A new conceptual design of intermediate heat exchanger (IHX) is proposed for application to the gas turbine high temperature reactor system (GTHTR300C) which is being developed by Japan Atomic Energy Agency (JAEA). The GTHTR300C cogenerates hydrogen using the iodine-sulfur (IS) hydrogen production process and electric power using gas turbine. The IHX is used to transport high temperature heat from the nuclear reactor to the hydrogen plant. The IHX proposed in this paper is a horizontal design as opposed to conventional vertical design. Therefore, JAEA investigated the advantage of the horizontal IHX and the economic evaluation when scaling up from conceptual design of high temperature engineering test reactor (HTTR) / IHX to GTHTR300C. To meet the performance requirement, both thermal and structural designs were performed to select heat transfer tube length, tube bundle diameter, insulation thickness, and the length of shell support in a horizontal pressure vessel. It is found that the length of the heat exchanger tube can be shortened and the superalloy-made center pipe structure can be eliminated, which results in reducing the quantity of construction steel by about 30%. Furthermore, the maximum stress concentration in the tubes is found to be significantly reduced such that the creep strength to withstand continuous operation is extended to 40 years, equaling the nuclear reactor life time, without replacement.

The Brayton cycle with supercritical carbon (S-CO 2 ) as working medium is one of the most promising new nuclear power systems. Turbine is the key device during the working process in the Brayton power cycle. The turbine structural presents small size and extremely high rotational speed for the special physical properties of S-CO 2 , which increase the difficulty for the structural design and strength safety significantly. According to the aerodynamic design and optimization results of 200 kW S-CO 2 radial inflow turbine, this paper proposes a detail structural design and analysis method for turbine impeller. Based on the three-dimensional blade profile data and meridional planes data, key structural design parameters are chosen and the parametric geometry model is established by CAD tools. On this basis, numerical simulation models of turbine are established to analyze the structural strength in detail. Then the influence of parameters on the turbine impeller strength is studied by a series of finite element numerical procedures. The influence mechanisms of key structural design parameters on impeller strength are discussed. Moreover, the final model of turbine impeller is obtained by parameter comparison and selection. The results show that for the initial model, the maximum von-Mises equivalent stress is 400.10 MPa, the maximum radial deformation is 0.0333 mm and the maximum axial deformation is 0.0770 mm. For the final model, the maximum von-Mises equivalent stress is 294.26 MPa, the maximum radial deformation is 0.0279 mm and the maximum axial deformation is 0.0769 mm. The maximum von-Mises equivalent stress and maximum radial deformation of structural decreases 26.45 % and 16.22 % respectively compared with the initial model. As a result, the impeller structural strength safety margin is obviously improved by the parameter analysis.

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.

Since first Nuclear Power Plant (NPP) in operation in early 1990s, China has built dozen of Light Water Reactor (LWR). Originated from Pressure Water Reactor (PWR) technology, almost all these plants, such as Qinshan, Dayawan, and Tianwan, were designed with a Pre-stressed Concrete Containment Vessel (PCCV). With the application of Third Generation (GEN III) reactor technology in recent years China is designing and building more Gen III NPPs. As the first kind of passive safety GEN III + technology, AP1000 uses double-layered containment structures comprising steel containment and shield building. GEN III EPR use similar double-layered containment structures, the outer layer is reinforced concrete shield structure, the inner layer is PCCV. In the traditional containment design, the design load and load combinations are determined based on different NPP operation conditions. This means the basic design principles of defense-in-depth and single failure criteria are incorporated into containment structure design through the consideration of all operation-related design basis loading conditions. However The Indian Ocean tsunami and the Fukushima earthquake and tsunami followed by the Fukushima Daiichi accident illustrated the hidden potential vulnerability in current NPP design when subjecting to Beyond Design Basis Event (BDBE) loading conditions. Containment, as the final physical shielding barrier to protect plant from potential radioactive releasing in accidental events, shall be designed to be not only functional during design basis accident conditions, but also be able to perform its containment and shielding functions, maintain the integrity of pressure boundary during BDBE and their loading conditions. Aiming on containment and shield structure design for GEN III NPPs, this paper summarized preliminary research results on BDBE design. Considering the occurrence frequency and impact severity to containment structure, Beyond Design Basis Seismic (BDBS) is taken as the investigation focus in this paper. The summary is based on an investigation of the BDBS loadings and their impact to safety-related nuclear containment and shielding structures of the NPPs. The purpose of this investigation is to understand the failure pattern and modes under BDBS loads; and how such failure impact the safety features of the plant and the safety functions of the systems (e.g. the impact to passive safety functions). This paper also covers ways to reduce the impact and mitigate the consequences of BDBS impact by pre-incorporated measures in the structural design.

