Nuclear data required for measurements of reactivity and nuclear - PowerPoint PPT Presentation

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 Tokyo Institute of Technology 2018 1

Introduction  Neutron multiplication is the most important process in a system where U or + Pu exits.  Neutron multiplication factor : k eff  Expected number of fission neutron in the next generation.  k eff = k inf x P NL  k inf : infinite multiplication factor ~ by composition  P NL : Neutron non-leakage probability ~Mainly by geometry  Reactivity: r = 1 – 1/k eff  Sometimes Dr is called reactivity.  Additivity of Dr is known  k eff is calculated accurately (uncertainty << 0.5% ） provided accurate info. on geometry +material. 2018 2018/12/3 2

Introduction ~ cont.  Except for k eff =1, k eff is in-directly measured quantity.  Near critical situation, core power increase or decreases with e t/T . T is converted to r & k eff with calculated kinetic parameters.  In deep sub-critical system, indices are measured.  a = dP/dt / P provided delayed-n can be neglected  g 2 : Square of spatial decay constant of neutron flux  k sub = neutron yield ratio (induced fission )/Total  To relate a , g 2 , k sub , etc. to r & k eff , 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

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

Point kinetics for near critical case 2018 2018/12/3 5

Point kinetics  Effective delayed neutron fraction per precursor, b eff,ij & the generation time L are called kinetic parameter  r is deduced from Reactor period T =P/(dP/dt) with calculated b eff,ij & L b b L r    eff , ij eff , ij Sum Sum Decay const. of     in ij ij T 1 T 1 T ij ij precursor. Listed in Data In reactivity measurement, this term must be small Files  In the definition of delayed neutron emission data Delayed neutron spectrum & emission have significant role. number, Listed in Data Files            Ω * , E , d dE r d , ij d , ij f , i i in in in in  b  b  ij     eff eff , ij        Ω * , E , d dE r ij t , i t , i f , i i in in in in i 6

Validation of sets if b eff,ij +  ij  In CROCUS cores, T is measured for several cases with the critical & perturbed geometries.  r in is deduced based on measured T with the listed  ij and calculated b eff,ij .  They are compared to r dir = D (1/k eff ) for the two geom. J.M.Paratte et al., ANE33(8)739-748,2006. Critical Perturbed U metal water water level 235 U level 0.95wt% Critical water Poison level UO 2 2013 7 235 U 1.81wt% Reactivity by raising level by 1.5~3cm Withdrawal of poison: B-water, Er 2 O 3 2013 7

Validation ~ r in and r dir in CROCUS  Light water moderated LEU, 0 < r <162 pcm Zoia, Nauchi, et al.,  Comparison varying nuclear data ANE 96(2016)377-388 b eff by JEFF- b eff with b eff with 3.1.1 ENDF/B-VII.0 JENDL-4.0 critical 758.5 737.4 743.1 Perturbed JEFF-3.1.1 ENDF/B-VII.0 JENDL-4.0 r in / r dir r in / r dir r in / r dir geometry 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  JEFF-3.1.1, JENDL-4.0=good, ENDF/B-VII.0 worse  Difference of r in / r dir due to nuclear data > that of b eff 8

Important component  Differences in b eff and r in repartition per precursor family is found between JENDL-4.0 & ENDF/B-VII.0. Comp mpari rison of f re reactivity r rep epart rtition n Comparison of effective delayed neutron fraction repartition per r fi fission n nuc uclide e and p prec recurs rsor r group up per fission nuclide and precursor grouop for r CR CROC OCUS US H4 H4 g geomet metry 300 70 ANE 96(2016)377-388 repartition per nuclide and precursor (pcm) ENDF/B-VII.0 JENDL-4.0 ENDF/B-VII.0 Effective delayed neutron fraction 60 250 JENDL-4.0 b Reactivity component (pcm) b eff , ij 50 200 eff , ij   40 1 T ij 150 30 100 20 50 10 0 0 235U_1 235U_2 235U_3 235U_4 235U_5 235U_6 238U_1 238U_2 238U_3 238U_4 238U_5 238U_6 9 235U_1 235U_2 235U_3 235U_4 235U_5 235U_6 238U_1 238U_2 238U_3 238U_4 238U_5 238U_6 L reactivity component beta_eff component  In the range of reactivity measurement, accuracy of the 2 nd group of 235 U is most important. 9

