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The water-cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for Chinese fusion engineering testing reactor (CFETR) and is being developed at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). This paper reviews design and evolution of the WCCB blanket for CFETR, and presents a new WCCB blanket design according to the latest CFETR core parameters (major and minor radii are R = 7.2 m and a = 2.2 m, respectively) and missions. This new design is expected to satisfy multiple CFETR operation modes of 0.2, 0.5, 1.0, and 1.5 GW fusion power and achieve tritium self-sufficiency. The feasibility of the updated blanket design is evaluated from the aspects of neutronics and thermo-hydraulics. Furthermore, the research and development (R&D) activities supporting to the WCCB blanket for CFETR are reported, including the design code, the water loop experiments, the pebble bed modeling and experiments, and the components fabrication technology. Keywords: Chinese fusion engineering testing, reactor, WCCB blanket, Research and development

Active cooling is essential for solar cells operating under high optical concentration ratios. A system comprises four solar cells that are in thermal contact on top of a copper tube is proposed. Water is flowing inside the tube in order to reduce solar cells temperature for increasing their performance. Computational Fluid Dynamics (CFD) simulation of such system has been performed in order to investigate the effect of water flow rate, tube internal diameter, and convective heat transfer coefficient on the temperature of the solar cells. It is found that increasing convective heat transfer coefficient has a significant effect on reducing solar cells temperatures operating at low flow rates and high optical concentration ratios. Also, a further increase of water flow rate has no effect on reducing cells temperatures.

This chapter reviews the main corrosion problems that have affected tubes in commercial PWR steam generators. For each of these main corrosion problems the types of steam generators that have been affected are first indicated. The locations in the steam generators affected by the corrosion problem are then identified together with a discussion of the design and operating factors that led to that specific corrosion problem. The likely future occurrence of the corrosion problem is then discussed, together with a discussion of actions that have been or could be taken to minimize occurrence of the problem.

In order to decrease the earth's energy imbalance degree, a R744 ground source heat pump system (GSHPS) with air-cooled and water-cooled gas coolers was designed, and the performances of this system were analyzed numerically. The results show that, for the proposed R744 system in different regions, the earth's energy imbalance degree can always be decreased to zero by optimizing the cooling load proportion of air-cooled gas cooler. Comparing with the system without air-cooled gas cooler, the investment cost of the proposed R744 system is decreased by 14.5%~25.5% due to the decreasing cost of ground heat exchanger, and the coefficient of performance is increased. With the increase of ground heat exchanger depth, the performance deteriorates, and the ground heat exchanger with shorter depth is preferred in the GSHPS. The investment and operation costs of the proposed R744 system are lower than that of the existing R134a system. © 2015. This is the authors’ accepted and refereed manuscript to the article. This manuscript version is made available under the CC-BY-NC-ND 4.0 license http://creativecommons.org/licenses/by-nc-nd/4.0/

Supercritical water-cooled reactors (SCWRs) are considered as a potential development target of well-established water-cooled reactors, such as pressurized water reactors (PWRs), boiling water reactors (BWRs), or pressurized heavy water reactors (PHWRs). Applying technologies of supercritical fossil-fired power plants, the efficiency of these nuclear power plants can significantly be improved, and the construction costs can even be reduced because of simplification of plant design because the simple, once-through steam cycle of the SCWR does not require steam generators, steam dryers, or recirculation pumps. Examples of the latest options are demonstrated for the SCWR in this chapter. Several challenges in various technology areas affecting the core design are described.

The single screw compressor is widely used in air and water cooled central air conditioning systems for various unique characteristics compared to other compressors. It is possible to control the cooling or heating capacity by adjusting the mechanical slides built inside this type of compressor. However, the efficiency of the compressor is drastically reduced at part load conditions if the capacity is controlled by mechanical slides. We have studied the effect of volume index on performance when the capacity is controlled by varying the speed of the compressor. The volume index is varied by changing the position of the mechanical slides and the optimum volume index is obtained for rated and part load conditions. This paper will present the results of this study and propose the optimum volume index for obtaining best annual performance of the compressor.

