Author(s): Madison Sellers, Tim Goorley Los Alamos ... · Emily Corcoran, Davbid Kelley, RMC. MCNP6...

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11-05868

MCNP6 Delayed Neutron Emission Validation withExperimental Measurements

Madison Sellers, Tim GoorleyLos Alamos National Laboratory

Emily Corcoran, Davbid Kelley, RMC

MCNP6 D l d NMCNP6 Delayed Neutron Emission Validation with

Experimental Measurements

Presented by: Madison Sellers

Supervisors: Dr. Tim Goorley, LANL

Dr. Emily Corcoran, RMCDr. David Kelly, RMC

LA-UR 11-05868

October 11 2011

OverviewOverview

• Introduction, Background & Theory, g y

• Experimentationp

• MCNP6 Model

• Results & Discussion

• Conclusions

The Delayed Neutron Counting System

Introduction Experimental MCNP6 Model of DNC System Results & Discussion Conclusions

The Delayed Neutron Counting System

C d i • Constructed in 2010

• Validated for 235U analysis, aqueous samples

• Current experiments include 233U samples

• Expanding analysis to include 239Pu analysis and mixtures of two or more fi il i t fissile isotopes

• Winter 2011/2012 will be updated with new hardware to increase sensitivity and efficiencyy

Delayed Neutron Generation

Introduction Background & Theory Experimental Results & Discussion Conclusions & Recommendations

• Many delayed neutron precursors

Group t1/2 [s] λ [s-1] βi = νi/νd [%]

1 55 6 0 014267 0 0328 ± 0 0042 Fissile Isotope1 55.6 0.014267 0.0328 ± 0.0042

2 24.5 0.028292 0.1539 ± 0.0068

3 16.3 0.042524 0.091 ± 0.009 γ

ss e sotopePrompt n

Delayed n

4 5.21 0.133042 0.197 ± 0.023

5 2.37 0.292467 0.3308 ± 0.0066

6 1.04 0.666488 0.0906 ± 0.0046

γDelayed neutron Precursor

7 0.424 1.634781 0.0812 ± 0.0016

8 0.198 3.554600 0.0229 ± 0.0095

• Eight (or 6) groups, denoted by t1/2 and production ratio

1 55 6 0 014267 0 0328 ± 0 0042

1081 55.6 0.014267 0.0328 ± 0.0042

2 24.5 0.028292 0.1539 ± 0.0068

3 16.3 0.042524 0.091 ± 0.009

tro103

105 Group 1Group 2Group 3Group 4Group 5Group 6Group 7

4 5.21 0.133042 0.197 ± 0.023

5 2.37 0.292467 0.3308 ± 0.0066

6 1.04 0.666488 0.0906 ± 0.0046

Time / (s)0 10 20 30 40 50 60

Total DN Production7 0.424 1.634781 0.0812 ± 0.0016

8 0.198 3.554600 0.0229 ± 0.0095

• Production ratios dictated by fission fragment yields

1 55 6 0 014267 0 0328 ± 0 0042

1081 55.6 0.014267 0.0328 ± 0.0042

2 24.5 0.028292 0.1539 ± 0.0068

3 16.3 0.042524 0.091 ± 0.009

tro103

105 Group 1Group 2Group 3Group 4Group 5Group 6Group 7

4 5.21 0.133042 0.197 ± 0.023

5 2.37 0.292467 0.3308 ± 0.0066

6 1.04 0.666488 0.0906 ± 0.0046

Time / (s)0 10 20 30 40 50 60

Total DN Production7 0.424 1.634781 0.0812 ± 0.0016

8 0.198 3.554600 0.0229 ± 0.0095

235U233U

βi = νi/νd [%]

0.0328 ± 0.0042Yie

8 233U239Pu

0.0328 0.0042

0.1539 ± 0.0068

0.091 ± 0.009

0 197 ± 0 023Prod

40.197 ± 0.023

0.3308 ± 0.0066

0.0906 ± 0.0046

0 0812 ± 0 0016

0.0812 ± 0.0016

0.0229 ± 0.009580 100 120 140 160

Fission Product Mass /(amu)

Each fissile isotope has a i t d l d t

Uranium - 235Plutonium - 239Uranium - 233

signature delayed neutron count rate curve

jfjAjj

idiirri eeeM

1))()(1()(

0 10 20 30 40 50 60

Time (s)

