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.

RELAP5/MOD3.2 Investigation of a VVER-1000 MCP Switching on Problem

2002 ◽  
Author(s):  
Pavlin Groudev ◽  
Malinka Pavlova
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Argonne National Laboratory ◽  
The United States ◽  
Reactor Power ◽  
Department Of Energy ◽  
The Other ◽  
United States Department ◽  
Coolant Pump ◽  
Bulgarian Academy

This paper provides a discussion of various RELAP5 parameters calculated for the investigation of the nuclear power reactor parameter behavior in case of switching on one main coolant pump (MCP) when the other three MCPs are in operation. The reference power plant for this analysis is Unit 6 at the Kozloduy Nuclear Power Plant (NPP) site. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. The event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which leads to insertion of positive reactivity due to the modeled feedback mechanisms. This investigation has been conducted by Bulgarian and Russian specialists on the stage when the reactor power was at 75% of the nominal level. The purpose of the experiment was the complete testing of reliability of all power plant equipment, testing the reliability of the main regulators and defining a jump of the neutron reactor power in case of switching on of one main coolant pump. The Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, and Kozloduy NPP have been developing a RELAP5/MOD3.2 model for Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios. This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of the VVER-1000 against the experimental transient data received from the Kozloduy NPP Unit 6. The comparisons between the RELAP5 results and the test data indicate good agreement. This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

scholarly journals Completion of Experimental Breeder Reactor-II Sodium Processing at Argonne National Laboratory

2002 ◽  
Author(s):  
Mary D. McDermott ◽  
Charles D. Griffin ◽  
Daniel K. Baird ◽  
Carl E. Baily ◽  
John A. Michelbacher ◽  
...  
Keyword(s):  
Sodium Hydroxide ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
Test Facility ◽  
United States Department ◽  
Safe Condition ◽  
And Control ◽  
Low Level Radioactive Waste

The Experimental Breeder Reactor - II (EBR-II) at Argonne National Laboratory - West (ANL-W) was shutdown in September 1994 as mandated by the United States Department of Energy. Located in eastern Idaho, this sodium-cooled reactor had been in service since 1964, and was a test facility for fuels development, materials irradiation, system and control theory tests, and hardware development. The EBR-II termination activities began in October 1994, with the reactor being maintained in an industrially and radiologically safe condition for decommissioning. With the shutdown of EBR-II, its sodium coolant became a waste necessitating its reaction to a disposal form. A Sodium Process Facility (SPF), designed to convert sodium to 50 wt% sodium hydroxide, existed at the ANL-W site, but had never been operated. The SPF was upgraded to current standards and codes, and then modified in 1998 to convert the sodium to 70 wt% sodium hydroxide, a substance that solidifies at 65°C (150°F) and is acceptable for burial as low level radioactive waste in Idaho. In December 1998, the SPF began operations. Working with sodium and highly concentrated sodium hydroxide presented some unique operating and maintenance conditions. Several lessons were learned throughout the operating period. Processing of the 330 m3 (87,000 gallons) of EBR-II primary sodium, 50 m3 (13,000 gallons) of EBR-II secondary sodium, and 290 m3 (77,000 gallons) of Fermi-1 primary sodium was successfully completed in March 2001, ahead of schedule and within budget.

scholarly journals The Component Test Facility: A National User Facility for Testing of High Temperature Gas-Cooled Reactor (HTGR) Components and Systems

2008 ◽  
Author(s):  
Vondell J. Balls ◽  
David S. Duncan ◽  
Stephanie L. Austad
Keyword(s):  
High Temperature ◽  
Large Scale ◽  
High Efficiency ◽  
Scale Effects ◽  
Nuclear Plant ◽  
Department Of Energy ◽  
Test Facility ◽  
Component Test ◽  
User Facility ◽  
High Temperature Gas

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

Liquefied Gaseous Fuels (LGF) Spill Test Facility Program

1985 ◽  
Vol 28 (5) ◽  
pp. 34-39
Author(s):  
Albert Dietz
Keyword(s):  
Large Scale ◽  
Atmospheric Dispersion ◽  
Atmospheric Stability ◽  
Test Site ◽  
Fire Effects ◽  
Department Of Energy ◽  
Test Facility ◽  
Plant Design ◽  
Liquefied Petroleum Gas ◽  
Toxic Materials