This chapter discusses the various designs of light water reactors currently in operation and future developments of these, some of which are already under construction. Light water reactors currently account for almost 90% of the world’s installed nuclear capacity. The principal design features of both pressurised water reactors and boiling water reactor are discussed including their safety systems. Operational requirements are outlined and fuel design and performance are discussed. The chapter then outlines future developments in terms of both the development of advanced light water reactors and small and medium reactors.

The basic issues arising in the planning and implementation of decommissioning of Russian-type water-cooled water-moderated nuclear reactors (WWERs) are summarised. In particular, specific construction and operation features that have a significant impact on decommissioning are discussed. Other aspects, including decommissioning technologies, especially those for the dismantling of highly activated reactor components and for smooth waste and material management are addressed. The ‘in-promptu’ Greifswald decommissioning project in Germany is used as a basic reference for other WWERs, and a discussion is held of how (even under such extraordinary circumstances) the planning and management of WWER decommissioning was successfully implemented. A summary of the challenges and lessons learned from more than 20 years’ decommissioning of five WWER 440 units at the Greifswald site is provided.

In this chapter, the main corrosion-erosion peculiarities to have emerged in the primary and secondary cooling circuits of WWER-type PWRs are presented and discussed. Special attention is paid to the description of the structural and technological specificity of WWER reactors. The comprehensive survey reveals that an optimized strategy for selecting the appropriate material quality and corresponding water chemistry provides the most direct tool, and consequently, plays a key role in corrosion control and mitigation both in the primary and secondary systems.

Radiofrequency and cryoablation are analgesic modalities which provide pain relief by interrupting nociceptive pathways. While water-cooled radiofrequency and cryoablation create a demonstrable tissue lesion, pulsed radiofrequency’s mechanism of action is obscure as no discernible tissue lesion is created. Yet, due to the lack of any increased postprocedural pain and neural dysfunction, the use of pulsed radiofrequency in clinical practice is extensive and a growing body of evidence suggests that pulsed radiofrequency may be effective in certain neuropathic pain conditions. Water-cooled radiofrequency creates a larger thermal tissue lesion, and it is used primarily for pain syndromes where the pain generator has numerous and variable sources of innervation. Cryoablation generates a reversible lesion and despite its potential for limited postprocedural pain and dysfunction, it has not gained the same popularity as pulsed radiofrequency. This latter trend could be explained partially by the initial lack of success of cryoablation in relieving pain immediately after thoracotomy and the subsequent lack of industry sponsorship.

Until 1993 it was assumed that swelling and irradiation creep were phenomena of little importance to light water cooled reactors. It is now recognized that swelling and irradiation creep are in progress in austenitic internals of pressurized water reactors (PWRs) especially, with boiling water reactors (BWRs) not being as vulnerable to these processes. Some manifestations of swelling and irradiation creep are already being observed in PWRs. Owing to the non-linear development of swelling with increasing neutron exposure, it is expected that consequences of swelling and irradiation creep will accelerate, especially as PWRs move beyond their original design lives of forty years.

Ageing and plant life management (AM and PLiM) approaches for pressurized light water cooled and moderated reactors (PWRs) are discussed with the focus on major systems, structures and components (SSCs) important for safe and reliable operation and overall plant life. An overview of the PWR nuclear steam supply system (NSSS), SSCs and selected examples of operational experience concerning SSC ageing degradation (SSC-AD) is given. Important SSC-AD mechanisms and their mitigation methods are explained. Requirements and operational strategies are discussed with the view to ensure that PWR NPPs always maintain sufficient SSC safety margins in the most cost-effective way. The importance of fundamental research into SSC-AD is emphasized, as is the necessity to robustly integrate science-based knowledge into all facets of NPP manufacture, operation and regulation.