*Li X, Henkelmann R, Baumgartner F (2004) Nucl Instrum Methods Phys Res B 215:246–251

Each fissile isotope has a

Uranium - 235

Each fissile isotope has a signature delayed neutron

count rate curve

Uranium - 233Plutonium -239

jfjAjj

idiirri eeeM

))()(1()(

Time / (s)

0 20 40 60 80 100

The Delayed Neutron Counting SystemIntroduction Experimental MCNP6 Model of DNC System Results & Discussion Conclusions

Diverter

Neutron Counter

Automatic Loader

Sample DisposalThe sample is irradiated in the SLOWPOKE-2 reactor

Sample Preparation

2.25 ‘’

0.67‘’

The Delayed Neutron Counting

Neutron Counting System Procedure

Diverter

Neutron Counter

Automatic Loader

The SLOWPOKE-2 Reactor

The SLOWPOKE 2 Reactor

20 kW research reactor enriched to 19.89%

Delayed Neutron Counting Arrangement

3He Detector

Sample Tubing

Paraffin

The Delayed Neutron Counting SystemIntroduction Experimental MCNP6 Model of DNC System Results & Discussion Conclusions

Diverter

Neutron Counter

Automatic Loader

Hardware & Software Control

The Fissile Analysis Program

The Fissile Analysis Program• Imports count excel file:

SLOWPOKE Test Data

11/04/2011 11:47:49 AM

Standard? YES

Sample # Cycle Time Counts Total Counts

1 A 1 015625 1863 793061 A 1.015625 1863 79306

1.546875 1821

2.0625 1759

• Corrects for background, dead time, normalizes to counts per second

2.59375 1691

• Outputs fissile (U-235) content in μg

Fissile Analysis Program Output

6.92 g 235U2 13 g 235U

2.13 g ULeast Squares Fit to Theory

))()(1()(i

fjAjjj

idiirri eeeMN

18Time (s)

0 10 20 30 40 50 600

MCNP6 Geometry

MCNP6 Plotter

19Visual Editor

MCNP6 Input Deck Summary

• tme = 0 -60e8 sh

• Sample is exposed to defined neutron flux• DN activity builds

• tme = 60e8 180e8 sh

• F8 tallies record delayed neutron activity

20Visual Editor

MS8 Describe the INPUT DECKMadison, 10/11/2011

3He Detectors & System Efficiency

He Detectors & System Efficiency

• F8 tallies (pulses) in active zone f d t tof detector

• User-defined delayed neutron energies and temporal behaviour g pusing 8-group model

))()(1()( tttfjAj idiirriN

• Experimental Efficiency: 34 ± 5%

))()(1()(i

idiirri eeeM

• MCNP6 Efficiency: 37%

Detector Wall Effects –Introduction Experimental MCNP6 Model of DNC System Results & Discussion Conclusions

Experimental Measurementsnn

E = 0 573 MeV

ETotal = 0.764 MeVaγ background

han EP = 0.573 MeV

EH 3 = 0.191 MeV

t EH-3 0.191 MeV

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

En r D p iti n in 3H D t t r (M V)

Energy Deposition in 3He Detector (MeV)

Detector Wall EffectsIntroduction Experimental MCNP6 Model of DNC System Results & Discussion Conclusions

Results have

Recombination Effects

Results have been scaled

Experimental WaveformMCNP6 with GEBPhoton Peak at 54 ngPh P k 534

0.010 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9

Photon Peak at 534 ng

Energy (MeV)

S(α,β) for Paraffin & Aqueous Solution Introduction Experimental MCNP6 Model of DNC System Results & Discussion Conclusions

1.00E-03

1.20E-03

8.00E-04

1.00E 03

) Without Thermal Tables

With Thermal Tables

4.00E-04

6.00E-04

2.00E-04

0.00E+006.00E+09 8.00E+09 1.00E+10 1.20E+10

Time (sh)

Molecular binding and crystalline effects are important at low neutron energies

Neutron Flux Estimate in SLOWPOKE-2Introduction Experimental MCNP6 Model of DNC System Results & Discussion Conclusions

1E+81E-9 1E-8 1E-7 1E-6 1E-5 1E-4 1E-3 1E-2 1E-1 1E+0 1E+1

• 69 Group Energy Spectrum from K. Khattab, I. Sulieman, (2010, Syrian MNSR)

