A parametric study of the steady-state operational characteristics of the Ohio State University natural circulation indirect-cycle, inherently safe boiling water reactor

Journal Article · Sat Jul 01 04:00:00 EDT 1995 · Nuclear Technology

OSTI ID:89646

Aybar, H S ^[1]; Aldemir, T; Christensen, R N ^[2]

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1. Turkish Atomic Energy Authority, Ankara (Turkey). Nuclear Computation and Design Group
2. Ohio State Univ., Columbus, OH (United States)

The Ohio State University Inherently Safe Reactor (OSU-ISR) is a conceptual design for a 340-MW(electric) [1,000-MW(thermal)], natural circulation, indirect-cycle, small boiling water reactor. All the OSU-ISR primary loop components are housed within a prestressed concrete reactor vessel (PCRV). The OSU-ISR performance has been investigated as a function of several design parameters in an attempt to better understand the interdependency among the system variables and hence to establish a knowledge base for the refinement of the conceptual design. The computational tool used in the study is a Dynamic Simulation for Nuclear Power Plants (DSNP) code whose predictions for the steady-state OSU-ISR performance compare favorably with RELAP5/MOD3 results for most of the operational characteristics of interest. The results show that (a) the key quantity that governs the OSU-ISR steady-state performance is the pressure difference between the primary and the secondary loops, (b) the magnitude of water-level swell (which occurs due to void formation in the core during operation and which affects the size of the steam separators that need to be used) can be more effectively controlled by varying the PCRV water level at cold shutdown rather than by varying the internal PCRV dimensions, (c) turbine inlet steam quality can be controlled without substantially affecting the other operational parameters by varying the secondary mass flow rate, and (d) the PCRV pressure and core exit steam quality are most sensitive to changes in the secondary loop pressure. The results also show that if there is a large drop in the secondary loop pressure (e.g., due to a steam line break), then although this pressure drop may induce a large drop in the PCRV pressure, the core flow, and hence core cooling capability, will not be appreciably affected.

• OSTI ID: 89646
• Journal Information: Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 1 Vol. 111; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
D CODES
DESIGN
NATURAL CONVECTION
PARAMETRIC ANALYSIS
PERFORMANCE
PRESSURE DEPENDENCE
PRIMARY COOLANT CIRCUITS
R CODES
REACTOR OPERATION
REACTOR SAFETY