WWER operator’s strategic goal is to ensure long-term operation in an economically optimized way via effective lifetime management while ensuring plant safety. after a brief overview of technical features of WWER design, lifetime management programmes (PLiM) for WWER plants will be presented by using examples of the most important systems, structures and components. The integration of the plant activities/programmes into coherent PLiM programmes will be demonstrated taking into account the frame of regulatory requirements regarding long-term operation (LTO). The role of the international research and co-operation environment affecting the lifetime management of WWER plants will also be presented.

Within the European Fusion Technology Programme, the Water-Cooled Lithium Lead (WCLL) DEMO breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decades, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be tested in ITER-FEAT. The present paper is focused on the study of the WCLL-TBM nuclear response in ITER-FEAT, being specifically oriented to the investigation of the local effects due to the typical C-shaped tubes of the breeder zone, since they could play a pivotal role in the module-relevant thermo–mechanical design. A 3D heterogeneous model of the WCLL-TBM, realistically simulating its new lay out and taking into account 9% Cr martensitic steel as reference structural material, has been set-up. A particular attention has been paid to the simulation of the characteristic ‘C’ shape of the breeder zone double walled tubes, which have been realistically reproduced. The WCLL-TBM model has been inserted into an existing ITER-FEAT 3D semi-heterogeneous model accounting for a proper D–T neutron source. Analyses have been performed by means of MCNP-4C code running on a cluster of four workstations through the implementation of a parallel virtual machine. For each analysis a large number of histories (>10.000.000) have been simulated, obtaining statistical uncertainties on the results lower than 3%. The main features of the WCLL-TBM nuclear response have been determined focusing the attention on power deposition density, material damage through DPA and He and H production rate, daily tritium production and tritium production rate radial distribution in the module. The obtained results are herewith presented and critically discussed.

This paper involves the results of some preliminary analyses of the conceptual design of an advanced divertor plates system, alternative to the current high heat flux loaded divertor plasma facing components for ITER. In proposed concept the divertor targets are cooled by liquid metal (LM) which is compatible with the main ITER divertor water coolant. At the same time LM design option is compatible with the space envelope, thermal and mechanical interfaces as for the base ITER water cooled divertor option.

This paper concerns the preliminary investigation of the thermal behaviour and of the afterheat removal capability of the water-cooled lithium-lead box-shaped DEMO blanket in case of Loss-Of-Coolant Accident (LOCA). The thermal behaviour of the outboard-segment first wall and breeder region has been determined in an hypothetical accidental situation where convection cooling of the outboard blanket is totally lost. The case of two separate water-cooling circuits has been considered, and three possible connection schemes have been analyzed. The minimum temperature increase has been obtained when a crossed inboard/outboard cooling-circuits connection is assumed (inboard segment-box and outboard breeder-box from one side, the reverse from the other side). In case of LOCA in the outboard breeder region, the obtained maximum temperature is 550°C reached after about 20 seconds. The corresponding thermal stresses remain below the allowed limit. On the contrary, if the LOCA happens in the outboard segment-box, the maximum temperature of 625°C is reached after 1 hour in the first wall and the thermal stresses are too high. Possible solutions are briefly discussed. A similar analysis has also been performed for a different water-cooled lithium-lead blanket concept, the cylindrical module concept. The comparison between the two concepts indicates that their thermal behaviour in case of LOCA is very similar, although the cylindrical modules concept reaches slightly higher temperatures.

The European technology program for the DEMO blanket is presently focused on two main blanket concepts: Helium Cooled Ceramic Blanket (HCCB) and Self Cooled or Water Cooled liquid Lithium-Lead Blanket (SC or WC-LiPbB). Both these concepts are affected by some critical feasibility issues, i.e.: need of beryllium (chemical toxicity, high cost) as a neutron multiplier in HCCB, corrosion and activation products in LiPbB concepts, as well as MHD effects specifically in SC-LiPbB, tritium permeation and marginal breeding capability in WC-LiPbB. On the basis of these considerations, use of pure liquid lithium was reconsidered as a possible alternative for DEMO blanket applications. Its properties, advantages and drawbacks have been reviewed and a conceptual blanket design has been developed. The results show that the Helium Cooled liquid Lithium Blanket (HCLiB) is an attractive concept from the thermohydraulic and neutronic points of view even though concerns remain with respect to the chemical reactivity of liquid lithium.

