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Nuclear data required for measurements of reactivity and nuclear material composition Nuclear Technology Research Laboratory Central Research Institute of Electric Power Industry Yasushi NAUCHI 2018 Symposium on Nuclear Data Nov. 30, 2018 at

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Central Research Institute of Electric Power Industry

Nuclear data required for measurements of reactivity and nuclear material composition

Nuclear Technology Research Laboratory

2018

2018 Symposium on Nuclear Data

  • Nov. 30, 2018 at Tokyo Institute of Technology

Yasushi NAUCHI

1

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SLIDE 2

Introduction

Neutron multiplication is the most important process in a system where U or + Pu exits. Neutron multiplication factor : keff

  • Expected number of fission neutron in the next generation.
  • keff = kinf x PNL

 kinf: infinite multiplication factor ~ by composition  PNL: Neutron non-leakage probability ~Mainly by geometry

Reactivity: r = 1 – 1/keff

  • Sometimes Dr is called reactivity.

 Additivity of Dr is known

keffis calculated accurately (uncertainty << 0.5%） provided accurate info. on geometry +material.

2018 2018/12/3 2

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Introduction ~ cont.

Except for keff =1, keff is in-directly measured quantity. Near critical situation, core power increase or decreases with et/T. T is converted to r & keff with calculated kinetic parameters. In deep sub-critical system, indices are measured.

  • a = dP/dt / P provided delayed-n can be neglected
  • g2: Square of spatial decay constant of neutron flux
  • ksub = neutron yield ratio (induced fission )/Total

To relate a, g2, ksub, etc. to r & keff, calculated quantities are more or less, necessary. Info. on geometry +material should be known for the

  • calc. ~ Which is accurate, calc. or experiment?

2018 2018/12/3 3

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Introduction

Among the reactivity measurement issues, items below are focused on.

  • Kinetic parameters to obtain r.
  • Subcritical multiplication factor by passive gamma

measurement

  • Neutron induced gamma ray spectroscopy (NIGS) for

determination of negative reactivity.

2018 2018/12/3 4

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Point kinetics for near critical case

2018 2018/12/3 5

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Effective delayed neutron fraction per precursor, beff,ij & the generation time L are called kinetic parameter r is deduced from Reactor period T =P/(dP/dt) with calculated beff,ij & L In the definition of delayed neutron emission data have significant role.

6

T Sum T Sum T

ij ij eff ij ij ij eff ij in

 b  b r     L  1 1

, ,

Point kinetics

In reactivity measurement, this term must be small

   

    

     

i in in in in i i f i t i t ij in in in in i i f ij d ij d ij ij eff eff

dE d E dE d E Ω r Ω r , , , ,

, , , * , , , * ,

          b b

Decay const. of precursor. Listed in Data Files Delayed neutron spectrum & emission number, Listed in Data Files

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SLIDE 7

In CROCUS cores, T is measured for several cases with the critical & perturbed geometries. rin is deduced based on measured T with the listed ij and calculated beff,ij . They are compared to rdir=D(1/keff) for the two geom.

2013 7 2013 7

Reactivity by raising level by 1.5~3cm Withdrawal of poison: B-water, Er2O3 Critical water level Perturbed water level U metal

235U

0.95wt% UO2

235U

1.81wt% Poison Critical water level

Validation of sets if beff,ij + ij

J.M.Paratte et al., ANE33(8)739-748,2006.

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SLIDE 8

Light water moderated LEU, 0 < r <162 pcm Comparison varying nuclear data

  • JEFF-3.1.1, JENDL-4.0=good, ENDF/B-VII.0 worse
  • Difference of rin / rdirdue to nuclear data > that of beff

8

beff by JEFF- 3.1.1 beff with ENDF/B-VII.0 beff with JENDL-4.0 critical 758.5 737.4 743.1 Perturbed geometry JEFF-3.1.1 rin/rdir ENDF/B-VII.0 rin/rdir JENDL-4.0 rin/rdir H2 1.025 0.842 0.971 H3 1.035 0.846 0.993 H4 1.028 0.855 0.979 B withdrow 1.048 0.865 1.017 Er withdrow 0.999 0.860 0.984

Validation ~rin and rdir in CROCUS

Zoia, Nauchi, et al., ANE 96(2016)377-388

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SLIDE 9

Differences in beff and rin repartition per precursor family is found between JENDL-4.0 & ENDF/B-VII.0. In the range of reactivity measurement, accuracy of the 2nd group of 235U is most important.

