march 19. 2009 aap09 - brazil - m. fallot et al. collaborations double chooz, nucifer and mure time...

March 19. 2009 AAP09 - Brazil - M. Fallot et

al.

Collaborations Double Chooz, Nucifer and MURE

Time evolution of reactor antineutrino energy spectrum and flux


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Outline

2 reasons why antineutrino spectrum and flux vary with time :

- Non equilibrium : U and Pu istope and energy spectrum simulations

- Variation of the fuel isotopic content of a reactor core : reactor antineutrino spectrum and flux simulation

proliferation scenario (taking into account out of equilibrium effects when rapid power changes)


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Integral -spectra measurements: [A. Hahn et al., Phys. Lett. B, 218,

(1989)] 235U, 239,241Pu targets @ILL, at better than 2% until 8 MeV

Summing individual -spectra: [Tengblad et al. (NPA503(1989)136)

111 nuclei @OSIRIS Studsvik and ISOLDE ]

Conversion - e : global shape uncertainty from 1.3%@3MeV to 9%@8MeV Measurement only related to thermal fission Irradiation time dependence (20 min & 1.5 d)

Don’t agree with the experimental integral spectra (important errors : 5% at 4MeV, 11% at 5MeV and 20% at 8MeV)

-spectra determinations

Remaining short-lived, high Q, unknown nuclei


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In agreement with Chooz and Bugey data (1.9% on the e flux)

€

N (Eν , t ) ≈ α f

f

∑ (t) ⋅ρ f (Eν )- i (t) : relative contributions to the total fission (i=1) - i (E) : -spectra 235U, 239Pu, 241Pu, and 238U (Schreckenbach et al. and P. Vogel

Don’t take into account -decay from products of radiative capture of neutron

• Determination of the -spectra :

Theoretical approach : Microscopic cal. of trans. mat. elts [H-V. Klapdor, Phys. Rev. Lett., 48, (1982)] Phenomenological model for unknown nuclei + databases [P. Vogel et al., Phys. Rev. C, 24, (1981)]

• Determination of the reactor -spectra :

V. Kopeikin : Resolution of the Bateman equations for selected set of fission products + fission rates from the power plants + neutron capture contributions (ENDF database) Antineutrino experiment approaches:

-spectra determinations


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Neutron capture, non equilibrium effects

« In the antineutrino energy range 1.8-3.5 MeV, the relative contribution of the additional radiation during the reactor operating period (Figs. 1, 5) is about 4%, which is somewhat greater than the error of the ILL spectra [5]. »

2 4

1.02

1.04

1

12

3

4

3

(,)/(,)t=0.5y t=0.5yconvers.E

Antineutrinoenergy,MeVE

E Ratio of the reactor -spectrum obtained by conversion + calculation methods to the - spectrum obtained by conversion method only convers. :1 – the additional contribution from fission product residual - activity of the previous two reactor cycles; 2 – from neutron capture by fission products; 3 – from increase of fission product - activity of the current reactor cycle from 1 day to 0.5 year; 4 – the sum (4=1+2+3)

V. Kopeikin et al. , arXiv:hep-ph/0110290


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Addition to the spectrum : 1, 2 and 3 correspond to the beginning, middle and end of the reactor operating cycle

1.5

23

3.52.50

0,02

1

2 3Δ,1/(MeVfission)

.

F

Antineutrino energy E , MeVν

0.01

0.03

Modification of the - and spectra associated with neutron capture by fission products

Neutron capture, non equilibrium effects

But also the evolution of the and spectra during operating (ON) and shutdown (OFF) periods

Solid line is the ratio of the spent fuel pool - spectrum to the reactor spectrum. Dashed lines are the ratios of the reactor - spectrum after the reactor is shut down to the reactor - spectrum at the end of the operating cycle.