A quality control plan was evolved for the NDE of Aluminium-Uranium (U233) alloy plate type nuclear fuel elements for the light water cooled and moderated research reactors PURNIMA III and KAMINI. Inspection was carried out at each stage of fabrication using x-ray radiography and microdensitometric scanning for fuel homogeneity evaluation; blister testing and immersion ultrasonic testing for gross and fine non-bond evaluation and metallography for microstructure evaluation. Conventional NDT methods were modified to suit the stringent quality requirements for fuel homogeneity and bond quality. Digital image processing, continuous motion radiography system and immersion ultrasonic testing system with scanner manipulator were developed and successfully employed for the quality control of the fuel plates.

Two CELIMENE modules have been operated during 3000 hours at the same maximum temperature (450°C) but the first one without magnetic field and the second one in the presence of a 1.4 T magnetic field. Comparison of dissolution and deposition rates obtained in parallel and perpendicular directions evidences that the magnetic field induces a dissymmetry for these two phenomena. Moreover the results obtained in the two CELIMENE tests show that the corrosion rate appears to be increased by the magnetic field.

Publisher Summary This chapter discusses the investigation of water cooled lithium–lead blankets jointly performed by CEA and the JRC of ISPRA in the framework of the European DEMO blanket study program. The reactor characteristics considered in the study were the following: DEMO-1 with double null plasma, neutron wall loading, outboard breeding blanket thickness, inboard breeding blanket thickness, no breeding elements behind diverters, unprotected first wall, and martensitic steel as structural material. Both inboard and outboard blankets were formed by 48 segment boxes made of water cooled steel structures that integrated the first wall and confined the breeding modules. These breeding modules consisted of cylindrical containers, whose axes were adapted to the toroidal D-shaped geometry; they are straight in the inboard region and curved in the outboard region. Each module contained lithium–lead and was cooled by a water tubes bundle. A continuous flowing of lithium–lead was imposed by tritium extraction outside the blanket.

Within the frame of the ITER conceptual design phase a Water Cooled Ceramic Blanket (WCCB) design has been developed by ENEA/ANSALDO, in close collaboration with the NET Team, for application as ITER Driver Blanket. Such a design has been selected as European design proposal for the ITER Driver Blanket and is now under comparison with the Japanese, United States and Soviet Union design proposals. An out of pile experiment is being performed to investigate the thermal-hydraulic and mechanical performance of a typical blanket module. Status of design and experimental activities is reported together with a summary of the main features and performances of the EC blanket, based on detailed analyses performed in several fields of investigation (Neutronics, Thermal-hydraulics, Thermomechanics, Tritium Dynamics, Safety, Industrial Feasibility, etc) during the past two years.

Publisher Summary This chapter reviews basic principles necessary for the prediction of response of a reactor to the transients that are of interest for an assessment of reactor safety. Public health and safety as well as safe operation of a power plant are the concern of all nuclear engineers, entailing special responsibilities because of the large inventory of fission products contained in the core of the reactor. Analysis of the consequences of operational transients and accidents takes two forms. The first is experimental, where the transient of interest and the consequences of the transient are observed. This approach has limitations when the transient being considered is expected to damage the plant. The second form of analysis relies upon the development of predictive capability to permit the analysis of the consequences of transients with the aid of digital computers or analog computers.

Experiments were performed on a 204mm diameter water cooled cyclone to measure the pressure drop and heat transfer in different sections of the cyclone. Hot gas at 250 °C entered the cyclone with and without suspended particles. Heat transfer and pressure drop in solid free gas were compared with those measured for suspensions of different solids.