The Department of Energy (DOE) is constructing a facility at the Nevada Test Site (NTS) which will be capable of the rapid release of large quantities of cryogenic or pressurized flammable or toxic materials. The facility is being built in concert with and in response to the needs of many industrial and governmental organizations. The facility is designed to satisfy the need for information for risk assessment, emergency response, regulation, plant design, plant siting, and hazard mitigation. It will be capable of spilling up to 200 m3 (53,000 gal) of cryogenic fluids such as Liquefied Natural Gas (LNG) or refrigerated ammonia at rates between 5 and 100 m3/min (1,000-26,000 gpm). In addition, up to 90 m3 (24,000 gal) of ambient temperature materials such as Liquefied Petroleum Gas (LPG) or chlorine, with pressures up to 2,000 kpa (300 psi), can be released at rates between 2 and 20 m3/min (500-5,000 gpm). An extensive sensor and data acquisition system is available to acquire data on spill characteristics such as rate, volume, temperature, and pressure; downwind gas concentration and aerosol characteristics; meteorological parameters; and blast or fire effects. The Frenchman Flat area of the NTS provides a uniquely favorable environment in which to perform large-scale atmospheric dispersion tests. Steady winds from the southwest occur with great regularity during the summer months for a variety of atmospheric stability conditions. Large-scale tests with toxic materials are possible because only limited access, federally controlled land is present for some 60 km (37 mi) downwind.

scholarly journals Post-combustion slipstream CO2-capture test facility at Jiangyou Power Plant, Sichuan, China: facility design and validation using 30% wt monoethanolamine (MEA) testing

Clean Energy ◽  
2020 ◽  
Vol 4 (2) ◽  
pp. 107-119
Author(s):  
Baodeng Wang ◽  
Qian Cui ◽  
Guoping Zhang ◽  
Yinhua Long ◽  
Yongwei Sun ◽  
...  
Keyword(s):  
Power Plant ◽  
Co2 Capture ◽  
Power Plants ◽  
Field Data ◽  
Co2 Storage ◽  
Test Facility ◽  
Plant Design ◽  
Future Studies ◽  
The World ◽  
Flexible Operation

Abstract Given the dominant share of coal in China’s energy-generation mix and the fact that >50% of the power plants in the country are currently <15 years old, efforts to significantly reduce China’s CO2 footprint will require the deployment of CO2 capture across at least part of its fleet of coal-fired power plants. CO2-capture technology is reaching commercial maturity, but it is still necessary to adapt the technology to regional conditions, such as power-plant design and flexible operation in the China context. Slipstream facilities provide valuable field data to support the commercialization of CO2 capture. We have built a slipstream facility at Jiangyou power plant in Sichuan that will allow us to explore China-relevant issues, especially flexible operation, over the next few years. We plan to share our results with the broader CO2-capture and CO2-storage (CCS) community to accelerate the deployment of CCS in China. This paper describes the design of the slipstream facility and presents results from our steady-state qualification tests using a well-studied benchmark solvent: 30% wt monoethanolamine (MEA). The results from our MEA tests compare favorably to results reported from other slipstream-test facilities around the world, allowing us to commission our system and establish a reference baseline for future studies.

Steam Generator for Advanced Ultra Supercritical Power Plants 700C to 760C

2011 ◽  
Author(s):  
Paul S. Weitzel
Keyword(s):  
Power Generation ◽  
Electric Power ◽  
Steam Generator ◽  
Power Plants ◽  
The United States ◽  
Plant Design ◽  
Levelized Cost Of Electricity ◽  
Cost Of Electricity ◽  
Generation Technology ◽  
The U.S