(E,toff) / (E,ton= 1 yr)

Spent Fuel Pool

0

0.02

0.04

2 32.5 3.5

toff = 1yr

toff = 1 d

V. Kopeikin et al. , arXiv:hep-ph/0110290


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Developed simulation toolsAnd results on U and Pu isotopes

spectra


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Principle of our strategy

weighted

Monte-Carlo Simulation :Evolution Code MURE

-branch database :BESTIOLE, …

Total e and - energy spectra with complete error treatment

- decay rates

€

Yi Z , A , t( )/e spectra

€

Sν ,i Z , A , Eν ( )

Two distinct studies

€

Nν (Eν , t ) = Yn Z , A , t( ) ⋅

n

∑ Sν ,n Z , A , Eν ( )

fissilemat. + FY

neutron flux

Core geometry nuclear

database

exp. spectrum

models

- e conversion : pas branch by branch method : no additional errorNeutron capture taken into account Long lived fission products accumulation Error treatment and propagation Nuclear database tests


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The e spectra formulation

Coulombcorrections

Phase spaceSpectral Shape factor(Well controlled for

allowed and forbidden unique

transitions)

€

P(E0

i, E

ν ) = F(Z , E) ⋅ pE(E0

i − E)2 ⋅Sν (E)with

depends on the transition : Branching Ratios, End-

Points,spin, parity of the mother and daughter nuclei

branching ratios individual spectra

S,n (Z,A, E) = bn,i(E0) . i

i P(E0, E)

i -- -

Remaining short-lived, high Q, unknown nuclei


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Bestiole Database

(ROOT and ASCII formats)

• Collect all available information : Nuclear Database : ENSDF

Experimental spectra 111 nuclei @ISOLDE [O. Tengblad et al., Nucl. Phys. A, 503, (1989)]

• 950 nuclei : ~ 10000 branches ~ 500 -n branches

• Tag all relevant information : Forbiddenness (spin & parity) Level of approximation

CEA/Saclay/SphN : D. Lhuillier, Th. Müller et al.


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MURE **MCNP Utility for Reactor Evolution, O. Méplan et al. ENC Proceedings (2005) Developed by CNRS/IN2P3/IPNO and LPSC

Geometry

Evolution

•Open source code : adapted to antineutrino needs (simple geometry

implementation, easy coupling to databases …) •Benchmarked with APOLLO2 code •Fuel Burnup •Fission product distributions •Refined effects : out of equilibrium spectra • Neutron capture on FPs …

proliferation scenario calculations

Possibility to simulate :

• Simple cases : pure U or Pu isotope fissions and associated spectra• Complexe cases : reactor and associated spectra


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Obtained spectra

X 4% agreement in the range 2 -6 MeV with Schreckenbach’s data , higher discrepancy at high energyX simulation errors (cf. Th. Müller’s talk) within the experimental error bars

Données :[A. Hahn et al., Phys. Lett. B, 218, (1989)]

Rudstam data + Bestiole + JENDL + Qb

Rudstam data + Bestiole

Ratio : (MURE - DATA)/DATA


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Several tracks to solve these problems :

- Use of Gross Theory in existing databases such as JENDL3.3 when the considered nuclei have been treated, other models… (QRPA)

- Inclusion of existing TAGS nuclear data and new measurements

- Sensitivity to fission yields databases

Origins of the discrepancy and solutions

- ?

ZAN

Z+1AN-1

γ

γ

γ2γ1

• Pandemonium effect : use of Ge detectors to measure the decay schemes : underestimate of b branches towards high energy excited states : overestimate of the high energy part of the FP b spectra

• Unknown fission products contribute importantly at E>5-6MeV

- an alternative : the « ratio method » (see Th. Müller’s talk) : relying on the very precise Schreckenbach’s team 235U and 239Pu b spectra measurements + corrections to be applied for time evolution and neutron capture


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decay database testing and fission yields

Treatment of Pandemonium with JENDL3.3 : Gross Theory spectra

Rudstam et al. dataJENDL totalJENDL exp. (ENSDF)BESTIOLE exp. (ENSDF)BESTIOLE beta-n branches (ENSDF)

142Cs83As 85As 89Br

Different databases for fission yields : important input for MURE simultion :

Yields from JENDL3.3

Yields from ENDFB

Yields from JEFF31

With beta spectra from :BESTIOLE + JENDL Gross Theory + Q approx.

