The design of the prestressed concrete reactor vessel for gas-cooled heating reactors

The Gas-cooled Heating Reactor (GHR) based on the pebble red reactor principle was developed by ABB/HRB. An essential part of this concept is the prestressed concrete reactor vessel in which the liner cooling system acts as a heat exchanger. As a main design feature the vessel is designed so that failure can be safely ruled out under all operating and accident conditions. It is of great advantage that the liner is not exposed to primary stresses and that corrosion can be excluded because of the environmental conditions. Relevant material flaws are ruled out by the considerably extent and level of quality assurance measures. A special heat-resistant concrete developed by HRB will be used for the prestressed concrete structure. Its strength behaviour is characterized by only a small reduction during normal operation and also under accident conditions. Even in the event of a hypothetical accident the integrity of the vessel remains intact. Thus the GHR offers a simple, safe and economic source of heat generation.     

基于有限元分析法的钠火事故下钢覆面完整性分析

中国实验快堆钠工艺间的钢覆面结构是防止泄漏的钠与混凝土接触的重要屏障,其完整性直接关系到钠火事故下建筑结构的完整性和可靠性。本文以中国实验快堆309管廊间作为研究对象,梳理了可能破坏钢覆面结构完整性的各项因素,基于池式钠火分析软件的输出结果,在建立钢覆面破损二维瞬态模型的基础上,利用ANSYS热-结构耦合功能系统地分析了钢覆面的应力集中区的产生规律。研究表明,在发生超设计基准的钠泄漏钠火事故时,钢覆面上的漏钠燃烧持续30min后,其厚度从3 mm腐蚀至1.53mm,并在两块钢板的焊接部两端出现应力集中区,最大应力超过了材料的屈服强度,材料发生塑性变形,存在断裂的风险。本文的研究结果对后续快堆钢覆面结构的设计和安全评定方面具有一定的参考价值。     

Excellox 6 and 7 Irradiated Fuel Transport Flasks

Abstract

The Sellafield nuclear fuels reprocessing plant in Cumbria has been receiving Light Water Reactor (LWR) fuel from European and Japanese power stations since the early 1970s. Virtually all of this has been delivered to Sell afield in flasks of composite design. This design of flask comprises a thick lead liner surrounding the fuel cavity and providing gamma radiation shielding, structural strength being provided by an enclosing steel shell. The composite flask has proved to be safe and efficient in operation but is now meeting the limits of its potential due to the trend towards higher burn up fuel High burnup fuel emits increasing levels of neutron radiation which the composite flask was not designed to accommodate; thus the need for new designs of flasks to carry the fuels of the future. The Excellox 6 and Excellox 7 irradiated fuel transport flasks have been developed over a period of five years. They are of monolithic construction and have been designed to complement existing composite flask types. The Excellox 6 flask is designed to carry high bumup PWR fuel up to 5 m in length from Europe, whilst the Excellox 7 flask is shorter and can carry high burnup PWR and BWR fuels up to 4.5 m in length from both Europe and Japan. As a consequence of meeting Japanese regulatory requirements the shielding design of the Excellox 7 differs from the Excellox 6.     

The plastic buckling of a steel liner encased in a concrete cylinder

The inelastic buckling and postbuckling performances of a steel liner encased in a rigid concrete containment vessel are studied by taking a strip of unit width from a pattern — either rectangular or diamond — along the circumferential direction in such a way that the strip will have the maximum deflection of a buckled panel. The complete load-deflection curves are obtained and the effect of initial imperfections is also included in the curves. From the discussion of failure modes, a design criteria can be obtained for the liner to maintain its integrity under accident conditions. A simplified spring model is used to calculate the maximum shear displacement and the corresponding shear force of studs at buckling. A design analysis procedure is developed from the limit and the ultimate design conditions and can be used to determine the liner thickness, studsizes and stud spacings in both the axial and circumferential directions.     

Containments design for the advanced and simplified boiling water reactor standard plants

The Advanced Boiling Water Reactor (ABWR) design is based on construction and operating experience of nuclear power plants in Japan, United States, and Europe. To optimize the plant arrangement of the Advanced Boiling Water Reactor (ABWR) and to verify the structural feasibility to carry design loads a study was conducted. To arrive at an optimized plant arrangement with a minimum size reactor building (RB), a circular cylindrical reinforced concrete containment vessel (RCCV) with a flat top slab and a monolithically connected diaphragm slab has been selected.The Simplified Boiling Water Reactor (SBWR) is being developed as a standardized 600 MWe Advanced Light Water Reactor. The design concept of the SBWR is based on simplicity and passive features to enhance safety and reliability, improve performance and increase economic viability. Due to the use of passive containment cooling, SBWR has features that are different from those of existing designs.The objectives of the study for the ABWR containment and RB are to perform a structural analysis of the containment and RB and to evaluate the structure for conformance to the U.S. NRC requirements. The main objective of the studies for the SBWR is to demonstrate the structural design feasibility of the containment for the pressure and the temperature response associated with the passive systems adopted for the SBWR.     

