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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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College students help develop waste-measuring device at Hanford
A partnership between Washington River Protection Solutions (WRPS) and Washington State University has resulted in the development of a device to measure radioactive and chemical tank waste at the Hanford Site. WRPS is the contractor at Hanford for the Department of Energy’s Office of Environmental Management.
Bismark Tyobeka, Andreas Pautz, Kostadin Ivanov
Nuclear Science and Engineering | Volume 168 | Number 2 | June 2011 | Pages 93-114
Technical Paper | doi.org/10.13182/NSE10-60
Articles are hosted by Taylor and Francis Online.
We introduce a new coupled neutronics/thermal-hydraulics code system for analyzing transients of high-temperature gas-cooled reactors (HTGRs), based on a neutron transport theory approach. At the heart of the coupled code system resides the DORT-TD code, a time-dependent extension of the well-known DORT discrete ordinates code. DORT-TD uses a fully implicit time integration scheme and is coupled via its generalized thermal-hydraulics interface to the THERMIX-DIREKT code, an HTGR-specific heat conduction/convection code for pebble bed-type reactor cores. Feedback is accounted for by interpolating multigroup cross sections from libraries pregenerated with appropriate spectral codes. These libraries are structured for user-specified discrete sets of thermal-hydraulic parameters, e.g., fuel and moderator temperatures. The coupled code system is applied to a pebble bed HTGR model case, i.e., the PBMR 268 MW design. Steady-state studies are performed, and several design-basis and beyond-design-basis transients are simulated in an effort to assess the adequacy of using neutron diffusion theory against the more accurate but yet computationally more expensive neutron transport approach. Relatively small but significant differences arise from using either theoretical approach, from which it is concluded that transport theory as the more versatile tool should be used as reference to quantify the effects of the approximations inherent in diffusion and to gain confidence in its predictive power, especially in safety analyses. In an effort to validate the DORT-TD/THERMIX code system, the neutronics stand-alone solver is benchmarked against available transient benchmark exercises, and the coupled code system is applied to the Organisation for Economic Co-operation and Development/Nuclear Energy Agency/Nuclear Science Committee PBMR 400 MW Coupled Neutronics Thermal Hydraulics Transient Benchmark, demonstrating its remarkable viability for a wide range of safety cases. The final product is a high-fidelity, highly flexible, and well-validated state-of-the-art computer code system, with multiple capabilities to analyze HTGR safety-related transients in an accurate and efficient manner.