ASME Boiler and Pressure Vessel Code Roadmap for Compact Heat Exchangers in High Temperature Reactors

2020 ◽  
Vol 6 (4) ◽  
Author(s):  
Robert B. Keating ◽  
Suzanne P. McKillop ◽  
Todd Allen ◽  
Mark Anderson
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Department Of Energy ◽  
Research Project ◽  
Compact Heat Exchangers ◽  
High Temperature Reactor ◽  
American Society

Abstract The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems will require compact heat exchangers (CHXs) for the next generation of nuclear reactors. The DOE is sponsoring research to support the development and deployment of CHXs for use in high temperature advanced reactors. The project is being executed by an Integrated Research Project (IRP) that includes university research institutes, national laboratories, manufacturers, and industry experts. The objective is to enable the use of CHX designs in advanced reactor service. A necessary step for achieving this objective is to ensure that the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 has rules for the construction of CHXs for nuclear service. However, construction rules alone are not sufficient to deploy a CHX in an advanced reactor. A strategy for ASME Boiler and Pressure Vessel Code, Section XI, Inservice Inspection (ISI) of a heat exchanger in an operating nuclear reactor will also be required. The purpose of this ASME Code Roadmap is to identify the research gaps impeding the development of suitable construction and ISI rules for CHXs for high temperature reactor service and to provide a framework to utilize the research project results consistent with the expectations and needs of the industry and future owners.

Welded Attachments and Hydrostatic Pressure Testing Rules for Diffusion Bonded Heat Exchangers

2020 ◽  
Author(s):  
Shaun R. Aakre ◽  
Ian W. Jentz
Keyword(s):  
Hydrostatic Pressure ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Department Of Energy ◽  
Alloy 800H ◽  
Compact Heat Exchangers ◽  
Pressure Testing ◽  
Full Penetration ◽  
Welding Processes

Abstract The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation’s energy, environmental, and energy security needs. Advanced high temperature reactor systems will require compact heat exchangers (CHX) for the next generation of nuclear reactor plant designs. A necessary step for achieving this objective is to ensure that the ASME Boiler and Pressure Vessel Code, Section III, Division 5 has rules for the construction of CHXs for nuclear service. Construction of Alloy 800H diffusion bonded Printed Circuit Heat Exchangers (PCHEs) involves multiple controlled welding processes. The primary diffusion bonding process creates a uniformly bonded PCHE body featuring a microchannel core. Secondary welding processes are needed to attach headers and nozzles to the PCHE body, forming a complete CHX. The quality of these welding processes is ensured by following the appropriate ASME Section IX weld qualification procedures. Experience in constructing both 316L and 800H PCHEs has given a set of acceptable attachment weld configurations and procedures. Headers were attached to the diffusion bonded block surface using full penetration welds, as required by Class A design. The integrity of these attachment welds was demonstrated through hydrostatic pressure testing.

Failure Mode Effects Analysis for Section III, Division 5 Class A, High Temperature Service of a Printed Circuit Heat Exchanger

2020 ◽  
Author(s):  
Ian Jentz ◽  
Suzanne McKillop ◽  
Robert Keating
Keyword(s):  
High Temperature ◽  
Failure Mode ◽  
Heat Exchangers ◽  
Failure Modes ◽  
Department Of Energy ◽  
Compact Heat Exchangers ◽  
Printed Circuit ◽  
Mode Effects ◽  
Class A ◽  
Asme Code

Abstract The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation’s energy, environmental, and energy security needs. Advanced high temperature reactor systems will require compact heat exchangers (CHX) for the next generation of nuclear reactor plant designs. A necessary step for achieving this objective is to ensure that the ASME Boiler and Pressure Vessel Code, Section III, Division 5 has rules for the construction of CHXs for nuclear service. Given their high thermal efficiency and compactness, expanding the use of Alloy 800H diffusion bonded Printed Circuit Heat Exchangers (PCHEs) beyond their current application in Section VIII, Division 1 to the high temperature nuclear applications is of interest. The research being completed under the Department of Energy project is focused on preparing a draft Code Case for consideration by the ASME Code Committees for high temperature nuclear components which must meet the requirements of Section III, Division 5, Subsection HB (Class A), Subpart B. Acceptance of a Code Case by the ASME Code Committees to use PCHEs in nuclear service requires a broad understanding of PCHE failure mechanisms. At the highest level, the ASME Code requirements prevent failures of structures and pressure boundaries. Historically, the approach is a process of understanding the known failure modes, such as overload failures, plastic collapse, progressive distortion (ratcheting) and fatigue, and then establishing rules for construction to preclude those failure modes in components. For Division 5 applications, attention to differential thermal expansion, creep life, and creep-fatigue must also be considered. Failure from these loadings is manifest within PCHEs both within the internal micro-channel geometry, and at substantially larger solid header and nozzle attachments. To address the adequacy of the PCHE, a Failure Mode Effects Analysis (FMEA) has been performed for standard etched channel PCHEs. This FMEA is linked to the proposed rules in the code case for compact heat exchangers in Section III, Division 5 Class A applications. The PCHE FMEA covers all design failures addressed by Section III.

