labr3 detector simulations and experiments

8/9/2019 LaBr3 Detector Simulations and Experiments

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LaBr3 detector simulations and

experimentsP.Srinivasan

Radiation Safety Systems DivisionBhabha Atomic Research Center


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LaBr3 system in RSSD, BARC• Of late LaBr3:Ce gaining popularity over the NaI (Tl) detector.•

LaBr3(Ce) has a density of 5.08 gm/cc, a light output of about61 photons/keV, fast decay time of 16 ns, and an emissionwavelength of 380 nm. The energy resolution is about 3%(FWHM) at 662 keV.

• In our laboratory, 2 no:s of 1.5” x 1.5” LaBr3 detector (Brillance-

380) coupled to the 51 mm R6231 Photo multiplier Tube wasprocured from Saint Gobain in 2009.• One of the units is installed as a fixed system inside a shield in

our Radiation Safety lab of the research reactor fuel fabricationfacility , Trombay.

• The other unit is attached to a compact USB MCA designed forfield and portable applications• We use the detectors for quick quantification of various radio

nuclides in gaseous & liquid effluents and other solid matriceslike slurry, sand, concrete, for day to day safety and regulatorypurposes in nuclear fuel processing facilities.


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LaBr3 spectrometry system, RSSD,BARC

Fixed System with tailor made shielding – for analysis of liquid

and solid sample matrices in various small geometries


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Portable LaBr3 spectrometry system

APSARACore

APSARA Reactor HallPURNIMA DT/DD neutron generator


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Detector Simulation and Calibration

• MCNP Monte Carlo simulations were performed tounderstand the spectral response of the 1.5”x1.5”LaBr3 detectors

• Full-energy-absorption efficiencies for point andcylindrical source geometries were estimated forsome routinely used source-detector geometryconditions by MCNP simulations.

• Experimental efficiency calibration of the detectorwas also carried out. A comparison of the full energypeak efficiency values obtained by both the MonteCarlo methods and by experimental measurementswas carried out.


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Internal Radioactivity of LaBr3 (Ce)

• LaBr3(Ce) crystal is known to have internal radioactivitydue to naturally occurring radioisotopes 138La and 227AcNaturally occurring lanthanum contains about 0.09% of138La, which has a half-life of 1.06 x10 11 -years. 138 Laemits two gamma rays: a 788.7-keV gamma ray from betadecay (34 %) to stable 138Ce and a 1435.8-keV gamma rayfrom electron capture (66 %) to stable 138Ba. There arealso Ba K x-rays from 31-38 keV.

• These gamma rays are present as a constant backgroundduring counting. Hence the detectors are more suitable for

short counting purposes than for long counting timeswhere the detector self activity pulses might interfere withthe sample activity. LaBr3 (Ce) detectors are also employedin hand-held survey devices for radio-isotope identificationin large structures like waste drums and cargo containers.


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Advantages and disadvantages of close contact geometry • Close contact geometry counting is advantageous for quick counting

purposes owing to the large solid angle subtended by the detector at thesource.

•

This results in higher efficiencies and the desired statistical accuracy in shortmeasurement times demanded during plant operations.• The major problem in this geometry is due to coincidence summing. The

effect of coincidence summing of gamma radiation occurs when two ormore gamma rays are emitted in coincidence from the decay of the sameradio-nuclide, and are recorded simultaneously within the resolving time of

a detector. The coincidence-summing effect depends on the decay schemeof the radio-nuclide and the solid angle subtended by the detector at thesource.

• For close geometries, point sources will exhibit larger coincidence-summingeffects than the extended sources as reported in literature (Debertin andSchotzig)

• The determination of energy dependence of efficiencies requiresconsiderable number of mono-energetic gamma-ray sources. Since, most ofthe gamma-ray emitters in the range of few hundred keV to few MeV emitmultiple gamma-rays, it is necessary to carry out coincidence summingcorrections for determining the true efficiencies.


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Monte Carlo Simulations • In the present work, the standard Monte Carlo package MCNP4C was used

to perform simulations of the spectral response of the LaBr3 (Ce) detectorand the estimation of full energy absorption efficiency at various energies.

• The Monte Carlo N-Particle transport code (MCNP) version 4C developed inthe Los Alamos National Laboratory was used for the computations.MCNP4C is a general purpose continuous energy generalized geometry,time dependent code, which deals with transport of neutrons, photons,and coupled electron photon transport, i.e., transport of secondary

electrons resulting from gamma interactions.• MCNP4C transports electrons and photons over an energy range of 1 keV

to 1 GeV using the ENDF VI (Evaluated Nuclear Data File – version VI) crosssection data.