Ratio : (MURE - DATA)/DATA

235U

[K. Takahashi and M. Yamada 1969]

-> comparer JENDL avec JEFF3 et ENDFBVII

M. Fallot et al. ND2007, L. Giot et al. Physor 2008


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Evolution of the spectrum shape with time

235U under monoenergetic n flux (no moderation) : evolution during 1 year

To be studied in the reactor framework : under realistic n spectrum

Bin per bin comparison with respect to 0.7day -spectrumfor 60 time steps during 1 year

0 2 4 6 8 10 12

Energy (MeV)


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World-wide initiatives

238U - spectrum integral measurement in Münschen (Niels Haag et al. @Garching ) : March-April 2009 (now !!!)

235U/239Pu ratio measurement in Moscow @Kurchatov institute (V. Kopeikin et al.)

Measurements of Pandemonium nuclei : Total Absorption Spectrometry collab. (Valence group)

D. Jordan, PhD Thesis

TAS measurements @ Univ. Jyvaskyla

Using large 4 scintillation detectors, aims to detect the full -ray cascade rather than individual -rays

12 BaF2 Crystals

New Surrey-Valencia Total Absorption Spectrometer


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Antineutrino energy spectrum and flux reactor simulations,

« gross » proliferation scenarios


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Scenarios and reactors of interest for IAEA ?

PWRs

BWR, FBR, CANDU reactors

Research reactor/isotope production reactors Pth >10MWth

Future reactors (PBMRs, Gen IV reactors, ADS, especially reactors using carbide, nitride, metal or molten salt fuels.

PWRs : PWR (full core) simulation and simple refuelling scenario studies

BWR, FBR, CANDU reactors : channel simulation, simplistic refuelling scenario study

Research reactor / isotope production reactors Pth >10MWth : started OSIRIS simulation for Nucifer, and high flux ILL reactor simulation

Future reactors (PBMRs, Gen IV reactors, ADS, especially reactors using carbide, nitride, metal or molten salt fuels.

An antineutrino measurement is directly related to the fission process in the reactor core.

An antineutrino measurement can provide in real time information on isotopic fission rates, which can be related to the thermal power and fissile inventory of the reactor.

International expert meeting organized by the Department of New Technologies of IAEA, October 26-28. 2008 :


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Chooz-B reactors (I, II)

• 2 core N4 series : 4.27 GWth

• 2 core N4 series : 4.27 GWth

13 m

4,5 m

Moderator/coolant

‣ pressurized borated water (155

bars)560 K < TH2O < 620 K

(ρ= 0.7 g.cm-3 )

Moderator/coolant

‣ pressurized borated water (155

bars)560 K < TH2O < 620 K

(ρ= 0.7 g.cm-3 )

3,8 m

4,8 m

Fuel

‣enriched UO2 pellet : 1.8, 2.4 and 3.1

%

(ρ= 10.85 g.cm-3 )

700 K < TUO2 < 1400 K

Fuel

‣enriched UO2 pellet : 1.8, 2.4 and 3.1

%

(ρ= 10.85 g.cm-3 )

700 K < TUO2 < 1400 K

9 mm 13 mm


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PWR N4 : Chooz-B (I, II)-like reactor

Fuel element

‣ 317 pellets / h = 4.2 m‣ Zircaloy coating/ 0.6 mm‣ 3 Enrichments : (1.8, 2.4, 3.1 %.)

12.6 mm

12.6 mm

‣ Zircaloy structure‣ 24 ‘guide’ tubes (poison, instrumentation,..)

214 mm

214 mm

Assembly

264

Core

‣ 3(4) enrichment zones‣ Refueling 1/3(4) 11 months

205

PWR : full core simulation. Approx. : No control rod reactor driving yet : constant power, Boron diluted into water and mean keff =1

PWR : full core simulation. Approx. : No control rod reactor driving yet : constant power, Boron diluted into water and mean keff =1


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Neutronics inputs

Systematic effects : ✓ Temperature : Thermalisation & Doppler effect

✓ Neutron Absorbant : Bore

Systematic effects : ✓ Temperature : Thermalisation & Doppler effect

✓ Neutron Absorbant : Bore

• A matter of neutronics : neutron flux & interaction cross sections (n,x)

• A matter of neutronics : neutron flux & interaction cross sections (n,x)

