Neutronics Simulations of 237 - American Nuclear...

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

Neutronics Simulations of 237Np Targets to Support Safety-Basis and 238Pu Production Assessment Efforts at the High Flux Isotope Reactor

David Chandler and R. J. Ellis [email protected] Oak Ridge National Laboratory Nuclear and Emerging Technologies for Space 2015 (NETS) Aerospace Nuclear Science and Technology Albuquerque, New Mexico, USA February 23-26, 2015

2 Chandler, NETS 2015, February 23-26

Presentation Overview.

• Brief description of 238Pu, its production path, and its uses

• Brief description of the High Flux Isotope Reactor

• Purpose of safety-basis calculations performed to support target irradiations

• Neutronics computational toolkit and methods

• Safety-basis analyses and results • 238Pu production assessment analyses

and results • Final remarks, summary, and

conclusions


238Pu is a unique isotope that is used as a heat source in radioisotope thermoelectric generators (RTG).

238Pu Quick Facts: • Half-life of 87.7 years • 100% alpha decay • Power output of about 560 W/kg

238Pu Supply Chain History: • Produced at the Savannah River Site (SRS) via neutron irradiation in

their large reactors until the mid-to-late 1980s • Obtained from foreign sources since the SRS reactors shutdown • US DOE currently tasked to reestablish domestic production capability

• A technology demonstration subproject has been initiated to develop and implement the technology required to establish a new domestic supply chain

Multi-Mission Radioisotope Thermoelectric Generator –

MMRTG (NASA)

~5.5 MeV

Pu-238 U-234 α (He-4)

87.7y


The US DOE and NASA have undertaken a program to reestablish a domestic 238Pu production program. Storage of 237Np

Target fabrication at ORNL REDC

INL

Irradiation of NpO2/Al pellets ATR at INL and HFIR at ORNL

Chemical processing

237Np 2.14E+06 Y

238Np 2.12 D

238Pu 87.7 Y

239Pu 2.41E+04 Y

ORNL

Pu powder PuO2

Power source (i.e., MMRTG)

Robotic rover (i.e., Curiosity)

LANL


The US DOE and NASA have undertaken a program to reestablish a domestic 238Pu production program. Storage of 237Np

Target fabrication at ORNL

Irradiation of NpO2/Al pellets ATR at INL and HFIR and ORNL

Chemical processing

237Np 2.14E+06 Y

238Np 2.12 D

238Pu 87.7 Y

239Pu 2.41E+04 Y

Power source (Multi-Mission Radioisotope

Thermoelectric Generator, etc.)

Robotic rover (Curiosity, etc.) or

satellite

INL

ORNL

The focus of this presentation is on the

target irradiations in the High Flux Isotope

Reactor.


Materials Irradiation • Fusion energy • Fission energy • National security

Isotope Production • Californium-252 • Plutonium-238 • Tungsten/Rhenium-188 • Selenium-75 • Nickel-63

Neutron Scattering • Study biology, physics, chemistry, materials

science, engineering • Cold neutrons to 7 state-of-the-art instruments • Thermal neutrons to 8 state-of-the-art instruments

The High Flux Isotope Reactor is located on the Oak Ridge National Laboratory campus and has served a broad range of science and technology

communities since it reached full power in 1966.

Gamma Irradiation • Qualify materials and components for the

nuclear industry • Characterize material behaviors in a

radiation environment

Neutron Activation Analysis • Nuclear and criminal forensics • Impurity analysis • Geology • Environmental studies


HFIR is a versatile 85 MW research reactor. • Operates with an average power density of 1.7 MW/L and a peak thermal neutron flux of

2.5x1015 neutrons/cm2-s (highest in the western world)

• Two fuel elements (inner and outer fuel elements) contain 540 involute-shaped fuel plates

• Highly enriched uranium (~93 wt% 235U) fuel in the form of U3O8-Al cermet in Al-6061 clad

• Pressurized, light-water-cooled, light-water-moderated, beryllium reflected, flux trap design

• Reactivity controlled by two concentric poison-bearing control elements

• Fuel cycles typically vary between 24 and 26 days depending on the experiment loading

• Cycle 459 scheduled to startup on 2/24/2015


Safety-basis calculations are performed to support experiment irradiations.

• Will the irradiation experiment increase the probability of occurrence of an accident previously evaluated in the SAR?

• Will the experiment increase the consequences of an accident previously evaluated in the SAR?

