale three-dimensional finite element model under existing hardware condition was demonstrated, so some technological reserves for dynamic analysis on complicated equipments or systems in nuclear engineering are provided.