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90-Th-232 MINSK EVAL-JUN01 V.M. Maslov et al.
DIST-JAN09 20090105
----JEFF-311 MATERIAL 9040
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
*************************** JEFF-3.1.1 *************************
** **
** Original data taken from: JEFF-3.1 **
** **
******************************************************************
***************************** JEFF-3.1 *************************
Original data taken from: ENDF/B-VI.8 + New eval.
New evaluation: Maslov, Minsk. MF=12-15 from ENDF/B-VI.8
MT=458(ENERGY RELEASE IN FISSION) evaluation from ENDF/B-VI.4
05-01 NEA/OECD (Rugama) 8 delayed neutron groups
Jefdoc-976(Spriggs,Campbel and Piksaikin,Prg Nucl Eng 41,223(2002)
----B-404-ISTC MATERIAL 9040
-----INCIDENT NEUTRON DATA
-----ENDF/B-VI FORMAT
*****************************************************************
UNRESOLVED RESONANCE PARAMETERS FOR 4-150 KEV REGION,
TOTAL, ELASTIC SCATTERING, INELASTIC SCATTERING, FISSION,
CAPTURE,(N,2N), (N,3N) AND (N,4N) CROSS SECTIONS AS WELL AS
ANGULAR AND ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS WERE
EVALUATED BY V.M. MASLOV, Yu.V. PORODZINSKIJ, M. BABA,
A. HASEGAWA, N.V. KORNILOV, A.B. KAGALENKO AND
N.A. TETEREVA/1/.
MF=1 GENERAL INFORMATION
MT=451 DESCRIPTIVE DATA AND DIRECTORY RECORDS
MT=452 NUMBER OF NEUTRONS PER FISSION
SUM OF MT'S= 455 AND 456
MT=455 DELAYED NEUTRON DATA TAKEN FROM REF./2/.
MT=456 NUMBER OF PROMPT NEUTRONS PER FISSION
CALCULATED WITH THE EMISSIVE FISSION MODEL TO FIT
MEASURED DATA /3-9/, BUT BASICALLY DATA BY FREHAUT
ET AL./5/ AND HOWE /9/, IRREGULARITY AROUND(n,nf)
REACTION THRESHOLD IS REPRODUCED.
MF=2 RESONANCE PARAMETERS
MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS
(RESOLVED RESONANCE REGION = 1.0E-5 EV TO 4 KEV),
(UNRESOLVED RESONANCE REGION = 4 KEV TO 150 KEV)
1) RESOLVED RESONANCE PARAMETERS (Breit-Wigner)WERE ADOPT
FROM ENDF/B-VI AS EVALUATED BY OLSEN /10/.
2) ENERGY-DEPENDENT UNRESOLVED RESONANCE PARAMETERS COVER
ENERGY RANGE FROM 4.0 TO 150 KEV. PARAMETERS WERE OBTAINE
TO REPRODUCE SMOOTH TOTAL AND CAPTURE CROSS SECTIONS,
CALCULATED WITH STATISTICAL MODEL. ENDF/B PROCESSING
CODES /11,12/ IGNORE DIRECT INELASTIC SCATTERING.
TO COMPENSATE THAT DEFICIENCY WE INCREASED AVERAGE
INELASTIC SCATTERING WIDTHS, CAPTURE WIDTHS ABOVE 50 KEV
ALSO WAS SLIGHTLY INCREASED TO KEEP CAPTURE CROSS SECTION
UNDISTORTED AS COMPARED WITH CALCULATED BY PHYSICALLY
CORRECT CODES (PCC). CALCULATED WITH PCC CODES CROSS
SECTIONS FIT MEASURED DATA ON TOTAL, CAPTURE AND INELASTI
SCATTERING. AS A RESULT, TOTAL,ELASTIC SCATTERING AND
CAPTURE CROSS SECTIONS, CALCULATED WITH THESE PCC CODES,
ARE REPRODUCED WITH CONVENTIONAL ENDF PROCESSING CODES
USING AVERAGE RESONANCE PARAMETERS GIVEN IN MF=2 MT=151.
2200-M/S CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS.
