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Radiation Protection & Shielding
The Radiation Protection and Shielding Division is developing and promoting radiation protection and shielding aspects of nuclear science and technology — including interaction of nuclear radiation with materials and biological systems, instruments and techniques for the measurement of nuclear radiation fields, and radiation shield design and evaluation.
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Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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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Latest News
NRC issues subsequent license renewal to Monticello plant
The Nuclear Regulatory Commission has renewed for a second time the operating license for Unit 1 of Minnesota’s Monticello nuclear power plant.
Chang H. Oh, Hong S. Lim, Eung S. Kim
Nuclear Technology | Volume 166 | Number 1 | April 2009 | Pages 101-112
Technical Paper | Special Issue on Nuclear Hydrogen Production, Control, and Management | doi.org/10.13182/NT09-A6972
Articles are hosted by Taylor and Francis Online.
The very high temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 1173 K (900°C) and operational fuel temperatures above 1523 K (1250°C). The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the high-temperature gas-cooled reactor concepts have sufficiently high temperatures to support process heat applications, such as hydrogen production, tar sands, oil shale, desalination, or cogenerative processes, the VHTR's higher temperatures can be detrimental to safety if a loss-of-coolant accident occurs and causes the mechanical strength degradation of the supporting graphite in the lower plenum. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion or density-gradient-driven stratified flow phenomena and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Therefore, there was a need to develop a computer code that can be used for VHTR air ingress-related graphite oxidation analyses. Prior to the start of the Republic of Korea/United States International Nuclear Energy Research Initiative collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate an air ingress phenomenon in the VHTR. Therefore, we have worked for the past 3 yr on developing and validating advanced computational methods for simulating air ingress in the VHTR. The Idaho National Laboratory is developing a system integration model of VHTR and hydrogen production plant. GAMMA code is being considered to be an integrated computer tool to analyze the thermal hydraulics of the coupled plant. Computer models for a high-temperature steam electrolysis (HTSE) process were developed and were implemented in an overall system process optimization code, HYSYS. The HTSE model will be implemented into GAMMA code as the integrated computer tool.This paper describes the governing equations and numerical methods used in GAMMA code and presents a portion of verification of the GAMMA code along with turbomachinery models and HTSE models that will be linked to GAMMA code.