Tritium permeation barriers (TPBs) are essential to reduce the tritium permeation through blanket structural materials and cooling tubes. The pores and cracks are the most important problem of thermally sprayed TPBs. Irradiation swelling is another serious problem in NiAl which used as adhesive layer in traditional Al 2 O 3 coatings. A self-healing coating TiC+mixture (TiC/Al 2 O 3 )+Al 2 O 3 without NiAl had been designed and manufactured by our group. It can cure 90% porosity after a certain heat treatment under normal atmosphere. The adhesive strength is as good as the traditional coating NiAl+Al 2 O 3 . The thermal shock cycles were 300, 210 and 123 at 600°C, 700°C and 800°C respectively. On the basis, the self-healing treatment parameters were explored further, and a more effective and efficient treatment was found. Structural design and component design were also researched. The thermal shock results showed that the thicknesses of TiC layer and mixture (TiC/Al 2 O 3 ) layer both had a significant effect while the Al 2 O 3 layer hadn’t. The interface of TiC and mixture (TiC/Al 2 O 3 ) was most easily damaged, so other proportion of TiC and Al 2 O 3 was applied, and several kinds of new coatings were designed, manufactured and tested. A better coating has been found, and it will probably be a good candidate for TPBs in fusion reactors.

Co-ordinated international activities are underway to complete an EU fusion demonstration power plant conceptual design by 2020, for construction in the early 2030s. However, many engineering design challenges still remain. For example, the design of in vessel components is technically challenging, with many design constraints needing to be mutually satisfied. One of the key challenges is the design validation of in-vessel structural components operating within the unique fusion environmental conditions. It has been recognized that these components would benefit from design criteria that have been specifically developed, providing improved accuracy and appropriate conservatism. This would provide DEMO designers with a much needed increase in design space and guidance. This paper highlights the justification behind developing fusion specific structural design criteria, discusses the key technical gaps of existing criteria and elaborates on the need to develop non-technical aspects of the documentation including consideration of human factors in design validation. Finally, a structured design criteria development plan is presented, highlighting planning work by the European fusion community to address these gaps in step with the current design milestones for the realisation of an EU DEMO.

The first Japanese spent fuel interim storage facility away from a reactor site is about to be commissioned in Mutsu City, Aomori Prefecture. In designing, licensing and construction of the Dual Purpose Casks (DPCs, for transport and storage) for this facility, codes and standards established by the Atomic Energy Society of Japan (AESJ) and by the Japan Society of Mechanical Engineers (JSME) have been applied. The AESJ established the first standard for DPCs as “Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facilities” in 2002 (later revised in 2010). The standard provides the design requirements to maintain the basic safety functions of DPCs, namely containment, heat removal, shielding, criticality prevention and the structural integrity of the cask itself and of the spent fuel cladding during transport and storage. Inspection methods and criteria to ensure maintenance of the basic safety functions and structural integrity over every stage of operations involving DPCs including pre-shipment after storage are prescribed as well. The structural integrity criteria for major DPC components refer to the rules provided by the JSME. JSME completed the structural design and construction code (the Code) for DPCs as “Rules on Transport/Storage Packagings for Spent Nuclear Fuel” in 2001 (later revised in 2007). Currently, the scope of the rules cover the Containment Vessel, Basket, Trunnions and Intermediate Shell as major components of DPCs. Rules for these components are based on those for components of nuclear power plants (NPP) with similar safety functions, but special considerations based on their shapes, loading types and required functions are added. The Code has differences from that for NPP components with considerations to DPC characteristics; - The primary stress and the secondary stress generated in Containment Vessels shall be evaluated under Service Conditions A to D (from ASME Sec III, Div.1). - Stress generated in the seal region lid bolts of Containment Vessels shall not exceed yield strength under Service Conditions A to D in order to maintain the containment function. - Fatigue analysis on Baskets is not required, and Trunnions can be designed only for Service Conditions A and B with special stress limits consistent with conventional assessment methods for transport packages. - Stress limits for earthquakes during storage are specified. - Ductile cast iron with special fracture toughness requirements can be used as a material for Containment Vessels. DPC specific considerations in standards and rules will be focused on in this paper. Additionally, comparison with the ASME Code will be discussed.