• Study potential impacts on reactor performance – Power tilts, reactivity penalty, neutron fluxes, etc.

• Time-dependent radionuclide inventories to support post-irradiation examination, storage, transportation, and dose consequence analyses

• Calculate needed input for follow-on thermal-structural (TS) and thermal-hydraulic (TH) analyses – Heat generation (during operation and post-shutdown), – Fission product gases (He, Kr, and Xe), and – Accumulated fission densities


The neutronics toolkit includes the MCNP, VESTA, and SCALE code packages.

• MCNP5 for Monte Carlo-based neutron and photon transport – Los Alamos National Laboratory – Multi-group neutron fluxes and cross-sections for activation calculations – Heat generation rates and fission reaction rates

• VESTA 2.0.2 for depletion calculations – Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) – Couples MCNP to ORIGEN 2.2 – Multi-group binning approach

• SCALE 6.1, 6.1.2, and 6.1.3 for activation and source term analyses – Oak Ridge National Laboratory – ORIGEN for decay heat, delayed gamma sources, and nuclide inventories – CSAS/COUPLE/ORIGEN for cross-section processing and activation

calculations

• Python, FORTRAN, MATLAB, and EXCEL – Interlink codes, post-process results, and plot results


Simplified flow of neutronics calculations.

VESTA

MCNP

ORIGEN 2.2

Time-dependent MCNP inputs

Time-dependent nuclide inventories

ORIGEN-S

Decay heat

Delayed gamma source

Post-shutdown inventories

KCODE Fixed Source

H(delayed γ)

H(neutron + FP KE)

H(prompt + capture γ)

Fission rates

Time-dependent fission power Time-dependent flux

H(delayed β)


Detailed MCNP model used for neutron and gamma transport calculations.

A A

Section A-A

• Pellet construction based on ceramic oxide with Al powder

– 20 vol.% NpO2

– 70 vol.% Al – 10 vol.% void

• Pellet stack length is about 20 inches

• 7 targets loaded in a small VXF

• 19 targets likely to be loaded in a large VXF

Inner small Vertical Experiment Facility (VXF)

Outer small VXF

Large VXF

Pin 1 Pin 2

Pin 3

Pin 4 Pin 5

Pin 6

Pin 7


Neutron flux profiles.

HFIR R-Z Thermal Flux Distribution

Target X-Y Thermal Flux Distribution


Nuclear heat generation rates are calculated at various times into the cycle to support T-H and T-S calculations.

• Heat generation rates increase with increasing time into the cycle due to the production of the fissile 238Np and 239Pu isotopes

• Follow-on thermal-hydraulic and thermal-structural calculations are performed to demonstrate the pellets melting temperature is never exceeded, the target capsule surface temperatures never exceed the adjacent coolant saturation temperature, and other limits are not reached during bounding events defined in the SAR

60 65 70 75 80 85 90 95 100 105 110 115 120 125 13025

75

125

175

225

275

325

375

4255

Hea

t Gen

erat

ion

Rat

e (W

/g)

Day 5Day 10Day 15Day 20EOC

• Heat generation due to:

– Fission product – Neutron – Beta + alpha – Prompt gamma – Capture gamma – Delayed gamma

Cycle 2 results shown in VXF-15

Pin 1 Pin 2

Pin 3

Pin 4 Pin 5

Pin 6

Pin 7

Pin 1

Pin 2

Pin 3

Pin 4 Pin 5

Pin 6

Pin 7


VESTA-calculated fission powers are converted into fission rates which are integrated over time to estimate

fission densities.

-1 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14x 1019

-25

0

25

50

75

100

125

150

175

200

225

250

275

300

325

350

375

400

fission density (fissions/cm3)

VE

STA

fiss

ion

pow

er (W

/g)

red circles: pin 1 material 1

green asterisks: pin 1 material 2

pink diamonds: pin 4 material 9

End-of-cycle 1

Day 10 into2nd cycle

End-of-cycle 2

Beginning-of-cycle 2

Beginning-of-cycle 1

VXF-3

• Fission densities are used to help characterize the pellets swelling properties as a function of irradiation time

• 238Np and 239Pu are produced during irradiation – 238Np reaches

equilibrium ~10 days into the cycle

– 239Pu continuously increases in concentration

– 238Np decays away during outages between cycles


Gases are produced in the pellets due to fission events.

