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Burnup benchmarks


1) NPD. 19-rod Fuel Clusters

D2O-moderated uranium oxide lattices.

Analysis of Isotopic Composition as function of burnup

Laboratory: AECL Atomic Energy of Canada Limited (Canada)

Facility: NPD Nuclear Power Demonstration reactor

NPD is a Canadian demonstration PHWR (25 MW electrical power) shut
down on 1987. The moderator and coolant were heavy water. The fuel
was in the form of 19 natural UO2 rod CANDU type clusters.

Isotopic measurements of irradiated 19 element fuel bundles have been done
in 1971 at the CEA in France. The measurements included a series of eight
bundles with irradiation in the range 1000-10000 MWd/TU which were analysed
by mass-spectroscopy.

Pitch (cm) 26.035 (square)
Coolants D2O
Moderator D2O
Number of rods 19 (1/6/12)
Radius of rod centers (cm) 0.0/1.65/3.18755
Angular dist.from reference axis (radians) 0.0/0.0/0.2617994
Fuel material UO2-nat
Density of fuel material (g/cm2) 10.0704
Radius of fuel rods (cm) 0.71247
Sheath material Zry-2
Density of sheath material (g/cm3) 6.57
Internal radius of sheath (cm) 0.71505
Thickness of sheath (cm) 0.04191
Material of pressure tube Zry-2
Density of pressure Tube (g/cm3) 6.556
Internal radius of pressure tube (cm) 4.1402
Thickness of pressure tube (cm) 0.4318
Material of calandria tube 57S Al alloy
Density of calandria Tube (g/cm3) 2.68
Internal radius of calandria tube (cm) 5.08
Thickness of calandria tube (cm) 0.12827
Effective fuel Temperature (K) 778.2
Effective coolant Temperature (K) 531.7
Temperature of moderator (K) 311.0
Experimental buckling (1/cm2) 0.000173

References

[1] DURET, M.F. et.al., Plutonium Production in NPD. A Comparison Between
Experiment
and Calculation, AECL-3995, 1971.

[2] INTERNATIONAL ATOMIC ENERGY AGENCY, In-core fuel management
benchmarks for PHWRs, IAEA-TECDOC-887, Task 9.3,Pag.122-127, 1996.


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2) NEA Burnup Credit Criticality Benchmark

LWR-Burnup credit criticality benchmark. Isotopic composition prediction

Laboratory: Pacific Northwest Laboratory, Richland, Washington (USA)

Facility: ATM-104 Approved Testing Material, Combustion Engineering (CE) 14x14
assembly

This benchmark compares the computed nuclide inventories for a simple pin cell
calculation. The fuel and operating specifications are based on data given
in the references for the Combustion Engineering (CE) 14 x 14 assembly designated
as Approved Testing Material ATM-104, for a series of experiments designed to
characterize spent fuel for light water reactors. The chemical assay data measured
in these experiments are of particular value in validating the isotopic predictions used
in burnup credit.

Data

Cell radius (cm) 0.879346
Moderator H2O
Moderator Density (g/cm3) 0.7569
Moderator Temperature (K) 558
Fuel material UO2 (3% at. U235 enrichment)
Density of fuel material (g/cm3) 10.045
Effective fuel Temperature (K) 841
Radius of fuel rods (cm) 0.47815
Clad material Zry-2
Density of clad material (g/cm3) 6.55
Internal radius of clad (cm) 0.493
Thickness of clad (cm) 0.066
Clad Temperature (K) 620
Cycle 1 avg boron
concentration (ppm)
331

Operating history data for isotopic calculation

OPERATING CYCLE BURN
[days]
DOWN
[days]
BORON
[% cycle 1]
1 306.0 71.0 100.0
2 381.7 83.1 141.9
3 466.0 85.0 152.3
4 461.1 1870.0 148.8

BURN is the fuel irradiation time

DOWN is the downtime between cycles except for cycle 4 where it
includes the decay time from reactor to measurement (cooling time)

BORON is the cycle-average boron concentration as a percent of
the cycle 1 concentration