Under the combined accident thermal and seismic loadings, the structural response of the AP1000 Auxiliary and Shield Building (ASB) is numerically investigated. A nonlinear Finite Element Model (FEM) of the AP1000 ASB is developed, in which the rebar in the reinforced concrete is explicitly described and the nonlinear behavior of the concrete is considered. The numerical modeling method and material models used by the reinforced concrete are validated by the testing results published in the literature. The propagation of the thermal loading-induced concrete cracks along the wall thickness is studied. Furthermore, the effects of thermal cracks on the wall stiffness, the development of the thermal stress and the axial forces acting on the reinforcement are fully discussed. The impact of thermal concrete cracks on the design demand of the rebar is also investigated. It is worthy of being further studied how to incorporate the appropriate reduction factor caused by concrete cracks into the linear structural design.

Following China’s road map of nuclear technology development, the development of self-reliant nuclear design codes becomes one of the most significant steps in the plan. Among the nuclear design codes, thermal-hydraulic analysis code is indispensable because it is the foundation of reactor safety analysis and reactor design. Recently, China Guangdong Nuclear Power Group has launched a series of R&D projects of reactor design code development. The sub-channel analysis code-LINDEN becomes one of the key subprojects. Since the sub-channel code is developed for thermal-hydraulic design and safety analysis of pressurized water reactors (PWRs), the basic requirements for the LINDEN code are reliability and stability. Therefore, the mathematical model and numerical method developed in the code are based on the matured approaches that have been used in various industrial applications. These models and methods includes: four-equation drift framework model of two-phase flow; the classical heat transfer model and fuel rod model (Poisson equation) as well as the constitutive relations; explicit formulation and stepping algorithms for equation solutions. The solver module of the code is programmed using object-oriented C/C++ language with the structural design.. With all these features, the code was developed to be stable, scalable and compatible. The code’s applicability has been further improved after model improvement and design optimization according to characteristics of China’s proprietary type of reactor. COBRA-IV and LINDEN have been used to conduct the thermal-hydraulics analysis for the Daya bay unit 1 and 2 nuclear plants at the steady-state conditions. The results demonstrate that the two codes agree well with each other. The preliminary tests show that the LINDEN code should be suitable for thermal-hydraulics analysis of large PWRs.

CFETR is an “ITER-like” China Fusion Engineering Test Reactor. The BIT (Breeder insider tube) HCCB (Helium cooled ceramic breeder) blanket has been designed as one option for CFETR. Its unique feature is that a BIT structure is adopted for the blanket tritium breeder unit. The breeder unit is an assembly of three coaxial coil-pipes, internal coil-pipe is filled with the tritium-breeding material, and the coaxial coil-pipe is embeded in the beryllium pebble. In order assess the feasibility of the BIT HCCB blanket design, a 3D finite element model slice of the BIT HCCB blanket is developed, and the thermo-hydraulic calculation was performed, and thermo-mechanical computations were carried out. And then the primary stress and secondary stress were evaluated according to Structural Design Criteria for ITER In-vessel Components (SDC-IC). This paper presents relevant analyses and results. The preliminary results shows current design satisfies with allowable limit of material and requirement of SDC-IC code.

Based on the structural design of the Chinese Helium Cooled Solid Breeder (CH HCSB) test blanket module (TBM), the thermal-hydraulic behavior and safety performance of the CH HCSB TBM cooling system has been studied using the RELAP5/Mod3.4 code. According to the accident sequences of Accident Analysis Specification for TBM, two design basis accidents including loss of off-site power and multiple TBM FW (first wall) HTS (heat transfer system) pipe failure are investigated. The results show that natural circulation is established in the helium cooling circuit to cool the TBM effectively after loss of off-site power. In addition, after multiple TBM FW HTS pipe failure, the TBM can be cooled down because of the thermal radiation. The maximum vacuum vessel (VV) pressure and the mass of not condensable gas spilling into the VV are within the limits for ITER design. Temperature difference between the break and intact FW pipes is also found.

Based on the structural design of the Chinese ITER Dual Functional Lithium-Lead Test Blanket Module (DFLL-TBM), Three Unprotected Loss of Flow Accidents (ULOFAs) were investigated preliminarily, assuming that the whole nuclear heat in TBM was carried away by the flowing lithium-lead (LiPb). The results show that the temperature of the first wall (FW) increases rapidly and the maximum temperature appears at the lower part of FW. In the analysis of ULOFAs, the maximum temperature might exceed the melting point of structure material steel. This event must be avoided by the fusion power shutdown system that terminates plasma burn.

