Simulation of βn
Emission From Fission
Using Evaluated Nuclear
Decay Data
May 2, 2013
Ian Gauld
Marco Pigni
Reactor and Nuclear Systems
Division
1
Managed by UT-Battelle
for the U.S. Department of Energy
Nuclear decay data from an end-user
perspective.
• Evaluated decay data have major importance to areas of
reactor safety and nuclear fuel cycle analysis
• Reactor safety applications include analysis of energy
release (decay heat) and beta-delayed neutron emission
after fission
• Decay heat impacts safety studies for irradiated nuclear fuel
during reactor operation, fuel handling, storage, and
disposal
• Delayed neutrons play an important role in reactor control
and behavior during transients
• Our group is an end user of decay data
2
Managed by UT-Battelle
for the U.S. Department of Energy
Material processing and fabrication
Commercial and research reactors
SCALE is a nuclear systems modeling
and simulate code used worldwide
for reactor and fuel cycle applications
• Criticality safety
• Radiation shielding
• Cross-section processing
• Reactor physics
• Sensitivity and uncertainty analysis
• Spent fuel and HLW characterization
Disposal
3
Managed by UT-Battelle
Reprocessing
for the U.S. Department of Energy
Interim storage
Transportation and storage
Simulation of Nuclear Fuel
• ORIGEN – Oak Ridge Isotope GENeration and Depletion code
• Irradiation and decay
• Calculates
–
–
–
–
–
Time dependent isotopic concentrations
Radioactivity
Decay heat (based on summation)
Radiation sources (neutron/gamma)
Toxicity
• Explicit simulation of 2228 nuclides using evaluated nuclear data
• Fast: 0.02 s per time step
• ENDF/B-VII.1 nuclear data for:
– 174 actinides
– 1151 fission products
– 903 structural activation materials
4
Managed by UT-Battelle
for the U.S. Department of Energy
ENDF/B-VII.1 Nuclear Data Libraries
Decay half lives, branching fractions, energy release
− 2226 nuclides
Cross sections
− ENDF/B-V, -VI, -VII
− JEFF-3.0/A special purpose activation file
Fission product yields
− Energy-dependent yields for 30 actinides
Gamma ray production data
− X-ray and gamma ray emissions per decay
Neutron production data from LANL SOURCES code
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−
−
−
−
Alpha decay energies
Stopping powers
α,n yield cross sections
Spontaneous fission spectral parameters
Delayed neutron spectra for 105 precursor nuclides
Alpha and beta spectra included in next release
5
Managed by UT-Battelle
for the U.S. Department of Energy
ENDF/B-VII.1 Decay Sublibrary
Improvements
• Decay data based on the Evaluated Nuclear Structure
Data File (ENSDF), translated into ENDF-6 format
• 3817 long-lived ground state or isomer materials
• More thorough treatment of the atomic radiation
• Improved Q value information
• Recent theoretical calculations of the continuous
spectrum from beta-delayed neutron emitters
• New TAGS (Total Absorption Gamma-ray Spectroscopy)
data
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for the U.S. Department of Energy
Decay Heat Standards
• ANS-5.1-2005 and ISO 10645 (1992) widely adopted in reactor
safety codes
• Experimentally-based curves developed using groups, fit to
experimental data at short decay times
• Groups developed to represent decay times from 1 second to 300
years after fission
• Necessitated because nuclear decay data inadequate for short
decay data times at the time of standard development (ANS-5.11971 draft, issued 1979)
• Parameters for exponential fits available for four fissionable
nuclides,
(MeV/s/fission)
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Managed by UT-Battelle
for the U.S. Department of Energy
Code Calculations using Evaluated
Nuclear Data
Alternate approach to standards-based methods using nuclear
decay data and fission yields for all fission products generated
by fission
– Simulate all fission products explicitly
– Provides greater insight into system performance
– Contributions from important nuclides, and gamma, beta, and
alpha components
– Gamma spectrum for determination of non-local energy
deposition
– Provides values for isotopes not considered by the current
Standards
– Can evaluate the impact of changes in fission energy (e.g.,
fast reactor systems)
8
Managed by UT-Battelle
for the U.S. Department of Energy
235U
9
thermal fission
Managed by UT-Battelle
for the U.S. Department of Energy
239Pu
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thermal fission
Managed by UT-Battelle
for the U.S. Department of Energy
241Pu
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thermal fission
Managed by UT-Battelle
for the U.S. Department of Energy
238U
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fast fission
