Summary of benchmark efforts for FFTF initial isothermal physics tests.

1. Benchmark Evaluation of
the Initial Isothermal
Physics Measurements at
the Fast Flux Test Facility
John Darrell Bess
R&D Staff Engineer
Reactor Physics Analysis and Design
Presenter: David W. Nigg
Directorate Fellow
Nuclear Science and Engineering Division
PHYSOR 2010
May 10, 2010
This paper was prepared at Idaho National Laboratory for the U.S.
Department of Energy under Contract Number (DE-AC07-05ID14517)

2. Objective
• Perform a benchmark analysis of the initial
isothermal physics test at the Fast Flux Test
Facility (FFTF)
–Support Fuel Cycle Research and Development
(FCR&D) and Generation-IV activities at Idaho National
Laboratory
–Submit completed benchmark for publication in the
International Handbook of Evaluated Reactor Physics
Benchmark Experiments
2

3. 3

4. Evaluation Process: ICSBEP & IRPhEP
4

5. Fast Flux Test Facility (FFTF)
• 400 MW, Na-cooled, MOX-fueled fast reactor
• Prototypic of a Liquid Metal Fast Breeder Reactor
• National research test facility
– Nuclear power plant operations and maintenance protocols
– Advanced nuclear fuels
– Materials and components
– Reactor safety
design
– Radioisotope
production
• Opportunity for
computational
validation of
methods
5

6. FFTF Core
• MOX Fuel
–73 Assemblies
–~23% Pu (~90% 239Pu)
–Natural-U
–SS-316 Cladding
• Absorbers
–Natural-B4C
• Reflectors
–Inconel-600
• Various Test
Assembly Locations
6

7. Summary of Evaluated Isothermal Physics Tests
• Fully-loaded critical with 73 fuel assemblies
• Two neutron spectra measurements
• 32 reactivity measurements
–21 control rod worths, two control rod bank worths, six
differential rod worths, two shutdown margins, one
excess reactivity
• Isothermal temperature coefficient
• Low-energy electron and gamma spectra
measurements
Measurements performed at
isothermal core temperature of 400ºF
7

8. Challenges
• Insufficient public data to
perform a detailed core
analysis
• Data-mining efforts at
PNNL to identify and
obtain core component
drawings and additional
experimental data
• HEX-Z Homogenization of
many reactor components
was necessary to
complete the benchmark
assessment
8

9. Benchmark Model – Driver Fuel Pins
SS316 cladding
2
84
Inconel 600
0.5
14.478 8
reflector 0 .50
OD
ID
O
D
0.
4 81
Void
33
2.032 UO2 insulator
pellet
Void
2
84
0.5
08
0.5
OD
ID
OD
124.46
0.4
82
6
91.44
(U,Pu)-O2
fuel pellet stack
Fuel/clad gap
2
84
0 .5
8
0 .50
OD
ID
OD
0.
49
40
2.032 UO2 insulator
3
pellet
14.478 Inconel 600
reflector
Dimensions in cm
09-GA50001-120-1
9

10. Benchmark Model – Driver Fuel Assemblies
SS316 duct
Fuel pin
Gas Plenum Region
109.22
298.45
Fuel Pin Lattice Region 124.46
Sodium coolant
SS316 duct
Sodium coolant
with homogenized
SS316 wire wrap
Fuel pin (217)
OD 0.5842
Fuel Pin Attachment Region
10.16
Pitch
0.72644 Lower Axial Shield Region 54.61
Detail of Fuel Pin
0.3048
Lattice Region
Dimensions in cm
09-GA50001-120-3
12.051
Drawing not to scale
11.0109
11.6205
Dimensions in cm
12.051
Fuel assembly pitch 09-GA50001-121-1
10

11. Benchmark Model – Absorber Assemblies
Sodium coolant
Driveline Region SS316 outer duct
SS316 inner duct
Sodium coolant
with homogenized
SS316 wire wrap
79.5528
Absorber pin (61)
OD 1.20396
SS316 ducts
Absorber pin
Above Poison Region
46.1772
298.45
Absorber Pin Lattice Region
91.44
Pitch
1.26492
Below Poison Region
19.35226
0.11176
Lower Shield Region 61.92774
0.3048 Dimensions in cm
09-GA50001-120-4
Detail of Absorber 10.20064
Pin Lattice Region
10.42416
12.051 11.0109
Drawing not to scale 11.6205
Dimensions in cm 12.051
09-GA50001-121-2 Fuel assembly pitch
11