Fast neutronsFission spectrum

~2MeV

Fast neutronsFission spectrum

~2MeVSlow

neutronsthermalisation

~0.025eV

Slow neutrons

thermalisation ~0.025eV

Epithermal domain

+moderator 1eV<En<1MeV

Epithermal domain

+moderator 1eV<En<1MeV

‣ Φ ~ 3,5 .10 14 n.cm-2.s-1 | <En> ~ 0.7 MeV‣ Φ ~ 3,5 .10 14 n.cm-2.s-1 | <En> ~ 0.7 MeV

Burnup effect

Burnup effect

σ(n,f) 238U

σ(n,γ) 238U

σ(n,f) 235U

σ(n,γ) 235U


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‣Thermal bump displacement :

Systematic effects • Thermalization :• Thermalization : • Doppler effect :• Doppler effect :

0 K

1800 K

• Boron Cross sections 1/v• Boron Cross sections 1/v

‣ Harder neutron spectrum

Tfuel ➚ ⇒ Resonant captures➚

• Criticity control with soluble boron :

‣ Cbore adjusted at each time step t , <keff (t+1) >= 1

• Criticity control with soluble boron :

‣ Cbore adjusted at each time step t , <keff (t+1) >= 1


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Systematic effects

Guessed uncertainties on input parameters

Inventory : ΔN (%) Nfission : Δ(Nf/s) (%)

235U 238U 239Pu 241Pu 235U 238U 239Pu 241Pu

ΔTm = 30 K 0.5 < 0.1 1. 0.8 0.4 < 0.1 < 0.1 1.5

ΔTf = 200 K 1. < 0.1 2. 2.5 1. 0.5 0.8 1.

ΔcB = 200 ppm 0.7 < 0.1 2. 1.8 2.4 2. 0.5 0.8

Δ(mean keff=1 or keff=1)

0.2 4.4 0.8 2.2

• Systematic study : ‣ T moderator : 300, 600,... 1200K‣ T fuel : 300, 600,.... 1500 K‣ Boron Concentration : 0, 500,... 3000 ppm mass.

• Systematic study : ‣ T moderator : 300, 600,... 1200K‣ T fuel : 300, 600,.... 1500 K‣ Boron Concentration : 0, 500,... 3000 ppm mass.

n, En, keff(<>, Inventory, Nf/s.

n, En, keff(<>, Inventory, Nf/s.

Also studied : - self-shielding effect on inventories and fission rates, - influence of the Monte-Carlo seed

Studied effects : of the order of 2% or

lower on main fission rates (exc. 238U)


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PWR refueling simulation

Constant power : 4.27GWth

Boron concentration 1000ppmMean keff = 1

PWR refuelled every year :250kg 239Pu retrieval900kg 235U adjunction

Constant power simulation of N4 PWR

6.4%

Folded by Nucifer response @25m :

Preliminary

Antineutrino rate/s in Nucifer


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Filling all control rods with 238U

25*205 control rods :

+11.5t 238U=> 238U mass

increase of 10%

Folded by Nucifer response @ 25m :

235U : +150kg239Pu :+100kg

235U and 239Pu fission rates change by -5.5% and +2.5% resp.

Antineutrino flux :Change of 0.5% !

Because of 238U fission rate increase +6-8% :

mimick 235U

238U inventory

235U and 238Pu inventory

Fission rates

Antineutrino rate/s in NuciferPreliminary

PreliminaryPreliminary

Preliminary


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Removal of the control rods

Corresponds to 65kg 239Pu removal, without changing 238U and 235U masses as rods are replaced by fresh Unat

Amount of 238U stays ~ the same, so no compensation by 238U of 235U and 239Pu fission

rates variations

Preliminary

1.3%

Normal refuelling

Replacement of the control rods


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CANDU reactor specifications(CANada Deuterium Uranium)

http://canteach.candu.org

Heavy water as moderator and coolant and natural uranium as fuel

Spatial separation of coolant (in force tubes) and moderator (between force tubes)

On-line refuelling :

Plutonium proliferation

Antineutrino flux and spectra : refuelling and 239Pu proliferation scenarii

Calandria

moderator

Force tubesV.M. Bui PhD, Collaboration with A. Nuttin (LPSC)

(*)A. Nuttin, Physor-2006, Study of CANDU Thorium-based Fuel Cycles by Deterministic and Monte Carlo Methods.