2200 M/S(B) RES. INTEG.(B)
TOTAL 12.077
ELASTIC 9.360
FISSION 11.8E-06 1.72
CAPTURE 2.717 277
MF=3 NEUTRON CROSS SECTIONS
BELOW 4 KEV, BACKGROUND CROSS SECTIONS WERE GIVEN FOR TOTAL
CROSS SECTION ONLY.
ABOVE 4 KEV UP TO 150 KEV EVALUATED CROSS SECTIONS WERE
REPRESENTED WITH THE UNRESOLVED RESONANCE PARAMETERS.
MT= 1, 2, 4, 51-81, 91 - TOTAL, ELASTIC AND INELASTIC
SCATTERING CROSS SECTIONS.
TOTAL, ELASTIC AND DIRECT INELASTIC FOR ROTATIONAL GROUND
STATE BAND LEVELS MT=51,52,53,54 (COUPLED LEVELS)
AS WELL AS OPTICAL TRANSMISSION COEFFICIENTS ARE OBTAINED
IN A RIGID ROTATOR MODEL COUPLED CHANNELS CALCULATIONS.
DIRECT EXCITATION OF VIBRATIONAL, OCTUPOLE AND K=2+ QUADRUPOL
BAND LEVELS,MT=55,56,57,58,59,61,62,63,64,65,66,67,68,72,75,
78,79, ARE OBTAINED IN A SOFT ROTATOR MODEL COUPLED
CHANNEL CALCULATIONS, FOR NORMALIZATION PURPOSES THESE DIRECT
INELASTIC CROSS SECTIONS WERE SUBTRACTED FROM MT=2 ELASTIC
SCATTERING CROSS SECTION. DIRECT INELASTIC CONTRIBUTIONS
WERE ADDED INCOHERENTLY TO HAUSER-FESHBACH CALCULATIONS
OF COMPOUND NUCLEUS INELASTIC SCATTERING CROSS SECTIONS.
THE DEFORMED OPTICAL POTENTIAL USED: ENERGY DEPENDENCE
OF VR AND WD VARIED TO FIT MEASURED TOTAL DATA /13-19/,
ELASTIC SCATTERING DATA /20-22/ AND EVALUATED VALUE OF
S0=(.94+-.02)x10-4(EV)-1/2
VR=(45.722-0.334xE) MEV; RR =1.2668 FM; AR =.6468 FM;
WD=(3.145+0.455xE)MEV; E< 8 MEV RD =1.25 FM;
WD= 6.785 MEV; E>= 8 MEV AD =.5246 FM;
VSO= 6.2 MEV; RS0=1.12 FM; ASO=.47 FM;
B2= .188; B4=.071;
FISSION, CAPTURE AND COMPOUND INELASTIC SCATTERING CROSS
SECTIONS WERE CALCULATED WITH HAUSER-FESHBACH-MOLDAUER/23/
APPROACH, AT INCIDENT NEUTRON ENERGIES HIGHER THAN 1.21 MEV
(LEVEL OVERLAPPING ENERGY) TEPEL ET AL./24/ THEORY WAS
EMPLOYED.
MEASURED DATA ON DISCRETE LEVEL EXCITATION CROSS SECTIONS
/25-31/, GROUPS OF LEVELS OF VIBRATIONAL, OCTUPOLE AND K=2+
BAND LEVELS /27/ AS WELL AS TOTAL INELASTIC CROSS SECTION
/29-31/ WERE TAKEN INTO ACCOUNT.
ADOPTED LEVEL SCHEME OF Th-232 FROM NUCLEAR DATA SHEETS /32/.
