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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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2025 ANS Annual Conference
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Chicago, IL|Chicago Marriott Downtown
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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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Webinar: MC&A and safety in advanced reactors in focus
Towell
Russell
Prasad
The American Nuclear Society’s Nuclear Nonproliferation Policy Division recently hosted a webinar on updating material control and accounting (MC&A) and security regulations for the evolving field of advanced reactors.
Moderator Shikha Prasad (CEO, Srijan LLC) was joined by two presenters, John Russell and Lester Towell, who looked at how regulations that were historically developed for traditional light water reactors will apply to the next generation of nuclear technology and what changes need to be made.
Te-Chuan Wang, Shih-Jen Wang, Jyh-Tong Teng
Nuclear Technology | Volume 156 | Number 3 | December 2006 | Pages 347-359
Technical Note | Thermal Hydraulics | doi.org/10.13182/NT06-A3796
Articles are hosted by Taylor and Francis Online.
Chinshan is a Mark-I boiling water reactor nuclear power plant (NPP) located in north Taiwan. It incorporates several severe-accident-mitigating features, especially two raw-water tanks in the mountain. According to a probabilistic risk assessment (PRA) of Chinshan NPP, station blackout (SBO) sequences are the most dominant sequences in internal core damage frequency. No credit is taken for the raw-water system in the development of a Chinshan PRA. Therefore, two dominant sequences (T3UTERDGX and T3UTERDG) of the SBO in the Chinshan PRA are cited as reference cases to evaluate the capacity of the raw-water system in the PRA and severe accident. The T3UTERDGX sequence is initiated by loss of off-site power (T3) followed by failure of both diesel generators (DG), failure of gas turbine generators, and failure to recover alternating current (ac) power (ER). That results in loss of all on- and off-site ac power. The high-pressure injection systems fail (UT) initially and timely reactor depressurization fails (X). The T3UTERDG sequence is the same as the T3UTERDGX sequence, except for failure of timely reactor depressurization (X). The MAAP4 code is used as a tool to evaluate the effectiveness of the raw-water system. Based on MAAP4 analysis, the raw-water system cannot cool down the core in the T3UTERDG sequence after introducing severe-accident-management guidelines. The raw-water system cannot flood dry-well water level above minimum debris submerge level (MDSL) in the T3UTERDGX sequence after reactor pressure vessel (RPV) breach. Sensitivity studies show that raw-water injection before the vessel water level reaches level 2 (L-2) can keep core coolability in the T3UTERDG sequence. Three times the raw-water injection rate is the minimum flow rate to flood the dry-well water level above MDSL and cool down the corium on the dry-well floor in the T3UTERDGX sequence. A raw-water system can be used as a mitigating measure, especially in an SBO. The RPV should be depressurized as quickly as possible if a raw-water system is the only mitigation measure in the accident. It is worthwhile to increase the raw-water flow rate to cool down the debris in the dry well after RPV breach.