Neutron Energy (MeV)

9 p gy p b, , ( , y )

• Magnitude of flux determined experimentally in DNC irradiation site

Neutron Flux & DN BehaviorIntroduction Experimental MCNP6 Model of DNC System Results & Discussion Conclusions

6,00069 Group Energy Spectrum

Thermal Neutrons Only

2,000Cou

00 10 20 30 40 50 60

C Ti ( )

Count Time (s)

Each fissile isotope has a

Uranium - 235

Each fissile isotope has a signature delayed neutron

count rate curve

Uranium - 233Plutonium -239

jfjAjj

idiirri eeeM

))()(1()(

Time / (s)

0 20 40 60 80 100

Temporal DN Behavior

U233 MCNP

U235 MCNP

PU239 MCNP

0.050 10 20 30 40 50 60

0 10 20 30 40 50 60

Count Time (s)

Experimental Measurements & MCNP6

U235 Experimental

U233 Experimental

U233 MCNP

U235 MCNP

0 10 20 30 40 50 60

Count Time (s)

Irradiation DurationIntroduction Background & Theory Experimental Results & Discussion Conclusions & Recommendations

5s Irradiation60s IrradiationMCNP6 5s

-1) MCNP6 5s

MCNP6 60s Irradiation20s Irradiation20s MCNP6

00 10 20 30 40 50 60

Time Following Irradiation (s)

CINDER vs. ACE Model

MCNP6 Initial Release Notes:

PHYS:N EMAX EMCNF IUNR DNB unused* FISNU COILF CUTN

DNB =delayed neutron control-1001 Analog delayed neutrons from ACE tables (if available, otherwise from CINDER tables))-101 Analog delayed neutrons from CINDER only–1 Analog delayed neutrons from ACE tables only0 No delayed neutrons produced1 15 biased number of neutrons produced per fission1-15 biased number of neutrons produced per fission.101-115 Biased number of delayed neutrons from CINDER tables1001-1015 Biased number of delayed neutrons from ACE tables, otherwise from CINDER tables

MCNP6 Initial Release Notes:

PHYS:N EMAX EMCNF IUNR DNB unused* FISNU COILF CUTN

ACE CINDER JEFF [*]

-1001 -101

ACE CINDER JEFF [*]

Prompt n 11992 11990

Delayed n 60 96

Weight Prompt n 1.52e-5 1.52e-5 1.58e-5

Weight Delayed n 7.61e-8 1.22e-7 1.06e-7

Un-normalized Comparisons – 235U (±1σ)Introduction Background & Theory Experimental Results & Discussion Conclusions & Recommendations

Experimental Measurement

MCNP6 Cinder

ACE DN Production Ratios

5 15 25 35 45 55 65

5 15 25 35 45 55 65Elapsed Time Since Irradiation(s)

Future Work

Introduction Background & Theory Experimental Results & Discussion Conclusions & Future Work

Future Work

i d i i d d li• Continued DNC experimentation and MCNP6 modeling

• Longer Count Times• Plutonium-239 & Uranium-233Plutonium 239 & Uranium 233• Mixtures of multiple fissile isotopes

• Delayed Gamma Emission from Special Nuclear Materials

• System to be built at RMC late 2011/2012• To be modeled in MCNP

• Work to be presented and submitted to ANS Conference in 2012

Thank you

Un-normalized Comparisons – 235UIntroduction Background & Theory Experimental Results & Discussion Conclusions & Recommendations

ExperimentalCINDER MCNP6ACE

0 10 20 30 40 50 60Count Time (s)

2.5x109

Supplemental Informationps

1 5 109

2.0x109Uranium - 235Plutonium - 239Uranium - 233

1.0x109

1.5x109

Each fissile isotope has a signature delayed

neutron count rate curveC

500.0x106

neutron count rate curve

Time (s)

0 10 20 30 40 50 600.0

Time (s)

jfjAjj

idiirri eeeM

mNtS ))()(1()(

jtStS )()( BackgroundtStC )()(

Isotope j Count Rate: n isotopes: Total Count Rate:

j M 1j 1

ε = system efficiency vj = delayed n production NA = Avogadro’s number σf = fission cross section ϕ neutron flux mj mass of isotope Mj Molar Mass of isotope βi production ratio for group iλi lifetime of group i tirr irradiation time td decay time t count time