LEVEL SCHEME:
--------------------------------------------------------
NO. ENERGY(MEV) SPIN-PARITY K-PARITY
--------------------------------------------------------
G.S. 0.0 0 + 0+
0.4936900-01 2 + 0+
0.1621200+00 4 + 0+
0.3332000+00 6 + 0+
0.5569000+00 8 + 0+
0.7142500+00 1 - 0-
0.7303500+00 0 + 0+
0.7741000+00 2 + 0+
0.7744000+00 3 - 0-
0.7853000+00 2 + 2+
0.8270000+00 10 + 0+
0.8296000+00 3 + 2+
0.8730000+00 4 + 2+
0.8836000+00 5 - 0-
0.8901000+00 4 + 0+
0.9604000+00 5 + 2+
0.1023100+01 6 + 2+
0.1042900+01 7 - 0-
0.1049900+01 6 + 0+
0.1053600+01 2 +
0.1072900+01 2 +
0.1077500+01 1 -
0.1078700+01 0 + 0+
0.1094400+01 3 +
0.1105700+01 3 -
0.1121800+01 2 + 0+
0.1137100+01 12 +
0.1143300+01 4 -
0.1146000+01 7 +
0.1148300+01 4 + 0+
0.1182500+01 3 -
0.1208900+01 5 -
OVERLAPPING LEVELS ARE ASSUMED ABOVE 1.21 MEV
MT=16,17,37. (N,2N), (N,3N) AND (N,4N) CROSS SECTION FROM
STATISTICAL MODEL CALCULATIONS /1/ WITH THE ACCOUNT OF
PRE-EQUILIBRIUM NEUTRON EMISSION (MODIFIED STAPRE CODE/33/
WAS USED). MEASURED (N,2N) DATA /34-46/ WERE CONSISTENTLY
REPRODUCED, WHILE CALCULATED FISSION CROSS SECTION, WHICH
DESCRIBES MEASURED DATA BASE, WAS USED AS MAJOR CONSTRAINT.
MT=18, 19, 20, 21. FISSION CROSS SECTION IS CALCULATED WITHIN
STATISTICAL MODEL /1/. FOR FISSION DATA ANALYSIS MEASURED
DATA /47-51/ WERE USED.
THE CONTRIBUTION OF EMISSIVE FISSION TO THE TOTAL FISSION
CROSS SECTION IS FIXED ACCORDING TO CONSISTENT DESCRIPTION
OF(N,F) AND (N,XN) REACTION DATA
MT=102 CAPTURE
CAPTURE CROSS SECTION DATA /52-55/ ARE DESCRIBED WITHIN A
STATISTICAL MODEL. ABOVE NEUTRON ENERGY 5 MEV CAPTURE IS
ASSUMED TO BE CONSTANT. COMPETITION OF (N,GF) AND (N,GN')
REACTIONS IS TAKEN INTO ACCOUNT.
RADIATIVE STRENGTH FUNCTION SGO = 0.XXX WAS ADJUSTED
TO FIT CAPTURE CROSS SECTION DATA ABOVE 4 KEV. CAPTURE CROSS
SECTION DATA AT HIGHER ENERGIES WERE DESCRIBED VARYING
LEVEL DENSITY OF 233-Th COMPOUND NUCLIDE.
MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS
FOR MT=2,51,52,53 AND 54 FROM COUPLED CHANNEL CALCULATIONS
(RIGID ROTATOR MODEL),
FOR MT=55,56,57,58,59,61,62,63,64,65,66,67,68,72,75,
78,79, FROM COUPLED CHANNEL MODEL (SOFT ROTATOR MODEL)
WITH ADDED ISOTROPIC STATISTICAL CONTRIBUTION.
MT=16, 17, 18-21, 37,38, 60, 69-71,73,76,77,80,81 AND 91 ARE
ISOTROPIC IN THE LAB SYSTEM.
MF=5 ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS
ENERGY DISTRIBUTIONS FOR MT=16,17,38,91 WERE CALCULATED BY
STATISTICAL MODEL OF CASCADE NEUTRON EMISSION TAKING INTO
ACCOUNT THE HISTORY OF THE DECAY WITH THE ALLOWANCE OF PRE-
EQUILIBRIUM EMISSION OF THE FIRST NEUTRON. MEASURED NEUTRON
EMISSION SPECTRA DATA BY BABA ET AL./56,57/ ARE REPRODUCED.
MT=18,19,20,21,37
PROMPT FISSION NEUTRON SPECTRA (PFNS)WERE CALCULATED WITH THE
SEMI-EMPIRICAL MODEL/1/, PRE-FISSION NEUTRON EMISSION IN
(N,XNF) REACTION, EITHER EQUILIBRIUM AND PRE-EQUILIBRIUM
MODES ARE INCLUDED. BASICALLY PFNS FROM FISSION
FRAGMENTS (FF) WERE CALCULATED AS A SUPERPOSITION OF TWO
WATT DISTRIBUTIONS FOR LIGHT AND HEAVY FF WITH EQUAL
CONTRIBUTIONS, BUT DIFFERENT TEMPERATURE PARAMETERS.