Managed by UT-Battelle
for the U.S. Department of Energy
239Pu
thermal fission γ energy
The effect of
introducing
TAGS data
from Algora,
(2010) to
JEFF-3.1.1
decay data
Testing JEFF-3.1.1 and ENDF/B-VII.1, Cabellos et al., ND2013
13
Managed by UT-Battelle
for the U.S. Department of Energy
OECD/NEA WPEC 25
Decay Heat Analysis
• International Working Party on Evaluation Co-operation of the
NEA Nuclear Science Committee NEA/WPEC-25
• VOLUME 25 - Assessment of Fission Product Decay Data for
Decay Heat Calculations (2007)
http://www.nea.fr/html/science/wpec/volume25/volume25.pdf
• Important to –
– Reactor LOCA analysis
– Delayed gamma analysis from active
neutron interrogation
• Known problems with data
• WPEC-25 developed a priority list of
isotopes for re-evaluation
14
Managed by UT-Battelle
for the U.S. Department of Energy
Electromagnetic decay heat following thermal
fission burst of 239Pu
Beta Delayed Neutron Emission
• Current methods in reactor physics analysis rely on a delayedneutron group representation (Keepin)
• ENDF/B 6-group; JEFF 8-group
• Based on theoretical-experimental approach to delayed neutron
emission
• Isotopes with similar characteristics combined with an effective
group half life and emission spectra
• Ability of nuclear decay data to simulate neutron emission rate
and temporal energy spectra is limited
(n/s/fission)
15
Managed by UT-Battelle
for the U.S. Department of Energy
βn Emission Simulation with ORIGEN
• Neutron methods in ORIGEN are based on the LANL SOURCES
code
• ORIGEN tracks production and decay of 1151 fission product
isotopes
• However, the neutron library currently has precursor data for
only 105 fission products – in this implementation, delay
neutrons are only calculated for the limited number of isotopes in
the neutron library (from SOURCES)
• ENDF/B-VII.1 has more than 500 n-emitters
• Delayed neutron energy spectra included for each fission
product – stored as multigroup representation used in ENDF/B
bins
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Managed by UT-Battelle
for the U.S. Department of Energy
ORIGEN βn Calculation –
235U
fission
Keepin
ORIGEN
0.001
0.0001
0.00001
0.000001
0.01
0.1
Delayed Neutron Yield [n/sec]
Neutron yield (n/fission/sec)
0.01
10000
0.5 s
1.0 s
8000
1.5 s
2.0 s
6000
2.6 s
4000
1
10
100
1000
Time (s)
2000
0
0
0.5
1
Energy [MeV]
17
Managed by UT-Battelle
for the U.S. Department of Energy
1.5
Recent Studies at UPM
 Calculations performed with JEFF-3.1.1 and ENDF/B-VII.1
JEFF 3.1.1: 241 n-emitters, 18 2n-emitters and 4 3n-emitters
ENDF/B-VII.1: 390 n-emitters, 111 2n-emitters, 14 3n-emitters and 2 4nemitters
At t=0 s, >100% difference between ENDF/B-VII.1 6-group data and
summation calculations using ENDF/B-VII.1 decay and yield data
Comparison of
delayed neutron
emission rate
calculated using
Keepin 6/8-group
formula and
Decay&FY Data
after a fission
pulse in 235U
Neutron emission by 1 fission (n/s)
1.0E-01
1.0E-02
1.0E-03
1.0E-04
Keepin-JEFF-3.1.1
1.0E-05
n_emiss_rate (JEFF-3.1.1)- My work
Keepin-ENDF/B-VII.1
n_emiss_rate (ENDF/B-VII.1)- My work
1.0E-06
0.01
0.1
1
10
100
Time after fission burst (s)
18
Managed by UT-Battelle
for the U.S. Department of Energy
Testing JEFF-3.1.1 and ENDF/B-VII.1, Cabellos et al., ND2013
1000
New Developments in Uncertainty
Analysis
A stochastic nuclear data sampling approach is implemented
in the next release of SCALE
• Defines uncertainty distributions and correlations for all nuclear
data
• Reaction cross sections
• Fission yields
• Nuclear decay data
• Executes any SCALE code using perturbed data parameters for
uncertainty analysis
• Performs parallel computations using MPI or OpenMP
• Response uncertainty computed by automated statistical analysis
of output response distribution
19
Managed by UT-Battelle
for the U.S. Department of Energy
Frequency Distributions of
Sampled Values
Kinf ; 0 GWD/T
Group 1 nu-fission ; 30 GWD/T
20
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for the U.S. Department of Energy
Kinf ; 60 GWD/T
Tc-99 concentration; 50 GWD/T
Uncertainty analysis –
235U
fission
300 years
21
Managed by UT-Battelle
for the U.S. Department of Energy
Summary and Conclusions
MTAS
• New detectors are being used to obtain improved
nuclear decay data
– Gamma calorimeter
– Neutron detectors
3Hen
• Improved data impact delayed energy release
(total and gamma decay heat) and delayed
neutron emission
• Work initiated to integrate new measurements
with the ORIGEN simulation code
• Planned performance evaluation using
comparisons with benchmarks and other
measurement data
• Complete uncertainty analysis now possible
22
Managed by UT-Battelle
for the U.S. Department of Energy
VANDLE