International Thermonuclear Experimental Reactor (ITER) TF feeder systems convey the cryogenic supply and electrical power to the TF coils. The Cryostat Feed-through (CFT) includes the straight feeder part from the cryostat wall to the S-Bend Box (SBB). It is the bottleneck of the feeders. The huge Lorentz-force is a challenge for the CFT design. So the reasonable distribution and structural design of the internal and external supports are important. The CFT include the cold (cryogenic) to warm (room temperature) transitions. It is highly integrated with the cryo-pipes, the busbars, the superconductor joints, the thermal radiation shield and the instrumentation pipes and so on. The cryogenic and electrical requirements, the vacuum and mechanical requirements, and so on are considered when the CFT is designed. This paper presents the functional requirements on the TF CFT, gives its structure. The supports are designed and arrayed according to their mechanical or thermal function separately to stand the huge mechanical loads and isolate the conducting heat load from room temperature respectively. The assembly scheme is also described. Mid-joint and cryostat joint are designed to give the facility for the assembly on location. The mechanical analysis result shows the stress in the stainless steel and G10 material both are within the materials stress safety margin. The heat load to the cryogenic pipes and busbars are also less than the requirement 15W. Transient thermal analysis of global feeder model indicates that 32 days are needed for the feeder components to cool down to the required condition.

On the premise that ensuring the reliability of structural strength, and in order to get the optimal structural parameters and the number of reinforcement rib of the ITER feeder S-bend box (SBB), so that its stress distribution is more uniform and reasonable under the critical pressure load, and the cost of materials is relatively smaller, this article, through theoretical calculations and inferences on the basis of the parametric modeling and static analysis and checking of SBB, setting up different reinforcement rib numbers, optimizes through the ANSYS optimization design module, with the maximum stress as the objection function, with and reasonable quality and displacement as state variables, with the size of wall thickness and the space of reinforcement rib as design variables. And it regulates true stress at stress concentration through submodeling of ANSYS and local solid modeling. The results of optimization analysis show that: SBB reached the optimal solution when the reinforcement rib number N of SBB takes 3, the maximum equivalent stress is 117 Mpa and the weight is 6417 Kg . Finally, SBB structural parameters, which obtained through optimization design, are rounded according to GB. These meet the design requirements, correspond to the practical applications and provide technical parameters and basis for the future development of SBB.

The System Based Code (SBC) concept has been proposed to achieve compatibility in matters of reliability, safety, and cost of Fast Breeder Reactors (FBR). This code extends the present structural design standard to include the areas of load setting, fabrication, inspection, maintenance, and so on. Therefore, a quantitative index which can connect different areas is required. In addition, the determination of its target value is also one of the key points to substantiate the SBC concept. Failure probability is one of the candidate indexes. We have proposed a new method to determine the reliability targets for the structures and components in FBR plants from the safety point of view by utilizing analysis models of a probabilistic safety assessment. In this study, the effectiveness of the failure probability as an index and the compatibility of the reliability targets derived by the new method were investigated through a trial setting of In-Service Inspection (ISI) request on the reactor vessel near the sodium surface based on the SBC concept. The failure probability due to creep-fatigue interaction was calculated by the Monte-Carlo simulation. In response, the reliability targets for fracture related to the risk from internal initiating events were derived. Cumulating the failure probability and the reliability targets up to the end of in-service period enables us to compare them directly, and we obtained a result that the reactor vessel has enough reliability even without ISI. Through this trial, we showed that the failure probability is a promising index, and the reliability targets derived by the new method are compatible with the SBC concept.

In the design of seismic category 1 buildings in nuclear power plants (NPP) or, outside the nuclear domain, in the conventional structural design of buildings, the seismic evaluation of these buildings may be done. In the occurrence of an earthquake in a NPP or in the case of changing the use of a conventional building, the seismic levels are modified. Then a new analysis need to be performed. This paper focusses on the situation where reinforcing the concrete building is needed and it also analyses how an extern reinforcement performed using polymers can be carried out to fulfill the new seismic requeriments. We present two main results: a) the resulting momentum-curvature diagrams obtained reinforcing standard segments embraced with polymers; b) the evaluation of the structure capacity on the basis of the modified diagrams. Finally, a modal pushover analysis is selected to perform the seismic evaluation of two types of concrete columns, those having a polymer reinforcement and those without it. This paper presents the basis of the subject in a theoretical form.

