A probability of rupture for WWER-1000 main piping was calculated based on the Failure Assessment Diagram (FAD), treating material properties of welds (the most likely zone for crack growing and its nucleation) and crack morphology parameters as stochastic values. In order to perform probabilistic calculations, Critical Temperature of Brittleness (CTB, WWER’s analogue of PWRs transition temperature which used to index the Fracture Toughness curve) and Yield (Ultimate) Strength as well were fitted by normal distribution, based on experimental data taken from the manufacture documentation of Ukrainian Nuclear Power Plants (NPPs). A set of calculations were conducted for Normal Operating conditions (NOC), Safe Shutdown Earthquake (SSE) and several emergency situations like: shaft jamming of a reactor cooling pump and break of the piping’s, connected to the considered ones (a set of LOCA events). Based on static and dynamic calculations, the most loaded zones were selected, where the cracks were postulated. Crack opening area was calculated according to original developed procedure, which accounts for membrane and linear stress components through the wall thickness. The Henry-Fauske flow model is used with modified parameters accounting for crack morphology as a normally distributed random variable. It is an important part of analysis, since different crack types have great differences in friction, bend protrusion and flow length parameters. The rupture probabilities for Main Circulating Piping were calculated with accounting for thermal aging effect. It was proven, that crack morphology parameters highly affect the leak rate and its distribution becomes more scattered. Among the mechanical characteristic, a Fracture toughness has more influence rather than Ultimate of Yield strength.

Safety-related items in nuclear power plants are now generally placed separately from the non-safety-related items, but it was not strictly required before. Therefore, it is very important to study whether the non-safety-related items will affect the safety-related items when they are dropped down in an earthquake situation, which determines the safety of a nuclear power plant and its future life extension applications. This research was based on the cooling water system room with the safety and non-safety related items installed together, as an example to study whether the non-safety-related items such as vent pipes and DN50 fire fighting pipes arranged above will damage the DN300 pipes and valves arranged below when earthquakes occur. For the experiments, the relative positions of objects in the room was reproduced by 1: 1. The pressure-holding performance of the pipe was used as a criterion for the damage. The research results of the experiments show that when the 10-meter-long DN50 pipe was dropped from the position of 8 meters height and the 8-meter-long vent dropped from position of 3.6 meters height, they do not affect the integrity of the DN300 valve and pipe below. After the experiment, pressure drop in two hours for the pipe is less than 0.1%. The main body of the valve does not fail neither. The numerical simulation study also shows that there is no failure phenomenon in the simulation as well. Compared with the test results, the impact acceleration and the vent deformation both have the same trend.

In Japan, most structures on the ground surface need seismic countermeasures because of frequently earthquakes. On the other hand, vibration isolation devices are applied to precision or important equipment in several facilities that dislikes vibration in order to reduce daily vibration. In general, vibration isolation devices are intended for high frequency and small amplitude range. However, it is difficult to cut off both vibration region caused by flying object collision and seismic motion with existing technologies. The authors propose insulation of equipment and vibration transmitted through the floor by floating equipment, and have. We have devised and built an air floating device that operates when a trigger input is applied to save the energy of this dynamically acting device. It was estimated by numerical calculation that the aero floating device keeps lifting stably in the condition with the air pressure in the auxiliary air chamber about 75 to 80 kPa. The performance specifications of the proposed device were verified from shaking table test. As a result, the effect of reducing the maximum acceleration by about 1/5 against the seismic motion of El Centro NS, Taft NS, Tohoku NS, and Hachinohe EW was confirmed by floating the mass on the frame assuming the equipment. From the obtained power spectrum diagram (PSD) of the response acceleration, it was confirmed that all frequency components up to 25 Hz is reduced by using proposed aero floating base isolation device.

The performance of the Reactor Core Isolation Cooling (RCIC) system under beyond design basis event (BDBE) conditions is not well-characterized. The operating band of the RCIC system is currently specified utilizing conservative assumptions, with restrictive operational guidelines not allowing for an adequate credit of the true capability of the system. For example, it is assumed that battery power is needed for RCIC operation to maintain the reactor pressure vessel (RPV) water level — a loss of battery power is conservatively assumed to result in failure of the RCIC turbopump system in a range of safety and risk assessments. However, the accidents at Fukushima Daiichi Nuclear Power Station (FDNPS) showed that the Unit 2 RCIC did not cease to operate following loss of battery power. In fact, it continued to inject water into the RPV for nearly 3 days following the earthquake. Improved understanding of Terry turbopump operations under BDBE conditions can support enhancement of accident management procedures and guidelines, promoting more robust severe accident prevention. Therefore, the U.S. Department of Energy (DOE), U.S. nuclear industry, and international stakeholders have funded the Terry Turbine Expanded Operating Band (TTEXOB) program. This program aims to better understand RCIC operations during BDBE conditions through combined experimental and modeling efforts. As part of the TTEXOB, airflow testing was performed at Texas A&M University (TAMU) of a small-scale ZS-1 and a full-scale GS-2 Terry turbine. This paper presents the corresponding efforts to model operation of the TAMU ZS-1 and GS-2 Terry turbines with Sandia National Laboratories’ (SNL) MELCOR code. The current MELCOR modeling approach represents the Terry turbine with a system of equations expressing the conservation of angular momentum. The joint analysis and experimental program identified that a) it is possible for the Terry turbine to develop the same power at different speeds, and b) turbine losses appear to be insensitive to the size of the turbine. As part of this program, further study of Terry turbine modeling unknowns and uncertainties is planned to support more extensive application of modeling and simulation to the enhancement of plant-specific operational and accident procedures.