0 2.5 5 7.5 10 12.5 15x 1019

0

1

2

3

4

5

6

7

8

9x 10-7

fissions/cm3

He

(mol

es/c

m3 )

0 2.5 5 7.5 10 12.5 15x 1019

0

0.6

1.2

1.8

2.4

3

3.6

4.2

4.8x 10-6

fissions/cm3

Kr (

mol

es/c

m3 )

0 2.5 5 7.5 10 12.5 15x 1019

0

1

2

3

4

5

6

7

8x 10-5

fissions/cm3

Xe

(mol

es/c

m3 )

Cycle 1 Cycle 2

• Fission gases released from the pellets enter the He-filled plenum region

• Reduces the thermal conductivity of the gas between the pellet and clad

• Reduces the ability of the pellet to transfer heat to the clad and out to the coolant


Impacts on reactor core performance. • Two holders with fully loaded targets have

– Negligible impact on core relative power profile

– No statistical change in core reactivity • No impact on estimated startup control

element position • No impact on cycle length

• More analyses will need to be performed for production level irradiations – Impact on core reactivity, – Impact on gas production in beryllium

reflector, – Impact on beam tube neutron fluxes, – Impact on pressure vessel embrittlement, – etc.

10 15 20

-25

-20

-15

-10

-5

0

5

10

15

20

25

Radius (cm)

Dis

tanc

e fro

m c

ore

mid

plan

e (c

m)

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

IFE OFE

Relative power profile in fuel elements at beginning-of-cycle


Modeling and simulation of fully loaded targets in all VXFs for 238Pu production assessments.

Brief model description:

• Pellets subdivided into 3 equal volume rings (self-shielding in pellet)

• Pellet stack subdivided into 6 axial regions (axial cosine flux shape)

• 70 target rods in the 10 ISVXFs

• 35 target rods in the 5 OSVXFs

• 114 target rods in the 6 LVXFs

• Total of 219 target rods modeled

A few notes:

• Other irradiation facilities exist for potential use (RBs, SPBs, etc.)

• Some VXFs may need to be left vacant for other users, to maintain neutron fluxes to the scattering instruments, etc.


0 50 100 150 200 250 300 350 4000

100

200

300

400

500

600

700

800

900

1000

time (days)

Pu2

38 (g

ram

s)

ISVXFs (x10)OSVXFs (x5)LVXFs (x6)Sum of 21 VXFs

Total: 904 g

ISVXFs: 384 g LVXFs: 365 g

OSVXFs: 155 g

VESTA and ENDF/B-VII.0 cross-sections used. The targets were irradiated for 8 cycles. No target shuffling,

replacement, or rotation modeled. One neutron transport and depletion step per cycle.


0 50 100 150 200 250 300 350 4000.7

0.75

0.8

0.85

0.9

0.95

1

time (days)

Pu2

38 to

Pu

ratio

ISVXFs (x10)OSVXFs (x5)LVXFs (x6)

0 50 100 150 200 250 300 350 4000

2

4

6

8

10

12

14

16

18

time (days)

100x

(Pu2

38(t)

/Np2

37(t=

0))

ISVXFs (x10)OSVXFs (x5)LVXFs (x6)

Irradiating 237Np produces 238Pu and other Pu isotopes. Efficiency and product purity are key.

Percent Conversion: • Production efficiency decreases with

irradiation time due to feed material consumption

• More target irradiations (replacing irradiated targets with fresh) increases waste, operator time, etc.

238Pu Purity (238Pu to Pu ratio): • 85% criteria • Higher purity the better • Purity decreases with irradiation time due

to subsequent neutron captures


Modeling and simulation of 6 cycles with fully loaded targets in all VXFs and target replacement. 13 neutron transport and depletion steps per cycle. • Core fuel, control elements, and NpO2/Al target depleted as a function of time to

assess the impact of fuel depletion, control element withdrawal, and time-varying neutron fluxes on 238Pu production

• Target replacement after every two cycles of irradiation

• A total of 394 pins modeled and 486 materials activated over the 6 cycles

Cycles 1 and 2 Cycles 3 and 4 Cycles 5 and 6

ISVXF set 1 ISVXF set 2 ISVXF set 3

OSVXF set 1

OSVXF set 2 LVXF set 1

ISVXF set 1 ISVXF set 2 OSVXF set 1


A total of about 0.96 kg 238Pu can be produced in HFIR’s permanent beryllium reflector in 6 nominal

cycles of irradiation with target replacement.