Anchoring structures, systems and components to concrete is a significant activity in the design and construction of a nuclear power plant. Early in this decade the Concrete Capacity Design method (CCD) was adopted by the American Concrete Institute (ACI) for use in the structural design for both commercial and nuclear facilities. This design method and associated qualification tests brings new challenges to designing efficient means for anchoring to concrete structures. Although the CCD method provides guidance on many aspects of concrete anchorage there are a few areas, pertinent to nuclear power plant construction, that are not covered or require significant interpretation of the most recent codes. This paper will focus on the design of shear lugs used to resist significant lateral loads. Results from laboratory tests of shear lugs are presented. These full scale tests considered the interaction of tension and shear loads on the performance of shear lug assemblies. Recommendations for the efficient use of shear lugs are provided.

The reactor coolant pump in nuclear power plant is the only revolving equipment in the nuclear power plant. Its functional stability will directly affect the security of nuclear power plant. The coolant pump of a very nuclear plant is examined by using response spectrum analysis to analysis dynamic characteristics and responses aiming at finding the natural frequencies of vibration, modes of vibration and seismic responses, and any possible step which may cause damage of the whole system. The favorable spectrum and unfavorable one are investigated as well. The paper focuses on avoiding the detrimental effects caused by earthquakes, therefore may lay down a theoretical foundation for structural design and installation.

The spacer grid is one of the structural components for the fuel assembly. In order to increase or extend the fuel life cycle, a spacer grid which has a much higher performance from the thermal/hydraulic and mechanical/structural point of view will be needed. From the thermal/hydraulic viewpoint, the CHF margin is very important in order to extend its life. Particularly, the mixing flow or cross flow between the subchannels have to be reinforced for this purpose. From the mechanical/structural viewpoint, the critical strength and the fuel rod support behaviour of a spacer grid are the same as the TH performance improvement for the next generation fuel. A computational fluid dynamics (CFD) analysis was performed to investigate the coolant mixing in a nuclear fuel bundle that is promoted by the mixing vane on the grid spacer. Single and multiple subchannels of one grid span of the fuel bundle were modeled to simulate a 5by5 rod array experiment with the mixing vane. The three-dimensional CFD models were generated by a structured multi-block method. The standard k-ε turbulence model was used in the current CFD simulation since it is practically useful and converges well for the complex turbulent flow in a nuclear fuel bundle. The CFD predictions of the axial and lateral mean flow velocities showed a somewhat larger difference from the experimental results near the spacer but represented the overall characteristics of the coolant mixing well in a nuclear fuel bundle with the mixing vane. Comparison of the single and multiple subchannel predictions shows a good agreement for the flow characteristics in the central subchannel of the rod array. The simulation of the multiple subchannels shows a slightly off-centered swirl in the peripheral subchannels due to the external wall of the rod array. It also shows no significant swirl and crossflow in the wall subchannels and the corner subchannels. In addition to this, the impact and the stress analysis of a spacer grid are accomplished by the FE method. The FE model was created using I-DEAS [4], and the ABAQUS/explicit version 6.3 commercial code was used for the solver. The FE analysis procedure was established, the FE analyses results were verified by the experimental method. The developed spacer grid will be evaluated from the thermal/hydraulic and mechanical/structural design criteria.

The intent of this paper is the presentation and discussion of a methodology for the evaluation and analysis of seismic loads effects on a nuclear power plant. To help in focussing the presented methodology, a preliminary simplified analysis of an integral, medium size next generation PWR reactor structure (IRIS project, an integral configuration PWR under study by an international group) was considered as an application example also for models/codes evaluation. The performed preliminary seismic analysis, even though by no means complete, is intended to evaluate the method of calculating the effects of dynamic loads propagation to the reactor internals for structural design as well as geometrical and functional optimisation purposes. To this goal, finite element method and separated (sub) structures approaches were employed for studying the overall dynamic behaviour of the nuclear reactor vessel. The analysis was set up by means of numerical models, implemented on the MARC FEM code, on the basis of Design Response Spectra as indicated on the relevant rules for Nuclear Power Plants (NRC 1.60) design. The seismic analysis is indented to evaluate the dynamic loads propagated from the ground through the Containment System and Vessel to the Steam Generator’s tubes.