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.     

CARR衰变箱及堆水池钢衬里设计难点

中国先进研究堆(CARR)的衰变箱和堆水池钢衬里是CARR中的关键设备之一,本文阐述了CARR衰变箱和堆水池钢衬里的设计,在焊接、运输、大型薄壳设备制造等方面存在的难点问题及解决方案。     

Theoretical analyses of thermal post-buckling problems of liner shells

The buckling of liner shells is a major problem in reactor containment design. In this paper, the local liner shell is considered as the liner plate with special initial imperfection form, based on which, Koiter's asymptotic theory of initial post-buckling is used to obtain expressions for the post-buckling equilibrium paths and post-buckling minimum loads for perfect and imperfect liner shells. The post-buckling behaviors of three liner shell models (four point-supported liner shell, clamped liner shell, and five-point-supported liner shell) are investigated. The influence of the initial imperfection stud spacing and liner thickness on the post-buckling minimum loads is also given.     

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.     

Control strategies for transients of hydrogen production plant in VHTR cogeneration systems

Operability of Very High Temperature Reactor (VHTR) hydrogen cogeneration systems in response to abnormal transients initiated by the hydrogen production plant is one of the important concerns from economical and safety points of views. The abnormal events in the hydrogen production plant could initiate load changes and induce temperature variations in a primary cooling system. Excessive temperature increase in the primary cooling system would cause reactor scrams since the temperature increase in the primary cooling system is restricted in order to prevent undue thermal stresses from reactor structures. Also, temperature decrease has a potential propagation path for reactor scrams by reactivity insertions as a consequence of the reactivity feedbacks. Since suspensions of reactor operation and electricity generation should be avoided even in case of abnormal events in the hydrogen production plant from an economical point of view, an establishment of a control scheme against abnormal transients of hydrogen production plant is required for plant system design.In the present study, basic controls and their integration for the GTHTR300C, a VHTR cogeneration system designed by JAEA with a direct Brayton cycle power conversion unit and thermochemical Iodine-Sulfur process hydrogen production plant (IS hydrogen production plant), against abnormal transients of IS hydrogen production plant are presented. Transient simulations for selected load change events in the IS hydrogen production plants are performed by an original system analysis code which enables to evaluate major phenomena assumed in process heat exchangers of the IS hydrogen production plant.It is shown that abnormal load change events are successfully simulated by the system analysis code developed. The results demonstrated the technical feasibility of proposed controls for continuous operation of the reactor and power conversion unit against load change events in the IS hydrogen production plant.     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

CANDU design features that influence fuel modelling for transients

The development of models to predict fuel behaviour under transient conditions is influenced by the design features of the CANDU-PHW reactor system. This paper describes those aspects of the CANDU design, and their influence on fuel behaviour and modelling activities, that are essentially different from other reactor systems.In our safety analysis we emphasize predictions by verified models rather than by experimental simulation of a large number of hypothetical transients. Consequently, mechanistic equations are developed and incorporated into the model to reflect the history dependent mechanical properties acquired before the transient as well as physical changes that occur during the transient. Because of the horizontal orientation of the fuel channels, the models are being developed to assess the effect of gravity on fuel behaviour during transients and any potential interaction between the fuel and the pressure tube.     

Response of liquid metal fast breeder reactor containment to a hypothetical core meltdown accident

This paper presents an analysis of the response of the containment building of a 2500-MWt liquid metal fast breeder reactor to a hypothetical reactor core meltdown. Although not mechanistically justifiable, this type of event is chosen for analysis as a basis for risk assessments. Containment space atmosphere compositions, temperatures, pressures, and structural temperatures are calculated, based on decay energy release and chemical reactions associated with the incident. The CACECO containment analysis code, which was used to make the calculations, is described in detail.Results of the study show that by utilizing the passive heat absorption capability of structures normally present in containment design, reactor plant containment integrity can be maintained for more than a day, even for extreme hypothetical events.     

Design and analysis on tritium system of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

As early application of fusion technology, the fusion–fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion–fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.     

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.     

