The natural circulation small modular reactor (NCSMR) is a 330 MW reactor which has no reactor coolant pumps (RCP) and no active safety injection systems at all. The reactor is mainly comprised of the reactor pressure vessel (RPV) with integral pressurize r and steam generator. RPV is enclosed by a vacuumed pressure containment vessel (PCV) and the PCV is submerged in the underground containment pool. A MELCOR model and corresponding input deck are developed for the RPV, PCV, and containment pool. The containment pool takes the role of ultimate heat sink (UHS) in accident situations. The containment pool may crack and leak in some critical accidents as the earthquake, leading to the severe accident of the reactor. A TMI-2 like SBLOCA in the RPV (stuck open RVVs) along with the containment pool crack (loss of ultimate heat sink) is simulated in the work. So me key parameters as the RRVs stuck open fraction, the PCV-SRVs open or not, the containment pool crack position would have large influence on the severe accident sequence. The sensitivity of these parameters to the accident sequence is analyzed in the work. According to the simulation results, the RPV pressure decreased with the RRVs stuck open. The depressurization of RPV accelerated with the RPV-SRV open fraction increase. The PCV pressure increased after that. Two cases as the PCV-SRV open after PCV pressure increase to 5 MPa, and PCV break while the RV d id not open, are analysis. The coolant discharge mass flo wrate in RPV and PCV were different in two cases, leading to the different degradation situation of the core. Since the containment pool is so important for the accident mitigation, sensitivity analysis is done for the containment pool crack position in the pool. The work will be meaningful in gaining an insight into the detailed process involved. One of the final goals of this work would be to identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents during the design process of NCSMR.

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.

The thimble tube, which is made of Zircaly-4, is one of the main components of a PWR fuel assembly. The thimble tube has an important role as a structural member of the skeleton. Another role of the thimble tube is to guide a rod cluster control assembly (RCCA) for insertion during the reactor operation, and the function has to be assured not only in normal operation but in a seismic event. In a horizontal seismic event, the fuel assembly vibrates laterally, which gives bending moment to the thimble tube. In addition, axial compressive force acts on the thimble tube in a vertical seismic event. The integrity of the thimble tube has to be maintained while this force and moment act. Mitsubishi has confirmed by the elastic stress analysis that the stress of the thimble tube is lower than the limit value requested for the seismic event. The stress evaluation method is based on the ASME code. The ASME code also describes the limit analysis which is available when the predicted stress is beyond elastic region of the material. In the analysis, the material is assumed to be elastic-perfectly plastic, and the maximum load that the structure can carry is calculated. For the reason mentioned above, the allowable limit of the thimble tube should be determined as a function between the force and the moment. We are planning to examine the allowable limit experimentally. As a step before testing, an analytical approach for the limit is discussed in this paper. Firstly, the allowable limit is calculated by a beam model assuming elastic-perfectly plastic material, based on the ASME code. Secondly, a 3D model analysis with elastic-plastic material is performed to predict the practical strength. Based on the comparison with the analysis using elastic-perfectly plastic material, ASME based limit is considerably conservative compared with the one with the actual stress-strain curve. Conversely, this means there is enough room to rationalize the allowable limit. As the future work, the experiment will be conducted to obtain the practical limit of the thimble tube and to verify the analysis results.

Sloshing is one of important issues for both loop-type and tank-type fast reactors with free liquid surface. Periods of seismic vibration are lengthened by base isolation systems installed to prevent damages to facilities during earthquakes, and get close to the natural periods of sloshing. As a result, sloshing is promoted. Sloshing waves are assumed to even reach a roof slab of a reactor vessel in severe seismic conditions. It is important to evaluate structural integrity for sloshing impacts on roofs. However, there are not any established evaluation methods for impact pressure on flat roofs yet. Therefore, in this study, a simple evaluation method is proposed based on Wagner’s theory. The effectiveness of the proposed method is illustrated using computational fluid dynamics analysis and literature data of sloshing experiments.

Fukushima Daiichi Nuclear Power Station accident occurred by the Great East Japan Earthquake on March 11, 2011. After that, continuous enhancement of nuclear safety is being required in Japan. The accident of Fukushima was caused by the seismic induced tsunami event, namely, multiple events. The other examples of multiple events due to the seismic event are such as internal fire and internal flooding in the nuclear power plants. In addition, structures, such as a building, and piping might be damaged by the seismic event, which could impact component failure dependently. In order to consider these kinds of events, the development of PRA procedures for multiple events caused by the seismic events will be highly demanded. So, we developed a basic PRA methodology for seismic induced tsunami events using “Direct Quantification of Fault Tree using Monte Carlo simulation (DQFM) methodology”. And we verified its applicability through the evaluation.

