Computational Analysis of the Thermal-Hydraulic Characteristics of the Encapsulated Nuclear Heat Source

Journal Article · Thu Dec 15 04:00:00 EST 2005 · Nuclear Technology

OSTI ID:20840331

Nishi, Yoshihisa ^[1]; Ueda, Nobuyuki ^[1]; Kinoshita, Izumi ^[1]; Greenspan, Ehud ^[2]

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1. Central Research Institute of Electric Power Industry (Japan)
2. University of California, Berkeley (United States)

The encapsulated nuclear heat source (ENHS) is a modular reactor that was selected by the 1999 U.S. Department of Energy Nuclear Energy Research Initiative program as a candidate Generation IV reactor concept. It is a fast neutron spectrum reactor cooled by lead-bismuth eutectic using natural circulation. One of the unique features of the ENHS is that the fission-generated heat is transferred from the primary coolant to the secondary coolant through rectangular intermediate heat exchanger (IHX) channels. The decay heat is removed by the reactor vessel auxiliary cooling system (RVACS).Events of protected loss of heat sink (PLOHS) and unprotected transient overpower (UTOP) have been analyzed for the ENHS using the CERES transient simulation code for liquid-metal-cooled reactors.It is found that the ENHS core is sufficiently cooled by the RVACS under the PLOHS condition. The core flow rate is affected by the growth and disappearance of temperature stratification in the primary plenum. It is also found that even under the inconceivable UTOP event considered, the ENHS reactor core is not catastrophically damaged. This is due to negative reactivity feedback from the radial expansion of the core, the grid plate, and the Doppler effect. The use of high-performance ferritic steel instead of HT-9 and proper design of the reactor control system could provide large safety margins against cladding damage.

• OSTI ID: 20840331
• Journal Information: Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 3 Vol. 152; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BISMUTH
CLADDING
FAST NEUTRONS
FERRITIC STEELS
FISSION
HEAT
HEAT SINKS
HEAT SOURCES
LEAD
LIQUID METAL COOLED REACTORS
NATURAL CONVECTION
PRIMARY COOLANT CIRCUITS
REACTIVITY COEFFICIENTS
REACTOR CONTROL SYSTEMS
REACTOR CORES
REACTOR VESSELS
SAFETY MARGINS
THERMAL HYDRAULICS
TRANSIENTS