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At least 127 records · Page 7

All sources NASA NTRS DOE OSTI

DOE OSTI2017

Ceramic oxygen transport membrane array reactor and reforming method

The invention relates to a commercially viable modular ceramic oxygen transport membrane system for utilizing heat generated in reactively-driven oxygen transport membrane tubes to generate steam, heat process fluid and/or provide energy to carry out endothermic chemical reactions. The system provides for improved thermal coupling of oxygen transport membrane tubes to steam generation tubes or process heater tubes or reactor tubes for efficient and effective radiant heat transfer.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2018

Phenomena Identification and Ranking Table (PIRT) study for metallic structural materials for advanced High-Temperature reactor

The Fluoride High-Temperature Reactor (FHR) technology promises many benefits including passive safety, proliferation-resistant waste forms, and improved economics. However, selection of reliable structural materials and identification of the possible degradation mechanisms for these is important for the licensure and the safe operation of FHRs. In order to address this task, the Georgia Tech led Integrated Research Project (IRP) hosted a Phenomena Identification and Ranking Table (PIRT) panel of experts to address degradation mechanisms and other materials related issues of importance to the FHRs. Futhermore, materials, ones that come in contact with FLiBe or FLiNaK molten salts or other related environments like high temperature steam etc., were considered in this PIRT. Focus of this PIRT was the metallic alloys, especially the ones that are permitted for the construction of elevated temperature Class A components by the ASME code. Degradation mechanisms considered in this PIRT included chemical degradation, mechanical degradation, radiation degradation, and synergistic effect of these mechanisms that may negatively impact operations or cause some safety concerns for the major structural components of FHRs. Main components which were considered included vessel and primary piping, primary heat exchangers, steam generator vessel, steam generator tubes, intermediate loop piping, valves and pumps. Welds in all structural components were identified as an important class of material, which varies in composition and properties, and needs more attention. Importance of impurity control in molten fluorides considered for FHR was highlighted throughout PIRT panel discussions. As a result, this paper gives a summary of important results from the PIRT panel discussions and report.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Impact on Cycle Efficiency of Small CHP Plants from Increasing Firing Temperature Enabled by AM of Turbine Blades

Combined Heat and Power (CHP) systems are gaining popularity because of their potential for high overall thermodynamic efficiency and increased need for distributed power generation. Most CHP systems include a gas turbine for electricity generation and a heat recovery steam generator (HRSG) for steam generation. The steam can be used to power a steam turbine for additional electricity generation or to drive rotating equipment, for space heating, for absorption chillers, etc. A CHP cycle configuration is often driven by the ‘principal’ utility for the facility that it services; steam or electricity. In either case, performance improvements in the gas turbine have the potential to increase the steam and power output from the cycle, which is a direct result of improvements in the gas turbine efficiency and power output. One possible opportunity to improve an existing gas turbine’s performance is to increase the firing temperature with improved turbine cooling and increased compression ratio.<br>In this study, the impact on CHP cycle performance from increasing the turbine firing temperature by 100 °C and improving the turbine blade cooling for a 6-MW scale gas turbine is estimated using an analytical cooled gas turbine model and a steam cycle model. A sensitivity analysis was performed to understand the impact of increasing the internal cooling effectiveness, thermal barrier coating performance and blade material upgrades on gas turbine and CHP cycle efficiency. The impacts of turbine blade cooling improvements were studied for three common CHP cycle configurations identified from the literature. Various definitions for CHP cycle efficiency from the literature are used in the comparisons. The results showed that a 100 °C increase in firing temperature can increase the gas turbine efficiency by 1 percentage point without improving cooling effectiveness and add 2 additional percentage points in efficiency by using advanced turbine blades with higher internal cooling efficiency. Studied engine upgrades showed potential for increasing the CHP cycle efficiency by 3 percentage points while increasing the steam generation rate by 8%.<br>

