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Proceedings Papers
Proc. ASME. HTR2008, Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 1, 733-738, September 28–October 1, 2008
Paper No: HTR2008-58199
Abstract
The Next Generation Nuclear Plant (NGNP) will most likely produce electricity and process heat, with both being considered for hydrogen production. To capture nuclear process heat, and transport it to a distant industrial facility requires a high temperature system of heat exchangers, pumps and/or compressors. The heat transfer system is particularly challenging not only due to the elevated temperatures (up to ∼ 1300K) and industrial scale power transport (≥50 MW), but also due to a potentially large separation distance between the nuclear and industrial plants (100+m) dictated by safety and licensing mandates. The work reported here is the preliminary analysis of two-phase thermosyphon heat transfer performance with alkali metals. A thermosyphon is a device for transporting heat from one point to another with quite extraordinary properties. In contrast to single-phased forced convective heat transfer via ‘pumping a fluid’, a thermosyphon (also called a wickless heat pipe) transfers heat through the vaporization / condensing process. The condensate is further returned to the hot source by gravity, i.e. without any requirement of pumps or compressors. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. Two-phase heat transfer by a thermosyphon has the advantage of high enthalpy transport that includes the sensible heat of the liquid, the latent heat of vaporization, and vapor superheat. In contrast, single-phase forced convection transports only the sensible heat of the fluid. Additionally, vapor-phase velocities within a thermosyphon are much greater than single-phase liquid velocities within a forced convective loop. Thermosyphon performance can be limited by the sonic limit (choking) of vapor flow and/or by condensate entrainment. Proper thermosyphon requires analysis of both.
Proceedings Papers
Proc. ASME. HTR2008, Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 1, 767-771, September 28–October 1, 2008
Paper No: HTR2008-58274
Abstract
The paper deals with the issue of increasing efficiency of nuclear power plants with the modular high-temperature helium reactor (HTGR) and direct gas turbine cycle. It should be noted that only this combination can highlight the advantages of the HTGR, namely the ability to heat helium to about 1000°C, in comparison with other reactor plants for electricity generation. The HTGR has never been used in the direct gas turbine cycle. At present, several designs of such commercial plants are at the stage of experimental validation of main technical features. In Russia, “OKB Mechanical Engineering” together with “General Atomics” (USA) are developing the GT-MHR project with the reactor power of 600 MW, reactor outlet helium temperature of 850 °C, and efficiency of about 45.2%; the South African Republic is developing the PBMR project with the reactor power of 400 MW, reactor outlet helium temperature of 900 °C, and efficiency of about 42%; and Japan is developing the GTHTR-300 project with the reactor power of 600 MW, reactor outlet helium temperature of 850°C, and efficiency of about 45.6%. As it has been proven by technical and economic estimations, one of the most important factors for successful promotion of reactor designs is their net efficiency, which must be not lower than 47%. A significant advantage of a reactor plant with the HTGR and gas-turbine power conversion unit over the steam cycle is considerable simplification of the power unit layout and reduction of the required equipment and systems (no steam generators, no turbine hall including steam lines, condenser, deaerator, etc.), which makes the gas-turbine power conversion unit more compact and less costly in production, operation and maintenance. However, in spite of this advantage, it seems that in the projects currently being developed, the potential of the gas-turbine cycle and high-temperature reactor to more efficiently generate electricity is not fully used. For example, in modern reactor plants with highly recuperative steam cycle with supercritical heat parameters, the net efficiency of electricity generation reaches 50–55%. There are three methods of Brayton cycle carnotization: regeneration, helium cooldown during compression, and heat supply during expansion. These methods can be used both separately and in combination, which gives a total of seven complex heat flow diagrams. Besides, there are ways to increase helium temperature at the reactor inlet and outlet, to reduce hydraulic losses in the helium path, to increase the turbomachine (TM) rotation speed in order to improve the turbine and compressor efficiency, to reduce helium leaks in the circulation path, etc. The analysis of GT-MHR, PBMR and GTHTR-300 development experience allows identification of the main ways of increasing the efficiency by selecting optimal parameters and design solutions for the reactor and power conversion unit. The paper estimates the probability of reaching the maximum electricity generation efficiency in reactor plants with the HTGR and gas turbine cycle with account of the up-to-date development status of major reactor plant components (reactor, vessels, turbocompressor (TC), generator, heat exchange equipment, and structural materials).
