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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2010

Next Generation Nuclear Plant Steam Generator and Intermediate Heat Exchanger Materials Research and Development Plan

DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Today’s high-temperature alloys and associated ASME Codes for reactor applications are approved up to 760°C. However, some primary system components, such as the Intermediate Heat Exchanger (IHX) for the NGNP will require use of materials that can withstand higher temperatures. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge. Examples include materials for the core barrel and core internals, such as the control rod sleeves. The requirements of the materials for the IHX are among the most demanding. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. A number of solid solution strengthened nickel based alloys have been considered for application in heat exchangers and core internals for the NGNP. The primary candidates are Inconel 617, Haynes 230, Incoloy 800H and Hastelloy XR. Based on the technical maturity, availability in required product forms, experience base, and high temperature mechanical properties all of the vendor pre-conceptual design studies have specified Alloy 617 as the material of choice for heat exchangers. Also a draft code case for Alloy 617 was developed previously. Although action was suspended before the code case was accepted by ASME, this draft code case provides a significant head start for achieving codification of the material. Similarly, Alloy 800H is the material of choice for control rod sleeves. In addition to the above listed considerations, Alloy 800H is already listed in the nuclear section of the ASME Code; although the maximum use temperature and time need to be increased.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Evaluation of the Effectiveness of a New Technology for Extraction of Insoluble Impurities from Nuclear Power Plant Steam Generators with Purge Water

An experimental technology for the removal of insoluble impurities from a horizontal steam generator with purge water during planned shutdowns of the power generating unit is improved through a more representative determination of the concentration of impurities in the purge water ahead of the water cleanup facility and a more precise effective time for the duration of the purge process. Tests with the improved technique at power generating unit No. 1 of the Rostov Nuclear Power Plant show that the efficiency with which insoluble impurities are removed from the steam generator volume was more than two orders of magnitude greater than under the standard regulations.

42 ENGINEERING↗
DOE OSTI2015

Eddy current signal comparison for tube identification

Inspection of nuclear power plant steam generator tubes is required to justify continued safe plant operation. The steam generators consist of thousands of tubes with nominal diameters of 15 to 22mm, approximately 1mm wall thickness, and 20 to 30m in length. The tubes are inspected by passing an eddy current probe through the tubes from tube end to tube end. It is critical to know exactly which tube identification (row and column) is associated with each tube's data. This is controlled by a precision manipulator that provides the tube ID to the eddy current system. Historically there have been some instances where the manipulator incorrectly reported the tube ID. This can have serious consequences including lack of inspection of a tube, or if a pluggable indication is detected, the tube is likely to be mis-plugged thereby risking a primary to secondary leak.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2011

New technology for purging the steam generators of nuclear power plants

A technology for removal of undissolved impurities from a horizontal steam generator using purge water is developed on the basis of a theoretical analysis. A purge with a maximal flow rate is drawn off from the zone with the highest accumulation of sludge in the lower part of the steam generator after the main circulation pump of the corresponding loop is shut off and the temperatures of the heat transfer medium at the inlet and outlet of the steam generator have equilibrated. An improved purge configuration is used for this technology; it employs shutoff and regulator valves, periodic purge lines separated by a cutoff fixture, and a D{sub y} 100 drain union as a connector for the periodic purge. Field tests show that the efficiency of this technology for sludge removal by purge water is several times that for the standard method.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2010

Power Plant Replacement Study

This report represents the final report for the Eastern Illinois University power plant replacement study. It contains all related documentation from consideration of possible solutions to the final recommended option. Included are the economic justifications associated with the chosen solution along with application for environmental permitting for the selected project for construction. This final report will summarize the results of execution of an EPC (energy performance contract) investment grade audit (IGA) which lead to an energy services agreement (ESA). The project includes scope of work to design and install energy conservation measures which are guaranteed by the contractor to be self-funding over its twenty year contract duration. The cost recovery is derived from systems performance improvements leading to energy savings. The prime focus of this EPC effort is to provide a replacement solution for Eastern Illinois University's aging and failing circa 1925 central steam production plant. Twenty-three ECMs were considered viable whose net impact will provide sufficient savings to successfully support the overall project objectives.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2011

Modeling a Helical-coil Steam Generator in RELAP5-3D for the Next Generation Nuclear Plant

Options for the primary heat transport loop heat exchangers for the Next Generation Nuclear Plant are currently being evaluated. A helical-coil steam generator is one heat exchanger design under consideration. Safety is an integral part of the helical-coil steam generator evaluation. Transient analysis plays a key role in evaluation of the steam generators safety. Using RELAP5-3D to model the helical-coil steam generator, a loss of pressure in the primary side of the steam generator is simulated. This report details the development of the steam generator model, the loss of pressure transient, and the response of the steam generator primary and secondary systems to the loss of primary pressure. Back ground on High Temperature Gas-cooled reactors, steam generators, the Next Generation Nuclear Plant is provided to increase the readers understanding of the material presented.