Japan Nuclear Safety Institute had recently reported the pioneering deterministic evaluation approach for nuclear power plant under seismic induced fault displacement. But the uncertainty of fault displacement based on probabilistic hazard analysis is described to be greater than that of other natural phenomena e.g. earthquake ground motions or seismic acceleration vibration in the report. Furthermore, for plant-wide risk assessment against fault displacement hazards beyond design basis displacement level, it is seriously necessary to promote a series of fundamental studies and develop the standard procedures regarding not only accident sequence analysis but also fragility analysis of buildings and structures as well as components and piping systems. Based on the above background, the objective of this study is focusing to obtain basic fragility data for the aleatory and epistemic uncertainties of structural responses for nuclear power plant buildings against fault displacement. A number of nonlinear soil-structure finite element analyses against relatively large fault displacement are performed with the randomness of soil and building material properties, the uncertainly of contact parameters relating to friction between soil and building, and also the uncertainty of fault hazards such as fault types and geometries. Their quantitative results for fragility data are shown in this paper.

Free standing spent fuel storage racks are submerged in water contained with spent fuel pool. During a postulated earthquake, the water surrounding the racks is accelerated and the so-called fluid-structure interaction (FSI) is significantly induced between water, racks and the pool walls[1]. The added mass is an important input parameter for the dynamic structural analysis of the spent fuel storage rack under earthquake[2]. The spent fuel storage rack is different even for the same vendors. Some rack are designed as the honeycomb construction, others are designed as the end-tube-connection construction. Therefore, the added mass for those racks have to be measured for the new rack’s design. More importantly, the added mass is influenced by the layout of the rack in the spent fuel pool. In this paper, an experiment is carried out to measure the added mass by free vibration test. The measured fluid force of the rack is analyzed by Fourier analysis to derive its vibration frequency. The added mass is then evaluated by the vibration frequency in the air and water. Moreover, a two dimensional CFD model of the spent fuel rack immersed in the water tank is built. The fluid force is obtained by a transient analysis with the help of dynamics mesh method.

High Energy Arcing Faults (HEAF) have the potential to cause extensive damage to the failed electrical components and distribution systems along with adjacent equipment and cables within the zone of influence (ZOI). Furthermore, the significant energy released during HEAF event can act as an ignition source to other components within the area of the HEAF. In Japan, during the Great East Japan Earthquake occurred in 2011, the seismic induced HEAF fire event, which induced the whole damage of the multiple high voltage switchgears, was observed in Onagawa Nuclear Power Plant (NPP). In response, in August 2017, the NRA (Nuclear Regular Authority) in Japan amended the safety requirement for the power supply to consider the influence of the successive fire due to the HEAF event (hereinafter HEAF fire event). Therefore, it is urgently necessary to establish the design criteria to prevent the HEAF fire event, and enhance the experiment data of the HEAF fire event. In order to estimate the total arc energy during the HEAF event and obtain the threshold value to prevent the HEAF fire for the existed non-arc proof electrical cabinets, several series of three-phase internal arc tests with high (6.9kV class) and low (480V class) voltage electrical cabinets were executed. We executed internal arc tests with full scale high/low voltage metal-enclosed switchgear components (non-arc proof type, copper bus conductor), and evaluated arc energy, the mechanical damage of the cabinet and the surrounding equipment due to the impulsive pressure and the possibility of successive fire occurrence. In case of high voltage switchgear, when the arcing energy exceeded 25.3MJ, successive fire was identified. Especially, in the case where the arc flash was discharged in the circuit breaker room, a 2-second arcing duration in a three-phase short-circuit current with 18.9kA (measured arcing energy over 40MJ) caused successive fire which required extinguishment. On the other hand, in case of low voltage power center, when the arcing energy exceeded 19MJ, successive fire was identified. According to these demonstrative tests, this paper presents the evaluation method to estimate total arc discharge energy during the HEAF event for high and low voltage electrical cabinets.

