scholarly journals Component Test Facility (Comtest) Phase 1 Engineering For 760°C (1400°F) Advanced Ultrasupercritical (A-USC) Steam Generator Development

2016 ◽  
Author(s):  
Paul Weitzel
Keyword(s):  
Steam Generator ◽  
Phase 1 ◽  
Test Facility ◽  
Component Test ◽  
Generator Development

Component Test Facility (ComTest) Phase 1 Engineering for 760C (1400F) Advanced Ultra-Supercritical (A-USC) Steam Generator Development

2015 ◽  
Author(s):  
Paul S. Weitzel
Keyword(s):  
Power Plant ◽  
Engineering Design ◽  
Steam Generator ◽  
The United States ◽  
Department Of Energy ◽  
Test Facility ◽  
Plant Design ◽  
Front End ◽  
Operation And Maintenance ◽  
Component Test

Babcock & Wilcox Power Generation Group, Inc. (B&W) has received a competitively bid award from the United States (U.S.) Department of Energy to perform the preliminary front-end engineering design of an advanced ultra-supercritical (A-USC) steam superheater for a future A-USC component test program (ComTest) achieving 760C (1400F) steam temperature. The current award will provide the engineering data necessary for proceeding to detail engineering, manufacturing, construction and operation of a ComTest. The steam generator superheater would subsequently supply the steam to an A-USC intermediate pressure steam turbine. For this study the ComTest facility site is being considered at the Youngstown Thermal heating plant facility in Youngstown, Ohio. The ComTest program is important because it would place functioning A-USC components in operation and in coordinated boiler and turbine service. It is also important to introduce the power plant operation and maintenance personnel to the level of skills required and provide initial hands-on training experience. Preliminary fabrication, construction and commissioning plans are to be developed in the study. A follow-on project would eventually provide a means to exercise the complete supply chain events required to practice and refine the process for A-USC power plant design, supply, manufacture, construction, commissioning, operation and maintenance. Representative participants would then be able to transfer knowledge and recommendations to the industry. ComTest is conceived as firing natural gas in a separate standalone facility that will not jeopardize the host facility or suffer from conflicting requirements in the host plant’s mission that could sacrifice the nickel alloy components and not achieve the testing goals. ComTest will utilize smaller quantities of the expensive materials and reduce the risk in the first operational practice for A-USC technology in the U.S. Components at suitable scale in ComTest provide more assurance before applying them to a full size A-USC demonstration plant. The description of the pre-front-end engineering design study and current results will be presented.

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

HTGR Component Test Facility (CTF) Feasibility Study for NGNP

2008 ◽  
Author(s):  
Jan P. van Ravenswaay ◽  
Jacques Holtzhausen ◽  
Jaco van der Merwe ◽  
Kobus Olivier ◽  
Riaan du Bruyn ◽  
...  
Keyword(s):  
Heat Transfer ◽  
High Temperature ◽  
Oil Recovery ◽  
Recovery Process ◽  
Initial Energy ◽  
Test Facility ◽  
Design Data ◽  
Component Test ◽  
Maintenance And Inspection ◽  
High Temperature Gas

The Next Generation Nuclear Plant (NGNP) Project is a US-based initiative led by Idaho National Laboratories to demonstrate the viability of using High Temperature Gas-Cooled Reactor (HTGR) technology for the production of high temperature steam and/or heat for applications such as heavy oil recovery, process steam/cogeneration and hydrogen production. A key part of the NGNP Project is the development of a Component Test Facility (CTF) that will support the development of high temperature gas thermal-hydraulic technologies as applied in heat transport and heat transfer applications in HTGRs. These applications include, but are not limited to, primary and secondary coolants, direct cycle power conversion, co-generation, intermediate, secondary and tertiary heat transfer, demonstration of processes requiring high temperatures as well as testing of NGNP specific control, maintenance and inspection philosophies and techniques. The feasibility of the envisioned CTF as a development and testing platform for components and systems in support of the NGNP was evaluated. For components and systems to be integrated into the NGNP full scale or at least representative size tests need to be conducted at NGNP representative conditions, with regards to pressure, flow rate and temperature. Typical components to be tested in the CTF include heat exchangers, steam generators, circulators, valves and gas piping. The Design Data Needs (DDNs), Technology Readiness Levels (TRLs) as well as Design Readiness Levels (DRLs) prepared in the pre-conceptual design of the NGNP Project and the NGNP lifecycle requirements were used as inputs to establish the CTF Functional and Operating Requirements (F&ORs). The existing South African PBMR test facilities were evaluated to determine their current applicability or possible modifications to meet the F&ORs of the CTF. Three concepts were proposed and initial energy balances and layouts were developed. This paper will present the results of this CTF study and the ongoing efforts to establish the CTF.