Advanced ultra-supercritical (A-USC) is a term used to designate a coal-fired power plant design with the inlet steam temperature to the turbine at 700 to 760C (1292 to 1400F). Average metal temperatures of the final superheater and final reheater could run higher, at up to about 815C (1500F). Nickel-based alloy materials are thus required. Increasing the efficiency of the Rankine regenerative-reheat steam cycle to improve the economics of electric power generation and to achieve lower cost of electricity has been a long sought after goal. Efficiency improvement is also a means for reducing the emission of carbon dioxide (CO2) and the cost of capture, as well as a means to reduce fuel consumption costs. In the United States (U.S.), European Union, India, China and Japan, industry support associations and private companies working to advance steam generator design technology have established programs for materials development of nickel-based alloys needed for use above 700C (1292F). The worldwide abundance of less expensive coal has driven economic growth. The challenge is to continue to improve the efficiency of coal-fired power generation technology, representing nearly 50% of the U.S. production, while maintaining economic electric power costs with plants that have favorable electric grid system operational characteristics for turndown and rate of load change response. The technical viability of A-USC is being demonstrated in the development programs of new alloys for use in the coal-fired environment where coal ash corrosion and steamside oxidation are the primary failure mechanisms. Identification of the creep rupture properties of alloys for higher temperature service under both laboratory and actual field conditions has been undertaken in a long-term program sponsored by the U.S. Department of Energy (DOE) and the Ohio Coal Development Office (OCDO). Ultimately, the economic viability of A-USC power plants is predicated on the comparable lower levelized cost of electricity (LCOE) with carbon capture and sequestration (CCS) using either oxy-combustion or post-combustion capture. Using nickel alloy components will drive the design and configuration arrangement of the steam generator relative to the plant. A-USC acceptance depends on achieving the higher functional value and lowering the perceived level of risks as this generation technology appears in a new form.

Transient Reactor Test Facility Restart 23 Years Later

2018 ◽  
Author(s):  
Bradley K. Heath ◽  
Cody C. Race ◽  
Lee O. Nelson
Keyword(s):  
The United States ◽  
Program Management ◽  
Department Of Energy ◽  
Test Facility ◽  
Personnel Training ◽  
National Environmental Policy Act ◽  
Management Procedure ◽  
Age Related ◽  
Operational Readiness ◽  
Reactor Test

The Transient Reactor Test (TREAT) Facility, located at the Idaho National Laboratory (INL), is a versatile test facility able to subject experimental specimens to various transient nuclear conditions. TREAT was placed in standby after operating from February 1959 through April 1994, resulting in the loss of nearly all transient testing capability in the United States. Recently, the US Department of Energy (DOE) determined this capability was again needed. After DOE completed National Environmental Policy Act actions in February 2014, INL established the Resumption of Transient Testing Program (RTTP). RTTP was a multi-year effort to restart TREAT to reestablish a domestic transient testing capability. After 23 years of standby operations, the RTTP completed restart activities on August 31, 2017, 13 months ahead of schedule and nearly $20 million under budget. RTTP activities included an Environmental Assessment that resulted in a Finding of “No Significant Impact” associated with restarting TREAT, establishment of a compliant Safety Analysis Report (SAR), refurbishment and/or replacement of key reactor systems and components, key system knowledge recovery, reestablishment of configuration management, procedure updates, personnel training and qualification, and demonstration of operational readiness for reactor operations. Several noteworthy factors that contributed to the restart of TREAT include: • Funding to acquire personnel and material resources provided in a timely fashion. • Close coordination with the regulator’s (DOE) nuclear safety program during updates, interactive review, and approval of safety documentation provided for timely update of the TREAT SAR and implementing documents. • Effective management control enabled by utilization of standard outage management techniques with a focus on age-related degradation and updated standards and requirements. • DOE program management ensured efficient implementation of program management tools. These tools focused on clear high-level milestones and spend plans allowing flexibility for the contractor to respond to evolving facility conditions and information in a near-real time manner and with minimal program overhead. This approach enabled efficient execution of work in an environment where determination of required work scope was dependent on performance of inspection, testing, analysis, and evaluation activities. • Implementation of the Contractor Assurance System, with frequent internal and externally-led assessments that facilitated process improvements and corrective actions to ensure the operational readiness for required contractor and DOE readiness assessments and safe nuclear operations. • The RTTP benefited from archived plant documentation and maintenance performed while the plant was in a safe-standby status. • Unique methods of reactivity control allowed for individual and integrated reactor system functional testing, procedure vetting, and personnel training while maintaining the reactor in a safe state.

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