Proceedings Papers
Proc. ASME. HTR2008, Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 1, 773-782, September 28–October 1, 2008
Paper No: HTR2008-58276
Abstract
This paper compared the performance of very high temperature reactor (VHTR) plants with direct and indirect closed Brayton Cycles (CBCs) and investigated the effect of the molecular weight of the CBC working fluid on the number of stages in and the size of the single shaft turbomachines. The CBC working fluids considered are helium (4 g/mole) and He-Xe and He-N 2 binary mixtures (15 g/mole). Also investigated are the effects of using LPC and HPC with inter-cooling, cooling the reactor pressure vessel with He bled off at the exit of the compressor, and changing the reactor exit temperature from 700°C to 950°C on the plant thermal efficiency, CBC pressure ratio and the number of stages in and size of the turbo-machines. Analyses are performed for reactor thermal power of 600 MW, shaft rotation speed of 3000 rpm, and IHX temperature pinch of 50 °C.
Proceedings Papers
Proc. ASME. HTR2008, Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 2, 391-404, September 28–October 1, 2008
Paper No: HTR2008-58258
Abstract
Since the last decade, Tractebel Engineering has been involved in several consecutive projects in the field of High Temperature Gas Reactor (HTGR). The objectives of the present project called RAPHAEL ( www.raphael-project.org ) is to provide R&D results in order to consolidate available data on generic V/HTR technologies and to develop innovative solutions to further contribute to the improvement of HTR performances. One of the objectives of the RAPHAEL Sub-project Safety is to qualify tools for performing safety analyses and supporting the safety approach and demonstration. One of the work packages concerns the validation of the existing thermal-hydraulic system codes capabilities needed to perform transient analysis in V/HTR. This validation is carried out by benchmarking against experimental data and by comparing simulation results given by several codes. The current paper presents the work performed at Tractebel Engineering on the simulation of the HE-FUS3 experimental loop — ENEA facility, Brasimone (Italy) — with the MELCOR v.1.8.6 code. The HE-FUS3 loop contains a wide range of components characteristic of a V/HTR like compressor, pipes, diffusers, valves, heaters and heat exchangers. Even if the loop characteristics/configuration is not prototypical of a V/HTR design, the loop is useful to assess the objectives identified by the Project, i.e. helium operating fluid, design pressure and temperature set respectively at 10.5 MPa and 530 °C. The experimental data of the HE-FUS3 loop made available for the benchmark are a set of steady state tests for the thermal-hydraulic characterization of the loop and two transient tests — Loss Of Flow Accidents (LOFA). Moreover, to assess the characteristics of the compressor, data have also been provided from a compressor test campaign. From the code-to-experiment comparison the ability of MELCOR v.1.8.6 to reproduce the experimental results is judged.
Proceedings Papers
Proc. ASME. HTR2008, Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 2, 287-292, September 28–October 1, 2008
Paper No: HTR2008-58212
Abstract
The Pebble Bed Modular Reactor (PBMR), under development in South Africa, is an advanced helium-cooled graphite moderated high-temperature gas-cooled nuclear reactor. The heat output of the PBMR is primarily suited for process applications or power generation. In addition, various desalination technologies can be coupled to the PBMR to further improve the overall efficiency and economics, where suitable site opportunities exist. Several desalination application concepts were evaluated for both a cogeneration configuration as well as a waste heat utilization configuration. These options were evaluated to compare the relative economics of the different concepts and to determine the feasibility of each configuration. The cogeneration desalination configuration included multiple PBMR units producing steam for a power cycle, using a back-pressure steam turbine generator exhausting into different thermal desalination technologies. These technologies include Multi-Effect Distillation (MED), Multi-Effect Distillation with Thermal Vapor Compression (MED-TVC) as well as Multi-Stage Flash (MSF) with all making use of extraction steam from backpressure turbines. These configurations are optimized to maximize gross revenue from combined power and desalinated water sales using representative economic assumptions with a sensitivity analysis to observe the impact of varying power and water costs. Increasing turbine back pressure results in a loss of power output but a gain in water production. The desalination systems are compared as incremental investments. A standard MED process with minimal effects appears most attractive, although results are very sensitive with regards to projected power and water values. The waste heat utilization desalination configuration is based on the current 165 MWe PBMR Demonstration Power Plant (DPP) to be built for the South African utility Eskom. This demonstration plant is proposed at the Koeberg Nuclear site and utilizes a direct, single shaft recuperative Brayton Cycle with helium as working fluid. The Brayton Cycle uses a pre-cooler and inter-cooler to cool the helium before entering the low-pressure compressor (LPC) and the high-pressure compressor (HPC) respectively. The pre-cooler and intercooler rejects 218 MWt of waste heat at 73°C and 63°C, respectively. This waste heat is ideally suited for some low temperature desalination processes and can be used without negative impact on the power output and efficiency of the nuclear power plant. These low temperature processes include Low Temperature Multi-Effect Distillation (LT-MED) as well Reverse Osmosis (RO) with pre-heated water. The relative economics of these design concepts are compared as add-ons to the PBMR-DPP and the results include a net present value (NPV) study for both technologies. From this study it can be concluded that both RO as well LT-MED provide water at reasonable production rates, although a final study recommendation would be based on site and area specific requirements.