9 9

50 100 150 200 250 300 235U_1 235U_2 235U_3 235U_4 235U_5 235U_6 238U_1 238U_2 238U_3 238U_4 238U_5 238U_6 Comparison of effective delayed neutron fraction repartition per fission nuclide and precursor grouop ENDF/B-VII.0 JENDL-4.0 Effective delayed neutron fraction repartition per nuclide and precursor (pcm) beta_eff component 10 20 30 40 50 60 70 L 235U_1 235U_2 235U_3 235U_4 235U_5 235U_6 238U_1 238U_2 238U_3 238U_4 238U_5 238U_6 Comp mpari rison of f re reactivity r rep epart rtition n per r fi fission n nuc uclide e and p prec recurs rsor r group up for r CR CROC OCUS US H4 H4 g geomet metry ENDF/B-VII.0 JENDL-4.0 Reactivity component (pcm) reactivity component

T

ij ij eff

 b  1

, ij eff ,

b

Important component

ANE 96(2016)377-388

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SLIDE 10

Current status of nuclear data

It is reported that covariance data in j, d,j brings 6% error for rin.

  • JENDL&JEFF are essentially based on Keepin’s data
  • Re-measurements and re-evaluation are preferable.

Extensive validation with integral experiments

  • CRIEPI+CEA’s studies for FUBILA MOX.

Common set of ij for every actinide as in JEFF-3.1.1 (8 groups) is preferable. For negative rin, the difference

  • f minimum ij is very sensitive in irradiated fuel.

2018 2018/12/3 10

A.D.Santos, Proc. PHYSOR2018 Half life (s) of longest precursors in JENDL-3.3 Nauchi, Proc. PHYSOR2018

234U 235U 236U 238U 237Np 238Pu 239Pu 240Pu 241Pu 242Pu

52.91 55.72 51.80 52.39 56.82 52.12 54.28 52.12 54.15 50.97

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SLIDE 11

Comparison of T-r curve

For FUBILA MOX core T-r curves of JENDL-4.0u (6- precursor groups) and JEFF-3.1.1(8-g) is compared Reactivity difference is enhanced for the large negative reactivity.

  • Evaluation of d,i1 & i1 and 6 & 8 groups are significant

2018 2018/12/3 11

0.9 0.95 1 1.05 1.1

  • 200
  • 150
  • 100
  • 50

50 100 150 200

ratio of deduced reactivity 8g/6g reactivity (pcm)

vertical: reactivity(8g) / reactiviy(6g)

T1/2 of the 1st group is 55.71s for JENDL-4.0u & 55.599s for JEFF-3.1.1

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Passive measurement for ksub

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Sub-critical multiplication factor

In sub-critical system, neutron emission is done by

  • uter source (Nprim) + induced fission (N2nd).

ksub = N2nd / (Nprim + N2nd) Yield ratio of g ray to neutron, (G/N) is different for radio active decay (prim) & induced fission (2nd). Total yield ratio: (G/N)=(G/N)prim(1-ksub)+(G/N)2ndksub To measure the total yield of g ray and neutrons might give ksub.  Analogously, Yield ratio of (Xe, Kr) is considered to give ksub in 1F-1,2,3.

2018 2018/12/3 13

Nauchi, Proc. PHYSOR2002

  • Y. Naito P2012-133052, 2013
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Parameters for spent FA

For PWR spent fuel assemblies, (G/N) is studied.

  • ORIGEN-2 +ORLIB-J32 lib for composition
  • Composition + Original ORLIB for g ray
  • MCNP-5 calc with ENDF/B-VI.6 lib.

for 2nd. Short lived FP g ray is taken into account with FPGS-90 (JNDC v2)

  • As BU increases, prim is dominated

by 244Cm & (G/N)prim becomes stable.

  • (G/N)2nd varies with BU due to the

change of the dominant fissile 235U=> Pu.

  • (G/N)prim separates from (G/N)2nd

ksub by G/N is owe to (G/N)prim, 2nd.

  • Verbeke’s evaluation, not so separated, large uncertainty.