• MCNP has variety of options to describe various types of sources anddifferent estimators for fluence, energy deposition, kerma etc. The photonphysics option of the code utilized in this work includes coherentscattering, incoherent (Compton) scattering, photoelectric effect (with Kand L shell fluorescence) and pair production. An energy cut of value of 1keV was used to terminate particle transport.

• MCNP gives the users the added flexibility to define various geometryregions through predefined macro bodies in addition to the conventionalBoolean operator based combinatorial geometry methods


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MCNP Model of the detector• An MCNP model, including the geometrical representation of the

locations of the LaBr3(Ce) crystal with the aluminium casing

were modelled as per the details provided by the manufacturer.• Self -attenuation inside the source matrix and attenuation and

scattering in the surrounding materials were considered in theMonte Carlo computations.

• The manufacturer specified source strength was considered incomputing the weight of starting source photons in the MCNPmodel for each of the standard sources used in the experiments.

• The gamma ray photon energies and corresponding emissionprobabilities of the sources was taken from literature forsimulation.

• The photon energy pulse height distribution in the detectormedium was estimated using the F8 tally option of the MCNP.The pulse height obtained from F8 tally in units of pulses-s -1 wasappropriately normalised to the number of photons emitted perunit time from the source radionuclide under consideration.


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•


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Experimental Measurements• In order to minimise the external background photons entering

the detector, a specially designed Lead shield lined withaluminium and copper was designed and fabricated for housingthe detector-PMT assembly. A butterfly door made of 25 mmMild Steel resting on bearings is provided on top to enablesample loading. The detector is positioned vertically enabling thesample to be placed on contact with the top face of the detector.Suitable thin plastic frames are also available, if required, forplacing the sample at a desired distance from the detector.

• Experimental efficiency calibration of the LaBr3(Ce) detector wasperformed using calibrated point sources of 133Ba, 137Cs, 60Coprocured from the Board of Radiation and isotope Technology(BRIT) and an a cylindrical source of 137Cs in grass. (IAEA intercomparison source) placed on contact with the detector topsurface. A concrete powder matrix spiked with severalradionuclides in a standard vial provided by the NPL, UK was also

used.


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Results: Insitu gamma spectrometry -FEP Efficiency of 1.5"x1.5"LaBr3 side-on contact with Apsara Al-Liner 2.5 m x 1.5 mx 32 mm


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In-situ gamma spectrum acquired using the portableLaBr3 system in the shield cubicle near beam hole-10

on the APSARA pool wall


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Monte Carlo FEPE chart LaBr3-AFD as a function ofconcrete packet thickness

Ht .4cm

Ht .5 cm

Ht .6 cm

Ht .7 cm


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Radiological monitoring systems for Thorium Fuel cycle


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Radiological monitoring systems forThorium Fuel cycle


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Assessment of Calibration parameters for the Aerial GammaSpectrometry System of RSSD

Radioactive Sources & Source Distribution Pattern • (Cs-137 -2.0 Ci) , (Co–60 - 0.6 Ci), (Ir-192 -1.4 Ci) & (I-131-1.0 Ci)• 20 × 5 array distributed over a rectangular area 950 m Χ 200 m witha pitch distance of 50 m. 100 grid points for each sourceDetector Dimension & Locations • 16” Χ 4”Χ 4” NaI(Tl) detector was placed at heights of 80 m, 100 m,

120m

MCNP simulation parameters • 6,00,000 photon histories / run X 60 runs – P- III,64 MB Ram- 6 min /

run• Energy Bin width -5.7 keV / channel.• The relative error in each energy bin of the fluence spectrum varies

from 6 % to 10 % except in the source bin where the relative error isof the order of 0.2 % owing to the adequacy of sampling in the sourceenergy bin.

• Out put from MCNP F2 / F5 tally formed the input to MARTHA• MARTHA folded the F5 tally results with the NaI(Tl) response

functions to arrive at the full spectrum.


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Aerial Gamma Spectrometry--• The photo-peak counts in the respective energy window• The Air To Ground Correlation Factors (AGCF) at various

altitudes required to arrive at the ground contamination level inkBq/m 2 from the counts observed at a particular height in therespective window of interest.

• The dose rates at 1m from the contaminated surface due to agiven radioisotope

(Contamination level in kBq/m2)• AGCF = -----------------------------------------

CPS in the respective energy window

• The CPS in the respective energy window can be obtained bymodifying the fluence (obtained from MCNP3 - F2 or F5 tally) by

i) The transmission factor of the helicopter bodyii) The total area of the detector seeing the

groundiii) The photo-peak efficiency of the detector


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Comparison of calculated and recorded photo peak counts at twoheights above the central line of the source area -AGSS

(Cs-137 - 2 Ci in a 20x5 matrix)


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Co-60 spectrum at 100 m altitude inside ahelicopter


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labr3 detector simulations and experiments

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