Design and analysis of multicavity prestressed concrete reactor vessels

A critical state-of-the-art review of multicavity prestressed concrete reactor vessel (PCRV) design and analysis practice is presented. Included are discussions of basic design concepts, the behavior of liners and penetrations, and the various tests required and/or employed to demonstrate acceptance of new vessel geometries and innovations. Brief reviews are given of the influences of design codes such as ACI/ASME Section III, Division 2, and BS4975; analysis methods including elastic, inelastic, and time-dependent techniques; the constituve equations that are essential to the satisfactory use of these techniques; and semi-empirical methods for calculating ultimate strengths of multicavity vessels.Tests conducted on liner plates, liner anchorage systems, and cooling tubes are reviewed together with the methods of analysis used in the design of anchorage systems. The adequacy and economy of present liner systems are considered and possible modifications in design are suggested.Design code requirements and methods of analysis for penetrations are discussed. The various types of closure designs that have been proposed and in some cases employed are evaluated on the basis of overall PCRV design philosophy. Several methods of prestressing PCRVs are considered with respect to relative advantages and disadvantages; existing overall vessel in-service inspection requirements are evaluated.     

The Austrian PCRV project with a hot liner — A status report

The analysis of the problem PCRV safety leads to a new vessel concept in which the pressure transmitting insulation with its cooling system is outside the hot linear. A control and adjustment system for vessel wall temperature ensures that, under all working conditions, the liner suffers only elastic compression. Outside the thermal insulation (special concrete) a cold steel barrier is placed, which can be constructed as a second liner, so that leaks are limited and can be detected and evacuated. After preliminary development work in the field of high temperature insulating concrete, prestressed concrete technology and instrumentation at elevated temperature, a large experimental ring has been constructed and tested extensively. As part of an experimental high temperature gas loop a large scale model vessel is being built at the Research Center, Seibersdorf, Austria. Construction is nearly complete and a pressure test is scheduled for early 1975. It will be followed by tests under working conditions of PWRs and later of HTGRs. The reference design of such a PCRV with a hot liner for a 1500 MW(e) pressurized water reactor will be finished at the same time as the model vessel.     

Source term research and prognosis

The work sponsored by EPRI on source term technology is discussed (source terms describe the fission product releases to the environment in a severe hypothetical accident). The experimental programs include (1) fission product release from fuel, (2) fission product transport in the reactor primary circuit, (3) aerosol behavior in reactor containment, (4) aerosol scrubbing by water pools, and (5) hydrogen combustion in the containment. Code development work is also included.     

Danger of Burning of Sodium Coolant

The properties of sodium as a substance that can burn are described. The radiation characteristics of sodium as a first-loop coolant in a fast nuclear reactor are presented. An assessment is made of the consequences of sodium burning in various situations. First and foremost, an unanticipated accident with burning of sodium in the first loop adopted in the BN-800 design, is examined. Next, situations with hypothetical scenarios are examined to obtain the limiting data charcterizing the potential fire hazard of radioactive sodium coolant. Specifically, a hypothetical situation where all of the sodium contained in the first loop of the reactor burns is examined. The computational results are analyzed from the standpoint of the role sodium plays in the overall problem of nuclear power plant safety.     

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.     

Parametric evaluation of large-scale high-temperature electrolysis hydrogen production using different advanced nuclear reactor heat sources

High-temperature electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800-950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an intermediate heat exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies.The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO₂ reactor concepts). Sensitivity studies were then performed with the objective of evaluating the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency of the integrated plant design for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered.     

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.     

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A&M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR.This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900–1000 °C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900–1000 °C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and discussed.     

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.     

Development of an evaluation method for the HTTR-IS nuclear hydrogen production system

A thermochemical water splitting hydrogen production system based on the iodine sulphur (IS) process is presently under development in JAEA. The hydrogen production system is to be connected to the HTTR operating test reactor in JAEA. An important development goal for the HTTR-IS system is design and construction of the IS process to the standards of a conventional chemical industrial plant in order to simplify the cost and operation of the overall nuclear hydrogen production.