Proof pressure tests of the first reactor vessels at Hartlepool and Heysham nuclear power stations

The nuclear stations currently nearing completion at Hartlepool and Heysham are the world's first design of Advanced Gas Cooled Reactor to use podded boiler construction. The proof pressure tests on the first reactor pressure vessels at both stations were carried out during the early part of 1980. The object of the proof pressure test is to satisfy statutory and contractual requirements by demonstrating the integrity of the completed concrete pressure vessel, liners and pressurised penetrations when subjected to a pneumatic test pressure of 740 lbf/in² g, equivalent to 1.15 times the design pressure of 644 lbf/in² g. Testing was carried out in the latter period of the construction phase of each station with the majority of the plant and buildings in an advanced state of construction and the reactor internals essentially complete. This paper examines the results of both reactor 1 vessel tests and compares the behaviour of the vessels with one another and with analytical predictions. Possible sources of instrumentation error are identified and discussed. The conclusions are drawn from correlation of the test results with theoretical predictions.The vessel pressure was raised and lowered in six stages to and from the test pressure with a hold period between each stage to read instruments, examine the vessel surfaces and leak check the penetrations and closures. To obtain the relevant information at each pressure hold, the vessels' permanent instrumentation, consisting of vibrating wire strain gauges and their associated electrical resistance thermometers, and the liner and reactor internal thermocouples were monitored. In addition, strain gauges fixed to the concrete face of the liner during construction were monitored and liner strains measured during the test were compared with predictions. Vessel deflections were monitored by special equipment mounted on the reactor building to enable vessel dilation to be measured. The vessel top and bottom caps were instrumented by manometric systems in order that the cap profiles and deflections could also be determined.The measurements and observations made during the proof pressure tests adequately demonstrated that the vessels behaved in accordance with predictions. The concrete strains and vessel external deflections confirmed that the vessels behaved in a linear and elastic manner throughout the tests, with no cracking being observed in the concrete during the tests. Liner strains compared favourably with predicted values, exhibiting a linear behaviour under increasing pressure. The strain levels recorded gave complete confidence in the liner design.The tests confirmed the integrity of the vessels, thus enabling them to go forward to the engineering run stage in the commissioning programme leading ultimately to the raising of power.     

Safety considerations for large GCFRs

The paper reviews the work that is currently being done in the Federal Republic of Germany on the design and safety aspects of a large 1000 MW(e) Gas Cooled Fast Reactor. The design studies comprise the layout of both the primary (helium) and the secondary (water-steam) system, the latter to the extent necessary to determine its feed back on the primary system. Safety studies include the analysis of the transient behaviour of the plant under various accident conditions including the design basis accident. It is shown that the GCFR system can be designed to meet the safety requirements currently in use in the Federal Republic of Germany for the Pressurized Water Reactor. Although up to the present time no realistic chain of events has been detected that would lead to accidents beyond the design basis accident, some work is being carried out in the field of hypothetical accidents. Included have been studies on problems associated with handling of gross core melting.     

Concept study for a combined reinforced concrete containment

A variety of different types of steel and concrete containments have been designed and constructed in the past. Most of the concrete containments had been pre-stressed, offering the advantage of small displacements and a certain leak-tightness of the concrete itself. However, considerable stresses in concrete as well as in the tendons have to be maintained during the whole lifetime of the plant in order to guarantee the required pre-stressing. The long-time behaviour and the ductility in the case of beyond-design-load cases must be verified. Contrary to a pre-stressed containment a reinforced containment will only be significantly loaded during test conditions or when needed in case of an accident. It offers additional margins which can be used especially for dynamic loads such as impacts or for beyond-design events.The aim of this paper is to show the feasibility of a so-called combined containment which means a containment capable of resisting both severe internal accidents and external hazards, mainly the aircraft crash impact as considered in the design of nuclear power plants in Germany.The concept is based on a lined reinforced containment without pre-stressing. The mechanical resistance function is provided by the reinforced concrete and the leak-tightness function is provided by a so-called composite liner made of non-metallic materials. Some results of tests performed at Siemens laboratories and at the University of Karlsruhe which show the capability of a composite liner to bridge over cracks at the concrete surface will be presented in the paper.The study shows that the combined reinforced concrete containment with a composite liner offers a robust concept with high flexibility with respect to load requirements, beyond-design events and geometrical shaping (arrangement of openings, an integration of adjacent structures). The concept may be further optimized by partial pre-stressing at areas of high concentration of stresses such as at transition zones or at disturbances around large openings.     