Kashiwazaki-Kariwa nuclear power station of TEPCO is the largest nuclear power station in the world, and it has seven nuclear power plants. As the experience at Fukushima Daiichi nuclear power station accident in March 2011 involving concurrent core damage at multiple units, it is considered that the risk derived from hazards of Earthquakes and Tsunamis is relatively significant in Japan, and these events have a high likelihood of damaging multiple units simultaneously. Therefore, it is very important to grasp the multi-unit specific risk. Although there are some unique accident scenarios of Multi-Unit PRA, this paper focuses on the influence of radioactive materials released outside the containment vessel on the accident management of the adjacent unit. The events including core damage and loss of containment function should be considered as the causes of the release of radioactive substances, and operator’s operation or the like should be considered as objects to be adversely affected by them. It is necessary to incorporate that into PRA to confirm the effect on risk. It is very difficult in terms of the maturity of evaluation method and the calculation load to accurately incorporate consequences derived from time series of various events and complicated interaction into PRA model. Therefore, as the first step in evaluating the risk of influence of radioactive material release on the accident management, some streamlining efforts are implemented according to the purpose. For example, Kashiwazaki-Kariwa unit 6 and unit 7 were set as the target units for model simplification. We also assume the earthquake as the initiating event due to the strong common factor for multi units. Whether or not to be operable in the adjacent plant is set conservatively based on deterministic evaluation. PRA taking into consideration the radiation influence by multi-unit accident is compared with normal PRA. Some kind of Core Damage Frequency (CDF) such as CDF 1 (Core Damage Frequency at which the damage of one or the other of two unit occur), CDF 2 (CDF at which the damage of both of units occur) and CDF TOTAL (CDF at which the damage of one or more units occur: CDF 1 + CDF 2 ) are quantified, and the degree of this issue is provided. Although the change of CDF TOTAL was insignificant, the necessity of further study was shown from the viewpoint of the amount and timing of radioactive substance released due to an approximately 1.5-fold increase in CDF 2 .

The NSRR (Nuclear Safety Research Reactor) is a research reactor of TRIGA-ACPR (Annular Core Pulse Reactor) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity initiated accident (RIA) conditions. Under the new regulation standards after the Fukushima Daiichi accident, the research reactors are being regulated according to the risk of the facility. Graded approach is introduced in the regulation. In order to apply the graded approach, the radiation effects of residents living around the NSRI under the external hazards were evaluated and the level of the risk of the NSRR facility was investigating. This report is summarized for the result of the evaluation in case the safety functions were lost by the tornado, earthquake and following tsunami. As the result, the risk is confirmed to be low, since the effective dose of the residents has been below 5 mSv per event due to the loss of the safety functions by the tornado, earthquake and following tsunami.

Large amounts of highly contaminated water over 800,000 m 3 accumulated in the reactor, turbine building and the trench in the facility were generated from the nuclear accident of Fukushima NPS (BWR) caused by the Great East Japan Earthquake. At present, the cold shutdown is completed stably by the circulating injection cooling system (SARRY, KURION) for the decontamination of radioactive nuclides such as 134 Cs and 137 Cs using zeolites and crystalline silicotitanate (CST). Further, the Advanced Liquid Processing System (ALPS) is under operation for the decontamination of 62 nuclides such as 90 Sr, 129 I and 60 Co, etc. However, the adsorption behaviors of actinoids through the decontamination systems are complicated, and especially their adsorption properties for zeolites and CST, major inorganic adsorbents, are not yet clarified. In near future, the decontamination of actinoids leached from the crushed fuel debris will be an important subject. In this study, the practical adsorption properties of U(VI) for various inorganic adsorbents were evaluated under different solution conditions. The adsorption properties (distribution behaviors and adsorption kinetics) were evaluated by batch adsorption method; 19 kinds of inorganic adsorbents including zeolites and CST (crystalline silicotitanate) were contacted with U(VI)) solutions. The conditions of 5 kinds of U(VI) solutions were as follows; Solution 1: [U(VI)] = 50 ppm, initial pH = 0.5 ∼ 5.5 Solution 2: [U(VI)] = 50 ppm, [NaCl] = 0.1 M, initial pH = 4.0 Solution 3: [U(VI)] = 50 ppm, [CaCl 2 ] = 0.1 M, initial pH = 4.0 Solution 4: [U(VI)] = 4.84 mM, [NaCl] = 0.1 M, initial pH = 3.18 Solution 5: [U(VI)] = 4.86 mM, 2,994 ppm boric acid/30% seawater, initial pH = 4.25 The uptake (%) and distribution coefficient ( K d . cm 3 /g) were estimated by counting the radioactivity using NaI(Tl) scintillation counter and liquid scintillation counter. In the simple Solution 1, the K d values for zeolites increased linearly with equilibrium pH up to pH 7. The K d value for tin hydroxide had a maximum profile around pH 7 and a relatively large K d value above 10 4 cm 3 /g was obtained. In the presence of NaCl and CaCl 2 (Solution 2 and 3), relatively large K d values above 10 2 cm 3 /g were obtained, other than mordenite and clinoptilolite, and the effect of [Ca 2+ ] on U(VI) uptake was larger than that of [Na + ]. In Solution 4 containing high concentration of U(VI), the uptake(%) was considerably lowered, while that for zeolite A, X and Y was estimated over 20%. Similar tendency was observed in Solution 5, and, in the case of granulated potassium titanate, yellow precipitate was observed on the surface due to the increase of equilibrium pH up to 5.25. The adsorption behavior of U(VI) on inorganic adsorbents is mainly governed by three steps; ion exchange, surface precipitation of hydrolysis species and sedimentation depending on equilibrium pH, and hence it should be noted the change of U(VI) chemical species. These basic adsorption data are useful for the selection of inorganic adsorbents in the Fukushima NPS decontamination process.

