SCALE 6.2.4 Validation: Reactor Physics Applications

ORNL is managed by UT-Battelle LLC for the US Department of Energy

SCALE 6.2.4 Validation: Reactor Physics ApplicationsGermina IlasJoseph Burns Briana HiscoxUgur Mertyurek

Oak Ridge National Laboratory

SCALE Users’ Group Workshop, August 4-6, 2021

Oak Ridge National Laboratory

22 SCALE Users Group Workshop, August 4-6, 2021

Background

• A four-volume report, pending publication in Summer 2021, documents SCALE 6.2.4/ENDF-B/VII.1 validation:1. Validation overview2. Nuclear criticality safety applications3. Reactor physics applications [nuclide inventory, decay heat, full core

analysis] 4. Radiation shielding applications

• This presents selected results from Vol. 3.


Objective

• Validate SCALE 6.2.4 capabilities and associated ENDF/B-VII.1 nuclear data libraries for predicting source terms in LWR spent nuclear fuel and to support full-core LWR and non-LWR analyses.

• Compare SCALE 6.2.4/ ENDF/B-VII.1 and SCALE 6.1/ ENDF/B-VII.0 results.


Relevance

• Quantifying the bias in predicting key metrics of interest to reactor physics applications (e.g., eigenvalue, nuclide inventory, decay heat) is essential for understanding applicability of capabilities and associated nuclear data in SCALE 6.2.4 to lattice and full-core physics safety-related analyses.

• Impact goes well beyond reactor physics and back-end of fuel cycle, to support modeling challenges in other areas where simulating nuclide transmutations and decay in nuclear fuel during and post irradiation is of great importance (e.g., isotope production, national security and nonproliferation applications).

• Bias in nuclide inventory can serve as indirect means for assessing computational bias in those integral quantities important to safety that are mainly driven by fuel composition at discharge from reactor (e.g., decay heat, used fuel reactivity).


Experimental Data Used as Validation Basis

• Nuclide inventories– PWR: radiochemical assay data for 92 fuel sample measurements, 40 nuclides – BWR: radiochemical assay data for 76 fuel sample measurements, 35 nuclides

• Decay heat– Full-assembly decay heat experiments for 91 PWR and 145 BWR assemblies– Fissile materials irradiations to determine energy release from fission of U-233, U-235, U-

238, Pu-239, Pu-241, and Th-232

• Full-core analysis – LWR: Watts Bar Unit 1 startup tests, eigenvalue and boron worth– non-LWR: HTR-10 and HTTR, eigenvalues, benchmarks in IRPhE Handbook


Experimental data: Nuclide inventory for PWR fuel burnup and enrichment space covered

1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.00

10

20

30

40

50

60

70

80 Calvert CliffsGKNGosgenHB RobinsonObrigheimTMI1TakahamaTrino VercellesseTurkey Point

Bur

nup

(GW

d/M

TU)

Enrichment (wt % U-235)

1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.50

5

10

15

20

25

30

35

40

45

50

Num

ber

of s

ampl

es


0 10 20 30 40 50 60 70 800

5

10

15

20

25

Num

ber

of s

ampl

es

Burnup (GWd/MTU)


Experimental data: Nuclide inventory for BWR fuelburnup, enrichment, and average void fraction space covered

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.00

10

20

30

40

50

60

70

80Fukushima Daini 1, 9x9-9Fukushima Daini 2, 8x8-2Fukushima Daini 2, 8x8-4Forsmark 3, SVEA-100Forsmark 3, GE14 Leibstadt, SVEA-96Limerick, GE11Dodewaard, 6x6

Bur

nup

(GW

d/M

TU)


0 10 20 30 40 50 60 700

5

10

15

20

25

Num

ber o

f sam

ples

Burnup (GWd/MTU)

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.00

5

10

15

20

25

30

35

Num

ber o

f sam

ples


0 10 20 30 40 50 60 70 800

5

10

15

20

25

Num

ber o

f sam

ples

Average void (%)


Experimental data: nuclide inventory for PWR fuelmeasured nuclides and uncertainties

Actinide Number ofsamples Application Exp. uncert. (%)

234U 55 BC, WM 0.5 – 5.2235U 92 BC, WM 0.4 – 3.8236U 77 BC, WM 0.4 – 2.4238U 92 BC, WM 0.1 – 4.3

238Pu 77 BC, RS, WM 0.3 – 14.3239Pu 92 BC, RS, WM 0.3 – 2.4240Pu 92 BC, RS, WM 0.3 – 2.7241Pu 92 BC, WM 0.3 – 2.5242Pu 91 BC, WM 0.3 – 5.3237Np 36 BC, WM 1.9 – 10.0241Am 39 BC, RS, WM 1.8 – 20.0243Am 38 BC, WM 1.8 – 100.0244Cm 57 RS 0.9 – 28.0245Cm 24 BC, WM 2.0 – 10.1246Cm 14 RS 5.0 – 10.1

