Environmentally-Friendly HTGR: MHR-50/100—Concept and Characteristics

2011 ◽  
Author(s):  
Isao Minatsuki ◽  
Tomomi Otani ◽  
Katsusuke Shimizu ◽  
Tetsuo Saguchi ◽  
Sunao Oyama ◽  
...  
Keyword(s):  
High Temperature ◽  
Conceptual Design ◽  
Financial Analysis ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Power ◽  
Primary System ◽  
Plant Design ◽  
Plant Dynamics ◽  
High Temperature Gas

A business plan and a new concept of the Mitsubishi small-sized High temperature gas-cooled modular Reactors (MHR-50/100) had been developed as reported in a paper at the HTR-2010 conference in Prague. The present paper reports the results of ensuing conceptual design study including updated market researches, improved safety features of the plant, and the plant dynamics analysis. Market researches on Japan, the USA, Southeast Asia and the Middle East have been updated applying the latest energy outlook data. The result shows that the potential market share for the type of HTGR (high temperature gas reactor) reactors is expected to be 10–20% in new construction of heat source plants in those market areas. A financial analysis made based on the results of the updated market research and the plant cost evaluations indicates that the feasibility of an HTGR business potentially exists. Concerning about the conceptual design, as main themes of the study, a plant design, safety design and plant dynamics have been carried out. The MHR-50/100 high safety characteristics have been confirmed based on the results of the following studies as reported in the present paper: (1) An investigation of a safety scenario during occurrence of a Total Black Out event; (2) An analysis of the reactor decay heat removal via a natural circulation. Lastly, the control methods for the reactor and associated steam cycle system for the MHR-50 have been studied. The results show that the reactor power changes can be effectively achieved by controlling the primary system helium flow rate. The ASURA code developed by MHI is used for simulation of such typical plant transients as 10% step load reduction and full load rejection. The results confirm the easy operability and controllability of the plant.

Goals, Requirements, and Design Implications for the Advanced High-Temperature Reactor

2006 ◽  
Author(s):  
Charles W. Forsberg
Keyword(s):  
High Temperature ◽  
Power Output ◽  
Nuclear Power ◽  
Heat Removal ◽  
Security Requirements ◽  
Electricity Production ◽  
Plant Design ◽  
High Temperature Reactor ◽  
High Temperature Gas ◽  
Electrical Output

The Advanced High-Temperature Reactor (AHTR), also called the liquid-salt-cooled Very High-Temperature Reactor (LS-VHTR), is a new reactor concept that has been under development for several years. The AHTR combines four existing technologies to create a new reactor option: graphite-matrix, coated-particle fuels (the same fuel as used in high-temperature gas-cooled reactors); a liquid-fluoride-salt coolant with a boiling point near 1400°C; plant designs and decay-heat-removal safety systems similar to those in sodium-cooled fast reactors; and a helium or nitrogen Brayton power cycle. This paper describes the basis for the selection of goals and requirements, the preliminary goals and requirements, and some of the design implications. For electricity production, the draft AHTR goals include peak coolant temperatures between 700 and 800°C and a maximum power output of about 4000 MW(t), for an electrical output of ∼2000 MW(e). The electrical output matches that expected for a large advanced light-water reactor (ALWR) built in 2025. Plant capital cost per kilowatt electric is to be at least one-third less than those for ALWRs with the long-term potential to significantly exceed this goal. For hydrogen production, the peak temperatures may be as high as 950°C, with a power output of 2400 MW(t). The safety goals are to equal or surpass those of the modular high-temperature gas-cooled reactor with a beyond-design-basis accident capability to withstand large system and structural failures (vessel failure, etc.) without significant fuel failure or off-site radionuclide releases. These safety goals may eliminate the technical need for evacuation zones and reduce security requirements and significantly exceed the safety goals of ALWRs. The plant design should enable economic dry cooling to make possible wider nuclear-power-plant siting options. Uranium consumption is to be less than that for a LWR, with major improvements in repository performance and nonproliferation characteristics.

Simulation and Analysis of Helium Circulator Trip ATWS Test at Full Power on the HTR-10

2008 ◽  
Author(s):  
Yujie Dong ◽  
Fubing Chen ◽  
Zuoyi Zhang ◽  
Shouyin Hu ◽  
Lei Shi ◽  
...  
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Power Level ◽  
Reactor Core ◽  
Heat Removal ◽  
Reactor Power ◽  
Code System ◽  
Safety Features ◽  
Forced Cooling ◽  
High Temperature Gas

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.

scholarly journals The Design of Turbomachinery for the Gas Turbine (Direct Cycle) High Temperature Gas Cooled Reactor Power Plant

