MSRE DESIGN AND OPERATIONS REPORT. PART III. NUCLEAR ANALYSIS
Preliminary considerations of the effects of core size and fuel-to- moderator ratio on critical mass and fuel concentration led to the specification of a core about 4.5 ft in diameter by 5.5 ft high. The average fuel fraction was set at 0.225. The nuclear characteristics of the reactor were examined for three combinations of fissile and fertile material (UF/sub 4/ and ThF/sub 4/) in a carrier salt composed of Li, Be, and ZrF/sub 4/. The radial distribution of the thermal neutron flux is strongly influenced by the presence of three controlrod thimbles near the axis of the core, with the result that the radial thermal flux maximum occurs about 8 in. from the axis. The axial distribution is essentially sinusoidal. The magnitude of the thermal flux depends on the choice of the fuel. Both the fael and the moderator temperature coefficients of reactivity are substantially negative, leading to prompt and delayed negative power coefficients. Reactivity coefficients were also calculated for changes in uranium concentration, Xe/sup 135/ concentration, and fuel-salt and graphite densities. Temperature distributions in the fuel and graphite in the reactor were calculated for the design power level. With the fuel inlet and outlet temperatures at 1175 and 1225 deg F, respectively, the fuel and graphite reactivityweighted average temperatures are 1211 and 1255 deg F, respectively. Fuel permeation of 2% of the graphite volume would increase the graphite weighted average temperature by 7 deg F. The power coefficient of reactivity with the reactor outlet temperature held constant is --0.006 to --0.008% delta k/k per Mw. Circulation of the fuel at 1200 gpm reduces the effective delayed neutron fraction from 0.0067 to 0.0036. Xenon poisoning is strongly dependent on the major competing mechanisms of stripping from the fuel in the pump bowl and transfer into the bare graphite. The equilibrium poisoning at 10 Mw is expected to be between --1.0 and --1.7% delta k/k. The fuel contains an inherent neutron source of over 10/sup 5/ n/sec due to alpha ,n reactions in the salt. This meets all the safety requirements of a source, but an external source will increase the flux for convenient monitoring of the subcritical reactivity. The total worth of the three control rods ranges from 5.6 to 7.6% delta k/k, depending on the fuel salt composition. Shutdown margins at 1200 deg F are 3.5% delta k/k or more in all cases. Calculations were made for conceivable reactivity accidents involving uncontrolled rod withdrawal, cold slugs,'' abnormal fuel additions, loss of graphite, abnormal filling of the reactor and primary flow stoppage. No intolerable conditions are produced if the reactor ssfety system (rod drop at 150% of sign power) functions for two of the three rods. The bio igical shield, with the possible addition of stacked concrete blocks in some areas, reduces the calculated radiation dose rates to permissible levels in all accessible areas. (auth)
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- DOE Contract Number:
- W-7405-ENG-26
- NSA Number:
- NSA-18-009650
- OSTI ID:
- 4114686
- Report Number(s):
- ORNL-TM-730
- Resource Relation:
- Other Information: Orig. Receipt Date: 31-DEC-64
- Country of Publication:
- United States
- Language:
- English
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ABSORPTION
ACCIDENTS
ADSORPTION
ALPHA PARTICLES
BERYLLIUM
BUILDINGS
CONCRETES
CONFIGURATION
CONTROL ELEMENTS
CONTROL SYSTEMS
COOLANT LOOPS
CRITICALITY
DEGASSING
DELAYED NEUTRONS
DENSITY
DISTRIBUTION
ENRICHMENT
EQUATIONS
EXCURSIONS
FAILURES
FERTILE MATERIALS
FISSION PRODUCTS
FISSIONABLE MATERIALS
FUELS
FUSED SALTS
GASES
GRAPHITE
GRAPHITE MODERATOR
HEAT TRANSFER
HIGH TEMPERATURE
LIQUID FLOW
LITHIUM
LOSSES
LOW TEMPERATURE
MASS
MATERIALS TESTING
MODERATORS
MONITORING
MSRE
MULTIPLICATION FACTORS
NEUTRON FLUX
NEUTRON SOURCES
NEUTRONS
NUCLEAR REACTIONS
OPERATION
PERSONNEL
PLANNING
POISONING
POWER PLANT