[PPT] - Joint ICTP-IAEA Workshop on the Evaluation of Nuclear Joint PowerPoint Presentation

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Joint ICTP-IAEA Workshop on the Evaluation of Nuclear Reaction Data for Applications 2-13 Oct 2017, ICTP, Trieste, Italy Joint ICTP-IAEA Workshop on the Evaluation of Nuclear Reaction Data for Applications 2-13 Oct 2017, ICTP, Trieste, Italy

PCA benchmark analysis using the ADVANTG3.0.1/MCNP6.1.1b codes PCA benchmark analysis using the ADVANTG3.0.1/MCNP6.1.1b codes

Mario Matijević University of Zagreb, Faculty of Electrical Engineering and Computing, Department of Applied Physics Unska 3, 10000 Zagreb, Croatia Mario Matijević University of Zagreb, Faculty of Electrical Engineering and Computing, Department of Applied Physics Unska 3, 10000 Zagreb, Croatia

mario.matijevic@fer.hr

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Contents: Contents:

1. PCA benchmark definition 2. PCA benchmark description 3. ADVANTG3.0.1/ MCNP6.1.1b codes 4. PCA response functions 5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b 6. Summary and conclusions

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Pool Critical Assembly Pressure Vessel (PCA) benchmark

Pool Critical Assembly Pressure Vessel (PCA) benchmark

– – A well known benchmark from the SINBAD database A well known benchmark from the SINBAD database – – Based on the PCA facility experiments at the ORNL Based on the PCA facility experiments at the ORNL – – A small pool A small pool-

type highly enriched experimental reactor

type highly enriched experimental reactor – – Measured and calculated (DORT) equivalent fission fluxes Measured and calculated (DORT) equivalent fission fluxes – – Validation of transport theory codes and XS libraries Validation of transport theory codes and XS libraries – – Prediction of the in Prediction of the in-

vessel neutron flux gradients

vessel neutron flux gradients

Scope of the PCA benchmark

Scope of the PCA benchmark

– – Validation of the methodology to predict the reaction rates in e Validation of the methodology to predict the reaction rates in ex x-

core region

core region – – Qualification of the pressure vessel fluence calculation methodo Qualification of the pressure vessel fluence calculation methodology logy – – Simulation of the neutron flux gradient inside carbon steel (RPV Simulation of the neutron flux gradient inside carbon steel (RPV) ) – – RPV surveillance programs of existing USA RPV surveillance programs of existing USA NPPs NPPs (RPV damage) (RPV damage) – – Measured RR inside RPV (A4, A5, A6) and water gap in Measured RR inside RPV (A4, A5, A6) and water gap in-

front (A3, A2)

front (A3, A2) – – Well defined neutron source, materials, and simple geometry Well defined neutron source, materials, and simple geometry – – DORT libraries in the PCA benchmark: BUGLE DORT libraries in the PCA benchmark: BUGLE-

93, SAILOR

93, SAILOR-

95, BUGLE

95, BUGLE-

96

96 – – Computational requirements by the U.S. NRC Regulatory Guide 1.19 Computational requirements by the U.S. NRC Regulatory Guide 1.190

1. PCA benchmark definition
1. PCA benchmark definition

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2. PCA benchmark description
2. PCA benchmark description

PCA pressure vessel wall benchmark facility MC model of PCA benchmark facility (water removed)

PCA benchmark “12/13” configuration

– – PCA core with components mock up the core PCA core with components mock up the core-

to

to-

cavity region in

cavity region in PWRs PWRs – – Al plate, Thermal shield (TS), pressure vessel simulator (PVS), Al plate, Thermal shield (TS), pressure vessel simulator (PVS), void box (VB) void box (VB) – – Water gap between Al plate and TS: 12 cm Water gap between Al plate and TS: 12 cm – – Water gap between TS and RPVS: 13 cm Water gap between TS and RPVS: 13 cm – – PCA facility is immersed in a large pool of water (coolant and m PCA facility is immersed in a large pool of water (coolant and moderator)

derator)

– – PCA core PCA core has has 25 material test reactor (MTR) plate 25 material test reactor (MTR) plate-

type elements (e=93%)

type elements (e=93%)

