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7.719300+4 1.913050+2 1 0 2 1
0.000000+0 0.000000+0 0 0 0 6
1.000000+0 2.000000+7 0 0 10 31
0.000000+0 0.000000+0 0 0 411 1
77-Ir-193 ORNL EVAL-MAR95 R.Q.WRIGHT, R.R.SPENCER
DIST-MAY05 REV1-MAY05 20050504
----JEFF-31 MATERIAL 7731
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
***************************** JEFF-3.1 *************************
** **
** Original data taken from: ENDF/B-VI.8 **
** **
******************************************************************
****************************************************************
ENDF/B-VI MOD 1 evaluation (R.Q. Wright, ORNL, March 1995)
Resolved resonance range:
-------------------------
Resolved resonance range upper limit is 300 eV; the highest
energy resonance included is at 336 eV. There are 45 s-wave
(1 bound level) resonances. The last five (highest energy)
resonances are fictitious resonances above the upper cutoff of
the resolved resonance range. The positive resonances are taken
from ref. [1] and are unchanged from the BROND natural iridium
evaluation. See BROND evaluation comment cards below for
additional details. The MLBW resonance formalism is used in
this evaluation. The scattering radius is 11.3 fm which gives a
thermal scattering cross section of 19.34 barns.
********* note the following change--
The parameters for the bound level (-21.7 eV) are the same as
the Mughabghab evaluation [1] and are not the same as those in
the BROND natural iridium evaluation. This change gives a
thermal capture cross section of 112 b which is the value
adopted for the current evaluation.
Unresolved resonance range:
---------------------------
The unresolved resonance range is 300 eV to 10 keV. Unresolved
parameters for s-wave and p-wave resonances are included. (The
BROND natural iridium evaluation did not have unresolved
resonance parameters). The average gamma-gamma is 0.093 eV.
D0 = 5.54 eV, S0 = 2.0E-4, S1 = 0.4E-4, SG0 = 167.87 and the
scattering radius = 9.0 fm. Unresolved parameters are energy
independent but are specified using the energy dependent format
(energies given in the file). The SESH program, ref. [2] was
used to determine the total, elastic, and capture cross sections
in the unresolved range in order to compare calculated cross
sections with measurments.
Smooth cross section (file 3) changes:
--------------------------------------
MT = 1 Total cross section was modified from 10 to 150 keV.
The total is 17% higher than natural iridium at 10 keV
and unchanged above 150 keV.
MT = 2 Elastic = total - nonelastic
MT = 4 Inelastic = sum of MT = 51, 52, 53, 54, and 91
MT = 51, 52, 53, 54 discrete inelastic levels-- these levels
correspond to MT = 51, 52, 55, and 58, respectively, in
the BROND natural iridium evaluation.
MT= 91 Inelastic continuum is same as the BROND natural iridium
evaluation.
MT=102 The 193Ir evaluation is based on the measured capture of
Macklin, et al, Ref. [3] from 10 keV to 2 MeV. From 2
to 20 MeV the 193Ir capture was obtained by
renormalizing the BROND natural iridium capture to the
Macklin 193Ir capture value at 2 MeV.
MT= 16, 17, 103, and 107 data are unchanged from the BROND
natural iridium evaluation.
File 12, mt 102 (capture gammma rays) changes:
----------------------------------------------
The 193Ir evaluation is based on the BROND natural iridium
evaluation. The separation of the original natural element
evaluation into 191Ir and 193Ir is based on information given in
the file 1 comments for MF = 12, MT = 102 in the original BROND
evaluation (see file 1 comments below) and on ref. [4].
Other sections (MF, MT values) not explicitly mentioned above
are unchanged from the original BROND natural iridium evaluation.
2200 m/s cross sections, barns
total = 131.34 barns
elastic = 19.34 barns
capture = 112.00 barns
capture resonance integral = 1373.0 barns
References:
-----------
1. S. F. Mughabghab, Neutron Cross Sections, Vol. 1, Part B,
(Academic Press, Orlando, Florida, 1984).
2. F. H. Frohner, "SESH-- a fortran-IV code for calculating the
self-shielding and multiple scattering effects for neutron
cross-section data interpretation in the unresolved resonance
region," Gulf General Atomic Report GA-8380 (1968).
3. R. L. Macklin (ORNL), Los Alamos Report LA-7479-MS (1978).
4. H. Kruger, et al., "Neutron Capture gamma-rays from Ir-192
and Ir-194," Nucl. Phys. A169, pp. 363-384 (1971).