FF KINETIC ENERGY, ONE MORE MODEL PARAMETER, MIGHT BE LOWER
THAN TKE, WHICH REFLECTS IT'S DEPENDENS ON THE MOMENT OF
NEUTRON EMISSION. THIS EFFECTIVELY REDUCES AVERAGE
ENERGY OF PFNS FOR INCIDENT NEUTRON ENERGIES ABOVE EMISSIVE
FISSION THRESHOLD. DATA ON PFNS OF /58-62/ ARE FITTED.
**************************************************************
MF=12-15 is from ENDFB-VI.7 evaluation
(OECD/NEA Data Bank, Rugama)
**************************************************************
MF = 12
MT = 18
FOR ALL INCIDENT ENERGIES THE MULTIPLICITY OF PHOTONS FROM
FISSION WAS DERIVED FROM THE DATA OF REF. 63 WHICH REPORTED THE
PHOTON SPECTRUM AND TOTAL PHOTON ENERGY FOR THERMAL FISSION OF
U-235. IT WAS ASSUMED THAT THE MULTIPLICITY IS INDEPENDENT OF
INCIDENT NEUTRON ENERGY.
MT = 102
FOR ALL INCIDENT NEUTRON ENERGIES PHOTON PRODUCTION FROM
NEUTRON CAPTURE IS REPRESENTED BY AN ENERGY DEPENDENT MULTIPLI-
CITY TO BE APPLIED TO THE (N,GAMMA) CROSS SECTION AND BY AN
ENERGY INDEPENDENT SPECTRUM. THE SPECTRUM USED WAS BASED ON AN
UNDOCUMENTED MEASURED SPECTRUM FOR U-238 WITH A MINOR ADJUSTMENT
FOR SMALL Q-VALUE DIFFERENCE BETWEEN TH-232 AND U-238. THE
AVERAGE ENERGY OF TE ASSUMED SPECTRUM WAS THEN DIVIDED INTO THE
Q-VALUE TO OBTAIN A MULTIPLICITY AT ZERO NEUTRON ENERGY. THE
MULTIPLICITY AT 20 MEV WAS OBTAINED (REF 64) BY USE OF THE
FORMULA M(E) = M0(EN+Q)/Q WHERE M0 IS THE MULTIPLICITY AT ZERO
NEUTRON ENERGY AS DESCRIBED ABOVE.
MF = 13
MT = 3
EXPLICIT REPRESENTATION OF THREE PHOTONS (.04971, .1632 AND
.3344 MEV) FROM INELASTIC SCATTERING WAS DERIVED FROM FILE 3
DATA AND KNOWN BRANCHING RATIOS. FOR INCIDENT NEUTRON ENERGIES
GT .7251 MEV THE METHOD OF REF. 36 WAS USED TO CALCULATE PHOTON
PRODUCTION CROSS SECTIONS AND SPECTRA FROM ALL REACTIONS EXCEPT
PHOTONS FROM THE FIRST THREE INELASTIC GROUPS, NEUTRON CAPTURE
AND NEUTRON-INDUCED FISSION.
MF = 14
MT = 3,18,102 ALL PHOTONS WERE ASSUMED TO BE ISOTROPIC
MF = 15
MT = 3
PHOTON SPECTRA WERE OBTAINED USING THE METHOD OF REF. 65.
MT = 18
THE MEASURED PHOTON SPECTRUM OF REF. 63 WAS USED FOR ALL
INCIDENT NEUTRON ENERGIES BECAUSE THERE ARE NO EXPERIMENTAL DATA
FOR TH-232.
MT = 102
THE SECTRUM USED WAS BASED ON AN UNDOCUMENTED MEASURED
PHOTON SPECTRUM AT THERMAL NEUTRON ENERGY FOR U-238 WITH A MINOR
ADJUSTMENT FOR THE SMALL Q-VALUE DIFFERENCE BETWEEN U-238 AND TH
-232. IT WAS ASSUMED THAT THE SPECTRUM REMAINED UNCHANGED FOR
ALL INCIDENT NEUTRON ENERGIES.
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