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2. PCA benchmark description
2. PCA benchmark description

Cross sectional view through the control element PCA standard MTR fuel element (water included) PCA control rod (water included) 5

PCA benchmark “12/13” configuration

Lead (11.34 g/cc) Boron carbide (1.6 g/cc) Aluminum Al-U alloy (fuel plate)

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2. PCA benchmark description
2. PCA benchmark description

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PCA benchmark “12/13” configuration

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2. PCA benchmark description
2. PCA benchmark description

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PCA benchmark “12/13” configuration

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2. PCA benchmark description
2. PCA benchmark description

reaction rates

eq i

φ σ =

Equivalent fission fluxes at detector locations:

Experimental measurements: A1 – A7

( ) ( ) ( ) ( ) ( )

i E eq i E E

E E dE E E dE E dE σ φ φ σ ϕ ϕ = ∫

∫ ∫

Maxwell 1/E slowing fission

8

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2. PCA benchmark description
2. PCA benchmark description

9

Reactions of interest (

Reactions of interest (shielding calculations shielding calculations) )

– –

237 237Np(n,f)

Np(n,f)137

137Cs,

Cs, 238

238U(n,f)

U(n,f)137

137Cs,

Cs, 103

103Rh(n,n')

Rh(n,n')130m

130mRh,

Rh, 115

115In(n,n')

In(n,n')115m

115mIn,

In, 58

58Ni(n,p)

Ni(n,p)58

58Co,

Co,

27 27Al(n,

Al(n,α α) )24

24Na

Na – – Results are given per unit PCA core neutron source (normalized) Results are given per unit PCA core neutron source (normalized) – – Calculated Calculated-

to

to-

measured (C/M) ratios of equivalent

measured (C/M) ratios of equivalent 235

235U fission fluxes

U fission fluxes – – Measurements Measurements in in core midplane (z=0 and y=0) core midplane (z=0 and y=0) at at locations A1 to A7 locations A1 to A7 – – Experimental access tubes: steel in PV and Plexiglas in water lo Experimental access tubes: steel in PV and Plexiglas in water locations cations – – Minimization of the perturbations of the neutron field Minimization of the perturbations of the neutron field – – DORT reaction rates (/atom/s) are given for dosimeters A1 DORT reaction rates (/atom/s) are given for dosimeters A1-

A8

A8

Critical configuration (

Critical configuration (eigenvalue calculations eigenvalue calculations) )

– – Fully inserted control rod reaches bottom of the fuel Fully inserted control rod reaches bottom of the fuel – – Withdrawn length is measured from the bottom of the fuel Withdrawn length is measured from the bottom of the fuel – – Safety rods (S1, S2 and S3) critical positions: 48.26 cm Safety rods (S1, S2 and S3) critical positions: 48.26 cm – – Regulating rod (RR) position: 38.43 cm Regulating rod (RR) position: 38.43 cm – – Total critical mass of Total critical mass of 235

235U: 3336.01 g

U: 3336.01 g

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3. ADVANTG3.0.1/ MCNP6.1.1B codes
3. ADVANTG3.0.1/ MCNP6.1.1B codes
ADVANTG3.0.1 code (ORNL)

ADVANTG3.0.1 code (ORNL)

– – Automated mesh Automated mesh-

based tool for generating VR parameters for MCNP code

based tool for generating VR parameters for MCNP code – – Approximate 3 Approximate 3-

D multigroup SN forward/adjoint transport solutions

D multigroup SN forward/adjoint transport solutions – – Denovo SN solver developed at ORNL with CADIS/FW Denovo SN solver developed at ORNL with CADIS/FW-

CADIS formalism

CADIS formalism – – VR parameters: space VR parameters: space-

energy weight

energy weight-

windows (WW) and biased source distributions

windows (WW) and biased source distributions (SB) cards for (SB) cards for the the MCNP input MCNP input

MCNP6.1.1b code (LLNL)

MCNP6.1.1b code (LLNL)