238U Fission Contributions

Supplemental Information

U Fission Contributions

U d i/ h l i • Used epi/thermal ratio from Andrews thesis

• Account for flux, delayed ccou t o u , de ayedneutron production, isotopic composition of U samples

• Determined that for natural U < 0.001% of DN recorded are from the fission of 238U

238U Fission Contributions

Supplemental Information

U Fission Contributions1600

• Samples contain identical amounts of 235U 1200

• DU sample has ~1.35x the ratio of 238U to 235U

• Appears to have no effect on 600

• Appears to have no effect on the magnitude of the count rate curve

00 10 20 30 40 50 60

Time / (s)

Fission Fragment Yield & Delayed Introduction Background & Theory Experimental Results & Discussion Conclusions & Recommendations

Neutron PrecursorsFissile Isotope

Prompt n

235U233U239Pu

Prompt n

Delayed n

γDelayed neutron Precursor

87Br (t1/2 = 55 s)

Fission Product Mass /(amu)

80 100 120 140 1600

88Kr + delayed

87Kr (ground state)

delayed neutron

System Reproducibility – Natural Uranium

Syste ep oduc b ty Natu a U a u

8Ideal: Slope = 1 0 Y-Intercept = 0 0

7 Set ISet IISet III Set IV Set V

SetSlope

(2σ uncertainty)

Y-Intercept (μg)

(2σ uncertainty)

I 0.98 ± 0.01 0.08 ± 0.04

Ideal: Slope = 1.0 Y Intercept = 0.0

5Set VSet VISet VII Set IIX

I 0.98 ± 0.01 0.08 ± 0.04

II 0.96 ± 0.08 0.1 ± 0.4

III 1.00 ± 0.02 -0.02 ± 0.06

IV 1.0 ± 0.1 0.04 ± 0.06

3 V 0.98 ± 0.01 0.08 ± 0.04

VI 0.96 ± 0.08 0.1 ± 0.3

VII 1.00 ± 0.02 0.0 ± 0.1

Actual 235U Mass / (g)

1 2 3 4 5 6 7 81 IIX 0.99 ± 0.01 0.04 ± 0.06

System V lid i

SampleActual 235U

Mass (μg)

DNC System 235U

Determination (μg)Relative Error

Validation –Depleted U i

1 5.52 ± 0.06 5.3 ± 0.2 -3.45%

2 5.56 ± 0.06 5.3 ± 0.2 -5.35%

3 5.50 ± 0.06 5.3 ± 0.2 -3.01%

Uranium 4 5.56 ± 0.06 5.3 ± 0.2 -4.32%

5 5.53 ± 0.06 5.3 ± 0.2 -4.78%

6 5.62 ± 0.06 5.4 ± 0.2 -4.45%

7 5.59 ± 0.06 5.4 ± 0.2 -4.24%

8 5.53 ± 0.06 5.3 ± 0.2 -4.47%Relative Error

Actual 235U mass

Total Solution

Mass (g)

Experimental

Mass (g)

Relative Error

1.54 ± 0.04 0.290 ± 0.003 1.5 ± 0.1 -3.2

1.54 ± 0.04 0.379 ± 0.003 1.5 ± 0.1 -3.2

1.54 ± 0.04 0.568 ± 0.003 1.5 ± 0.1 -2.6

1.55 ± 0.04 0.660± 0.003 1.6 ± 0.1 -0.6

Error independent of solution nitric

acid solution volume

1.53 ± 0.04 0.819 ± 0.003 1.5 ± 0.1 -3.9

1.46 ± 0.04 0.738 ± 0.003 1.5 ± 0.1 -1.4

1.59 ± 0.04 0.943 ± 0.003 1.5 ± 0.1 -3.8

Determining 235U Content

ete g U Co te t10

Ideal DNC System OutputDU 5 4 ppm 235U

( g) 1

DU 5.4 ppm UDU 0.054 ppm 235UNat U 0.073 ppm 235UNat U 7.3 ppm 235U (calibration)

0 001 0 01 0 1 1 100.001

43Actual 235U Content / (g)

0.001 0.01 0.1 1 10

Author(s): Madison Sellers, Tim Goorley Los Alamos ... · Emily Corcoran, Davbid Kelley, RMC. MCNP6...

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