Application of Nuclear Decay Data to Reactor Modeling and

Report
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
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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
−
−
−
−
−
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
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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
6
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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)
7
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
10
thermal fission
Managed by UT-Battelle
for the U.S. Department of Energy
241Pu
11
thermal fission
Managed by UT-Battelle
for the U.S. Department of Energy
238U
12
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
16
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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
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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

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