– – general general-

purpose Monte Carlo N

purpose Monte Carlo N-

Particle code with arbitrary 3D geometry

Particle code with arbitrary 3D geometry – – neutron, photon, electron, or coupled neutron, photon, electron, or coupled n/p/e n/p/e transport transport – – XS libraries: continuous, discrete, multigroup, XS libraries: continuous, discrete, multigroup, S(a,b S(a,b) law, dosimetry, ) law, dosimetry,… … – – powerful general source, rich collection of VR techniques, flexi powerful general source, rich collection of VR techniques, flexible tallies ble tallies – – Pointwise XS data with MAKXSF for XS libraries with Doppler broa Pointwise XS data with MAKXSF for XS libraries with Doppler broadening dening

PCA model with MCNP in z=0 cm plane PCA model with MCNP in y=30.84 cm plane 10

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11

MCNP model of the ¼ cask ADVANTG SN mesh for VR parameters Adjoint function (1.8-2.4) MeV point detector = adjoint source

forward flux adjoint flux detector source source detector forward transport: adjoint transport: source detector

†

1 ˆ( , ) ( , ) ( , ) q r E q r E r E R φ =

†

( , ) ( , ) R w r E r E φ =

†

( , ) ( , ) ˆ( , ) ( , ) q r E R w r E q r E r E φ = =

†

( , ) ( , ) ( , ) ( , )

d d

r E q r E r E r E σ σ φ =

∫

Target

weight Initial weight Weighted adjoint source Biased source CADIS FW- CADIS

Hybrid shielding

†( ,

) ( , )

d

q r E r E σ =

Adjoint

source

3. ADVANTG3.0.1/ MCNP6.1.1B codes
3. ADVANTG3.0.1/ MCNP6.1.1B codes

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4. PCA response functions
4. PCA response functions
Cross section

Cross sections s from the IAEA NDS service from the IAEA NDS service

– – ENDF/B retrieval service ( ENDF/B retrieval service (www www-

nds.iaea.org

nds.iaea.org) ) – – IRDF IRDF-

2002 and IRDFF

2002 and IRDFF-

2014 dosimetry libraries

2014 dosimetry libraries

Selection of XS library and reaction type XS reaction explorer XS reaction data in the ENDF-6 format XS reaction plotting and extracting for MCNP Cross sections with reconstructed resonances and applied Doppler broadening at the temperature 293 K IRDFF-1.05 library Al-27(n,a)Na-24

12

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4. PCA response functions
4. PCA response functions
Cross section

Cross sections s from the IAEA NDS service from the IAEA NDS service

– – IRDFF IRDFF-

2014 and IRDF

2014 and IRDF-

2002 dosimetry library

2002 dosimetry library – – Neutron excitation ( Neutron excitation (n,n n,n’ ’) of the first isomeric state ) of the first isomeric state Rh Rh (56.12 min ) and In (4.486 h) (56.12 min ) and In (4.486 h) – – Metastable Metastable isomers isomers 103m

103mRh and

Rh and 115m

115mIn with reactions MF/MT = 3/51

In with reactions MF/MT = 3/51

[1] Joel A. Kulesza & Roger L. Martz (2017) Evaluation of the Pool Critical Assembly Benchmark with Explicitly Modeled Geometry Using MCNP6, Nuclear Technology, 197:3, 284-295, DOI: 10.1080/00295450.2016.1273711 [1]

13 DE/DF cards define the response spectrum for point detector tallies

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5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b

14

MCNP6.1.1b eigenvalue results:

keff = (0.99924 ± 0.00100) c c ---------

criticality source

criticality source -----------

kcode

kcode 2000 1.0 50 350 1e4 2000 1.0 50 350 1e4 ksrc ksrc 20.22 31.46 0 $ starting point 20.22 31.46 0 $ starting point c c ----------------------------------------------

keff cycle number

keff cycle number Entropy of source Average keff

keff neutron flux in z=0 cm plane keff neutron flux RE in z=0 cm plane

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5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b

15

ADVANTG3.0.1 parameters:

ADVANTG3.0.1 parameters:

– – FW FW-

CADIS method for one reaction on all detectors (A1

CADIS method for one reaction on all detectors (A1-

A8)

A8) – – S4/P1, SC spatial differencing, S4/P1, SC spatial differencing, eps eps=1e =1e-