Proceedings Papers
Proc. ASME. HTR2008, Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 2, 303-310, September 28–October 1, 2008
Paper No: HTR2008-58015
Abstract
A power-generating unit with the high-temperature helium reactor (GT-MHR) has a turbomachine (TM) that is intended for both conversion of coolant thermal energy into electric power in the direct gas-turbine cycle, and provision of helium circulation in the primary circuit. The vertically oriented TM is placed in the central area of the power conversion unit (PCU). TM consists of a turbocompressor (TC) and a generator. Their rotors are joined with a diaphragm coupling and supported by electro-magnetic bearings (EMB). The complexity and novelty of the task of the full electromagnetic suspension system development requires thorough stepwise experimental work, from small-scale physical models to full-scale specimen. On this purpose, the following is planned within the framework of the GT-MHR Project: investigations of the “flexible” rotor small-scale mockup with electro-magnetic bearings (“Minimockup” test facility); tests of the radial EMB; tests of the position sensors; tests of the TM rotor scale model; tests of the TM catcher bearings (CB) friction pairs; tests of the CB mockups; tests of EMB and CB pilot samples and investigation of the full-scale electromagnetic suspension system as a part of full-scale turbocompressor tests. The rotor scale model (RSM) tests aim at investigation of dynamics of rotor supported by electromagnetic bearings to validate GT-MHR turbomachine serviceability. Like the full-scale turbomachine rotor, the RSM consist of two parts: the generator rotor model and the turbocompressor rotor model that are joined with a coupling. Both flexible and rigid coupling options are tested. Each rotor is supported by one axial and two radial EMBs. The rotor is arranged vertically. The RSM rotor length is 10.54 m, and mass is 1171 kg. The designs of physical model elements, namely of the turbine, compressors, generator and exciter, are simplified and performed with account of rigid characteristics, which are identical to those of the full-scale turbomachine elements.
Proceedings Papers
Proc. ASME. HTR2008, Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 2, 57-68, September 28–October 1, 2008
Paper No: HTR2008-58166
Abstract
In the frame of the international forum GenIV, CEA has selected various innovative concepts of Gas cooled Nuclear Reactor. Among them, an indirect-cycle gas reactor is under consideration. Thermal hydraulic performances are a key issue for the design. For transient conditions and decay heat removal situations, the thermal hydraulic performance must remain as high as possible. In this context, all the transient situations, the incidental and accidental scenarii must be evaluated by a validated system code able to correctly describe, in particular, the thermalhydraulics of the whole plant. As concepts use a helium compressor to maintain the flow in the core, a special emphasis must be laid on compressor modelling. Centrifugal circulators with a vaneless diffuser have significant properties in term of simplicity, cost, ability to operate over a wide range of conditions. The objective of this paper is to present a dedicated description of centrifugal compressor, based on a one dimensional approach. This type of model requires various correlations as input data. The present contribution consists in establishing and validating the numerical simulations (including different sets of correlations) by comparison with representative experimental data. The results obtained show a qualitatively correct behaviour of the model compared to open literature cases of the gas turbine aircraft community and helium circulators of High Temperature Gas Reactors. Further work on modelling and validation are nevertheless needed to have a better confidence in the simulation predictions.