Analysis of an AP600 intermediate-size loss-of-coolant accident

Technical Report · Fri Sep 01 04:00:00 EDT 1995

DOI:https://doi.org/10.2172/107770· OSTI ID:107770

Boyack, B. E.; Lime, J. F. ^[1]

---

1. Los Alamos National Lab., NM (United States)

A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; American Nuclear Society, La Grange Park, IL (United States); American Inst. of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers, New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); European Nuclear Society (ENS), Bern (Switzerland); Atomic Energy Society of Japan, Tokyo (Japan); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 107770
• Report Number(s): NUREG/CP--0142-Vol.2; CONF-950904--Vol.2; ON: TI95017078
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
DIAGRAMS
EXPERIMENTAL DATA
HYDRAULICS
LOSS OF COOLANT
N CODES
PWR TYPE REACTORS
R CODES
REACTOR SAFETY
SCALE MODELS
T CODES
THEORETICAL DATA