0 50 100 150 200 250 3000

100

200

300

400

500

600

700

800

900

1000

time (days)

Pu2

38 (g

ram

s)

ISVXFs (x10)ISVXFs (x10)ISVXFs (x10)OSVXFs (x5)OSVXFs (x5)LVXFs (x6)Sum of 21 VXFs

Total: 956 g (per 6 cycles)

LVXFs: 293 g (per 6 cycles)

ISVXFs: 172 g (per 2 cycles)

OSVXFs: 98 g

(per 4 cycles)


0 50 100 150 200 250 3000.84

0.86

0.88

0.9

0.92

0.94

0.96

0.98

1

time (days)

Pu2

38 to

Pu

ratio

ISVXFs (x10)ISVXFs (x10)ISVXFs (x10)OSVXFs (x5)OSVXFs (x5)LVXFs (x6)

0 50 100 150 200 250 3000

1

2

3

4

5

6

7

8

9

10

time (days)

100x

(Pu2

38(t)

/Np2

37(t=

0))

ISVXFs (x10)ISVXFs (x10)ISVXFs (x10)OSVXFs (x5)OSVXFs (x5)LVXFs (x6)

Target replacement keeps the purity levels above 85% and 8‒9% conversion is achieved.

Percent Conversion: • ISVXF: ~8.0% conversion in 2 cycles

• OSVXF: ~9.0% conversion in 4 cycles

• LSVXF: ~8.5% conversion in 6 cycles

238Pu Purity (238Pu to Pu ratio): • 2, 4, and 6 cycles in the ISVXFs,

OSVXFs, and LVXFs keep the purity

above 85%


Cycle ISVXF set 1

ISVXF set 1

ISVXF set 1

OSVXF set 1

OSVXF set 2 LVXF

1 97.1 - - 30.5 - 61.4 2 171.9 - - 56.6 - 116.7 3 - 97.0 - 78.9 - 167.0 4 - 171.9 - 98.2 - 213.0 5 - - 97.1 - 30.5 254.8 6 - - 172.0 - 56.5 293.2

238Pu production per VXF set per cycle (grams)

Cycles/year 6 7 238Pu or PuO2

238Pu PuO2 238Pu PuO2

ISVXF 0.52 0.69 0.60 0.80 OSVXF 0.15 0.20 0.17 0.23 LVXF 0.29 0.39 0.34 0.46 SUM 0.96 1.28 1.12 1.49

Calculated annual 238Pu production (kg) and estimated PuO2 production (kg)

A total of about 0.96-1.12 kg 238Pu can be produced annually in HFIR’s permanent beryllium reflector.


Final remarks, summary, and conclusions. • A technology demonstration subproject has been initiated to

develop a safe and efficient 238Pu production infrastructure

• Neutronics calculations are being performed at HFIR to support the goals of this subproject to – Design and qualify target irradiations – Assess annual 238Pu production capabilities

• It is estimated that 0.96-1.12 kg 238Pu (~1.3-1.5 kg PuO2) can be produced in HFIR’s permanent beryllium reflector per year

• Validation/benchmark studies are needed to confirm results – Post-irradiation 237Np and Pu vector appear to be in good agreement – Fission product inventory appears to be over-estimated

• Studies and experiments are being performed and planned to – Evaluate Np and Pu cross-sections and fission yields – Assess ENDF/B-VII.0, ENDF/B-VII.1, and JEFF 3.1 cross-section data with

the VESTA and SCALE 6.1.3 ORIGEN depletion codes


• Cold and thermal neutron scattering • Materials irradiation • Isotope production • Neutron activation analysis • Gamma irradiation

Thank you. The authors would like to thank C. Bryan, R. W. Hobbs, and R. M. Wham for their support and contributions to this work. This work has been sponsored by NASA’s Science Mission Directorate and the US DOE Office of Space and Defense Power Systems.

• The High Flux Isotope Reactor is located on the Oak Ridge National Laboratory campus. • Oak Ridge National Laboratory is managed by UT-Battelle for the US DOE. • The High Flux Isotope Reactor is a U.S. DOE Office of Science User Facility.

Date post:	30-Sep-2020
Category:	Documents
Upload:	others
View:	3 times
Download:	0 times

Neutronics Simulations of 237 - American Nuclear...

Documents