Ian Gauld Marco Pigni Reactor and Nuclear Systems Division May 2 2013 Nuclear decay data from an enduser perspective Evaluated decay data have major importance to areas of reactor safety and nuclear fuel cycle analysis ID: 729157
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Simulation of βn Emission From Fission Using Evaluated Nuclear Decay Data
Ian Gauld
Marco PigniReactor and Nuclear Systems Division
May 2, 2013Slide2
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 fissionDecay heat impacts safety studies for irradiated nuclear fuel during reactor operation, fuel handling, storage, and disposalDelayed neutrons play an important role in reactor control and behavior during transientsOur group is an end user of decay dataSlide3
3
SCALE is
a nuclear systems modeling and simulate code used worldwide
for reactor and fuel cycle
applications
Disposal
Material processing and fabrication
Commercial and research reactors
Interim storage
Transportation and storage
Reprocessing
Criticality safety
Radiation
shielding
Cross-section processing
Reactor physics
Sensitivity and uncertainty analysis
Spent fuel and HLW characterizationSlide4
ORIGEN – Oak Ridge Isotope GEN
eration and Depletion code
Irradiation and decayCalculates Time dependent isotopic concentrationsRadioactivityDecay 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
Simulation of Nuclear FuelSlide5
ENDF/B-VII.1 Nuclear Data LibrariesDecay half lives, branching fractions, energy release2226 nuclides
Cross sectionsENDF/B-V, -VI, -VII
JEFF-3.0/A special purpose activation fileFission product yieldsEnergy-dependent yields for 30 actinidesGamma ray production
dataX-ray and gamma ray emissions per decayNeutron 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 releaseSlide6
ENDF/B-VII.1 Decay Sublibrary ImprovementsDecay data based on the Evaluated Nuclear Structure Data File (ENSDF), translated into ENDF-6 format3817 long-lived ground state or isomer materialsMore thorough treatment of the atomic radiationImproved Q value informationRecent theoretical calculations of the continuous spectrum from beta-delayed neutron emitters
New TAGS (Total Absorption Gamma-ray Spectroscopy) dataSlide7
Decay Heat StandardsANS-5.1-2005 and ISO 10645 (1992) widely adopted in reactor safety codesExperimentally-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-5.1-1971 draft, issued 1979)Parameters for exponential fits available for four fissionable nuclides,
(
MeV
/s/fission)Slide8
Code Calculations using Evaluated Nuclear DataAlternate approach to standards-based methods using nuclear decay data and fission yields for all fission products generated by fissionSimulate all fission products explicitlyProvides greater insight into system performanceContributions from important nuclides, and gamma, beta, and alpha componentsGamma spectrum for determination of non-local energy deposition
Provides values for isotopes not considered by the current StandardsCan evaluate the impact of changes in fission energy (e.g., fast reactor systems)Slide9
235U thermal fissionSlide10
239Pu thermal fissionSlide11
241Pu thermal fissionSlide12
238U fast fissionSlide13
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 Slide14
OECD/NEA WPEC 25Decay Heat AnalysisInternational Working Party on Evaluation Co-operation of the NEA Nuclear Science Committee NEA/WPEC-25VOLUME 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
Electromagnetic decay heat following thermal fission burst of
239
Pu Slide15
Beta Delayed Neutron EmissionCurrent methods in reactor physics analysis rely on a delayed-neutron group representation (Keepin)ENDF/B 6-group; JEFF 8-groupBased on theoretical-experimental approach to delayed neutron emissionIsotopes with similar characteristics combined with an effective group half life and emission spectraAbility of nuclear decay data to simulate neutron emission rate and temporal energy spectra is limited
(n/s/fission)Slide16
βn Emission Simulation with ORIGENNeutron methods in ORIGEN are based on the LANL SOURCES codeORIGEN tracks production and decay of 1151 fission product isotopesHowever, 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-
emittersDelayed neutron energy spectra included for each fission product – stored as multigroup representation used in ENDF/B binsSlide17
ORIGEN βn Calculation – 235U fissionSlide18
Recent Studies at UPMCalculations performed with JEFF-3.1.1 and ENDF/B-VII.1 JEFF 3.1.1: 241 n-
emitters, 18 2n-emitters and 4
b3n-emitters ENDF/B-VII.1: 390 n-
emitters, 111 2n-emitters, 14
b
3n-emitters and 2
b
4n-emitters
Testing JEFF-3.1.1 and ENDF/B-VII.1, Cabellos et al., ND2013
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
235
USlide19
New Developments in Uncertainty AnalysisA stochastic nuclear data sampling approach is implemented in the next release of SCALE
Defines uncertainty distributions and correlations for all nuclear data
Reaction cross sectionsFission 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 distributionSlide20
Frequency Distributions of Sampled Values
Group 1 nu-fission ;
30 GWD/T Kinf
; 60 GWD/T
K
inf
;
0 GWD/T
Tc-99 concentration;
5
0 GWD/T Slide21
Uncertainty analysis – 235U fission300 yearsSlide22
Summary and ConclusionsNew detectors are being used to obtain improved nuclear decay dataGamma calorimeterNeutron detectorsImproved data impact delayed energy release (total and gamma decay heat) and delayed neutron emissionWork 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
MTAS
3Hen
VANDLE