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.

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 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.

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.

The reactor scram function realized by the rapidly dropping of control rods ensures safety when the reactor accidents (loss of electricity and earthquake, etc.) happen. In the thorium base molten salt reactor (TMSR - SF1), the rod drop time is obviously affected by the resistance which produced in the molten salt as its high density and viscosity. In this paper, the drop time of the control rod is obtained by the theoretical and experimental methods for comparison. Firstly, the drop time is analyzed both in air and water condition with calculation and experiment. And the method used for the resistance calculation of the rod during dropping is verified. Secondly, the similarity criterion is adopted to calculate the drop time in molten salt condition. The study shows that: 1) In air and water condition, the calculation is coincidence with the experimental results within the maximum error less than 2 %. 2) The drop time of the rod in molten salt is 2.8 s with a dropping height 2.4m in reactor, which satisfy the safety requirement of the control system. 3) It is necessary to use another buffer beside the disc spring to protect the driving mechanism of the rod during the rod dropping.

The storage tanks in nuclear facilities has a significant impact on the safety of the reactor and the radiation shielding, so its mechanical property analysis has been widely concerned in the field of engineering and scientific research. Meanwhile, the storage tank is usually filled with gas and liquid medium. In the presence of external disturbances (such as external force, displacement, earthquake etc.), the position and structure of the vessel changes, that lead to changing of the gas-liquid interface. This characteristic can make the storage tank system as a tightly fluid-structure coupling system. In this paper, a storage tank which stored radioactive gas liquid medium is choosing to study such fluid-structure coupling system phenomenon, and a typical dynamic seismic condition is assumed. A two-way fluid-structure coupling method is used with CFD (Computational Fluid Dynamics) and FEM (Finite Element Method) numerical method. The study considered interaction between structure and two phase turbulent fluid. In FEM calculation, the time history seismic acceleration load is applied to the support of tank, and the flow loading coming from fluid medium is applied to the wall of tank which is send from CFD code. Then, the structure displacement which is calculate by FEM is transferred to CFD code. In CFD calculation, multiphase fluid numerical model is applied to simulate the flow characteristics of gas-water two phase fluid, and the turbulent properties are also considered in the calculation. Mesh deformation method is used to simulate the displacement of flow passage boundary which is send by FEM code. After CFD calculation, flow loading is transferred to the tank wall of FEM code again. Such loop of FEM and CFD calculation continues to go on with the seismic time history, the response characteristics of the tank will be solved. In order to evaluate the difference between the above method and the traditional analysis method. An independent calculation used added mass approach is carrying out, in which the effect of steady state water is applied to the wall of the vessel, and this load will not change with the earthquake. All others load and constraint mode are same with the above method. According to the two-way fluid-structure coupling analysis, the detailed characteristics of liquid free surface distribution and structural response of the vessel are obtained. The results show that the response vibration amplitude of the tank structure increases with the earthquake, and the response is mainly affected by the liquid sloshing. According to comparative analysis, the advantages of coupling method are proved. The method from this study can be used for the same type of analysis.

For nuclear power stations, the main function of Essential Service Water System (ESWS) is to discharge the waste heat from reactor core and spent fuel pool to the environment controllably, which is directly related to the safety and economy of nuclear power stations. Usually ESWS use open water from sea, rivers, lakes, reservoirs, as heat transfer medium. Extremely harsh environmental conditions may disable system functions and even lead to ESWS failure, directly reduce the safety and economy of nuclear power stations, and cause serious nuclear accidents. Failure of ESWS is one of the main reasons that lead to the Fukushima nuclear accident because of the loss of electricity after the earthquake and tsunami. Based on the typical ESWS configuration and conditions of serving nuclear power stations in China, the influence of environmental conditions on the function of water system is studied, and the corresponding measures are analyzed. These conditions can be divided into three categories: temperatures, water levels, and physical and chemical characteristics. Temperatures affect cooling characteristic of ESWS mainly. Nuclear power stations in tropical areas need to focus on cooling capacity might be reduced by high temperature. Those in cold region need attention to excessive cooling and freezing problems caused by low temperature. The influence of water levels is mainly fluid transport capacity and selection of equipment to ESWS. When the range of natural water level is too wide, designers shall consider measures to narrow it, such as the construction of highly reliable reservoir. Inland nuclear power stations shall try to ensure the reliability of ESWS; prevent water level changes beyond the scope of design caused by drought and flood disasters. The effects of physical and chemical properties are derived from the open water characteristics, including high salinity, high chloride ion concentration, carrying solid particles, suspended solids, and aquatics, and so on. These characteristics will cause the equipment and pipeline eroded or even damaged, aqueducts of intake and output jammed, heat exchangers of the final heat sink weakened and other negative effects, resulting in ESWS performance decline. Some of these factors are the characteristics of station site natural environment, some others are changes caused by human activities. Some factors are sustained, long-term; some others may be sudden, temporary. Influence on these factors need to be taken measures from many aspects, including structure, biological disinfection, special materials and equipment, environmental protection measures around the nuclear power station, and so on. On the whole, the environmental factors that affect ESWS in the nuclear power stations are wide, and the influence mechanism is more complex. These factors ultimately act on ESWS, but most of them cannot be banished inside of ESWS or the final heat sink system. Against the negative effects from environmental conditions, it has to be considered from all steps in the engineering of nuclear power stations, including design, construction and operation. All the measures shall be suitable to local conditions, in order to ensure the safety and economy of nuclear power stations.

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.

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.