An Air-Brayton Nuclear-Hydrogen Combined-Cycle Peak- and Base-Load Electric Plant

2007 ◽  
Author(s):  
Charles Forsberg
Keyword(s):  
High Temperature ◽  
Power Output ◽  
Nuclear Reactor ◽  
Peak Load ◽  
Combined Cycle ◽  
Operating Conditions ◽  
Electricity Production ◽  
Spinning Reserve ◽  
High Temperature Reactor ◽  
Base Load

A combined-cycle power plant is proposed that uses heat from a high-temperature nuclear reactor and hydrogen produced by the high-temperature reactor to meet base-load and peak-load electrical demands. For base-load electricity production, air is compressed; flows through a heat exchanger, where it is heated to between 700 and 900°C; and exits through a high-temperature gas turbine to produce electricity. The heat, via an intermediate heat-transport loop, is provided by a high-temperature reactor. The hot exhaust from the Brayton-cycle turbine is then fed to a heat recovery steam generator that provides steam to a steam turbine for added electrical power production. To meet peak electricity demand, after nuclear heating of the compressed air, hydrogen is injected into the combustion chamber, combusts, and heats the air to 1300°C—the operating conditions for a standard natural-gas-fired combined-cycle plant. This process increases the plant efficiency and power output. Hydrogen is produced at night by electrolysis or other methods using energy from the nuclear reactor and is stored until needed. Therefore, the electricity output to the electric grid can vary from zero (i.e., when hydrogen is being produced) to the maximum peak power while the nuclear reactor operates at constant load. Because nuclear heat raises air temperatures above the auto-ignition temperatures of the hydrogen and powers the air compressor, the power output can be varied rapidly (compared with the capabilities of fossil-fired turbines) to meet spinning reserve requirements and stabilize the grid.

scholarly journals Thermodynamic analysis of high temperature nuclear reactor coupled with advanced gas turbine combined cycle

2019 ◽  
Vol 23 (Suppl. 4) ◽  
pp. 1187-1197 ◽  
Author(s):  
Marek Jaszczur ◽  
Michal Dudek ◽  
Zygmunt Kolenda
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Gas Turbine ◽  
Thermodynamic Analysis ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Coal Gasification ◽  
Nuclear Reactor ◽  
Combined Cycle ◽  
Oil Processing

One of the most advanced and most effective technology for electricity generation nowadays based on a gas turbine combined cycle. This technology uses natural gas, synthesis gas from the coal gasification or crude oil processing products as the energy carriers but at the same time, gas turbine combined cycle emits SO2, NOx, and CO2 to the environment. In this paper, a thermodynamic analysis of environmentally friendly, high temperature gas nuclear reactor system coupled with gas turbine combined cycle technology has been investigated. The analysed system is one of the most advanced concepts and allows us to produce electricity with the higher thermal efficiency than could be offered by any currently existing nuclear power plant technology. The results show that it is possible to achieve thermal efficiency higher than 50% what is not only more than could be produced by any modern nuclear plant but it is also more than could be offered by traditional (coal or lignite) power plant.

A Heuristic for Designing Hybrid Compact Heat Exchangers for Nuclear Applications

2021 ◽  
Author(s):  
Venkata Rajesh Saranam ◽  
Peter Carter ◽  
Kyle Rozman ◽  
Ömer Dogan ◽  
Brian K. Paul
Keyword(s):  
Heat Exchanger ◽  
Diffusion Bonding ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
Compact Heat Exchangers ◽  
Bonding Parameters ◽  
Heat Exchanger Design ◽  
Creep Fatigue ◽  
Gen Iv

Abstract Hybrid compact heat exchangers (HCHEs) are a potential source of innovation for intermediate heat exchangers in nuclear industry, with HCHEs being designed for Gen-IV nuclear power applications. Compact heat exchangers are commonly fabricated using diffusion bonding, which can provide challenges for HCHEs due to resultant non-uniform stress distributions across hybrid structures during bonding, leading to variations in joint properties that can compromise performance and safety. In this paper, we introduce and evaluate a heuristic for determining whether a feasible set of diffusion bonding conditions exist for producing HCHE designs capable of meeting regulatory requirements under nuclear boiler and pressure vessel codes. A diffusion bonding model for predicting pore elimination and structural analyses are used to inform the heuristic and a heat exchanger design for 316 stainless steel is used to evaluate the efficacy of the heuristic to develop acceptable diffusion bonding parameters. A set of diffusion bonding conditions were identified and validated experimentally by producing various test coupons for evaluating bond strength, ductility, porosity, grain size, creep rupture, creep fatigue and channel deviation. A five-layer hybrid compact heat exchanger structure was fabricated and tensile tested demonstrating that the bonding parameters satisfy all criteria in this paper for diffusion bonding HCHEs with application to the nuclear industry.

