Simulation of β n Emission From Fission Using Evaluated Nuclear Decay Data

1. Simulation of βnEmission From Fission Using Evaluated Nuclear Decay Data May 2, 2013 Ian Gauld Marco Pigni Reactor and Nuclear Systems Division

2. 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 emissionafter 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

3. 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 Interim storage Reprocessing Transportation and storage

4. 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

5. 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

6. 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

7. 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-5.1-1971 draft, issued 1979) • Parameters for exponential fits available for four fissionable nuclides, (MeV/s/fission)

8. 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)

9. 235U thermal fission

10. 239Pu thermal fission

11. 241Pu thermal fission

12. 238U fast fission

13. 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

14. OECD/NEA WPEC 25Decay 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 Electromagnetic decay heat following thermal fission burst of 239Pu

15. Beta Delayed Neutron Emission • Current methods in reactor physics analysis rely on a delayed-neutron 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)

16. βnEmission 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

17. ORIGEN βn Calculation – 235U fission

18. 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 b3n-emitters ENDF/B-VII.1: 390 n-emitters, 111 2n-emitters, 14 b3n-emitters and 2 b4n-emitters • 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&FYData after a fission pulse in 235U Testing JEFF-3.1.1 and ENDF/B-VII.1, Cabellos et al., ND2013

19. 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

20. Frequency Distributions of Sampled Values Kinf ; 60 GWD/T Kinf ; 0 GWD/T Tc-99 concentration; 50 GWD/T Group 1 nu-fission ; 30 GWD/T

21. Uncertainty analysis – 235U fission 300 years

22. MTAS Summary and Conclusions • New detectors are being used to obtain improved nuclear decay data • Gamma calorimeter • Neutron detectors • 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 3Hen VANDLE