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2026 Annual Conference
May 31–June 3, 2026
Denver, CO|Sheraton Denver
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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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Modernizing I&C for operations and maintenance, one phase at a time
The two reactors at Dominion Energy’s Surry plant are among the oldest in the U.S. nuclear fleet. Yet when the plant celebrated its 50th anniversary in 2023, staff could raise a toast to the future. Surry was one of the first plants to file a subsequent license renewal (SLR) application, and in May 2021, it became official: the plant was licensed to operate for a full 80 years, extending its reactors’ lifespans into 2052 and 2053.
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.