Uysal, Selcuk Can↗

DOE OSTI2020

Advanced Steam-Electric Power Cycle for Advanced Reactor Concepts

This report presents improvements made to RELAP-7 for modeling Rankine power cycles, which are used on the secondary side of some advanced reactor concepts. For example, for a High-Temperature gas-cooled Reactor Pebble bed Module (HTR-PM), in the primary loop, helium transfers heat from the fuel to a steam generator, and on the secondary side, dry steam from the steam generator applies work to a turbine, which is converted into electrical power by a generator. After passing through the turbine, the steam passes through a condenser and pumped to the steam generator inlet, completing a Rankine cycle on the secondary side. This HTR-PM concept is used as an example to demonstrate a steam-electric power cycle capability in RELAP-7. This power cycle requires a number of different physical components, including a pump, compressor, turbine, steam generator, and condenser. The steam generator is modeled using a heat exchange between primary and secondary loops, and the condenser is modeled using a cooling source term. Recent efforts demonstrated preliminary models for the turbomachinery components (pump/compressor and turbine), but more physical models were required. Furthermore, several improvements were made for the robustness of mixed single-phase/two-phase flow present in a Rankine cycle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Oxygen transport membrane reactor based method and system for generating electric power

A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2017

TRACE Analysis of Loss of Flow for the APR1400

The APR1400 is a pressurized water reactor (PWR). The APR1400 nuclear steam supply system consists of a reactor pressure vessel (RPV), which contains the reactor core, and two primary reactor coolant system loops connected in parallel to the RPV. Each coolant loop consists of a hot leg (HL), a u-tube steam generator (SG), two reactor coolant pumps (RCPs) and two cold legs (CLs). A pressurizer is connected to the hot leg on one of the coolant loops. During normal operation, the RCPs circulate high pressure water through the reactor core to remove the heat. The heated water exits the RPV and passes through the hot legs into the steam generator tubes. Heat is transferred through the walls of the steam generator tubes into a lower pressure secondary coolant system. Cooled primary system water from the steam generators is returned through the RCPs and cold legs to the RPV. The RCPs operate at constant speed; the primary coolant system is designed to provide a prescribed rate of coolant flow through the reactor core. The APR1400 SG secondary coolant system pressure is controlled through modulation of valves in the steam lines to maintain a prescribed turbine inlet pressure. The SG secondary coolant inventory is controlled by using a feedwater control system to maintain a prescribed liquid level in the steam generator downcomer. This water inventory level is sufficient to cover the outer surfaces of the steam generator tubes while preventing overfill that would cause liquid to pass from the steam generators through the steam lines into the turbine system. In the current work, the Nuclear Regulatory Commission's reactor systems analysis code, TRACE?2 was used to perform an analysis of a loss of flow anticipated operational occurrence (AOO). The loss of flow is initiated by a loss of off-site power (LOOP), resulting in a trip of the RCPs. The regulatory purpose of the current work is to confirm safety analyses documented in the APR1400 Design Control Document (DCD)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Application of European Large Component Technology in the United States - 12561

The European nuclear community currently has safer, more environmentally-friendly disposal options for large component disposal than the United States. The technology and innovation that makes these options viable will soon be available stateside. The removal and replacement of steam generators is a large and complicated task making volume reduction a vital service for the nuclear power industry. As plants age, the replacement of certain components is required in order to maintain plant efficiency and public safety. Currently in the United States, there are 3 options for the disposal of these items including: on-site storage, direct-disposal at a licensed (low-level radioactive waste) LLWR site, and volume reduction prior to disposal. If today's nuclear plants are to be used by future generations, waste stored on-site will eventually have to be disposed of to make room for newer, more advanced technology. Also, the space at LLWR disposal sites is finite, so all waste sent for disposal should be minimized whenever possible. Reducing the volume of decommissioned steam generators prior to final disposal shows the public that the nuclear industry is doing everything that it can to minimize its waste and conserve the environment. Additionally, many steam generators are massive and are not easily transported to a disposal site. When there is no practical transportation route, the size and weight of the component must be reduced. For various reasons including public safety, cost, and technical expertise, size reduction cannot be practically performed at the site of generation. However, since most facilities have some kind of access to a waterway, the components can be safely transported to another facility with barge access for size reduction. The size reduction process for steam generators has evolved over the past 5 years. Because of reductions to the amount of final waste, the dose to personnel, and the processing time, it is now an environmentally and economically friendly alternative to on-site storage and direct disposal. The major operations of this process include: - removal of the carbon steel steam generator dome; - removal of the channel head and majority of tube sheet to free the tubes from the tube sheet and gain access to all of the tubes for grit blasting; - grit blast decontamination of the steam generator tubes to allow disassembly; - removal of the U-Bend Section of the tubes to allow for tube pulling - tube pulling operations; - disassembly of the remainder of the steam generator shell and tube support assemblies; - size reduction of the steam generator resultant materials. All of the equipment for the American market is based on previous designs but has been adapted to meet the size requirements for processing of larger steam generators. This includes a large horizontal band saw, a tube pulling machine, tube blasting equipment, and a diamond wire saw frame. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Simulated performance of biomass gasification based combined power and refrigeration plant for community scale application