Dismantling Method of Fuel Cycle Facilities Obtained by Dismantling of the JRTF

2010 ◽  
Author(s):  
Fumihiko Kanayama
Keyword(s):  
Research And Development ◽  
Fuel Cycle ◽  
Phase 1 ◽  
Atomic Energy ◽  
Purex Process ◽  
Test Facility ◽  
Nuclear Facilities ◽  
Phase 3 ◽  
Energy Agency ◽  
Uranium Metal

The Japan Atomic Energy Research Institute Reprocessing Test Facility (JRTF) was the first reprocessing facility which was constructed by applying only Japanese technology to establish basic technology on wet reprocessing. JRTF had been operated since 1968 to 1969 using spent fuels (uranium metal/aluminum clad, about 600kg as uranium metal and 600MWD/T) from the Japan Research Reactor No.3 (JRR-3). Reprocessing testings on PUREX process were implemented at 3 runs, so that, 200g of plutonium dioxide were extracted. After JRTF was shut down at 1970, it had been used for research and development of reprocessing since 1971. The more mature research and development of nuclear are, the more opportunity of dismantling of old nuclear facilities would be. Japan Atomic Energy Agency (JAEA) has an experience of full scale of dismantling through decommissioning of Japan Power Demonstration Reactor (JPDR)1). On the other hand, we didn’t have that of fuel cycle facility. Moreover, it is considered that dismantling methods of nuclear reactor and fuel cycle facility are different for following reason, components contaminated TRU nuclide including Pu, and components installed inside narrow cells. Dismantling methods are important factor to decide manpower and time for dismantling. So, it is indispensable to optimize dismantling method in order to minimize manpower and time for dismantling. Considering the background mentioned above, the decommissioning project of JRTF was started in 1990. The decommissioning project of JRTF is carried out phase by phase. Phase 1; Investigation for dismantling of the JRTF2)3)4). Phase 2; R&D of decommissioning technologies for dismantling of the JRTF5)6)7)8). Phase 3; Actual dismantling of the JRTF9)10). There were several components used for reprocessing and a system for liquid radwaste storage, and those were installed inside of each of several thick concrete cells. The inner surfaces of each cell were contaminated by TRU nuclides including Pu. In phase 3, components used in reprocessing and a system for liquid radwaste storage were dismantled. Moreover, opening was made in concrete walls (including ceiling) for this work. Effective practices for dismantling fuel cycle facilities were obtained through these works. On this report, effective dismantle methods obtained by actual dismantling activities in JRTF are introduced.

The Design and Testing of a Molten Salt Steam Generator for Solar Application

1988 ◽  
Vol 110 (1) ◽  
pp. 38-44 ◽  
Author(s):  
W. A. Allman ◽  
D. C. Smith ◽  
C. R. Kakarala
Keyword(s):  
Molten Salt ◽  
Steam Generator ◽  
Test Facility ◽  
Process Conditions ◽  
Test Results ◽  
Overall Design ◽  
Central Receiver ◽  
And Performance ◽  
Design And Testing ◽  
Solar Application

This paper describes the design and testing of the Steam Generator Subsystem (SGS) for the Molten Salt Electric Experiment at Sandia Laboratories in Albuquerque, New Mexico. The Molten Salt Electric Experiment (MSEE) has been established at the Department of Energy’s five megawatt thermal Solar Central Receiver Test Facility, to demonstrate the feasibility of the molten salt central receiver concept. The experiment is capable of generating 0.75 megawatts of electric power from solar energy, with the capability of storing seven megawatt-hours of thermal energy. The steam generator subsystem transfers sensible heat from the solar-heated molten nitrate salt to produce steam to drive a conventional turbine. This paper discusses the design requirements dictated by the steam generator application and also reviews the process conditions. Details of each of the SGS components are given, featuring the aspects of the design and performance unique to the solar application. The paper concludes with a summary of the test results confirming the overall design of the subsystem.

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