3, 1.6e6 mesh cells

3, 1.6e6 mesh cells – – ANISN ANISN-

format coupled

format coupled n n-

g

g multigroup library BPLUS (updated BUGLE multigroup library BPLUS (updated BUGLE-

96)

96) – – BUGLE BUGLE-

96 library: LWR shielding, PV dosimetry, VITAMIN

96 library: LWR shielding, PV dosimetry, VITAMIN-

B6 collapsing

B6 collapsing – – 5 weighting spectra of 1 5 weighting spectra of 1-

D model of a reactor cavity and

D model of a reactor cavity and bioshield bioshield – – BPLUS library: 47n/20g, 393 isotopes, ENDF/B BPLUS library: 47n/20g, 393 isotopes, ENDF/B-

VII.0 evaluation

VII.0 evaluation – – Mixed 11 pure materials into 26857 macro materials Mixed 11 pure materials into 26857 macro materials – – Memory requirements (GB): 6 Memory requirements (GB): 6-

9 FW Denovo, 10

9 FW Denovo, 10-

14 ADJ Denovo

14 ADJ Denovo

MCNP6.1.1b parameters:

MCNP6.1.1b parameters:

– – Continuous XS for neutron transport with Doppler correction and Continuous XS for neutron transport with Doppler correction and S(a,b S(a,b) law ) law – – MCNP6.1.1b with distributed (PVM) calculation on Quad Core CPU MCNP6.1.1b with distributed (PVM) calculation on Quad Core CPU – – Using point detectors inside small void spheres at centers of A1 Using point detectors inside small void spheres at centers of A1-

A8

A8 – – Mesh tally for capturing the global MC neutron transport Mesh tally for capturing the global MC neutron transport – – card card “ “ctme ctme 720 720” ” $ cumulative time in min $ cumulative time in min – – About 2e6 neutron histories per MC simulation About 2e6 neutron histories per MC simulation – – Point detectors have Point detectors have RE RE on average 1

n average 1-
2 %

2 %

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Integrated forward flux in z=0 cm Integrated adjoint flux in z=0 cm WW first group in z=0 cm

5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
ADVANTG3.0.1

ADVANTG3.0.1 FW FW-

CADIS

CADIS results results for for 27

27Al(n,a)

Al(n,a)24

24Ni

Ni

Integrated adjoint flux with contours (A8) (A1) (A8) (A1) (core) (TS) (RPV) (VB)

ψ† w=R/ψ† ψ ψ†

16

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5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b

MC neutron flux solution for Al-27(n,a)Ni-24 in z=0 cm plane MC neutron flux RE on 1 sigma level for Al-27(n,a)Ni-24 in z=0 cm plane

Location

237Np(n,f) 238U(n,f) 27Al(n,a) 58Ni(n,p) 115In(n,n') 103Rh(n,n')

MCNP Avg ± sig DORT Avg ± sig A1 0.862

0.801

0.887 0.902 0.945 0.88 ± 0.02 0.91 ± 0.02 A2

0.852

0.938 0.986

0.93 ± 0.04

0.92 ± 0.01 A3 0.908

0.765

0.835 0.877

0.85 ± 0.03

0.96 ± 0.02 A4 0.869 0.882 0.874 0.898 0.935 0.905 0.90 ± 0.01 0.94 ± 0.03 A5 0.914 0.880 0.929 0.919 0.935 0.877 0.91 ± 0.01 0.92 ± 0.03 A6 0.878 0.884 0.968 0.957 0.969 0.881 0.92 ± 0.02 0.91 ± 0.04 A7 0.939

0.94 ± 0.00

0.89 ± 0.00 A8

MCNP6.1.1b results for all reactions

MCNP6.1.1b results for all reactions

Equivalent fission fluxes C/M ratios (“-” experimental result not provided in the PCA benchmark) RPV 17

[Lower quality results since ADVANTG was used with default eps=1e-3]

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5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
Isotopes of Fe and self

