Many empirical formulae have been proposed for evaluating the local damage to reinforced concrete structures caused by rigid projectile impact. Most of these formulae are based on impact tests perpendicular to the target structures. To date, few impact tests oblique to the target structures have been conducted. In this study, we aim to obtain a new formula for evaluating the local damage caused by oblique impacts based on previous experimental and simulation results. We analyze and simulate the local damage owing to impact by deformable projectiles. The experimental and simulation results were in good agreement and confirmed the validity of the proposed analytical method. Furthermore, the internal energy of the deformable projectile absorbed upon impact was approximately 60% of the total energy. In comparison to a rigid projectile, it is possible to reduce the impact load and consequently the damage to the target.

Investigations of corrosion behavior of austenitic 1.4970 (15–15 Ti) steel in Pb-Bi eutectic at 400–550 °C show an effect of structural state of material with identical composition on the depth of solution-based attack. Structural boundaries play a role of active paths along which the solution-based liquid-metal attack develops preferentially. In this view it is important to have quantitative and qualitative information about grain boundary type distribution in material and state of boundaries with respect to the accumulated strains. The EBSD analysis performed on 1.4970 steel in solution-annealed and cold-worked (40% reduction) states indicates that deformation increases substantially the total length of strained boundaries. The increase in fraction of active diffusion paths results in acceleration of corrosion attack on steel via solution.

This paper introduced the design and research of spent resin conical dryer which was based on the analysis of the thermal decomposition characteristics of resins. The drying experiment of non-radioactive cation exchange resins and anion exchange resins was also carried out in this study. The result showed that the water content of resins reduced from about 55%(wt) to 8.5%(wt) and the volume reduction ration reached 2.17 with a drying end temperature of 90°C, which preliminarily verified the feasibility of the vacuum drying process and conical dryer device for treating radioactive spent resins.

The UK advanced gas cooled reactors (AGRs) use a graphite core with carbon dioxide gas as the primary coolant. There is a diaphragm above the core which separates re-entrant gas at lower temperature and higher pressure from that leaving the channel guide tubes at reactor outlet temperature. This diaphragm is known as the hot box dome. The dome is perforated to facilitate the passage of fuel and control rods into the core. The dome is fabricated in carbon-manganese steel and incorporates a number of full penetration welds which are post-weld heat treated (PWHT). The dome’s upper surface is insulated to protect it from gas at high temperature, intended to maintain the dome at a temperature of below 380°C. Since dome failure could conceivably result in gas by-passing and, hence, failing to adequately cool the core the original safety case claims that gross failure of the dome is incredible. More recently potential failure modes of the dome have been reviewed and various dome weld failure scenarios have been analysed and assessed to demonstrate a tolerance to the consequences of complete failure of certain welds. On this basis the dome could be shown to satisfy a lesser classification of high integrity, although no claim to reclassify the region has been made. Through-life temperature monitoring is carried out to demonstrate that the peak dome temperature remains below 380°C. This has shown evidence of rising temperatures, believed to arise from a reduction in the effectiveness of the upper surface insulation, an effect that was acknowledged by the original design. Work to investigate this effect has developed the understanding of the dome thermal environment which is far more complex than previously thought. The hottest parts of the dome are far smaller and more localised than previously thought, and lie immediately above the monitored locations. In order to support a case to operate for an extended life, it is now proposed that the upper temperature limit could safely be increased to 390°C. Structural integrity analyses and assessments have been carried out to support the proposed increase to 390°C and include a demonstration of the absence of a cliff-edge effect by assessing cases with the hottest parts of the dome at temperatures of 400 and 410°C. The work seeks to demonstrate adequate margins of safety against all potential failure modes. ASME III code assessment against primary stress limits has been used to guard against failure by plastic collapse and/or creep rupture. Creep-fatigue initiation assessments have been used to demonstrate margins against the formation of defects using the EDF Energy high temperature assessment procedure, R5. This has enabled the consideration of potential defects to be confined to those that might have formed during welding or PWHT and have been missed by extensive pre-service inspections. Notwithstanding the low likelihood that any such defects exist, with high confidence it may be postulated that any that do would be located in welds and be of limited size. Defect tolerance assessments have been carried out, including the calculation of limiting defect sizes in accordance with the EDF Energy R6 procedure and the growth of postulated defects by creep and fatigue using R5. Other failure and degradations mechanisms have been considered and eliminated as a potential threat by drawing on reviews of relevant operating experience on other reactors with similar materials and environments, and material property data from long term tests. This paper describes how the multi-facetted programme of work, which proposes a modification to an existing safety case, has been devised to explicitly address all conceivable modes of failure and demonstrate a robust argument against each one.