2018 2018/12/3 14

0.00 0.02 0.04 0.06 0.08 0.10 0.12 10 20 30 40 50 60 (G/N)prim (G/N)2nd (G/N) (G/N) Assembly burn-up (MWd/kgHM)

CRIEPI Research Report T03063, 2003

g ray >4MeV 244Cm sp.fis 235U-fis 2.57±0.30 2.79±0.31

UCRL-AR- 228518-REV-1

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SLIDE 15

Source multiplication method by g ray

g ray ( >3MeV) were measured in KUCA. ksub estimation by Verbeke’s fission g ray yield of 252Cf spontaneous &

235U fis. + JENDL/FPD-2000 is good,

although overwrap of Al(n,g) is and neglected Accuracy of (G/N) for 239,241Pu &

244Cm + uncertainty-reduction

as well as JENDL/FPD&FPY-2011

2018 2018/12/3 15

BGO 3”x3”

geometry 2x1.5 2x2 3x2 4x2 Ksub by g ray 0.699 0.787 0.898 0.958 Ksub by calc. 0.726 0.792 0.897 0.956 C/E 1.039 1.006 0.999 0.998

100 101 102 103 104 2 4 6 8 10 4x2 geom g ray from source 252Cf count rates(cps) g ray pulse height (MeV) Al+Fe capture g ray H(n,g) Gprim+2nd fission g ray

252Cf prim. fission g ray

4x2 3x2 2x2 2x1.5 Cf

KURRI progress report 2009

UCRL-AR- 228518-REV-1

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SLIDE 16

Neutron Induced Gamma ray Spectroscopy

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Reactivity in Un-known material

Current measurement technique of keff & r is based

  • n the fact that the material composition is within a

predictable range. For unknown material represented by fuel debris accumulated in 1F-1,2,3, estimation of material (isotope level) is essentially required. Material composition near surface of the fuel debris is available by novel methods like LIBS, but depth information is essential in criticality problem. In CRIEPI, efficient storage of fuel-debris canister is focused on by measuring reactivity.

2018 2018/12/3 17

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Issues in Storage Volume

Fuel debris might be stored number of canisters Large space is required for them

  • Initial 235U enrichment increases .

Narrower canisters and looser storage of canisters for criticality Safety

  • Large volume of structural materials below

active part of a core were caught in melt fuel

CR, guide, shaft, mechanism, vessel, core support

Unless they are separated & distinguished from fuel, they have to be treated as “fresh fuel(5%EU)”.

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looser narrower CR guide tube Drive shaft & mechanism Active part

  • f core

Core support canister Vessel

  • Y. Nauchi et al. JNST52(7-8)1074-1083,2015
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Concept of Capture Credit (CapC)

Without any information on stored debris, the infinite multiplication factor of the canister should be estimated as With a measured data of reaction rate ratio of capture i to the total fission <Si,c>/<Sf>, the factor is estimated as Denser storage within the reactivity is OK= taking CapC

2018 19

  • pt

f

  • pt

c

  • pt

f t EU

k     S  S S 

 % 5 ,

N-production N-absorption fresh fuel, optimum geom.         S S  S  S S 

  a    

f c i

  • pt

f

  • pt

c

  • pt

f t i

k

, ,

1 a: 0<a<1, taking account of spectrum perturbation, measurement & calc error, extra safety margin, etc.

   a r

f c i t i EU

k k S S   

  , , % 5 ,

1 1

  • Y. Nauchi et al. JNST52(7-8)1074-1083,2015
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SLIDE 20

NIGS

Some neutron induced reactions radiate g rays

  • Capture: prompt, discrete energy, in SS, 7~10MeV
  • Fission prompt: continuum spectrum
  • Short lived FP 3~4.5MeV
  • De-excited g ray followed by emission of nucleon

☺Their energy is higher than FP g rays.

For CapC, target are fission and capture g rays. Also, 238U is focused on for determining fission / 238U capture for BUC.

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NIGS is very resemble to PGAA. But, in our experiment, g ray from short-lived FP is not excluded. Then the term NIGS is used for including PGAA+DGA.

Neutron source spectrometer Collimator Canister g ray Reaction

  • Y. Nauchi et al. JNST52(7-8)1074-1083,2015
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NIGS for Simulated Fuel Debris in KUCA-C

Simulated fuel in light water driven by 252Cf

  • debris = SS (red) +U-Al-fuel (green)

 g rays were measured by BGO As increment of SS loading ratio, capture(6-10MeV) components increase and fission component (3-5MeV) decreases.