Chromium Activity Measurements in Nickel Based Alloys for Very High Temperature Reactors: Inconel 617, Haynes 230, and Model Alloys

2009 ◽  
Vol 131 (6) ◽  
Author(s):  
Stéphane Gossé ◽  
Thierry Alpettaz ◽  
Sylvie Chatain ◽  
Christine Guéneau
Keyword(s):  
High Temperature ◽  
Heat Exchangers ◽  
Mass Spectrometric ◽  
Operating Conditions ◽  
Pure Substance ◽  
Haynes 230 ◽  
Inconel 617 ◽  
High Temperature Reactor ◽  
High Temperature Reactors ◽  
Very High

The alloys Haynes 230 and Inconel 617 are potential candidates for the intermediate heat exchangers (IHXs) of (very) high temperature reactors ((V)-HTRs). The behavior under corrosion of these alloys by the (V)-HTR coolant (impure helium) is an important selection criterion because it defines the service life of these components. At high temperature, the Haynes 230 is likely to develop a chromium oxide on the surface. This layer protects from the exchanges with the surrounding medium and thus confers certain passivity on metal. At very high temperature, the initial microstructure made up of austenitic grains and coarse intra- and intergranular M6C carbide grains rich in W will evolve. The M6C carbides remain and some M23C6 richer in Cr appear. Then, carbon can reduce the protective oxide layer. The alloy loses its protective coating and can corrode quickly. Experimental investigations were performed on these nickel based alloys under an impure helium flow (Rouillard, F., 2007, “Mécanismes de formation et de destruction de la couche d’oxyde sur un alliage chrominoformeur en milieu HTR,” Ph.D. thesis, Ecole des Mines de Saint-Etienne, France). To predict the surface reactivity of chromium under impure helium, it is necessary to determine its chemical activity in a temperature range close to the operating conditions of the heat exchangers (T≈1273 K). For that, high temperature mass spectrometry measurements coupled to multiple effusion Knudsen cells are carried out on several samples: Haynes 230, Inconel 617, and model alloys 1178, 1181, and 1201. This coupling makes it possible for the thermodynamic equilibrium to be obtained between the vapor phase and the condensed phase of the sample. The measurement of the chromium ionic intensity (I) of the molecular beam resulting from a cell containing an alloy provides the values of partial pressure according to the temperature. This value is compared with that of the pure substance (Cr) at the same temperature. These calculations provide thermodynamic data characteristic of the chromium behavior in these alloys. These activity results call into question those previously measured by Hilpert and Ali-Khan (1978, “Mass Spectrometric Studies of Alloys Proposed for High-Temperature Reactor Systems: I. Alloy IN-643,” J. Nucl. Mater., 78, pp. 265–271; 1979, “Mass Spectrometric Studies of Alloys Proposed for High-Temperature Reactor Systems: II. Inconel Alloy 617 and Nimomic Alloy PE 13,” J. Nucl. Mater., 80, pp. 126–131), largely used in the literature.

Assessment of Control Room Radiological Habitability of High-Temperature Reactor Pebble-Bed Module in Shidao Bay Multi-Reactor Nuclear Power Site

2018 ◽  
Author(s):  
Xinpeng Li ◽  
Sheng Fang
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Source Term ◽  
Meteorological Data ◽  
Dose Calculation ◽  
Dose Assessment ◽  
Specific Information ◽  
Control Room ◽  
Pebble Bed ◽  
High Temperature Reactor

The control room radiological habitability (CRRH) is important for staff safety in a nuclear power plant, which is also a licensing requirement of the High-temperature Reactor Pebble-bed Module (HTR-PM) in China. Meanwhile, the complexity of the dose assessment increases for the multi-reactor site, which put forward higher requirements for building layout. The CRRH is investigated comprehensively for the multi-reactor site at Shidao Bay in this study. For a large-break loss of coolant accident of HTR-PM and CAP1000 in Shidao Bay nuclear power site, this study estimates doses of body, thyroid and skin due to three exposure pathways using NRC-recommended ARCON96 and dose calculation method in RG 1.195. To perform a realistic evaluation, the latest design and site-specific information are utilized as the input parameters, including the unique accidental source term of HTR-PM and the RG1.183-recommended source term of CAP1000, the release and ventilation parameters, the final layout and the meteorological data in a whole year. The evaluation results demonstrate that the individual dose level of staff in the control room is far below the requirement of the regulatory guide, which guarantees the CRRH of HTR-PM. The contribution of primary radionuclides suggests that tellurium and iodine are primary contributors of the inhalation dose of body and thyroid, which is worthy of paying particular attention to the CRRH design in HTR-PM.

DOE-ASME Generation IV Materials Tasks

2006 ◽  
Author(s):  
Timothy E. McGreevy ◽  
Robert I. Jetter
Keyword(s):  
Nuclear Reactor ◽  
Elevated Temperatures ◽  
Future Generation ◽  
Department Of Energy ◽  
Design Codes ◽  
Technology Program ◽  
Program Plan ◽  
American Society

The Department of Energy (DOE) and the American Society of Mechanical Engineers (ASME) wish to update and expand appropriate materials, construction and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. The scope of interest addresses specific materials and design tasks, all of which are tied to the Generation IV Reactors Integrated Materials Technology Program Plan. Many of the tasks are directly applicable to ASME Section III Subsection NH. The tasks are summarized and discussed with respect to Generation IV needs.

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