Up to now, two kinds of filler metal with or without nickel element for submerged arc welding have been largely used in the reactor pressure vessel (RPV) manufacturing. In order to study the effect of nickel element on weld metal properties of SA-508 Gr.3 Cl.1, submerged arc welding material with nickel (AWS classification F8P4-EGN-F2N, F2 for short) and welding material without nickel (F8P4-EA3N-A3N, A3 for short) were used; and conventional mechanical properties, low-cycle fatigue test, and proton irradiation analysis of the two weld metals were studied. Results show that the mechanical properties of the two different weld metals are similar, except that the Charpy V-notch impact property of the weld metal with nickel is better than that without nickel; the micro-structures of F2 and A3 weld metals are both composed of ferrite base and granular bainite, but the columnar grain size of F2 weld metal is smaller relatively, which results in better impact property. In addition, the irradiated A3 weld metal has fewer dislocation loops than the irradiated F2 weld metal after the same proton irradiation dose; the irradiated weld metals both have higher micro-Vickers hardness than before.

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.

This paper studies the development method of integration between design and analysis. Development is based on the current design platform of nuclear power plant of SNERDI - PDS. Because of complexity and iterations, the calculation and analysis work often takes a lot of manpower and drags down the entire design progress. To improve efficient and accuracy, the integration between design and analysis became the trend of technology development. Some design systems, such as Ansys Workbanch, ProE, Catia and so on, can be integrated modeling and analysis in a unified platform. However, these systems can be only applied to product design, but not in nuclear power design. At present, CAP series NPP mainly uses PDS as the layout design platform of all disciplines. This paper describes the bidirectional data sharing between design and analysis, effectively expand the scope of data utilization, and enhance the efficiency of the analysis.

This paper proposes the experimental research for the performance of the air eductor used in main control room (MCR). The air eductor is used for emergency ventilating in advanced passive pressurized water reactor in accident. The compress air is supplied to the eductor as a power source and the indoor air is suctioned to the eductor. The performance of the eductor is related to the habitability of MCR. The entrainment ratio and the air pressure of discharge side are the main concerned performance. The entrainment ratio is a value that resulted from the compress air flow rate divided by the suction air flow rate. A test system was set up to test the performance of eductor. The experimental results show that the entrainment ratio of rectangle nozzle with compress air pressure 0.76MPa, 0.80MPa and 0.83MPa were 15.02, 15.04 and 15.06, respectively.

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.

The area radiation monitoring system can supply information of the abnormal radiation levels in a nuclear power plant quickly, give an alarm on site explicitly, and display the dose rate value and alarm information remotely. Thus, it can help the radiation protection officer to take proper actions timely to protect personnel from excess irradiation. According to the experience and technology from the pressurized water reactors (PWRs), the area radiation monitoring system in nuclear island of high temperature gas-cooled reactor pebbled-bed module (HTR-PM) which is a modular high temperature gas-cooled reactor demonstration power plant in China has been designed. It is used to monitor the external radiation exposure dose rate at the passageway in the control areas, operational areas, equipment rooms, etc. The design scheme of the area radiation monitoring system in nuclear island of HTR-PM is presented, which can be divided into two parts, the stationary area γ radiation monitoring subsystem and portable radiation dose rate monitoring subsystem. The operational principle of the system, the layout of the area radiation monitors, the main characteristics of the instruments, and the communication logical framework of the system are introduced in this paper.