More and more attentions have been paid to the impact of the earthquake on the reactor thermal hydraulics after the Fukushima accident. In order to figure out the seismic effect on the flow in primary loop, the natural circulation in primary loop of CPR1000 influenced by earthquake is modeled by learning from the study on natural circulation under ocean conditions. Natural circulation ability at earthquake, which includes longitudinal earthquake and transversal earthquake, is dynamically simulated by a program written with MATLAB, respectively. Besides, the safety of the system is discussed based on the calculated results. The research results show that, once the longitudinal earthquake occurs, in the earlier stage of earthquake, the peak and valley values of the coolant mass flow rate decrease gradually and the amplitude will be a constant as time goes on because of flow fluctuation in seismic condition. On the contrary, as for transversal earthquake, the peak and valley values of the coolant mass flow rate increase gradually in the earlier stage of earthquake. However, the amplitude will be a constant as time goes on, just like that in the longitudinal earthquake. For both of the two conditions, they have the same characteristics. The constant mentioned above grows up with the increase of the earthquake acceleration peak value and the period; the fluctuation period of mass flow rate is generally the same as the period of earthquake acceleration and the mass flow rate changes steadily; the average mass flow rate decreases slightly with the increase of the earthquake acceleration peak value and the period. Both of the vibration period and the peak of seismic wave have disturbing effect on reactor coolant system natural circulation. And the increase of the peak of the earthquake acceleration as well as the period of the earthquake acceleration will reduce the natural circulation ability. In the range of research parameters, the system is considered to be safe according to the calculation results of core coolant temperature.

In typical pressurized water reactor (PWR) plant, in case that one steam generator (SG) is dried out and cannot be credited for the primary cooldown, at least one reactor coolant pump (RCP) has to be operated in order to homogenize the primary coolant temperature distribution among loops when the plant is cooled down to the cold shutdown state. For example, an accident such as steam line break (SLB) and feedwater line break (FLB) leads to this situation. If the natural circulation condition is established due to unavailability of all the RCPs, the natural circulation in the primary loop connected to the affected SG would be interrupted in the plant cooldown phase. In this situation, the continuous cooldown disturbs the smooth depressurization because it leads to void generation at the top of the affected SG tube where the high temperature coolant is left. In addition, there is a possibility that all RCPs cannot be operated in case of the earthquake or the fire if the RCPs are not earthquake-proof and fire-resistant. Therefore the establishment of the cooldown procedure without RCPs operation under the temperature unbalanced condition among the primary loops can contribute to the safety enhancement for typical PWR plants. The several experiments have been already performed to observe the natural circulation phenomena under the temperature unbalanced condition. It has been reported that the plant can be continuously cooled down with smooth depressurization by stepwise cooling manner using MSRVs of the intact SGs. In this study, Mitsubishi Heavy Industries, Ltd. (MHI) performed the transient analyses to simulate the natural circulation cooldown test under the temperature unbalanced condition among loops performed by Large Scale Test Facility (JAEA ROSA/LSTF) using M-RELAP5, which was a modified plant system transient code by MHI based on RELAP5-3D. Based on the analysis results, the thermal hydraulic phenomena of natural circulation cooldown under the temperature unbalanced condition were investigated. As a result, the mechanism of natural circulation interruption was clarified, and this paper shows the outline of the cooldown procedure under the temperature unbalanced condition which could be applied to the PWR plants.