Thermal performance analysis and sizing of a biomass gasification based combined power and refrigeration plant (CPR) is reported in this study. The plant is capable of producing 100 kWe of electrical output while simultaneously producing a refrigeration effect, varying from 28-68 ton of refrigeration (TR). The topping gas turbine cycle is an indirectly heated all-air cycle. A combustor heat exchanger duplex (CHX) unit burns producer gas and transfer heat to air. This arrangement avoids complex gas cleaning requirements for the biomass-derived producer gas. The exhaust air of the topping GT is utilized to run a bottoming ammonia absorption refrigeration (AAR) cycle via a heat recovery steam generator (HRSG), steam produced in the HRSG supplying heat to the generator of the refrigeration cycle. Effects of major operating parameters like topping cycle pressure ratio (r{sub p}) and turbine inlet temperature (TIT) on the energetic performance of the plant are studied. Energetic performance of the plant is evaluated via energy efficiency, required biomass consumption and fuel energy savings ratio (FESR). The FESR calculation method is significant for indicating the savings in fuel of a combined power and process heat plant instead of separate plants for power and process heat. The study reveals that, topping cycle attains maximum power efficiency of 30%in pressure ratio range of 8-10. Up to a certain value of pressure ratio the required air flow rate through the GT unit decreases with increase in pressure ratio and then increases with further increase in pressure ratio. The capacity of refrigeration of the AAR unit initially decreases up to a certain value of topping GT cycle pressure ratio and then increases with further increase in pressure ratio. The FESR is found to be maximized at a pressure ratio of 9 (when TIT=1100°C), the maximum value being 53%. The FESR is higher for higher TIT. The heat exchanger sizing is also influenced by the topping cycle pressure ratio and GT-TIT.

09 BIOMASS FUELS↗

DOE OSTI2015

Waste remediation

A system including a steam generation system and a chamber. The steam generation system includes a complex and the steam generation system is configured to receive water, concentrate electromagnetic (EM) radiation received from an EM radiation source, apply the EM radiation to the complex, where the complex absorbs the EM radiation to generate heat, and transform, using the heat generated by the complex, the water to steam. The chamber is configured to receive the steam and an object, wherein the object is of medical waste, medical equipment, fabric, and fecal matter.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2017

Waste remediation

60 APPLIED LIFE SCIENCES↗

NASA NTRS1981

The small community solar thermal power experiment. Parabolic dish technology for industrial process heat application

Aspects of incorporating a thermal energy transport system (ETS) into a field of parabolic dish collectors for industrial process heat (IPH) applications were investigated. Specific objectives are to: (1) verify the mathematical optimization of pipe diameters and insulation thicknesses calculated by a computer code; (2) verify the cost model for pipe network costs using conventional pipe network construction; (3) develop a design and the associated production costs for incorporating risers and downcomers on a low cost concentrator (LCC); (4) investigate the cost reduction of using unconventional pipe construction technology. The pipe network design and costs for a particular IPH application, specifically solar thermally enhanced oil recovery (STEOR) are analyzed. The application involves the hybrid operation of a solar powered steam generator in conjunction with a steam generator using fossil fuels to generate STEOR steam for wells. It is concluded that the STEOR application provides a baseline pipe network geometry used for optimization studies of pipe diameter and insulation thickness, and for development of comparative cost data, and operating parameters for the design of riser/downcomer modifications to the low cost concentrator.

Polzien, R. E.↗

NASA NTRS1981

Steam Rankine Solar Receiver, phase 2

A steam rankine solar receiver (SRSR) based on a tubular concept was designed and developed. The SRSR is an insulated, cylindrical coiled tube boiler which is mounted at the focal plane of a fully tracking parabolic solar reflector. The concentrated solar energy received at the focal plane is then transformed to thermal energy through steam generation. The steam is used in a small Rankine cycle heat engine to drive a generator for the production of electrical energy. The SRSR was designed to have a dual mode capability, performing as a once through boiler with and without reheat. This was achieved by means of two coils which constitute the boiler. The boiler core size of the SRSR is 17.0 inches in diameter and 21.5 inches long. The tube size is 7/16 inch I.D. by 0.070 inch wall for the primary, and 3/4 inch I.D. by 0.125 inch wall for the reheat section. The materials used were corrosion resistant steel (CRES) type 321 and type 347 stainless steel. The core is insulated with 6 inches of cerablanket insulation wrapped around the outer wall. The aperture end and the reflector back plate at the closed end section are made of silicon carbide. The SRSR accepts 85 kwth and has a design life of 10,000 hrs when producing steam at 1400 F and 2550 psig.

Deanda, L. E.↗

DOE OSTI2019

Collaboration for the Management of the Bruce Power Major Component Replacement (MCR) Nuclear Waste - 19630

Bruce Power's Major Component Replacement (MCR) Program will generate large volume waste via steam generators (SGs) and intermediate-level nuclear waste, which will require specialized packaging and handling to safely receive and store it at Ontario Power Generation's (OPG) Western Waste Management Facility (WWMF). The refurbishment of a Candu reactor involves, in part, the removal and replacement of all of a reactor core's fuel channel assemblies. Bruce Power and OPG collaborated to develop nuclear waste management solutions to optimize the use of OPG's existing Steam Generator Storage Building (SGSB) and Retube Component Storage Building (RCSB). By designing specialized Retube Waste Containers (RWCs) and optimizing configurations for the waste to be stored in the buildings, we will reduce the interim nuclear waste storage needs at the OPG WWMF, while improving Bruce Power MCR process efficiencies and reducing the overall costs of the MCR Program (Units 3 through 8). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

CRUD Source Term Assessment and Development for VERA

In pressurized water reactors, corrosion of primary loop materials is an ongoing process, as primary loop components (steam generator, hot leg, cold leg, and core) are constantly exposed to the coolant. As a result of this corrosion, primary loop materials are released into the coolant in the form of aqueous metal ions and particulate oxides. These corrosion products are the source of materials that form crud deposits on fuel rod clad surfaces. The mass of corrosion products available to create crud will limit the rate and extent of crud buildup that occurs. This is commonly referred to as the crud “source term”. The major contribution to the source term mass comes from the steam generator, due to the large surface area exposed to the coolant. Steam generators are not identical; the wetted area exposed to the coolant and the alloy used in the steam generator tubes vary. Different alloys corrode at different rates; Inconel 690 corrodes at approximately one-third of the rate of corrosion of Inconel 600. The mass of corrosion products released from the steam generator also varies by the area exposed to the coolant. The release of these corrosion products depends on the local environmental factors within the primary loop, including temperature, pH, local fluid conditions, electrochemical potential, and thermodynamic solubilities of the species present within this system. As such, each PWR will have a different crud source term based on these plant-specific parameters. In order to accurately model and predict crud growth and issues that arise from the presence of crud (CIPS, CILC), the source of these corrosion product materials must be accurately modeled in a system mass balance. These models should be mechanistic, representative of the underlying physics phenomena, and account for the effects of the plant-specific design and operating parameters, including the alloys used, local temperatures, and fluid flow conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Development of Scaling Criteria for Modeling Condensation in Horizontal Tubes Using Reduced Pressure System and Simulant Fluid