This paper continues our recent discussion on piping vibration and practical measures for preventing vibration-related damage in nuclear power industries. Our emphasis is on an empirical approach, which attempts to estimate a so-called “dynamic susceptibility” at various locations in a piping system. This approach uses a dynamic susceptibility factor (DS), which is a quotient of the modal stress to the modal velocity, as an indicator of the risk levels of vibration to measure the vibration sensitivity for excitation sources in a given frequency interval of particular interest. In the present paper, Benchmark examples tested by the general purpose finite element program ANSYS, and commercial piping programs CAEPIPE and PIPESTRESS, are presented and the potential of using DS parameters as a screening tool for determining “potentially-large” alternating stresses is illustrated. It is demonstrated that, combined with knowledge of typical vibration sources, this is a practical and cost effective way for forming a base for the vibration control prior to installation and for the planning of post-installation vibration monitoring of a piping system under operation.

Nuclear radiation can induce hardening and embrittlement of RPV steel. Solute defect copper rich precipitate / clusters is considered harden the base steel. Reduce or delay the irradiation hardening and embrittlement of RPV steel could be effective to ensure the safe operation of nuclear power. Furthermore, the expected target of the improved RPV steel will assist consummate nuclear power evaluation to decide whether feasible to extend the service lifetime of nuclear power plant from 40 to 60 years facing at present in China nuclear industry. In order to study the precipitate phase Cu rich clusters, but different from the traditional methods, the present work employed a new technique, electromagnetic induction heating to produce a high temperature gradient along the model rod Fe-1.2wt.%Cu alloy sample. The results exhibit that this method can speed up the research process and is convenient to study the transformation in microstructure and hard / brittle defect phase formation of copper solute rich clusters. In addition, the change of Vickers hardness along the high temperature gradient assist denote the transformation in microstructure and the Cu rich clusters in quantity roughly. And large rolling deformation can promote formation of Cu rich cluster.

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.

A logarithmic relationship for the oxide film thickness of Nickel-based alloys in high temperature water versus immersion time or dissolved hydrogen (DH) is identified based on the compilation and analysis of the experimental data reported in some literature. A theoretical equation is used to quantify the correlations between thickness and time or DH. The logarithmic correlation between inner layer thickness and iron content in alloy matrix and the linear correlation between outer layer thickness and iron content in alloy matrix are primary identified by fitting the experimental data in our own work. The correlation of oxidation kinetics and stress corrosion cracking growth rate is discussed.

Additive Manufacturing (AM) can fabricate 3D near-net-shaped functional parts using unit materials, such as powder or wire. Additive manufacturing’s computer-aided design offers superior geometrical flexibility. The near-net shaping capability also reduces materials waste and increase manufacturing efficiency significantly. These benefits make AM desirable for critical industry applications, such as art, aerospace, ground transportation, and medical. Confident utilization of the technology requires thorough understanding of the AM materials, ensuring both structural integrity and performance requirements are met or exceeded. Safety and economics are essential to nuclear power plant. In this study, mechanical properties of a ferritic steel fabricated by electric melting additive manufacturing (EMAM) technique are studied and compared with ASME SA-336 Gr.F12, which applied to nuclear main steam line penetration, the results are systematically presented and discussed. Key technical issues of application of AM to manufacturing nuclear components are also discussed.

The guide tube (GT) representation in the reactor-internal system model has been crude and inaccurate in the past because the lack of experimental data on a) the rotary stiffness of the hold down bolt joint and b) the rotary stiffness of the support pins. In addition, the slotted enclosure and the complex interior in the continuous section further complicated the modeling accuracy. In this paper, the static and dynamic characteristic of the domestically manufactured CAP1000 guide tubes are studied both experimentally and analytically. Using the stiffness data obtained from the tests and combined with the enclosure stiffness calculated from the detailed 3D finite element models, i.e., one for the intermediate section and one for the continuous section, a simplified model of the GT was established. Using the simplified model, the dynamic characteristics (natural frequency and modal shapes) were compared to the dynamic test results in the aerial condition. It can be concluded that the simplified guide tube model is believed to be the most accurate one, by far, to be used in reactor-internal system analyses.