ASME Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components”, began as a joint effort of ASME, ANSI-N45 and the USAEC in 1966–67. The first draft version of the Code appeared in 1968 with the first official release in 1970. The original Subgroup was a part of Section III and consisted of thirteen members; four still are active in Section XI. Currently the membership exceeds 200. In the period 1970–1986, Divisions dealing with Gas-Cooled and Liquid-Metal Cooled reactors were added as well as major expansions in Division 1 on Light Water Cooled reactors. For example, there is Class 1 (IWB), Class 2 (IWC), Class 3 (IWD), Component Supports (IWF), Pumps (IWP), Valves (IWV), Steel Containments (IWE) and Concrete Containments (IWL). A review is under way to determine if a Subsection on Core Support Structures (IWG) should be incorporated. Section XI has embarked on several new activities in the period 1984–88. These include a major effort in modifying the Code to cover Plant Life Extension. Fracture mechanics criteria are being developed to handle cracked austenitic and ferritic steel piping and to establish criteria for handling reactor pressure-temperature transients. A major effort and the topic of this abstract has been in the areas of nondestructive examination and testing. These activities, starting with changes in the 1986 edition of the code, are discussed herein.

After dismantling of the titanium neutralizer plates inspection yielded satisfactory status of flat areas whereas edges and curved shapes were heavily melted. At the inner plates of the lower divertor strongly focused melting and cutting was found which is caused by fast electrons. These electrons are continuously produced. The production mechanism is not yet clear but runaway processes can be excluded.

The meteorological data recorded over nine years at Dhahran have been analysed to establish the design climatic parameters which have influences on the performance of the solar cooling system and the building. System model based on analysis and recent empirical results have been presented. The time dependent periodic performance have been presented for the building. Based on long term weather data, problem areas for the operation of water cooled lithiumbromide-water absorption cooling system have been identified.

The NET liquid breeder blanket was subjected to a major design revision and integrated in the new NET-DN machine configuration. In this paper briefly the most essential design features are summarized and some results from thermohydraulics and 1D as well as 3D neutronics analyses are presented. It is concluded that the performance meets well the requirements of NET but that the concept needs substantial improvement if applied to a reactor.

This chapter discusses various generation techniques and presents a broad concept of apparatus and methods involved in the production of thermal energy for distribution of heat services. It describes three-pass smoke tube economic boiler, twin furnace boiler, and reverse-flow furnace boiler. In considering three-pass smoke tube boiler plant for a particular service, it is most advisable to ensure provision of boilers specifically designed for the duty within scope of media to be generated and of requisite operational parameters. Various types of water tube boilers have in the past been developed to complement ranges of thermal generators within scope of largest three-pass economics and higher ratings and within the context of design to suit major heating and industrial applications. The water tube boiler is a rapid heat generator with quick response to changes in load by the virtue of its small water content and integral part of boiler circulation system being formed by water cooled tubular furnace walls, promoting efficient circulation. The larger sizes occupy floor space more economically than equivalent multiple smoke tube generators and the water tube boiler is easily adaptable for a wide variety of fuels.

The extraction of long pulse or DC-beams at high power level needs extraction grids which are cooled in a very intense and direct way.

The very low productivity of conventional methods employed for heat treating railroad rails limits the wider application of hardened rails. This paper gives a brief history and the current status of The Algoma Steel Corporation's process for heat treating railway rails in-line with the hot rolling mill. Property comparisons between in-line and off-line head hardened rails are given and tentative product specifications for the first commercial Accelerated Water Cooled products are presented.

The standard confinement field of the advanced stellarator WENDELSTEIN VII-AS is exclusively generated by a modularly designed coil set. Fourty-five twisted coils produce a toroidal field of 3 T during a pulse time of 5 sec. Water cooled copper wires carry a winding current of 40 kA at 3.3 kV. Force components in the radial, poloidal and toroidal directions are calculated for the individual coils using EFFI-code /1/. On the basis of an optimized support concept a stress analysis for the coils was carried out using SAP V (2) /2/.

