Modeling and performance of the MHTGR (Modular High-Temperature Gas-Cooled Reactor) reactor cavity cooling system

Technical Report · Sun Apr 01 00:00:00 EST 1990

DOI:https://doi.org/10.2172/7013277· OSTI ID:7013277

Conklin, J C ^[1]

---

1. Oak Ridge National Lab., TN (USA)

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (USA). Div. of Regulatory Applications; Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USNRC
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 7013277
• Report Number(s): NUREG/CR-5514; ORNL/TM-11451; ON: TI90010518; TRN: 90-026892
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING
HTGR TYPE REACTORS
REACTOR VESSELS
REACTOR COOLING SYSTEMS
COMPUTERIZED SIMULATION
CAVITIES
DYNAMICS
FLOW MODELS
FLUID FLOW
GEOMETRY
HEAT FLUX
HEAT TRANSFER
HYDRAULICS
M CODES
PERFORMANCE TESTING
POWER DENSITY
REACTOR SAFETY
STEAM
SURFACE PROPERTIES
THERMODYNAMICS
COMPUTER CODES
CONTAINERS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
MATHEMATICAL MODELS
MATHEMATICS
MECHANICS
REACTOR COMPONENTS
REACTORS
SAFETY
SIMULATION
TESTING
220900* - Nuclear Reactor Technology- Reactor Safety
210300 - Power Reactors
Nonbreeding
Graphite Moderated
420400 - Engineering- Heat Transfer & Fluid Flow