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Journal ofNuclear Materials 103 & 104(1981)1563-1566 North-HotiandPublishingCompany
INSTRUMENTED
IN-REACTOR
TEST CAPABILITIES
IN FFTF
R. E. Bauer
Hanford
Engineering Development Laboratory Richland, Washington
Fusion materials research and development requires high neutron fluxes coupled with Both of these needs can be provided by the Fast Flux a large irradiation volume. The Materials Open Test Assembly, the vehicle used for materials Test Facility. irradiations in the FFTF, is described.
1.
INTRODUCTION
A major problem in fusion materials research is the unavailability of neutron devices that produce high dpa rates for accelerated testing, yet provide enough irradiation volume for the The Fast Flux Test amount of testing needed. Facility near Richland, Washington can provide both the testing volume and high neutron damMaterials age rates on a near term basis. tests can be conducted solely in the FFTF or irradiations can be alternated between the FFTF and another facility such as High Flux Isotope Reactor to adjust He/dpa ratios to those exMaterial pected for first wall materials. irradiations in the FFTF are conducted using two instrumented assemblies called Materials Open Test Assemblies (MOTA). This paper describes the features of the MOTA. 2.
DESCRIPTION ASSEMBLY
2.1 Irradiation
OF THE ~TERI~S
OPEN TEST
Environment
MOTAs 1 and 2 will be located in row 4 positions of the sodium cooled Fast Test Reactor111 (FTR) core. MOTA 2 will be irradiated for two consecutive reactor cycles and then removed for 1 cycle for specimen examination. The peak damage accumulated in this two cycle irradiation (approximately 200 full power days) is 32 dpa (Figure 1). After examination, the specimens are placed in a new test vehicle and inserted for another two cycle irradiation. This pattern is repeated until the goal damage is reached. 2.2 Test Assembly
and Canisters
The test assembly consists of 48 canisters arranged in eight levels around a central Thirty of these canisters which restalk. present 2.411 of usable volume are located directly in the core region with the other 18 (1.5%) located just above the core. In addition to the 48 canisters, a larger (%l&) below core canister is located in the vicinity of the sodium coolant inlet. The in-core and below
core canisters
are shown in Figure 2.
The canisters have a cylindrical usable irradiation space 2.82 cm in diameter by 13.34 cm long. The temperature at the midplane of each canister is measured by a chromel-alumel thermocouple which projects from the bottom of the canister along the center axis to the middle of the The below core canister specimen compartment. is a non-instrumented weeper 9.19 cm in diameter by 13.3 cm long which operates at the inlet coolant temperature of 360°C. The simpler The canisters are of two designs. is a weeper which allows the sodium coolant to enter the specimen compartment and flow around and among the test specimens, therefore maintaining the specimen temperature at the ambient coolant temperature. Temperatures greater than the ambient coolant temperature are achieved by using the second capsule design, shown in Figure 2. A gas gap surrounding the specimen compartment insulates the specimens and the static sodium surrounding them from the reactor coolant. Utilizing gamma heating in the specimens and internal components of the canister as a heat source, canister temperature can be controlled by adjusting the Ar-He gas composition, hence thermal conductivity in the insulating gap. The canisters are fastened to the central stalk with a hinge so that they may be rotated out from the stalk during disassembly allowing the insulating end cap and specimens to be easily removed. The thermocouple and gas leads to the canister bend around this hinge to permit flexing, are routed up through the center stalk, penetrate the reactor pressure boundary, and are connected to the instrumentation and control system. 2.3
Temperature
Control and Data Acquisition
On line temperature control and data acquisition tasks are accomplished by a computerized system. The computer records the temperatures
R.E. Bauer /Instrumented
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in-reactor
test capabilities
in FFTI,‘
MOTA NEUTRON FLUX DISTRIBUTION DISTANCE FROM CORE CENTER Icm)
s.
10’3
s id
BC
10’2
I -30
1
L -20
2
3
I
I
I
I
I
1
I
I
-10
0
10
20
30
40
w
a0
DISTANCE FROM CORE CENTER (In.)
Figure 1.
MOTA Axial Fast Flux Distribution.
Figure 2.
MOTA Test Train
(30 in-core
canisters
shown) and Capsule
Configuration.
R.E. Bauer 1 Instrumented in-reactor test capabilities in FFTF
of all 48 canisters onto a removable rigid magnetic disk for later offline data reFor 30 of the canisters it compares duction. the current canister temperature with the If the target temperature for that canister. current temperature is not within ?5"C of the target temperature, then the computer calculates and then distributes a new gas blend to that canister to bring the temperature within limits. Data for rupture times of pressurized tube specimens in MOTA are obtained using the FTR Each specimen cover gas monitoring system. is filled with a special "tag" gas which differentiates if from all other specimens. The gas "burst" when the specimen fails can be detected and the type of gas released identified using the cover gas chromatograph. 3.
CONCLUSION
MOTA can be a powerful tool for conducting fusion materials irradiations. It can help accelerate fusion materials testing and development by providing a large test volume in a high neutron fast flux on a near term Detailed information on the neutron basis. spectrum in FFTF may be found in Reference 2. The MOTA canisters
are designed
for structural
materials testing at temperatures ranging from 37O"C, set bv the inlet coolant temperature, to 73O"C, a practical limit based on materials limitations and time of testing. The design and instrumentation configuration, however, lend themselves to modification for other types of materials, such as solid breeders. REFERENCES [ll
Fast Flux Test Facility Irradiation Services, U.S. Energy Research & Development Administration (August 1977).
[2]
Fleischman, R. M. and Nelson, J. V., Three Dimensional Neutronics Calculations for the Fast Test Reactor (FTR) and the FTR Engineering Mock Up Critical Assembly (EMC), HEDL-TME-72-42 (April 1972).
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