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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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Latest News
Vogtle-3 shuts down for valve issue
One of the new Vogtle units in Georgia was shut down unexpectedly on Monday last week for a valve issue that has since been investigated and repaired. According to multiple local news outlets, Georgia Power reported on July 17 that Unit 3 was back in service.
Southern Company spokesperson Jacob Hawkins confirmed that Vogtle-3 went off line at 9:25 p.m. local time on July 8 “due to lowering water levels in the steam generators caused by a valve issue on one of the three main feedwater pumps.”
Kuan-Fu Chen, Ching-Hui Wu, Min Lee
Nuclear Technology | Volume 161 | Number 2 | February 2008 | Pages 81-97
Technical Paper | Reactor Safety | doi.org/10.13182/NT08-A3915
Articles are hosted by Taylor and Francis Online.
Probabilistic safety assessment (PSA) employs a systematic approach to estimate the risk associated with the operation of nuclear power plants (NPPs). Severe accident management guidance (SAMG), which delineates the mitigation actions of core meltdown accidents of NPPs, is developed to support operators and staffs in the technical support centers during the emergency responses of core melt accidents. Proper execution of SAMG could lower the failure probability of containment and reduce the amount of radionuclides released to the environment during the accident. It can be expected that the implementation of SAMG will reduce the risk of NPPs. However, SAMG is not available when most of the conventional level-2 PSA analyses are performed. In the present study, the mitigation actions of SAMG are incorporated into the level-2 PSA model of the ChinShan Nuclear Power Station of the Taiwan Power Company. The NPP analyzed employs a General Electric-designed boiling water reactor-4 with Mark I containment.The effectiveness of the mitigation actions specified in SAMG to terminate the progression of the accident is verified and validated using the MAAP4 code. The containment system event trees and containment phenomenological event trees of the level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The Human Cognitive Reliability (HCR) and Technique for Human Error Rate Prediction (THERP) models are used to quantify the human error probability (HEP) of all the actions in the level-2 PSA model. The MAAP4 code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the total frequency of accident progression beyond vessel failure is reduced by 41% and the change in the probability of containment staying intact is not very significant because of the implementation of SAMG. After SAMG implementation, the frequency of containment early failure is reduced by 69.9%. The frequency of suppression pool venting is increased by 77.9%. The changes in the frequency of other containment failure modes are relatively insignificant. The most important human action is specified in Guideline RC/F of Severe Accident Guideline-1, i.e., In-Vessel Injection to Arrest Core Damage.