One of the most vulnerable components in a fusion reactor may be the first wall, which is facing the plasma and which must be water cooled.

The technological performance of the water cooled Nickel screens is described. The benefits of Beryllium as a screen material in terms of plasma radiation, mechanical reliability and thermal suitability are quantified. Design features of the new Beryllium screens are described.

The performance of Air cooled solar water pump of 2600 litres/day has been studied over a period of seven months and the results and operational problems are discussed. A 1,20,000 litres/day, 12 meter lift, water cooled pump has been commissioned and its operational problems are reported. A new design called ‘Direct cooled pump’ of 100 litres/cycle and 10,000 litres/cycle has been constructed and installed to operate for a number of cycles in a day by simple modification of air cooled pump.

Publisher Summary No treatment of nuclear power safety would be complete without including a discussion of WASH-1400, Reactor Safety Study, sponsored by the United States Nuclear Regulatory Commission. Risk is a commonly used word that can convey a variety of meanings to different people. Its estimates are made of potential fatalities and injuries to people and of property damage resulting from both nuclear power plant and non-nuclear accidents. The results suggest that the risks to the public from potential accidents in nuclear power plants are comparatively small. In addition to fatalities and property damage, a number of other health effects could be caused by nuclear accidents. These include injuries and long-term health effects such as cancers, genetic effects, and thyroid gland illness. The number of cases of genetic effects and long-term cancer fatalities is predicted to be smaller than the normal incidence rate of these diseases. Even for a large accident, the small increases in these diseases would be difficult to detect from the normal incidence rate. Thyroid illnesses that result from a large accident are mainly the formation of nodules on the thyroid gland, which lead to serious consequences.

Tore Supra possesses a completely water cooled first wall for 30 s long pulse duration at 230°C with 25 MW of additionnal heating power injection. The first wall is composed of: - water cooled stainless steel panels covering 65 m 2 of the external part of the inner vessel having a heat removal capacity of 12 MW - an actively cooled graphite wall of 13 m 2 , located on the inner part of the chamber having a heat removal capacity of 13 MW. For independant vessel baking, a 50 KW electrical resistor system located on the inner vessel wall is foreseen.

The object of this work is to study the effect of combustion air swirl, the swirler to furnace diameter ratio and the air-fuel ratio on flame radiation characteristics. Experiments were carried out on a cylindrical oil-fired furnace which consists of seven water cooled axial segments. Measurements of the red brightness temperature for different gas segments and the gas temperature for different points were conducted. The zonal exchange method was applied for flame radiation calculations and a computer program was made to calculate the different geomtrical and thermodynamic properties for each gas zone as well as its radiant energy transfer to the furnace walls. From the obtained results, it was found that both the intensity of swirl and the air-fuel ratio affect markedly the flame radiation characteristics, while the swirler to furnace diameter ratio has a relatively small effect.

The manufacturing feasibility of water cooled liquid breeder blanket modules for NET has been studied by STEIN INDUSTRIE under contract NET/84–708 and continuous contact with the NET-Team. According to the basic design proposal, the blanket modules consist of banana like bent stainless steel pressure tubes of about 7 m length and 0.15 – 0.20 m diameter. These tubes are filled with slowly circulating eutectic Li 17 Pb 83 which is cooled via helicoil coolant tubes.

A simple method for the measurement of the relative volume of non-condensable gases in vapour has been applied. The relative volume of air in the vapour discharged by the condenser of a refrigerating plant, without automatic removal of air, is nearly always higher than 10 to 20%. With an automatic system the relative volume of air can be kept down, for instance from 3 to 5%, thus reducing condenser water or the energy consumption of the compressor. To provide efficient and economical air removal and full condenser capacity, water cooled condensers should be constructed in accordance with the directives set out in this paper. The construction of air cooled condensers should also cater for efficient air removal and ensure continuous liquefied refrigerant flow through all parallel branches of the finned tubes. Comparative measurements show that, by improved construction, the weight and the bulk of condensers can be decreased by approximately 20 to 40%.