1977 ◽  
Author(s):  
R. G. Adams ◽  
F. H. Boenig
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Gas Turbine ◽  
Power Conversion ◽  
Reactor Power ◽  
Working Fluid ◽  
Primary Coolant ◽  
Reactor Power Plant ◽  
Direct Cycle ◽  
High Temperature Gas

The Gas Turbine HTGR, or “Direct Cycle” High-Temperature Gas-Cooled, Reactor power plant, uses a closed-cycle gas turbine directly in the primary coolant circuit of a helium-cooled high-temperature nuclear reactor. Previous papers have described configuration studies leading to the selection of reactor and power conversion loop layout, and the considerations affecting the design of the components of the power conversion loop. This paper discusses briefly the effects of the helium working fluid and the reactor cooling loop environment on the design requirements of the direct-cycle turbomachinery and describes the mechanical arrangement of a typical turbomachine for this application. The aerodynamic design is outlined, and the mechanical design is described in some detail, with particular emphasis on the bearings and seals for the turbomachine.

Analysis of Protected Accidental Transients in the EFIT Reactor With the RELAP5 Thermal-Hydraulic Code

2008 ◽  
Author(s):  
Giacomino Bandini ◽  
Maddalena Casamirra ◽  
Francesco Castiglia ◽  
Mariarosa Giardina ◽  
Paride Meloni ◽  
...  
Keyword(s):  
Proton Beam ◽  
Thermal Power ◽  
Heat Removal ◽  
Reactor Power ◽  
Lead Target ◽  
Primary System ◽  
Safety Behavior ◽  
Forced Circulation ◽  
Detailed Model ◽  
Decay Heat

The European Facility for Industrial Transmutation (EFIT) is aimed at demonstrating the feasibility of transmutation process through the Accelerator Driven System (ADS) route on an industrial scale. The conceptual design of this reactor of about 400 MW thermal power is under development in the frame of the European EUROTRANS Integrated Project of the EURATOM Sixth Framework Program (FP6). EFIT is a pool-type reactor cooled by forced circulation of lead in the primary system where the heat is removed by steam generators installed inside the reactor vessel. The reactor power is sustained by a spallation neutron source supplied by a proton beam impinging on a lead target at the core centre. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat in case of loss of secondary circuits heat removal capability. A quite detailed model of the EFIT reactor has been developed for the RELAP5 thermal-hydraulic code to be used in preliminary accidental transient analyses aimed at verifying the validity of the adopted solutions for the current reactor design with respect to the safety requirements, and confirm the inherent safety behavior of the reactor, such as decay heat removal in accidental conditions relying on natural circulation in the primary system. The accident analyses for the EFIT reactor include both protected and unprotected transients, on whether the reactor automatic trip, consisting in proton beam switch off, is actuated or not by the protection system. In this paper, the main results of the analyses of some protected transients with RELAP5 are presented. The analyzed transients concern the Protected Loss of Heat Sink (PLOHS), in which the DHR system plays a key role in bringing the reactor in safe conditions, and the Protected Loss of Flow (PLOF) transients with partial or total loss of forced circulation in the primary system.

Conceptual Design and Cooling Blade Development of 1700°C Class High-Temperature Gas Turbine

2005 ◽  
Vol 127 (2) ◽  
pp. 358-368 ◽  
Author(s):  
Shoko Ito ◽  
Hiroshi Saeki ◽  
Asako Inomata ◽  
Fumio Ootomo ◽  
Katsuya Yamashita ◽  
...  
Keyword(s):  
Heat Transfer ◽  
High Temperature ◽  
Gas Turbine ◽  
Conceptual Design ◽  
Thermal Efficiency ◽  
Internal Cooling ◽  
Closed Cycle ◽  
Steam Consumption ◽  
Flow Phenomena ◽  
High Temperature Gas

In this paper we describe the conceptual design and cooling blade development of a 1700°C-class high-temperature gas turbine in the ACRO-GT-2000 (Advanced Carbon Dioxide Recovery System of Closed-Cycle Gas Turbine Aiming 2000 K) project. In the ACRO-GT closed cycle power plant system, the thermal efficiency aimed at is more than 60% of the higher heating value of fuel (HHV). Because of the high thermal efficiency requirement, the 1700°C-class high-temperature gas turbine must be designed with the minimum amount of cooling and seal steam consumption. The hybrid cooling scheme, which is a combination of closed loop internal cooling and film ejection cooling, was chosen from among several cooling schemes. The elemental experiments and numerical studies, such as those on blade surface heat transfer, internal cooling channel heat transfer, and pressure loss and rotor coolant passage distribution flow phenomena, were conducted and the results were applied to the conceptual design advancement. As a result, the cooling steam consumption in the first stage nozzle and blade was reduced by about 40% compared with the previous design that was performed in the WE-NET (World Energy Network) Phase-I.

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