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Simulation inputs

A channelA bundle

Fresh fuel

1 32 Pool

4 5 1’

Irradiated fuel

Moderator

Mirror

Annular gas CO2

coolant

Calandria tube

37pins

Force tube

Refuelling of a channel 2/3 fresh fuel and 1/3 irradiated fuel

Different simulation steps :

Force tube →channel of 12 fuel bundles

1 bundle + mirror → full reactor, homogeneous, infinite (no leak)

Fit boundary mirror dimensions to obtain the CANDU moderation ratio

Temperature dependance : Tmod= 300 K Tcool= 600 K Tfuel= 1200 K

Spatial dependance of the neutron flux

Dwell time def. : threshold 1.05 (A. Nuttin* et al.) : leaks : 3000pcm, absorptions (Boron and impurities) : 2000pcm.

T= 300 K for all components

)(

)1(

tN

tNK

neutrons

neutrons +=∞

Correct mean temperatures

Dwell time : 200d


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Channel simulation & gross diversion scenario

3 channels (12 bundles) simulations with 100d, 200d and 300d refueling periods

Principle : refuel faster some channels to take Pu away and refuel slower the same number to mask diversion

Isotopic Vector Plutonium: Pu of military quality (VP>90%)

2 “full core” refuelling scenarii

- Standard : 400 channels refueled @200 days

- Proliferant : 200 channels refueled @ 100d + 200 channels refueled @ 300d to mask diversion

X-checks : collaboration with A. Nuttin et al., Physor 2006 Conf. Proc. , comparison between MURE and deterministic code DRAGON

Inventory @ refuelling period 100d




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A gross proliferation scenario e

in N

ucife

r (H

z)F

its a

nd s

ums

Folded with Nucifer response @ 25m for 1 channel with different refueling periods :

100d 200d 300d

0 100 200 300 4000 100 200 300

Time (days)

0 100 200 300 400

Time (days)Time (days)

400 channels (~ CANDU600-like) refueled every 200d, @ 2 channels per day : adequation between dwell time, daily number of refuelled channels and total number of channels : flat profile of the flux

« Normal refueling » :

+…+

Adding 400 channels with history shifted by 1 day

0 100 200 300 400

0.02588 Hz => 2236 per day in Nucifer

Preliminary

Time (days)


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Core refuelled @ 100d and 300d

Folded by Nucifer response @ 25m :

200 channels refuelled every 100d, 200 channels every 300d, @ 2 channels per day

3.5% discrepancy in the antineutrino rate !!!Nucifer reaches 1% statistic precision after 1 day

3.5% discrepancy in the antineutrino rate !!!Nucifer reaches 1% statistic precision after 1 day

Fission rates : Normal refuelling : 54.1% 235U - 45.9% 239PuDiversion scenario : 60,1% 235U - 39,9% 239Pu

Time (days)

0 100 200 300 400

+660kg 235U after 200x100d + 200x300d-430kg 239Pu including 152kg from 100d channels

0.02497 Hz => 2157 per day

0 100 200 300 400

Time (days)

0.02588 Hz => 2236 per day

+600kg 235U after 400x200d-464kg 239Pu


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Conclusions and outlooks

A set of performant tools (MURE+BESTIOLE+databases…) to compute the antineutrino energy spectrum and flux under various conditions : experiment and reactor simulations

Neutron capture and long-lived fission products contributions to the spectrum : need to be evaluated carefully and compared with Kopeikin’s et al. results

Simulation of different kinds of reactors, including Chooz-B ones for the Double Chooz experiment

First gross proliferation scenarios studied, with PWR and CANDU reactors, including the response of the Nucifer detector placed at 25m from the cores

Nucifer sensitivity in 239Pu content : ~+-65kg <=> 1% change in the measured antineutrino flux by Nucifer @25m of a PWR core,

1% statistical accuracy reached in 4days by Nucifer in these conditions


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Backup Slides

march 19. 2009 aap09 - brazil - m. fallot et al. collaborations double chooz, nucifer and mure time...

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