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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.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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
Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
Maolong Liu, Yuki Ishiwatari, Koji Okamoto
Nuclear Technology | Volume 186 | Number 2 | May 2014 | Pages 216-228
Technical Paper | Fission Reactors | doi.org/10.13182/NT13-57
Articles are hosted by Taylor and Francis Online.
As Units 1, 2, and 3 of the Fukushima Daiichi nuclear power plant (NPP) entered the phase of long-term station blackout following the huge tsunami, the decay heat could not be effectively removed from the reactor vessel and resulted in high in-vessel pressure and temperature. The Tokyo Electric Power Company announced that the safety relief valves of Fukushima Daiichi NPP Unit 1 (1F1) were never manually opened. However, the measured reactor pressure was decreased to ∼1 MPa at 2:43 on March 12, 2011. Such unanticipated depressurization might accelerate core uncovery and on the other hand delay containment failure caused by direct containment heating. In addition, the failure time and the failure path of the boiling water reactor pressure boundary before manual depressurization have a huge impact on the resulting source term. The authors modeled the creep failure of the stainless steel guide tubes of the source range monitor in the core and the main steam line and estimated the possible depressurization mechanism of 1F1 using the SAMPSON (Severe Accident Analysis Code with Mechanistic, Parallelized Simulations Oriented towards Nuclear Field) severe accident analysis code.