Reactivity Studies on the Advanced Neutron Source

Conference · Sun Sep 30 00:00:00 EDT 1990

OSTI ID:6491967

Ryskamp, John M. ^[1]; Redmond, II, Everett L. ^[1]; Fletcher, C. D. ^[1]

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1. Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States). EG&G Idaho, Inc.

An Advanced Neutron Source (ANS) with a peak thermal neutron flux of about 8.5 x 10¹⁹ m^-2s^-1 is being designed for condensed matter physics, materials science, isotope production, and fundamental physics research. The ANS is a new reactor-based research facility being planned by Oak Ridge National Laboratory (ORNL) to meet the need for an intense steady-state source of neutrons. The design effort is currently in the conceptual phase. A reference reactor design has been selected in order to examine the safety, performance, and costs associated with this one design. The ANS Project has an established, documented safety philosophy, and safety-related design criteria are currently being established. The purpose of this paper is to present analyses of safety aspects of the reference reactor design that are related to core reactivity events. These analyses include control rod worth, shutdown rod worth, heavy water voiding, neutron beam tube flooding, light water ingress, and single fuel element criticality. Understanding these safety aspects will allow us to make design modifications that improve the reactor safety and achieve the safety related design criteria.

• Research Organization: Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States). EG&G Idaho, Inc.
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 6491967
• Report Number(s): EGG-M-89519; CONF-900917-20; ON: DE91001916; TRN: 90-035992
• Resource Relation: Conference: American Nuclear Society Topical Meeting on Safety of Non-Commercial Nuclear Reactor Research and Irradiation Facilities, Boise, ID (United States), 30 Sep - 3 Oct 1990
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
NEUTRON SOURCE FACILITIES
RESEARCH REACTORS
REACTIVITY
REACTOR SAFETY
CONTROL ROD WORTHS
DESIGN
FUEL ELEMENTS
HEAVY WATER
M CODES
R CODES
REACTOR CORES
REACTOR PHYSICS
VOID COEFFICIENT
WATER INFLUX
COMPUTER CODES
HYDROGEN COMPOUNDS
OXYGEN COMPOUNDS
PHYSICS
REACTIVITY COEFFICIENTS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
WATER
Nuclear Criticality Safety Program (NCSP)
Thermal Neutron Flux
Monte Carlo Theory Models
Evaluated Nuclear Data File (ENDF)
MCNP-3B
RELAP-5
Heavy Water Reactor
Hafnium Control Rods
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors
220100 - Nuclear Reactor Technology- Theory & Calculation