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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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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Deep Space: The new frontier of radiation controls
In commercial nuclear power, there has always been a deliberate tension between the regulator and the utility owner. The regulator fundamentally exists to protect the worker, and the utility, to make a profit. It is a win-win balance.
From the U.S. nuclear industry has emerged a brilliantly successful occupational nuclear safety record—largely the result of an ALARA (as low as reasonably achievable) process that has driven exposure rates down to what only a decade ago would have been considered unthinkable. In the U.S. nuclear industry, the system has accomplished an excellent, nearly seamless process that succeeds to the benefit of both employee and utility owner.
Joseph O. Erb, James G. Miller
Nuclear Technology | Volume 83 | Number 3 | December 1988 | Pages 367-373
Technical Paper | Fifth International Retran Meeting / Heat Transfer and Fluid Flow | doi.org/10.13182/NT88-A34149
Articles are hosted by Taylor and Francis Online.
The rod ejection transient is a postulated Condition IV event initiated by the mechanical failure of a control rod mechanism pressure housing. Such a failure results in the rapid ejection of a rod cluster control assembly from the core, followed by a fast reactivity insertion. A severe asymmetric core power distribution may result, possibly leading to fuel rod damage. Reactor protection for the transient is provided by negative reactivity feedback effects and by reactor trips on high neutron flux levels. This transient has been modeled for Virginia Electric and Power Company’s Surry and North Anna nuclear power stations using RETRAN-02. The analysis is performed in two parts. First, the core average power history is calculated using a single-loop model with point kinetics and three axially stacked core control volumes. The ejected rod’s reactivity is inserted linearly over 0.1 s. The negative reactivity feedback effects due to Doppler and moderator temperature changes and the reactor trip are also modeled. The effect of the locally peaked core flux shape, omitted by the nominal point kinetics model, is approximated by applying a conservative power weighting factor to the Doppler reactivity feedback. The core average power history is adjusted to represent peak core power conditions and input to the hot spot thermal-hydraulic analysis model. The hot spot model, which represents a single fuel rod at the core’s peak power, predicts maximum fuel enthalpy and temperature transients. This model has two control volumes, one for the hot spot location and the second for a sink for flow from the hot channel. From these results, the amount of fuel damage and the radiological consequences can be assessed.