42 ENGINEERING↗
DOE OSTI2019

Review and categorization of existing studies on the estimation of probabilistic failure metrics for Reactor Coolant Pressure Boundary piping and steam generator tubes in Nuclear Power Plants

This article presents the first output of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP), titled Methodology for Assessing Pipe Failure Rates in Advanced Water-Cooled Reactors (AWCRs), launched in 2018. This IAEA CRP is aimed at developing a good-practices document for estimating the probabilistic failure metrics of piping in AWCRs. The reliability of piping that comprise the Reactor Coolant Pressure Boundary (RCPB) is important for maintaining safe and stable operations of Nuclear Power Plants (NPPs) because failure of those piping components could lead to undesirable consequences, such as plant shutdown, costly repair, the occurrence of Loss-of-Coolant Accidents (LOCAs) and, possibly, subsequent core damage or large release events. Probabilistic failure metrics (e.g., failure rate, failure frequency, or failure probability) of RCPB components are the key inputs to the Probabilistic Safety/Risk Assessment (PSA/PRA) and risk management of NPPs. The estimation of probabilistic failure metrics, however, is challenging, especially for AWCRs, due to the lack of operating experience. Therefore, as the first step of the IAEA CRP activities, this paper is developed to provide a literature review of the existing studies (from 2000 to April 2019) on the estimation of probabilistic failure metrics for RCPB piping and Steam Generator (SG) tubes of NPPs and to categorize them based on four criteria: (1) explicitness of incorporation of physical failure mechanisms; (2) types of failure characterization; (3) types of physical models for degradation; and (4) explicitness of consideration of maintenance. The existing studies are also analyzed from the viewpoint of the following key aspects: (i) uncertainty analysis, (ii) sensitivity analysis, (iii) validation strategies, and (iv) the areas of applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Risk analysis of heat recovery steam generator with semi quantitative risk based inspection API 581

Corrosion is a major problem that most often occurs in the power plant. Heat recovery steam generator (HRSG) is an equipment that has a high risk to the power plant. The impact of corrosion damage causing HRSG power plant stops operating. Furthermore, it could be threaten the safety of employees. The Risk Based Inspection (RBI) guidelines by the American Petroleum Institute (API) 58 has been used to risk analysis in the HRSG 1. By using this methodology, the risk that caused by unexpected failure as a function of the probability and consequence of failure can be estimated. This paper presented a case study relating to the risk analysis in the HRSG, starting with a summary of the basic principles and procedures of risk assessment and applying corrosion RBI for process industries. The risk level of each HRSG equipment were analyzed: HP superheater has a medium high risk (4C), HP evaporator has a medium-high risk (4C), and the HP economizer has a medium risk (3C). The results of the risk assessment using semi-quantitative method of standard API 581 based on the existing equipment at medium risk. In the fact, there is no critical problem in the equipment components. Damage mechanisms were prominent throughout the equipment is thinning mechanism. The evaluation of the risk approach was done with the aim of reducing risk by optimizing the risk assessment activities.

42 ENGINEERING↗
DOE OSTI2016

A simulation of a steam generator tube rupture in AP1000 nuclear power plant based on RELAP5-MV

In the operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, steam generator tube rupture (SGTR) is one of the main accidents. In this research, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SGTR accident in AP1000 nuclear power plant (NPP) is analyzed by Relap5-MV thermo-hydraulics code. The transient test scenario is modeled using the RELAP5-MV for the AP1000 NPP. The calculation was performed up to a transient time of 3,000.0 s. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially beginners. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Ultrasonic measurement of temperature distributions in extreme environments: Electrical power plants testing in utility-scale steam generators

Thermal heterogeneities within energy conversion and storage, material processing, nuclear processes, aerospace, and military applications are often inaccessible to characterization by insertion sensors. When sensor deployment is possible, conventional pointwise temperature probes quickly degrade when inserted into harsh environments typical of such processes. We developed spatially-resolved ultrasonic thermometry to noninvasively measure the spatial distributions of thermal properties in such applications, even when sizable thermal gradients are present. Our method divides the path of ultrasonic propagation into segments bound by echogenic features, which create echoes in pulse-echo mode, encoding the information about interior temperature distributions. We use the acquired ultrasonic responses to estimate the internal temperature distributions by solving an inverse problem or concatenating segmental estimates. This work describes the implementation and industrial testing of the developed method at a coal-fired electrical power generation plant. We inserted an echogenically segmented Inconel 625 waveguide into the combustion zone of the utility-scale boiler and continuously acquired ultrasonic data while keeping sensitive components away from the damaging combustion environment. The accuracy of the time-dependent temperature distributions reconstructed from the ultrasonic measurements was comparable to that of thermocouples. The resiliency of ultrasonic thermometry to harsh combustion conditions was far superior to conventional insertion sensors. The measurements obtained during plant operation captured daily steam generation cycles in response to changing customer demand and intermittent contributions of renewable power sources to the power grid. These measurements have revealed new insights into the relationship between the dynamic power generation load and the conditions inside the steam generator. As a result, the successful industrial testing of spatially-resolved ultrasonic thermometry in solids indicates that the developed technology has matured to become an attractive alternative to conventional sensing in solving challenging problems of long-term thermal characterizations in extreme environments.