12. Benchmark Model –Configuration
Outer radial shield
Sodium coolant
Inner radial shield
Radial reflectors in Row 8 and 9
Radial reflectors in Row 7
FS
8 Driver fuel assembly in the
Outer Enrichment Zone
9
230 Driver fuel assembly in the
S 3 Inner Enrichment Zone
7 S S
T In reactor thimble
2 T
6 F S FS
FS Fixed shim control rods
S 1 F
V S 4 S In-core shim assemblies
5
FS # Secondary control rods
# Primary control rods
F Fueled open test assembly
V Vibration open test assembly
Dimensions in cm
09-GA50001-122-1 12

13. Fully-Loaded Critical
Safety Rods Control Rod Control Rods Fixed Shim
1, 2, and 3 4 5, 6, 7, 8 and 9 Control Rods
(fully withdrawn) 3 Total
(fully inserted)
34.29
43.4368 43.9928
79.5528
46.1772 46.1772
91.44
46.1772
19.35226 91.44 91.44
91.44
91.44 19.35226 19.35226
36.116 35.56
19.35226
MCNP5
61.92774 61.92774 61.92774 61.92774
ENDF/B-VII.0
Lower Shield Region Below Poison Region Above Poison Region Drawing not to scale
Dimensions in cm
T = 480 K
Withdrawn Absorber Region Absorber Pin Lattice Region Driveline Region 09-GA50001-121-8
13

14. Neutron Spectra
• Near core center
– In Reactor Thimble (IRT)
• Two measurements
– Core midplane
– 80 cm below midplane
• Used Gaussian Energy
Broadening with tallies in
MCNP to simulate detectors
• Good agreement at core
midplane
• Homogenization of lower
assemblies believed to
impact below-core
measurements
14

15. Neutron Spectrum at Core Midplane
10
9
8
Relative Flux Per Unit Lethargy
7
6
5
4
3
2
1
0
1 10 100 1000 10000
Energy (keV)
Experimental Spectrum Calculated Spectrum
15

16. Neutron Spectrum 80-cm Below Core Midplane
10
9
8
Relative Flux Per Unit Lethargy
7
6
5
4
3
2
1
0
1 10 100 1000 10000
Energy (keV)
Experimental Spectrum Calculated Spectrum
16

17. Rod Worth Measurements
Individual rod
worths calculated
low by ~2-6% but
within 1σ
17

18. Rod Worth Measurements
• Differential rod
worths calculated
low by ~3-7%
(<2σ)
• Good agreement
for SDM and ER
measurements
18

19. Isothermal Temperature Coefficient
• Evaluated with MCNP5 and
ENDF/B-VII.0 cross section
data at 455 and 505 K ( 25 K
from 480 K)
• Model adjustments for
temperature, coolant
density, and cross sections
• Correlation between core
assembly pitch and
temperature unknown and
was estimated
• Calculated 8.7% lower than
the benchmark value
19

20. Low-Energy Electron and Gamma Spectra
• Near Core Center
– In Reactor Thimble (IRT)
– Core Midplane
• Measurement
uncertainty of 10%
assumed for SiLi
detectors at -30ºF
• Bad correlation for
electron spectrum and
good correlation for
gamma spectrum
(calculated ~37% high)
• Homogenization of IRT
believed to impact results
20

21. Low-Energy Electron Spectrum
10
1
Normalized Flux
0.1
0.01
0.001
0.0001
0 1 2 3 4 5 6
Electron Energy (MeV)
Experimental Spectrum Calculated Spectrum
21

22. Low-Energy Gamma Spectrum
10
1
Normalized Flux
0.1
0.01
0.001
0.0001
0 1 2 3 4 5 6
Photon Energy (MeV)
Experimental Spectrum Calculated Spectrum
22

23. Future Work
• Ongoing effort at PNNL to gather existing FFTF
resources into a database for DOE researchers
• Development of a fully heterogeneous benchmark
model of the FFTF
• Evaluation of Reactor Characterization Program
measurements
–Low power measurement of fission rates and spectra
–Passive sensor irradiation in simulated assemblies
–Eight full-power day irradiation of passive sensors
• Assessment of experimental data from 10 years of
operation
23

24. Conclusions
• Evaluation of the Initial Isothermal Physics Tests in
the FFTF has been completed
–Approved benchmark included in the 2010 edition (in
press) of the IRPhEP Handbook
• Good agreement for most reactor physics
measurements
–Homogenization effects believed to impact electron and
below-core neutron spectra calculations
• Future tasks have been identified for further
benchmark evaluation of FFTF experimental data
24

25. Acknowledgments
• David Wootan – PNNL
• Rich Lell, Dick McKnight, and Jim Morman –
ANL
• Sam Bays, Blair Briggs, Dave Nigg, and Chris
White – INL
25

26. Questions?
26