FissionProduct

Number ofsamples Application Exp. uncert. (%)

90Sr 15 RS, WM 1.5 – 8.099Tc 20 RS, WM 3.5 – 8.9

101Ru 7 BC 5.0 – 12.2106Ru 31 RS 3.0 – 12.2103Rh 8 BC 4.0 – 14.2109Ag 6 BC 5.0 – 9.1125Sb 18 RS 5.0 – 9.4133Cs 10 BC 1.0 – 2.5134Cs 59 RS 1.5 – 5.0135Cs 16 WM 1.5 – 14.0137Cs 73 BI, RS, WM 1.3 – 3.5143Nd 36 BC 0.1 – 5.1145Nd 36 BC 0.1 – 5.9148Nd 77 BI 0.1 – 6.7144Ce 32 RS 1.7 – 10.0147Sm 24 BC 0.1 – 10.6149Sm 20 BC 0.1 – 20.0150Sm 24 BC 0.1 – 4.2151Sm 24 BC 0.1 – 38.5152Sm 24 BC 0.1 – 3.2151Eu 12 BC 0.9 – 9.8153Eu 19 BC 0.9 – 5.6154Eu 44 RS 1.7 – 11.9155Eu 11 BC 3.2 – 16.1

155Gd 19 BC 1.4 – 6.7


Experimental data: Decay heat for PWR fuel assembliesburnup and cooling time space covered

15 20 25 30 35 40 45 50 550

10

20

30

40

50

N

umbe

r of m

easu

rem

ents

Burnup (GWd/MTU)

PWR Nmeasurements=91

0 5 10 15 20 25 300

5

10

15

20

25

30

35

Cooling time (years)

Num

ber o

f mea

sure

men

ts

PWR Nmeasurements=91


Experimental data: Decay heat for BWR fuel assembliesburnup and cooling time space covered

0 5 10 15 20 25 30 35 40 45 500

10

20

30

40

50

60

70

80

Burnup (GWd/MTU)

Num

ber

of m

easu

rem

ents

BWR Nmeasurements=145

0 5 10 15 20 25 300

10

20

30

40

50

60

70

80

Cooling time (years)

Nmeasurements=145BWR

Num

ber o

f mea

sure

men

ts


SCALE simulations for nuclide inventory

XS Processing(XSPROC)

2D Neutron Transport (NEWT)

Depletion(ORIGEN)

Time step loop

Input data252-gr ENDF/B-VII.1

TRITON

Output dataNuclide vectors

Sample irradiation and decay history modeling

Design and operation data

¼ assembly model for Gosgen GU3 sample

½ assembly model for Calvert Cliffs MKP109-CC sample


SCALE simulations for full-assembly decay heat


Results: Nuclide inventory in PWR fuel comparison calculation-experiment for actinides

SCALE 6.2.4/ENDF/B-VII.1Nuclide No. samples C/E average C/E stdev

234U 55 1.127 0.173235U 92 1.013 0.036236U 77 0.981 0.034238U 92 0.999 0.004

238Pu 77 0.959 0.074239Pu 92 1.019 0.034240Pu 92 1.005 0.035241Pu 92 0.978 0.047242Pu 91 0.959 0.064237Np 36 1.024 0.199241Am 39 1.043 0.192243Am 38 0.933 0.125244Cm 57 0.987 0.113245Cm 24 0.987 0.150246Cm 14 0.930 0.224

SCALE 6.1/ENDF/B-VII.0Nuclide No. samples C/E average C/E stdev

234U 55 1.124 0.176235U 92 1.012 0.035236U 77 0.981 0.035238U 92 0.999 0.004

238Pu 77 0.883 0.059239Pu 92 1.041 0.035240Pu 92 1.022 0.034241Pu 92 0.986 0.045242Pu 91 0.941 0.061237Np 36 1.039 0.195241Am 39 1.102 0.207243Am 38 1.029 0.140244Cm 57 0.956 0.111245Cm 24 0.985 0.156246Cm 14 0.956 0.255


Results: Nuclide inventory in PWR fuel comparison calculation-experiment for fission products