Performance validation on the prototype of control rod driving mechanism for the TRR-II project

This paper presents the case study of prototyping a control-rod driving mechanism (CRDM), which is a crucial safety system in the Taiwan Research Reactor (TRR-II). Concurrent engineering concept has been implemented in this work, which consists of iterative parallel procedures with design analysis and performance testing. Such kind of nuclear-grade system and/or components distinguish themselves from generic industrial products with the so-called fail safe design feature, which means that the system must still be safe even if components fail. Hence, a series of performance testing and design improvement have been interactively executed, to ensure the mechanical integrity and durability of the prototype. Functional testing results show that the overall performance of the CRDM meets the specification requirements.     

Present status of the High Temperature Engineering Test Reactor (HTTR)

In Japan, the research and development on the High Temperature Gas-cooled Reactors (HTGRs) had been carried out for more than fifteen years since 1969 as the multi-purpose Very High Temperature gas-cooled Reactor (VHTR) program for direct utilization of nuclear process heat such as nuclear steel making. Recently, reflecting the change of the social and energy situation and with less incentives for industries to introduce such in the near future, the JAERI changed the program to a more basic ‘HTTR program’ to establish and upgrade the HTGR technology basis.The HTTR is a test reactor with a thermal output of 30 MW and reactor outlet coolant temperature of 950°C, employing a pin-in-block type fuel block, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. Since 1986 a detailed design has been made, in which major systems and components are determined in line with the HTTR concept, paying essential considerations into the design for achieving the reactor outlet coolant temperature of 950°C. The safety review of the Government started in February 1989. By request of the Science and Technology Agency the Reactor Safety Research Association reviewed the safety evaluation guideline, general design criteria, design code and design guide for the graphite and the high-temperature structure of the HTTR.The installation permit of the HTTR was issued by the Government in November 1990.     

Progress of LWR structural safety research at JAERI

At JAERI, structural safety research, in which issues of aging and structural integrity of LWR components are involved, has been performed as one of the items of safety research prescribed in the national 5-Year Safety Research Program determined by the Nuclear Safety Commission. In order to achieve the aims of the program, research on aging mechanism and prediction, detection and evaluation of aging, and structural integrity evaluation of aged components are in progress. Reactor pressure vessel, concrete structures and electrical cables were identified as key components to be investigated in the research. This paper presents an overview of the progress of the research program. In addition, results of a leak-before-break (LBB) research program which was carried out as a part of the structural safety research program are described.     

Analysis of PCRV liners by the initial stress method

In the design of prestressed concrete reactor vessel (PCRV) liners with closely spaced anchors, analyses are conducted to determine forces on and displacements in the liner components under concrete-imposed strains. In this paper a new method of one-dimensional liner stress analysis is presented. The method uses a stiffness approach in which the final set of simultaneous equations are the equilibrium equations in terms of unknown nodal displacements. The solution of these nonlinear simultaneous equations is accomplished using the ‘initial stress method’. A parametric study has been conducted to investigate the effects of main liner design variables. Results of this study are presented and discussed.     

Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor – Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid–structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules.     

Development of ITER tritium transport package

For tritium supply to the fusion reactor of ITER (International Thermonuclear Experimental Reactor; the way to new energy) [1], tritium needs to be transported from tritium production sites, mainly the CANDU type reactor sites to the Tritium Plant building of ITER. Korea Atomic Energy Research Institute (KAERI) was commissioned the work of developing the transport package for tritium by ITER organization and the first stage of the development has been finished. The developed package was designed to carry 70 g of tritium and classified as a type B(U) package, which should comply with the requirements stipulated in IAEA regulations for the transportation of radioactive materials [2]. The package is composed of a storage vessel, a containment vessel, an overpack and an aluminum liner, which is a unique feature of the package. The aluminum liner between the storage vessel and the containment vessel is for containment control under the repetitive use of the package. The package has enough pressure resistance for 5 years in-site storage and the structural and thermal integrity under the hypothetical accident conditions has been demonstrated through a series of analyses.     

Some considerations to the failure analysis of a pointwise attached steel liner membrane under constraint load

In the application of reinforced or prestressed concrete reactor containments, the safety enclosure will be obtained through a steel liner membrane, which is attached pointwise to the interior concrete surface. It is the objective and aim of this study to analyse the overall structural behavior of the bonded system consisting of concrete containment, studs, and steel liner—especially under the aspect of extreme load and deformation conditions. The parametric analysis is carried out on the basis of the geometric length/depth ratio of a single liner field. In order to reduce the considerable computational effort to a minimum, it is necessary to decouple the overall system in its structural components, i.e., at first an imperfect predeflected ‘buckling’ field and the residual ‘plane’ liner field are considered separately. A further reduction enablesm the use of stud anchor characteristics which are based on experiments. Three-dimensional analyses are performed for the single ‘buckling’ field to obtain specific load-displacement functions; the residual plane system is considered with two- as well as one-dimensional models. For the comprehensive parametric evaluation of the overall system behaviour, a linear model is assumed to which these load-displacement functions are applied. Constraint temperatures are introduced as a unit scale—up to failure of the overall system; hereby partial structural failure might lead to temporary relief.     

CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 °C which is substantially lower than ∼627 °C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a crucial factor for consideration in safety design. This study provides future researchers with a guideline on designing safety measures for the fourth generation of the fast reactors with no particular reference to any specific manufacturer.     

Heat and mass transfers in a concrete wall with composite liner under accidental conditions

The aim of this work was to study the behaviour of a concrete wall, covered with a composite liner and exposed to accidental conditions leading to high temperatures and pressures on a face of the material. In the laboratory, two practical levels of accidental situations (beyond design) have been considered. Firstly, the “SC1” scenario (accidental conditions) consisted of a rise from ambient conditions to a saturation point of 160 °C, and a pressure of 0.75 MPa in 12 h, using the maximum increase possible with the apparatus. This rise was then followed by cooling, leading to 0.22 MPa and 120 °C in 24 h. These conditions were maintained for several days. Secondly, a “SC2” scenario (severe accident conditions) consisted of a rise to a saturation point of 173 °C and a pressure of 1 MPa, these conditions were maintained for 24 h before cooling.A cylindrical specimen of 1.3 m of thickness was used. Thermocouples, pressure taps and moisture gauges were implemented before concreting. These devices provided local information, and were mostly distributed in the first 0.30 m of the concrete. The concrete composition (high performance concrete) was the same as that used for the construction of the CIVAUX 2 nuclear power station.Typical experimental results for the evolution of temperature, pressure and water content as functions of time are shown for the two test conditions. The concrete attached to the back of the composite dried, and a mass transfer was induced towards colder zones in the centre of the specimen. The liner acted as a heat insulator and the pressure acting on the back of the composite remained lower than that applied on the composite. The residual adhesion of the liner to the concrete was measured. Finally, the overall results allowed the comparison of situations where the wall was lined and unlined, during exposure to SC1 and SC2 conditions.     

Summary report on new regulatory codes and criteria and guidelines: IAEA guidelines for WWER reactor coolant system integrity assessment

In the frame of the nuclear safety programme to assist the countries of Central and Eastern Europe, the IAEA identified and ranked in total 263 safety issues for WWER-440/230, WWER-440/213 and WWER-1000/320 nuclear power plants, related to both design and operation. In the area of reactor coolant system integrity, 24 safety issues were identified and 15 of them ranked as of high safety significance. These include: reactor pressure-vessel integrity and related aspects, primary and secondary high-energy piping integrity, steam generator integrity and reliability of the non-destructive testing for in-service inspection. In addition to obtaining international consensus on possible solutions to address the safety issues identified and to reviewing completeness of proposed safety improvements, the IAEA initiated development of guidelines to address the issues of highest safety concern. In the area of the reactor coolant system integrity, guidelines for the leak before break concept application, for the pressurized thermal shock analysis and for the in-service inspection systems qualification were developed. Further activities of the IAEA were focused on the implementation of guidelines developed in the Member States concerned. With this objective, a Co-ordinated Research Programme ‘Round-robin Exercise on WWER-440 Reactor Pressure Vessel Embrittlement, Annealing and Re-embrittlement’, a ‘WWER-440/213 Pressurized Thermal Shock Analysis Benchmark Exercise’ and a pilot study to implement the qualification approach to a real power plant component have been initiated by the IAEA and are well under way at present. In this paper, an overview of these IAEA activities related to reactor coolant system integrity is provided and the main principles and elements of guidelines developed discussed.     

Electromagnetic–thermal–structural coupling analysis of the ITER edge localized mode coil with fiexible supports

In a fusion reactor, the edge localized mode (ELM) coil has a mitigating effect on the ELMs of the plasma. The coil is placed close to the plasma between the vacuum vessel and the blanket to reduce its design power and improve its mitigating ability. The coil works in a high-temperature, high-nuclear-heat and high-magnetic-field environment. Due to the existence of outer superconducting coils, the coil is subjected to an alternating electromagnetic force induced by its own alternating current and the outer magnetic field. The design goal for the ELM coil is to maintain its structural integrity in the multi-physical field. Taking as an example the middle ELM coil (with flexible supports) of ITER (the International Thermonuclear Fusion Reactor),an electromagnetic–thermal–structural coupling analysis is carried out using ANSYS. The results show that the flexible supports help the three-layer casing meet the static and fatigue design requirements. The structural design of the middle ELM coil is reasonable and feasible. The work described in this paper provides the theoretical basis and method for ELM coil design.