In this paper, the structure configurations of the in-containment refueling water storage tank (IRWST) of Chinese 3 rd generation nuclear power plants (NPPs) was described firstly. Then, the general structural calculation for several loads, especially thermal load, were presented, as well as the stability evaluation of IRWST base-slab. The effect from fluid-structure interaction was also considered in the calculation to evaluate the design margin of IRWST. Finally, structure strength evaluation was performed for construction load case.

The determination of Design Ground Motion time history (or response spectra) is the primary and critical step to derive correct Seismic Design Inputs for a Nuclear Power Plant (NPP) design. Historically Design Ground Motion (design SSE input) for a NPP was determined by early version procedure provided in RG 1.60. It was based on a theory of deterministic approach; the resulting ground motion is given in acceleration response spectra located at free surface of a site. As a transition point, 1997 was the year where new procedure was developed and recommended in RG 1.165 based on the new theory of SSE ground motion probabilistic approach. RG 1.165 was authorized for application on all new NPPs’ design after 1997. With the advancing of PSHA approach, RG 1.165 was withdrawn and replaced with new RG 1.208 in 2008. RG 1.208 established an effective way through the similar probabilistic approach used in RG 1.165 by improving PSHA method. Both RG1.165 results and RG 1.208 results are focused on addressing site-specific design, its Ground Motion Response Spectra (GMRS) and Ground Motion Time History (converted from GMRS) are used as design inputs to specific Nuclear Island (NI) seismic design. To accomplish a Standard Design Certification, the RG 1.60 DRS is used to develop the Certified Seismic Design Response Spectra (CSDRS) by modifying control points on original RG 1.60 curves to broaden the spectra in higher frequency range. In reality, CSDRS serves as a good approach to define DRS and Design Ground Motion Time History for standard design of new NPPs in current timeframe, hence envelop the site-specific GMRS given in RG 1.208. In this paper, through the comparison of above US NRC regulatory requirements and Chinese regulatory requirements, gives recommendations on the determination of Design Ground Motion Response Spectra (or Time History), which serves as the basis for deriving seismic design inputs at required specific location (e.g. the bottom of NI foundation level) for potential “GEN III & Plus” plants in China.

The fire and explosion assessment is necessary for NPP to evaluate the effect caused by malicious large commercial aircraft impact. This report focuses on the fire and explosion effect on shield building structure of nuclear island according to the impact screening. With the investigation and analysis for the fire and explosion of aircraft impact, a simple and effective method was established. With this method the pool fire and fire explosion are analyzed and their effects on shield building structure are assessed based on the temperature and pressure distribution obtained from the FEA model.

Normal residual heat removal exchanger (RNS HX) is used to remove reactor residual heat during the second stage of reactor shutdown. A hypothetical over-pressurized accident, during which RNS HX would suffer the normal operating pressure and operating temperature of RCS should be considered during the design of RNS HX. This paper introduces Bolt-Flange-Gasket seal structures for RNS HX, considering the leakage-proof during over pressurized situation and some other operating-related aspects. At tube side, Metal-to-metal contact (MMC), with which a mass energy could be stored in metals during bolt preloading and could release when inner pressure increases, is used, so that the energy stored in metals could resist high pressure to prevent the joint disjoin. Another feather of tube side joint is that a slight angle is designed at the periphery of tubesheet. At shell side, a segmental gasket, which is divided into two or more pieces rather than a whole circle, is used to seal the joint. As the gasket pieces are installed from the side of tube bundle, the shell does not need to be removed away completely from the bundle in order to install a circle gasket, which could facilitate gasket changing process and save the operating space.