Spent fuel pool stores fuel assemblies that have been removed from the reactor in previous years, and the release of radioactive material to the external environment is easier than reactor core after fuel damage occurs. In the earthquake situation, the spent fuel pool itself and its accident corresponding mitigation measures are likely to fail at the same time, causing serious accident consequences. Based on the design and operation characteristics of typical CPR1000 units, this paper uses the probability safety analysis method to quantitatively evaluate the risk of spent fuel pool caused by seismic events, and identify the weak links in the design and operation of CPR1000 units. Quantitative analysis shows that the risk of the spent fuel pool caused by seismic events is relatively low relative to the core, with the main risks coming from the collapse of the fuel building and structural damage to the spent fuel pool. In addition, seismic events affect both the core and spent fuel pool, and the steam and radioactivity of spent fuel pools will have an impact on mitigation of core accidents.

For seismic safety evaluation method of nuclear power plant, nuclear power plant seismic margin analysis (SMA) and nuclear power plant seismic probability safety assessment (SPSA) are the most widely used methods. SMA is a method base on deterministic theory. Seismic capacity is valued by high confidence and low failure probability (HCLPF). Through the seismic failure logic of structure, system and components (SSCs), the method can calculated the HCLPF of the whole nuclear power plant, and verify whether the plant can withstand a SSE earthquake test. The SPSA method is the most widely used seismic safety assessment method based on probability theory. Through the analysis and quantification of earthquake accident sequence, a SPSA project can fully identify the seismic risk of nuclear power plant and seismic weak points. Also SPSA can guide the nuclear power plant seismic safety improvement. No matter which method is used to analyze the seismic safety of nuclear power plant, it is necessary to analyze and calculate the seismic fragility of the SSCs. SMA method needs to use a large number of HCLPF data, and seismic fragility analysis and calculation results is one of the main sources of HCLPF data. The SPSA method needs to use seismic fragility data of SSCs which are list in the seismic equipment list (SEL) as input data, so that it can support the quantitative analysis of the risk assessment model. Because of the existence of uncertainty, the seismic fragilities cannot be put to directly logic operation. This brings great difficulty to the using of fragility data. In the paper, the logic operation method and the uncertainty analysis method of seismic fragility is studied, and the calculation program is compiled based on the Monte Carlo method. In this paper, a program is used to calculate the case. The performance of the program is verified and the uncertainty of the system fragility is analyzed. Due to the existence of uncertainty, the fragility cannot put into the numerical calculation directly. In this paper, the calculation method of the failure frequency of components is studied, and the corresponding program is developed by using Monte Carlo method. In this paper, a program is used to calculate the failure frequency of the components under different ground motion levels, and the uncertainty of the failure frequency is also studied.

The present paper provide an available FEA method for the structural analysis and seismic design of pipelines coated with flexible shielding materials, in nuclear power plant, subjected to earthquake action. The modal and spectral analysis are carried out by using FEM software ANSYS. By comparing the numerical simulation results of dynamical behavior of the pipelines with or without flexible shielding materials, the effects of the flexible shielding material on the seismic performance of the pipeline system in nuclear power plant are discussed. Finally, a local supporting constraint for reducing seismic effects is put forward, which can significantly decrease the maximum stresses in the pipelines coated with flexible shielding materials.

The extreme hazards such as large floods, earthquakes, fires or explosion may be lead to an extensive damage or large scale incident in NPP, for instance, loss of command and control, loss of most instruments or equipment which are used for accident mitigation. The typical situations are loss of control room and alternate shutdown capabilities, or loss of AC and DC power, or all of them. Extreme Hazards Mitigation Guideline (EHMG) is developed as the coping strategy for the above extensive damage situations according to the requirements of the characteristic of CPR1000 NPP and the modifications implemented after Fukushima accident. The technical scheme and frame of EHMG includes two modules: EH-IRs (initial response) and EH-MGs (mitigation guideline). And then based on the typical accident analysis, the critical diagnostic parameters and mitigation strategies were made for the EHMG. EH-IRs contains the following contents: Rebuilding the command and control, Off-Site and On-Site Communications, Initial Operational Response Actions, Initial Damage Assessment. EH-MGs contains the following strategies: Manually operating auxiliary turbo pump, Manually depressurizing SGs and injecting into SG, Alternate makeup to RCS, Controlling containment conditions, Alternate Makeup to Spent fuel pool, Alternate Makeup to water storage tank, Containment flooding and so on. The strategies in EHMG can cover both prevention and mitigation phases of severe accident. The first EHMG in China has been validated and implemented in Hong Yan He NPP in September, 2016.