Current status of nuclear data  It is reported that covariance data in  j ,  d,j brings 6% error for r in . A.D.Santos, Proc. PHYSOR2018  JENDL&JEFF are essentially based on Keepin’s data  Re-measurements and re-evaluation are preferable.  Extensive validation with integral experiments Nauchi, Proc.  CRIEPI+CEA’s studies for FUBILA MOX. PHYSOR2018  Common set of  ij for every actinide as in JEFF-3.1.1 (8 groups) is preferable. For negative r in , the difference of minimum  ij is very sensitive in irradiated fuel. Half life (s) of longest precursors in JENDL-3.3 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 2018 2018/12/3 10

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 vertical: reactivity(8g) / reactiviy(6g) 1.1 ratio of deduced reactivity 8g/6g 1.05 1 T1/2 of the 1st group is 55.71s for JENDL-4.0u & 0.95 55.599s for JEFF-3.1.1 0.9 -200 -150 -100 -50 0 50 100 150 200 reactivity (pcm)  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

Passive measurement for k sub 2018 2018/12/3 12

Sub-critical multiplication factor  In sub-critical system, neutron emission is done by outer source (N prim ) + induced fission (N 2nd ).  k sub = N 2nd / (N prim + N 2nd )  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-k sub )+(G/N) 2nd k sub  To measure the total yield of g ray and neutrons Nauchi, Proc. might give k sub . PHYSOR2002  Analogously, Yield ratio of (Xe, Kr) is considered to give k sub in 1F-1,2,3. Y. Naito P2012-133052, 2013 2018 2018/12/3 13

Parameters for spent FA  For PWR spent fuel assemblies, (G/N) is studied. CRIEPI Research  ORIGEN-2 +ORLIB-J32 lib for composition Report T03063, 2003 g ray >4MeV  Composition + Original ORLIB for g ray (G/N)prim 0.12 (G/N)2nd (G/N)  MCNP-5 calc with ENDF/B-VI.6 lib. 0.10 for 2 nd . Short lived FP g ray is taken into 0.08 (G/N) account with FPGS-90 (JNDC v2) 0.06  As BU increases, prim is dominated 0.04 by 244 Cm & (G/N) prim becomes stable. 0.02 0.00  (G/N) 2nd varies with BU due to the 0 10 20 30 40 50 60 Assembly burn-up (MWd/kgHM) change of the dominant fissile 235 U=> Pu. UCRL-AR- 228518-REV-1  (G/N) prim separates from (G/N) 2nd 244Cm sp.fis 235U-fis  k sub by G/N is owe to (G/N) prim, 2nd . 2.57 ± 0.30 2.79 ± 0.31  Verbeke’s evaluation, not so separated, large uncertainty. 2018/12/3 2018 14

Source multiplication method by g ray 3x2 4x2  g ray ( >3MeV) were measured in KUCA. 2x2 2x1.5  k sub estimation by Verbeke’s UCRL-AR- 228518-REV-1 fission g ray yield of 252 Cf spontaneous & Cf 235 U fis. + JENDL/FPD-2000 is good, although overwrap of Al(n, g ) is and neglected BGO 3”  x3” geometry 2x1.5 2x2 3x2 4x2 Ksub by g ray 4x2 geom 0.699 0.787 0.898 0.958 g ray from source 252Cf 10 4 Ksub by calc. 0.726 0.792 0.897 0.956 H(n, g ) count rates(cps) Al+Fe C/E 1.039 1.006 0.999 0.998 10 3 capture Gprim+2nd g ray fission g ray 10 2  Accuracy of (G/N) for 239,241 Pu & 10 1 244 Cm + uncertainty-reduction 252 Cf prim. fission g ray 10 0 as well as JENDL/FPD&FPY-2011 0 2 4 6 8 10 g ray pulse height (MeV) 15 KURRI progress report 2009 2018 2018/12/3

Neutron Induced Gamma ray Spectroscopy 2018 2018/12/3 16

Reactivity in Un-known material  Current measurement technique of k eff & r is based on 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. 2018/12/3 2018 17