Specific power

OPERATING
CYCLE
Specific Power
[Kw/kgU]
  Sample 1 Sample 2 Sample 3
1 17.24 24.72 31.12
2 19.43 26.76 32.51
3 17.04 22.84 26.20
4 14.57 18.87 22.12
Cumulative
Burnup
[GWd/MTU]
27.35 37.12 44.34

Initial fuel number densities

Nuclide Number Density
atoms/(barn.cm)
U-234 6.15165E-06
U-235 6.89220E-04
U-236 3.16265E-06
U-238 2.17104E-02
C-12 9.13357E-06
N-14 1.04072E-05
O 4.48178E-02

Cycle 1 coolant number densities

Nuclide Number Density
atoms/(barn.cm)
H-1 5.06153E-02
O-16 2.53076E-02
B-10 2.75612E-06
B-11 1.11890E-05

REFERENCES


[1] NUCLEAR ENERGY AGENCY, Burnup Credit Criticality Benchmark, Isotopic

Composition Prediction, NEA/NSC/DOC (92)/10; NEA 1401/02, 1992.

[2] GUENTHER, R.J. et al, Characterization of Spent Fuel Approved Testing Material-

ATM-104, Pacific Northwest Laboratory report PNL-5109-104, Richland, Washington,
December 1991.

[3] BIERMAN, S.R., Spent Reactor fuel Benchmark. Composition Data for Code
Validation.Proceeding of the International Conference on Nuclear Criticality
Safety-ICNC91, Oxford. United Kingdom p.II-113, September 9-13, 1991.



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3) LWR-Pu recycling benchmarks

LWR-Pu recycling benchmarks

These numerical benchmarks compare the computed nuclide inventories for simple pin
cell calculations, for two cases:

A) highly degraded plutonium, and
B) normal recycled plutonium.


Reference results
Isotopic concentrations (mg/g fuel) on the fuel pin for:

U-234, U-235,U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241,Pu-242, Am-241,
Am-242m,Am-243, Cm-242, Cm-243,
Cm-244, Cm-245, Mo-95, Tc-99, Ru-101, Rh-103,
Pd-105, Pd-107,Pd-108, Ag-109, Xe-131, Xe-135, Cs-133, Cs-135, Nd-143, Nd-145, Pm-147,
Pm-148m,Sm-149, Sm-150, Sm-151, Sm-152, Eu-153, Eu-154, Eu-155,
for cases A and B, at 50 GWd/MTU burnup.


Calculated Parameters
Isotopic concentrations (mg/g fuel) on the fuel pin for:

U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241,Pu-242, Am-241,
Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Mo-95, Tc-99, Ru-101, Rh-103, Pd-105,
Pd-107, Pd-108, Ag-109, Xe-131, Xe-135, Cs-133, Cs-135, Nd-143, Nd-145, Pm-147,
Pm-148m, Sm-149, Sm-150, Sm-151, Sm-152, Eu-153,Eu-154, Eu-155
for cases A and B, at 50 GWd/MTU burnup.


Data

Pitch (cm)

1.3133 (square)

Moderator

H2O(+Bnat)

Moderator Temperature (K)

579

Fuel material

UO2-PuO2

Effective fuel Temperature (K)

933

Radius of fuel rods (cm)

0.4095

Clad material

Zr

Atomic density of clad material
(atoms/barn.cm)

0.043248

Radius of clad (cm)

0.4750

Clad Temperature (K)

579


Initial fuel number densities

Nuclide
Number Density
atoms/(barn.cm)
A
B
U-234
-
2.4626E-07
U-235
1.4456E-04
5.1515E-05
U-238
1.9939E-02
2.0295E-02
Pu-238
1.1467E-04
2.1800E-05
Pu-239
1.0285E-03
7.1155E-04
Pu-240
7.9657E-04
2.7623E-04
Pu-241
3.3997E-04
1.4591E-04
Pu-242
5.6388E-04
4.7643E-05
O
4.5851E-02
4.3100E-02

Coolant number densities

Nuclide Number Density
atoms/(barn.cm)
H-1 4.7716E-02
O-16 2.3858E-02
B-nat 1.98606E-05

References

[1] NUCLEAR ENERGY AGENCY, Package ID: NEA 1505/01


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