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U: max 1.7kg

235U: max1.6kg

0.1000 1.000 10.00 100.0 1000 104 105 2 4 6 8 10 12

BGO GO: Gamma ray s spect ctrum m (no not-Final v ver ersion) n) U-A

  • Al & SUS

US plates es mixed ed sub- ub-cr critical c core e

SUS90-U90 SUS60-U120 SUS45-U135 SUS30-U150 SUS0-U180 Pulse height spectrum (count/MeV/ch) Eγ (MeV) SUS60-U120 should be revised since gain-shift is found 252Cf decay correction are done for August & December experiments

  • Y. Nauchi et al.

JNST52(7-8)1074- 1083,2015

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SLIDE 22

Count rate ratio & <c,i>/<f>

Directly measured data and reaction rate ratio

  • Count rate ratio (black dot) of (6-10MeV) / (3-5MeV)

monotonously increases with is (SS-capture) / fission reaction rate ratio.

  • CapC by NIGS is feasible

 CapC of 157Gd is also possible

For CapC by NIGS, g ray emission data should be well evaluated

  • In CapGam, no absolute data

for 53Cr & 17% error for 56Fe.

  • 155Gd data for Eg > 4MeV is not available in CapGam.
  • Yield uncertainty of fission prompt g by Verbeke ~ 12~15%

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0.000 0.2000 0.4000 0.6000 0.8000 1.000 0.2 0.4 0.6 0.8 1 1.2

Measured ed g gamma ray em emission n and nd co count unt r rate d data against nst c calcul culated SUS c captur ure / /fi fission. n.

count rate ratio gamma emission ratio [count rate ratio] or [gamma ray emission ratio] (6-10MeV)/(3-10MeV) SUS captre/fission (calculation) Primary gamma ray from 252Cf is not eliminated from experimental data

  • Y. Nauchi et al.

JNST52(7-8)1074- 1083,2015

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NIGS for residual fissile enrichment

Determination of burn-up depletion of fissile is beneficial for efficient storage of fuel-debris. Fission prompt g can be measured. How to scale? 238U(n,g) is major reactions in fuel.

  • NIGS for 238U is of use to determine

residual enrichment

  • 4060 keV g ray is targeted.
  • Emission of the g ray is not

modeled in FXSLIB-J40.

NIGS in KUCA-A core

2018 2018/12/3 23

HP-Ge of 20%effici.

93%EU (U-Al) Nat-U metal Polyethylene Al sheath

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SLIDE 24

238U capture g and DGA

238U(n,g) 4060.6keV is identified 351.3count (±15.4%)

  • 90Rb(4061.7keV) contamination is at most 35 counts .

Short lived FP g rays were measured simultaneously.

  • Possibility of DGA for

235U:239Pu:241Pu

As for 238U(n,g), variation

  • f g ray emission / capture

between each resonance-level & thermal reaction is known, although it has not yet well evaluated.

  • To measure the other g rays (3583.1, 3982.8, 3991.4) gives

neutron spectrum info.

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400 600 800 1000 3000 3200 3600 4000 4400 4800

count rate (/ch//2.5h)

pulse height (keV) 88Br 16s 90Rb 158s 238U capture 91Rb 58s 95Y 10min 97Y 3.8s 136Te 17.5s C de-excitation followed by in-elastic scattering

KURRI progress report 2017

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Summary

2018 2018/12/3 25

Property Nuclide Purpose dij, dij, ij

235,238U,239,241Pu. For minor nuclide,

do not evaluate smaller 1. T-r curve Fission prompt g ray spectrum + yield Spontaneous: 244Cm, 252Cf Induced: 235U, 239, 241Pu (G/N) for ksub CapC by NIGS, Enrichment (BUC) by NIGS FP yield: Kr, Xe short lived Spontaneous: 244Cm, Induced: 235U, 239, 241Pu Criticality monitor in 1F (G/N) for ksub BUC by NIGS (DGA), Capture g ray emission thermal is prioritized SS materials: Fe,Cr,Ni,Mn

155, 157Gd, Si,

CapC by NIGS, Capture g ray emission, resonance & thermal

238U, 240Pu

BUC (Enrichment) by NIGS g ray emission followed by nucleon emission O, Zr Characterization for Fuel debris by NIGS