Isotopes of Fe and self-

shielding

shielding

– – Neutron spectrum Neutron spectrum “ “softening softening” ” inside the RPV simulator inside the RPV simulator – – Neutrons > 1 MeV are shifted to resonance region of Fe => self Neutrons > 1 MeV are shifted to resonance region of Fe => self-

shielding!

shielding! – – Important reaction becomes inelastic neutron scattering ( Important reaction becomes inelastic neutron scattering (n,n n,n’ ’) )

Total cross section (MT=1) of iron isotopes Inelastic scattering (MT=3) of iron isotopes 18

( )

2 2 2 2 , 2 2

( ) 2 sin 2( )sin 2 4 (2 1)sin ( ) 1/ 4

n t l l l l

N g E E E N l E E π σ δ δ π δ Γ = Γ − Γ + − + + − + Γ D D

Hard-sphere nuclear model

( / ) tan ( / )

l l l

J R N R δ = D D

l=0 (s-wave) neutrons l=1 (p-wave) neutrons

/ R δ = D

1 1

tan ( / ) R δ δ

−

= − D

(LWR) (FBR)

2 ,0 2 2

4( ) ( ) 1 4( )

t pot

E E E R E E E E σ σ σ − Γ ⎡ ⎤ = + + ⎢ ⎥ − + Γ Γ ⎣ ⎦ D

1/v interpolation Breit- Wigner Prob. tables

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5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b

0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 10 20 30 40 50 60 70 80 90 100 Detector position (cm) FOM factor 237Np(n,f) 238U(n,f) 27Al(n,a) 58Ni(n,p) 115In(n,n') 103Rh(n,n')

MCNP FOM factors for all reactions

MCNP FOM factors for all reactions

Computational model metric addressing memory and CPU time Figure-of-merit (FOM) factor: CPU time necessary to reach a given level RE FOM = 1/(RE2xT), where RE is MC rel.error and T is CPU time in min

FOM factors for different detector locations 19

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5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
ADVANTG3.0.1

ADVANTG3.0.1 CADIS CADIS optimization

ptimization of
f A3

A3 detector detector

MC neutron flux solution for Al-27(n,a)Ni-24 in z=0 cm plane MC neutron flux RE on 1 sigma level for Al-27(n,a)Ni-24 in z=0 cm plane Integrated adjoint flux in z=0 cm WW first group in z=0 cm

ψ† w=R/ψ†

20 (A3) (A3)

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5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
5. PCA results using ADVANTG3.0.1/ MCNP6.1.1b
ADVANTG

ADVANTG3.0.1 3.0.1 FW FW-

CADIS global flux solution

CADIS global flux solution

Integrated contributon flux in z=0 cm Integrated forward and adjoint flux in z=0 cm MC neutron flux global solution in z=0 cm plane MC neutron flux RE in z=0 cm plane (RPV) (TS)

ψ ψ†

more histories needed! average RE ≈ 30%

21

local SN mesh refinement?

ψc =ψψ†

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Commercial hybrid shielding codes today:

Commercial hybrid shielding codes today:

– – SCALE6.2 (CADIS SCALE6.2 (CADIS/ /FW FW-

CADIS in MAVRIC)

CADIS in MAVRIC) – – ADVANTG3.0.1 + MCNP5v1.6 (or MCNP6.1.1b) ADVANTG3.0.1 + MCNP5v1.6 (or MCNP6.1.1b) – – A A3

3MCNP (TORT + MCNP5v1.6)

MCNP (TORT + MCNP5v1.6) – – ATTILA + MCNP6.1.1B ATTILA + MCNP6.1.1B

6. Summary and conclusions
6. Summary and conclusions
Hybrid shielding:

Hybrid shielding:

– – FW FW-

CADIS for distributed detectors and global flux solution

CADIS for distributed detectors and global flux solution – – CADIS for local answers, would require 6 reactions x 8 locations CADIS for local answers, would require 6 reactions x 8 locations = 48 runs! = 48 runs! – – SN based VR technique removes burden of manual VR preparation SN based VR technique removes burden of manual VR preparation – – Computational trade Computational trade-

off between SN and MC simulations
ff between SN and MC simulations
PCA benchmark:

PCA benchmark:

– – Good agreement between calculated (C) and measured (M) results Good agreement between calculated (C) and measured (M) results – – Obtained results in accordance with the Regulatory Guide 1.190 Obtained results in accordance with the Regulatory Guide 1.190 – – Sensitivity of BPLUS shielding library on the iron XS (self Sensitivity of BPLUS shielding library on the iron XS (self-

shielding)

shielding) – – Differences due to XS libraries, weighting spectrum, core modeli Differences due to XS libraries, weighting spectrum, core modeling, ng,… … – – More neutron groups are preferred due to spectral effects More neutron groups are preferred due to spectral effects

22

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References: References:

1. 1. “ “Shielding Integral Benchmark Archive and Database Shielding Integral Benchmark Archive and Database, ,” ” SINBAD SINBAD Reactior Reactior Shielding Shielding Benchmark Experiments, OECD Benchmark Experiments, OECD-

NEA, 2011.

NEA, 2011. 2. 2. I.

I. Remec

Remec and F.B. and F.B. Kam Kam, , “ “Pool Critical Assembly Pressure Vessel Facility Benchmark Pool Critical Assembly Pressure Vessel Facility Benchmark, ,” ” NUREG/CR NUREG/CR-

6454 (ORNL/TM

6454 (ORNL/TM-

13205), August 1997.

13205), August 1997. 3. 3. International Atomic Energy Agency, Vienna, Austria, Internation International Atomic Energy Agency, Vienna, Austria, International Reactor Dosimetry al Reactor Dosimetry File 2002 (IRDF File 2002 (IRDF-

2002), 2006.

2002), 2006. 4. 4. U.S.NRC, U.S.NRC, Calculational and Dosimetry Methods for Determining Pressure Ves Calculational and Dosimetry Methods for Determining Pressure Vessel sel Neutron Fluence Neutron Fluence, Regulatory Guide 1.190, 2001. , Regulatory Guide 1.190, 2001. 5. 5. J.C. Wagner and A. J.C. Wagner and A. Haghighat Haghighat, , “ “Automated variance reduction of Monte Carlo Automated variance reduction of Monte Carlo shielding calculations using the discrete ordinates adjoint func shielding calculations using the discrete ordinates adjoint function, tion,” ” Nuclear Science Nuclear Science and Engineering and Engineering, vol. 128, no. 2, pp. 186 , vol. 128, no. 2, pp. 186-

208, 1998.

208, 1998. 6. 6. G.I. Bell and S. G.I. Bell and S. Glasstone Glasstone, , Nuclear Reactor Theory Nuclear Reactor Theory, Van , Van Nostrand Nostrand Reinhold Company, Reinhold Company, Litton Educational Publishing, 1970. Litton Educational Publishing, 1970. 7. 7. M.B Chadwick, P. M.B Chadwick, P. Oblo Oblož žinsky insky, M. Herman et al., , M. Herman et al., “ “ENDF/B ENDF/B-

VII.0: next generation

VII.0: next generation evaluated nuclear data library for nuclear science and technolog evaluated nuclear data library for nuclear science and technology, y,” ” Nuclear Data Nuclear Data Sheets Sheets, vol. 107, no. 12, pp. 2931 , vol. 107, no. 12, pp. 2931-

3060, 2006.

3060, 2006. 8. 8. A.

A. Trkov

Trkov, M. Herman, and D.A. Brown, , M. Herman, and D.A. Brown, “ “ENDF ENDF-

6 Formats Manual

6 Formats Manual” ”, BNL, 2011. , BNL, 2011. 9. 9. T.

T. Goorley

Goorley, "MCNP6.1.1 , "MCNP6.1.1-

Beta Release Notes", LA

Beta Release Notes", LA-

UR

UR-

14

14-

24680 (2014).

24680 (2014). 10. 10. S.W. Mosher, S.R. Johnson, A.M. S.W. Mosher, S.R. Johnson, A.M. Bevil Bevil et al., et al., “ “ADVANTG ADVANTG – – An Automated Variance An Automated Variance Reduction Parameter Generator Reduction Parameter Generator” ”, ORNL/TM , ORNL/TM-

2013/416, Rev. 1, August 2015.

2013/416, Rev. 1, August 2015.

23