42 ENGINEERING↗
DOE OSTI2012

CHARACTERIZATION OF ELEVATED TEMPERATURE PROPERTIES OF HEAT EXCHANGER AND STEAM GENERATOR ALLOYS

The Next Generation Nuclear Plant project is considering Alloy 800H and Alloy 617 for steam generator and intermediate heat exchangers. It is envisioned that a steam generator would operate with reactor outlet temperatures from 750 to 800 C, while an intermediate heat exchanger for primary to secondary helium would operate up to an outlet temperature of 950 C. Although both alloys are of interest due in part to their technical maturity, a number of specific properties require further characterization for design of nuclear components. Strain rate sensitivity of both alloys has been characterized and is found to be significant above 600 C. Both alloys also exhibit dynamic strain aging, characterized by serrated flow, over a wide range of temperatures and strain rates. High temperature tensile testing of Alloy 617 has been conducted over a range of temperatures. Dynamic strain aging is a concern for these materials since it is observed to result in reduced ductility for many solid solution alloys. Creep, fatigue, and creep-fatigue properties of Alloy 617 have been measured as well, with the goal of determining the influence of the temperature, strain rate and atmosphere on the creep fatigue life of Alloy 617. Elevated temperature properties and implications for codification of the alloys will be described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Characterization of Elevated Temperature Properties of Heat Exchanger and Steam Generator Alloys

The Next Generation Nuclear Plant project is considering Alloy 800H and Alloy 617 for steam generator and intermediate heat exchangers. It is envisioned that a steam generator would operate with reactor outlet temperatures from 750 to 800°C, while an intermediate heat exchanger for primary to secondary helium would operate up to an outlet temperature of 950°C. Although both alloys are of interest due in part to their technical maturity, a number of specific properties require further characterization for design of nuclear components. Strain rate sensitivity of both alloys has been characterized and is found to be significant above 600°C. Both alloys also exhibit dynamic strain aging, characterized by serrated flow, over a wide range of temperatures and strain rates. In general dynamic strain aging is observed to begin at higher temperatures and serrated flow persists to higher temperatures in Alloy 617 compared to Alloy 800H. Dynamic strain aging is a concern for these materials since it is observed to result in reduced ductility for many solid solution alloys. The role of dynamic strain aging in the creep-fatigue behavior of Alloy 617 at temperatures of 800°C and above has also been examined in detail. Serrated flow is found to persist in cyclic stress-strain curves up to nearly the cycle to failure in some temperature and strain regimes. Results of those experiments and implications for creep-fatigue testing protocols will be described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Novel rotating field probe for inspection of tubes

Inspection of steam generator tubes in nuclear power plants is extremely critical for safe operation of the power plant. In the nuclear industry, steam generator tube inspection using eddy current techniques has evolved over the years from a single bobbin coil, to rotating probe coil (RPC) and array probe, in an attempt to improve the speed and reliability of inspection. The RPC probe offers the accurate spatial resolution but involves complex mechanical rotation. This paper presents a novel design of eddy current probes based on rotating fields produced by three identical coils excited by a balanced three-phase supply. The sensor thereby achieves rotating probe functionality by electronic means and eliminates the need for mechanical rotation. The field generated by the probe is largely radial that result in induced currents that flow circularly around the radial axis and rotating around the tube at a synchronous speed effectively producing induced eddy currents that are multidirectional. The probe will consequently be sensitive to cracks of all orientations in the tube wall. The finite element model (FEM) results of the rotating fields and induced currents are presented. A prototype probe is being built to validate simulation results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Results and Comparison from the SAM Linear Fresnel Technology Performance Model: Preprint

This paper presents the new Linear Fresnel technology performance model in NREL's System Advisor Model. The model predicts the financial and technical performance of direct-steam-generation Linear Fresnel power plants, and can be used to analyze a range of system configurations. This paper presents a brief discussion of the model formulation and motivation, and provides extensive discussion ofthe model performance and financial results. The Linear Fresnel technology is also compared to other concentrating solar power technologies in both qualitative and quantitative measures. The Linear Fresnel model - developed in conjunction with the Electric Power Research Institute - provides users with the ability to model a variety of solar field layouts, fossil backup configurations, thermalreceiver designs, and steam generation conditions. This flexibility aims to encompass current market solutions for the DSG Linear Fresnel technology, which is seeing increasing exposure in fossil plant augmentation and stand-alone power generation applications.

14 SOLAR ENERGY↗
DOE OSTI2021

Co-Generation Wastewater Treatment at Coal-Fired Energy Plants (Final Report)

Steam electric energy generation typically results in the discharge of one or more effluent streams. Coal-fired energy plant air quality regulations have had the effect of transferring pollutants previously in the flue gas to water by means of the flue gas desulfurization (FGD) system. Combined physical-chemical and biological treatment solutions are currently the most economical approach for treating FGD wastewater. Multi-effect evaporators are significantly higher cost but can meet potential upcoming regulations that are more stringent.

01 COAL, LIGNITE, AND PEAT↗