This paper explains the strategy of our company (Tokyo Electric Power, TEPCO) regarding means of long-term heat removal from the primary containment vessel (PCV) of Units 6 and 7 (ABWR) of the Kashiwazaki-Kariwa Nuclear Power Station in a severe accident. If the PCV continues in a high-temperature state for a long time, the strength of the PCV concrete will decline, and the risk of being affected by an earthquake will increase. Therefore, it is crucial for safety to cool the PCV and reduce its temperature to the maximum working temperature or lower. TEPCO provides a means of cooling the reactor pressure vessel (RPV) and PCV called the alternative coolant circulation system (ACCS). This system uses the heat exchanger of the residual heat removal (RHR) system, the make up water condensate (MUWC) pump, and alternative heat exchanger vehicles. By using these measures, it is possible reduce temperature in the PCV over the long term to the maximum working temperature (design value) or less, even in severe accident scenarios such as a large LOCA + ECCS function failures + SBO (station blackout). This function has quite high reliability, but in a scenario where these measures cannot be used, expectations are placed on the filtered vent (FV). However, due to FV characteristics, it is impossible to reduce to below the saturation temperature of 100°C at atmospheric pressure using FV alone, and it will be necessary in the medium/long-term to cool the PCV while also restoring the cooling equipment. Therefore, the following restoration operation of PCV cooling and its dose evaluation were studied. (1) RPV heat removal by restoring the RHR system (2) RPV and PCV heat removal using a portable pump employing a portable heat exchanger (3) RPV and PCV heat removal using the suppression pool water clean up system (SPCU) employing portable heat exchangers (4) RPV heat removal using the clean up water system (CUW) By clarifying beforehand issues such as feasibility of these systems, the on-site environment for restoration measures, and the necessary gear/systems, the authors were able to secure means of long-term cooling of the PCV, and further enhance PCV reliability.

Since the responses of Liquid storage tanks (LST) include the liquid motion, structural motion, and fluid-structure interaction (FSI) under the earthquake ground motion, it is a difficult problem about shaking table test model for satisfying similarity ratio requirement. The two experimental models, namely satisfied FSI similarity ratio model (FSI model) and unsatisfied FSI similarity ratio (un-FSI) model, respectively, are presented in this paper. The PCS storage tank of AP1000 Nuclear Power Engineering, as the study project, is studied by Adina software, and the responses, such as stress, stain, and displacement, etc., of prototype model, FSI model, and un-FSI model are compared. The results provide that when researchers study the rigidity LST dynamic response parameters and vibration characteristics, such as displacement, acceleration and liquid wave height, the tank wall stress response parameters etc., FSI model should be used.

Based on the shock damage propagation distances and the median fragility limit of the equipments, the NEI 07–13 employs the shock damage rules for determining the potential for affecting safe shutdown and fuel cooling equipments. However, the NEI 07–13 does not provide more detailed guidance for performing the shock damage assessments, because both the shock damage distances and the methodology for developing the median fragility limit are not provided in NEI 07–13. This paper discussed methodology developed for performing simplified assessments for shock effects considering the material nonlinearity of the impact zone and the soil-structure interaction. Three different models (i.e., linear model, nonlinear mode, and SSI model) were developed to calculate the in-structure shock response. The results of the linear model show the shock response due to aircraft impact would completely propagate from the center of initial impact zone and then along a structure pathway (e.g. wall, floor, basemat) to the in-structure without any energy dissipation. As a result, the in-structure shock response spectra are considerably higher than the spectra associated with the design-basis earthquake in the high frequency range. In order to reduce the shock effects on the in-structure safety-related systems and equipments, the material nonlinearity of the impact zone and the soil-structure interaction were incorporated in the dynamic analysis. The numerical results show that both the material nonlinearity and the soil-structure interaction would obviously absorb the energy of the shock waves, so the in-structure shock response spectra were reduced due to these two factors. Finally, the representative shock response spectra were compared with those used in the seismic margin assessment in order to assess specific equipment survival.