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

Slide1

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