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Division Spotlight
Nuclear Installations Safety
Devoted specifically to the safety of nuclear installations and the health and safety of the public, this division seeks a better understanding of the role of safety in the design, construction and operation of nuclear installation facilities. The division also promotes engineering and scientific technology advancement associated with the safety of such facilities.
Meeting Spotlight
Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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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Latest News
BWXT will scout potential TRISO fuel production sites in Wyoming
BWX Technologies Inc. announced today that its Advanced Technologies subsidiary has signed a cooperation agreement with the state of Wyoming to evaluate locations and requirements for siting a potential new TRISO nuclear fuel fabrication facility in the state.
V. K. Chexal, W. H. Layman, W. W. Brown, G. B. Caldwell
Nuclear Technology | Volume 54 | Number 3 | September 1981 | Pages 332-341
First International Retran Meeting | Heat Transfer and Fluid Flow | doi.org/10.13182/NT81-A32778
Articles are hosted by Taylor and Francis Online.
The Nuclear Safety Analysis Center (NSAC) has performed a thermal-hydraulic analysis of the Crystal River Unit 3 nuclear plant incident that occurred on February 26, 1980. The incident was initiated at 2:23 p.m. by an instrument and control system electrical malfunction that resulted in loss of power on the nonnuclear instrumentation (NNI) “X” bus. This failure caused the loss of several control and indication parameters, including pressurizer and steam generator level, and all reactor coolant system (RCS) temperatures. The loss of control parameters fed erroneous signals to the integrated control system, which in turn initially increased reactor power level, terminated feedwater flow to the steam generators, and opened steam turbine throttle valves to maintain outlet steam conditions. In addition, the power-operated relief valve (PORV) opened prematurely and remained open as a result of faulty circuit design in the NNI. This transient culminated in a reactor trip, turbine trip, and an engineered safeguards actuation, discharging ≈40 000 gal of primary system coolant to the floor of the containment building. The thermal-hydraulic analysis of the above event was performed by NSAC, using the RETRAN computer code. The objectives were as follows: