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Division Spotlight
Human Factors, Instrumentation & Controls
Improving task performance, system reliability, system and personnel safety, efficiency, and effectiveness are the division's main objectives. Its major areas of interest include task design, procedures, training, instrument and control layout and placement, stress control, anthropometrics, psychological input, and motivation.
Meeting Spotlight
Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2023)
February 6–9, 2023
Amelia Island, FL|Omni Amelia Island Resort
Standards Program
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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Nuclear Science and Engineering
February 2023
Nuclear Technology
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January 2023
Latest News
Nuclear energy: enabling production of food, fiber, hydrocarbon biofuels, and negative carbon emissions
In the 1960s, Alvin Weinberg at Oak Ridge National Laboratory initiated a series of studies on nuclear agro-industrial complexes1 to address the needs of the world’s growing population. Agriculture was a central component of these studies, as it must be. Much of the emphasis was on desalination of seawater to provide fresh water for irrigation of crops. Remarkable advances have lowered the cost of desalination to make that option viable in countries like Israel. Later studies2 asked the question, are there sufficient minerals (potassium, phosphorous, copper, nickel, etc.) to enable a prosperous global society assuming sufficient nuclear energy? The answer was a qualified “yes,” with the caveat that mineral resources will limit some technological options. These studies were defined by the characteristic of looking across agricultural and industrial sectors to address multiple challenges using nuclear energy.
F. D’Auria, D. Bestion
Nuclear Technology | Volume 208 | Number 6 | June 2022 | Pages 990-1011
Technical Paper | doi.org/10.1080/00295450.2021.1997059
Articles are hosted by Taylor and Francis Online.
In the domain of reactor transient simulation, the identification of thermal-hydraulic phenomena (THPs) plays a major role. The system codes should model all influential THPs and should be validated against integral effect tests and separate effect tests which cover all influential THPs. The validation and the uncertainty quantification should cover every model related to an influential THP. A list of 116 THPs, recently established, covers all water-cooled reactors and design basis accident (DBA) analyses. It synthesizes more than 30 years of Organisation for Economic Co-operation and Development and International Atomic Energy Agency activities conducted by several safety specialists. A new tentative method to identify THPs was proposed based on two sources of information, the parameter evolutions in transients (depressurization, voiding, refill, heating, ...) and the set of balance equations with source and sink terms for convection, diffusion, interfacial transfers, and wall transfers. This method is just based on Gen-2 pressurized water reactor (PWR) transient analysis at the system, component, and basic process levels.
The comparison of the 116 THP list with the list of the tentative methods is made in this paper for the particular case of GEN-2 PWRs. No major contradiction was found. Both methods identified phenomena at the system, component, and process level. The 116 list better identified special components that require “special models” in system codes. The use of equations identified many more local process THPs, which may help ranking phenomena in a scaling analysis and identifying validation needs. The comparison confirms a potential synergy and complementarity between approach 1, which is based on the 116 THP list, and approach 2, which is based on the tentative method; the outcome of this comparison suggests further efforts to combine them and complement them in a new international collaborative context.
This analysis reports ongoing discussions between members of the Forum and Network of System Thermal Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics network of system code developers.
