Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly
- Idaho State University 921. S. 8th Avenue Pocatello, ID 83209 (United States)
The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)
- OSTI ID:
- 23042547
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 115; Conference: 2016 ANS Winter Meeting and Nuclear Technology Expo, Las Vegas, NV (United States), 6-10 Nov 2016; Other Information: Country of input: France; 3 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US); ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
CENTRIFUGAL PUMPS
CLADDING
COOLANTS
CORROSION
DEPLETED URANIUM
EBR-2 REACTOR
ENRICHED URANIUM
FUEL ASSEMBLIES
FUEL RODS
NEUTRON LEAKAGE
NEUTRON REFLECTORS
REACTIVITY COEFFICIENTS
REACTOR PHYSICS
REACTOR VESSELS
SENSITIVITY ANALYSIS
STAINLESS STEELS
STEADY-STATE CONDITIONS