90Sr 15 0.991 0.06699Tc 20 1.164 0.158

101Ru 7 1.056 0.113106Ru 31 1.059 0.218103Rh 8 1.115 0.106109Ag 6 1.764 0.683125Sb 18 1.988 0.450133Cs 10 1.023 0.017134Cs 59 0.898 0.070135Cs 16 1.021 0.036137Cs 73 0.989 0.032143Nd 36 1.008 0.021145Nd 36 1.002 0.011148Nd 77 1.000 0.003144Ce 32 0.968 0.075


90Sr 15 0.991 0.06999Tc 20 1.152 0.154

101Ru 7 1.058 0.123106Ru 31 1.079 0.227103Rh 8 1.091 0.109109Ag 6 1.773 0.746125Sb 18 1.996 0.466133Cs 10 1.019 0.017134Cs 59 0.930 0.071135Cs 16 1.027 0.037137Cs 73 0.993 0.031143Nd 36 1.008 0.032145Nd 36 0.995 0.022148Nd 77 1.006 0.014144Ce 32 0.979 0.081


Results: Nuclide inventory in PWR fuel comparison calculation-experiment for fission products (cont.)


147Sm 24 1.009 0.032149Sm 20 1.013 0.061150Sm 24 1.009 0.028151Sm 24 0.972 0.042152Sm 24 1.010 0.035151Eu 12 0.886 0.190153Eu 19 0.965 0.030154Eu 44 1.066 0.107155Eu 11 0.977 0.077

155Gd 19 0.936 0.142


147Sm 24 1.016 0.034149Sm 20 1.019 0.062150Sm 24 1.008 0.032151Sm 24 0.979 0.044152Sm 24 1.016 0.037151Eu 12 0.893 0.198153Eu 19 0.991 0.031154Eu 44 1.042 0.104155Eu 11 0.956 0.077

155Gd 19 0.916 0.144


Results: Nuclide inventory in BWR fuel comparison calculation-experiment for actinides

SCALE 6.2.4/ENDF/B-VII.1Nuclide No. samples C/E average C/E stdev234U 76 1.039 0.128235U 76 1.027 0.111236U 76 1.017 0.047238U 76 0.999 0.003238Pu 76 1.063 0.206239Pu 76 0.974 0.085240Pu 76 0.999 0.088241Pu 76 0.941 0.114242Pu 76 0.985 0.168237Np 29 0.986 0.120241Am 62 0.991 0.167243Am 62 1.053 0.344242Cm 48 0.906 0.679243Cm 48 0.632 0.526244Cm 51 0.983 0.456


Results: Nuclide inventory in BWR fuel comparison calculation-experiment for fission products

SCALE 6.2.4/ENDF/B-VII.1Nuclide No. samples C/E average C/E stdev95Mo 23 1.020 0.07799Tc 16 1.262 0.156101Ru 14 1.058 0.134103Rh 15 1.066 0.090109Ag 15 1.336 0.386133Cs 16 0.970 0.072137Cs 42 0.968 0.061143Nd 50 1.039 0.038145Nd 50 1.022 0.030146Nd 50 1.001 0.025

SCALE 6.2.4/ENDF/B-VII.1Nuclide No. samples C/E average C/E stdev148Nd 75 1.005 0.027147Sm 35 0.991 0.079149Sm 32 0.921 0.120150Sm 34 1.033 0.067151Sm 35 0.974 0.116152Sm 35 1.055 0.065151Eu 15 0.880 0.208153Eu 25 1.037 0.034155Eu 25 1.009 0.161155Gd 25 1.055 0.110


Results: major actinide 235U in PWR fuel

0 10 20 30 40 50 60 70 80-20

-15

-10

-5

0

5

10

15

20

mean-2σ

mean+2σ

mean=1.3%

235U

Calvert CliffsGKNGosgenHB RobinsonObrigheimTMI1TakahamaTrino VercellesseTurkey Point

(C/E

-1) (

%)

burnup (GWd/MTU)

-12 -10 -8 -6 -4 -2 0 2 4 6 8 10 120

5

10

15

20

25

30

Num

ber

of s

ampl

es

C/E-1 (%)

235U mean = 1.3% σ = 3.6%


Results: major actinide 239Pu in PWR fuel

0 10 20 30 40 50 60 70 80-20

-15

-10

-5

0

5

10

15

20

(C/E

-1) (

%)

burnup (GWd/MTU)

239Pu

Calvert CliffsGKNGosgenHB RobinsonObrigheimTMI1TakahamaTrino VercellesseTurkey Point

mean=1.9%

mean+2σ

mean-2σ

-12 -10 -8 -6 -4 -2 0 2 4 6 8 10 120

5

10

15

20

25

mean = 1.9% σ = 3.4%

C/E-1 (%)