Most of the currently operating nuclear power plants (NPPs), generation-II and III, are equipped with active safety systems, which raises concerns regarding those safety systems' operational reliability and the suitability of the accidents prevention measures. Following the severe accidents at nuclear power plants at Three Mile Island, Chernobyl, and recently in Fukushima some advance Pressurized Water Reactors (PWRs) have adopted passively cooled steam generator natural circulation system to remove the decay heat passively through the steam generator (SG). This is done by condensing steam from the steam generator inside a heat exchanger submerged in a tank of water or an air cooled system. The APR+ (Advanced Power Reactor Plus), which is currently being developed in South Korea, has adopted a Passive Auxiliary Feedwater System (PAFS), which removes the decay heat through the steam generator replacing the conventional active auxiliary feedwater system. The main part of the PAFS, in where the steam condenses, is the Passive Condensation Heat Exchanger (PCHX). The steam condenses under high pressure (7.4 MPa, 290 deg. C) in 8.4 m long horizontal tube, with 3 degrees inclination, with inner diameter of 44.8 mm and wall thickness of 3 mm. The PCHX consists of 4 bundles, each bundle has 60 condensation tubes. The performance of the PCHX performance should be evaluated. Condensation has been extensively studied; however, most previous experimental work was directed to low pressure (i.e., near atmospheric pressure) and small condensation tubes. Moreover, studies on the horizontal type heat exchanger are limited, since previous safety systems and experimentation work were directed to the vertical type heat exchangers. Due to the high construction cost of a full size high pressure condensation experiment, a reduced-size reduced-pressure condensation experiment is suggested to evaluate the heat removal performance of a condensation tube of the PAFS heat exchanger. A scaling methodology to be established to design a reduced-size reduced-pressure condensation experiment. This study attempts to develop a scaling criteria for the PAFS condensation system, and to apply it for modeling a high pressure water system by a low pressure Freon system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Degradation Mechanisms and Development of Protective Coatings for Thermal Energy Storage and High Temperature Fluid Containment Materials

This research addresses cost savings to achieve the SETO target of $15/kWh th for thermal energy storage and 6¢/kWh for levelized cost of energy. Next-generation solar power conversion systems require high-temperature, advanced fluids in the range of 600° to 900°C. The three major advanced fluid groups—supercritical CO 2 , liquid metal alloys, and molten salts—are very corrosive to common alloys used in vessels, heat exchangers, and piping at these temperatures. Understanding degradation mechanisms is essential for making recommendations on proper working conditions and suitable cost-effective coating materials for CSP applications at high temperatures. The implementation of the findings from this research will significantly reduce investment risk by increasing the lifetime of the CSP plant components. In the past, the majority of the hot corrosion evaluations have been performed for waste incinerators, gas turbine engines, and electric power generation (steam-generating equipment) applications for different materials, molten-salt systems, and coatings. Containment materials and resistant coatings for supercritical CO 2 , molten salts and liquid metal alloys have not being evaluated for CSP applications. We plan to understand the underlying degradation mechanisms to select protective approaches such as anodic passivation, cathodic protection, and coating designs.The novelty of the proposed research is the development of protective coatings and working conditions for materials in contact with high temperature TES and HTF. Current state-of-the-art materials and protection methods are non-existent. Novel approaches are: the use of graphite nanomaterials to enhance the non-wettability of the aggressive fluids with the coating; self-healing behavior of Si-coatings in liquid Al-Si alloys; anodic passivation in molten salts to protect metallic substrates and the modeling prediction of attack in liquid metal alloys.

14 SOLAR ENERGY↗

DOE OSTI2023

Dynamic Modeling of a Solar-To-Hydrogen Flexible High Temperature Steam Electrolysis Plant

Sustainble hydrogen production for use as a renewable combustible fuel and clean chemical feedstock is an important objective as the world moves towards a renewable energy future. High temperature steam electrolysis is a promising hydrogen production technology due to its reduced electric input that is offset by heat input into steam generation and steam superheating. An option to provide this heat is to use concentrating solar thermal technology that can sustainably provide heat input while renewable electricity is used for the electrolysis reaction. In this work, a solar-to-hydrogen high temperature steam electrolysis plant is designed and dynamically modeled, showing continuous hydrogen production by utilizing supplemental heating and efficient recuperative heating from the electrolysis product streams. Through this design, over 90% of the required heat input for the process can by met by a combination of solar and recuperative heat. Additionally, the plant can flexibility operate by ramping down hydrogen production and through flexible heat integration, which intelligently integrates solar heat based on solar conditions. Smooth operation with flexible hydrogen production is demonstrated which decreases electrical input during on-peak grid times and also decreases the total supplemental heat load over the course of a day from 26.1% to 24.5%. In addition, by using flexible heat integration, the plant can increase its solar heat usage by 4.1% relative to a base case. Both options for flexibility show efficient use of solar thermal energy to sustainably and continuously produce hydrogen.

Immonen, Jake (ORCID:0000000341231625)↗

DOE OSTI2012

The pre-conceptual design of the nuclear island of ASTRID

The CEA is involved in a substantial effort on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) pre-conceptual design in cooperation with EDF, as experienced Sodium-cooled Fast Reactor (SFR) operator, AREVA, as experienced SFR Nuclear Island engineering company and components designer, ALSTOM POWER as energy conversion system designer and COMEX NUCLEAIRE as mechanical systems designer. The CEA is looking for other partnerships, in France and abroad. The ASTRID preliminary design is based on a sodium-cooled pool reactor of 1500 MWth generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and the main components with regard to future commercial reactors. ASTRID lifetime target is 60 years. Two Energy Conversion Systems are studied in parallel until the end of 2012: Rankine steam cycle or Brayton gas based energy conversion cycle. ASTRID design is guided by the following major objectives: improved safety, simplification of structures, improved In Service Inspection and Repair (ISIR), improved manufacturing conditions for cost reduction and increased quality, reduction of risks related to sodium fires and water/sodium reaction, and improved robustness against external hazards. The core is supported by a diagrid, which lay on a strong back to transfer the weight to the main vessel. AREVA is involved in a substantial effort in order to improve the core support structure in particular regarding the ISIR and the connection to primary pump. In the preliminary design, the primary system is formed by the main vessel and the upper closure comprising the reactor roof, two rotating plugs - used for fuel handling - and the components plugs located in the roof penetrations. The Above Core Structure deflects the sodium flow in the hot pool and provides support to core instrumentation and guidance of the control rod drive mechanisms. The number of the major components in the main vessel, primary pumps, Intermediate Heat Exchangers, and Decay Heat Exchangers are now under consideration. Under normal conditions, power release is achieved using the steam/water plant (in case of Rankine steam cycle) or the gas plant (in case of Brayton gas cycle). The diverse design and operating modes of Decay Heat Removal systems provide protection against common cause failures. A Decay Heat Removal system through the reactor vault is in particular studied with the objective to complement Direct Reactor Cooling systems. At this stage of the studies, the secondary system comprises four independent sodium loops (two and three sodium loops configurations are also investigated). Each loop includes one mechanical pump (or a large capacity Annular Linear Induction Electromagnetic Pump), and three modular Steam Generator Units characterized by once through straight tube units with a ferritic tube bundle; nevertheless, helical coil steam generator with tubes made of Alloy 800, and inverted type steam generator with a ferritic tube bundle are also investigated. The limited power of each modular Steam Generator Unit allows the whole secondary loop to withstand a large water/sodium reaction consecutive to the postulated simultaneous rupture of all the heat exchange tubes of one module. The arrangement of the components is based on the 'Regain' concept, in which the secondary pump is situated at a low level in the circuit; conventional arrangement, as SUPERPHENIX type, is a back-up option. Alternative arrangements based on gas cycles are also studied together with Na-gas heat exchanger design. This paper presents a status of the ASTRID pre-conceptual design. The most promising options are highlighted as well as less risky and back-up options. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