239Pu

Num

ber o

f sam

ples


Results: Decay heat in PWR and BWR fuel assemblies

0.94 0.95 0.96 0.97 0.98 0.99 1.00 1.01 1.02 1.03 1.04 1.05 1.060

5

10

15

20

25

30

35

N

umbe

r of m

easu

rem

ents

C/E

PWR mean = 1.006 σ = 0.016

0.70 0.75 0.80 0.85 0.90 0.95 1.00 1.05 1.10 1.15 1.20 1.25 1.300

10

20

30

40

50

Num

ber

of m

easu

rem

ents

C/E

BWR mean = 0.984 σ = 0.077


Results: Fissile material irradiationscomparison experiment-calculation

0.1 1 10 100 1000 10000 1000000.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

MeV

/fiss

ion

time (s)

ORIGEN YAYOY Lowell ORNL

235U Fission

0.1 1 10 100 1000 10000 100000

0.0

0.2

0.4

0.6

0.8

1.0

1.2

MeV

/fiss

ion

time (s)

ORIGEN YAYOY Lowell ORNL

239Pu Fission


Results: Full-core analysis for non-LWRsHTR-10 IRPhE benchmark

k-eff σ ∆ k-eff (pcm)

Benchmark value 1.0000 0.0037 referenceSCALE/KENO-VI MG 1.00378 0.00008 378 ± 370 SCALE/KENO-VI CE 1.00271 0.00010 271 ± 370


Results: Full-core analysis for non-LWRsHTTR IRPhE benchmark

k-eff σ ∆ k-eff (pcm)

Benchmark value 1.0025 -0.0060, +0.0071

reference

SCALE6.2.4/KENO-VI CE 1.00755 0.00010 505 (-61,+72)SCALE6.2.4/KENO-VI MG (annular DH)

1.00760 0.00008 510 (-60,+71)

SCALE6.2.4/KENO-VI MG (cylindrical DH)

1.00679 0.00007 429 (-60,+71)

Permanent reflectorgraphite

R3R

RRRR

24 R

2RR 3 3RRI 3 R

13 I

42

2 4RRR1

1 R1

RR3 1 1

3R2

2 C 2 R23 1 1 3

RR

R1

1 R1

RR4 2 2 4

R3 3 R1 3R3 R

R3 3 R

RR2

4 R2R

RRRI


Impacts: SCALE 6.2.4/ENDF-B/VII.1 vs SCALE 6.1/ENDF-B/VII.0 for PWR nuclide inventories

• Comparison calculation-experiment for SCALE 6.2.4/ENDF-B/VII.1 indicates a good agreement for important nuclides, considering the uncertainties associated with the experimental data.

• No significant change for U between SCALE 6.2.4/ENDF-B/VII.1 and SCALE 6.1/ENDF-B/VII.0.

• Significant improvement in Pu nuclide predictions is noted– 239Pu C/E improves on average by ~2% (σ=5%) – 238Pu C/E improves on average by ~8% (σ=9%)

• Changes in Pu nuclides lead to changes in calculation of higher actinides Am and Cm.

• Sources of the observed C/E differences include:– differences in evaluated nuclear data (ENDF/B-VII.1 vs. ENDF/B-VII.0)– improvements in cross-section processing (252-gr vs 238-gr)


What’s next?

• Further investigate impacts and sources of differences between SCALE 6.2.4/ENDF-B/VII.1 vs SCALE 6.1/ENDF-B/VII.0

• Continue validation for SCALE 6.3/ENDF/B-VIII.0 and improve experimental data basis

– Bias and uncertainty values must be reassessed to keep pace with changes in fuel characteristics for fuel currently used or planned for commercial reactors in the future. Modern fuels are characterized by higher burnups, higher enrichments, complex and heterogeneous assembly designs.

– Bias and uncertainty values need be updated to account for changes in computational capabilities and evaluated nuclear data that are used for simulations.


References

• G. Ilas and B. Hiscox, “Validation of SCALE 6.2.4 and ENDF/B-VII.1 Data Libraries for Nuclide Inventory Analysis in PWR Used Fuel”, Transactions of the American Nuclear Society, vol. 124, p.552-554 (2021). https://www.ans.org/pubs/transactions/article-49653/

• G. Ilas and J. Burns, “SCALE 6.2.4 Validation for Light Water Reactor Decay Heat Analysis”, Nuclear Technology (2021) (in press)

• G. Ilas, J. Burns, B. Hiscox, U. Mertyurek, “SCALE 6.2.4 Validation: Reactor Physics, ORNL/TM/2020-1500/V3 (2021) (in press)

https://www.ans.org/pubs/transactions/article-49653/


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SCALE 6.2.4 Validation: Reactor Physics Applications