*****************************************************************
BROND natural iridium evaluation comments
-----------------------------------------
CONTENT OF THE FILE:
MF = 1 GENERAL INFORMATION:
MT = 451 COMMENTS OF EVALUATION, REFERENCES AND
DICTIONARY.
RESULTS OF EVALUATION FULFILLED BY NIKOLAEV
M.N. (MF = 1-5, 9) AND ZABRODSKAYA S.V.
(MF = 12, 14, 15) ARE GIVEN. DATA COMPILED
AND CHECKED BY SAVOSKINA G.V.
DESCRIPTION IS WRITTEN DOWN BY NIKOLAEV M.N.
0.0253 EV CROSS SECTIONS:
RADIATIVE CAPTURE IR-191 954.+-10 B
RADIATIVE CAPTURE IR-193 112.+-5 B
RADIATIVE CAPTURE IR-NAT 426.5+-2.4 B
ELASTIC SCATTERING IR-191 13.9 B
ELASTIC SCATTERING IR-193 14.0 B
ELASTIC SCATTERING IR-NAT 14.0+-2.8 B
MF = 2 RESONANCE REGION:
MT = 151 RESOLVED RESONANCE REGION EXTENDS FOR
IR-191 TO 160 EV (BOUND LEVEL +45
S-RESONANCES +5 FICTITIOUS S-RESONANCES OUT
OF THE REGION) AND FOR IR-193 EXTENTS TO 300
EV (BOUND LEVEL +39 S-RESONANCES +5
FICTITIOUS S-RESONANCES OUT OF REGION).
RESONANCE PARAMETERS ARE TAKEN FROM REF.[1].
FICTITIOUS RESONANCES USED FOR TAKING INTO
ACCOUNT UNRESOLVED RESONANCES CONTRIBUTIONS.
THEIR ENERGIES ARE TAKEN FROM REF.[1].
AVERAGE WIDTHS ADOPTED FOR FICTITIOUS
RESONANCES.
MF = 3 NEUTRON CROSS SECTION:
IN THE REGION 0.00001 -160 EV ALL CROSS SECTIONS
DETERMINED BY RESONANCE PARAMETERS. IN THE MF=3 FILE
ALL CROSS SECTIONS IN THAT REGION ARE EQUAL TO ZERO. IN
THE REGION 160 - 300 EV IN THE MF = 3 FILE CONTRIBUTION
OF UNRESOLVED RESONANCES OF IR-191 CALCULATED ON THE
BASIS OF AVERAGED RESONANCE PARAMETERS IS GIVEN IN THE
MF = 3 FILE. IN THE REGION 300 EV - 100 KEV CROSS
SECTION CALCULATED VIA THE AVERAGED RESONANCE
PARAMETERS ARE ADOPTED. AVERAGED PARAMETERS WAS CHOSEN
BY FITTING EXISTED EXPERIMENTAL DATA REF.[2] ON IR-191
AND IR-193 CAPTURE CROSS SECTIONS IN THE REGION 3 - 100
KEV INTERVAL AND TOTAL CROSS SECTION DATA IN THE REGION
300 EV - 100 KEV.
-------------------------------------------------
USED AVERAGED PARAMETERS ! IR-191 ! IR-193
-------------------------------------------------
SCATTERING RADIUS, FERMI 5.7 6.0
.................................................
NEUTRON STRENGTH FUNCTION
(MULTIPLIED BY 10-4)
L=0 3.37 3.59
L=1 1.0 1.0
L=2 1.0 1.0
.................................................
RADIATIVE STRENGTH FUNCTION
(MULTIPLIED BY 10-4)
L=0 240. 134.
L=1 100. 100.
L=2 100. 100.
-------------------------------------------------
CALCULATIONS ARE FULFILLED BY EVPAR CODE REF.[3].
MT = 1 TOTAL CROSS SECTION ABOVE 100 KEV IS ADOPTED
ON THE BASIS OF EXPERIMENTAL DATA COMPILED IN
REF.[2].
MT = 2 ELASTIC SCATTERING CROSS SECTION ABOVE 100
KEV IS EQUAL TO DIFFERENCE BETWEEN TOTAL
CROSS SECTION AND SUMMARIZED NONELASTIC CROSS
SECTION.
MT = 4 INELASTIC SCATTERING CROSS SECTION IS THE SUM
OF MT = 51 - 91.
MT = 16 (N,2N) CROSS SECTION IS ADOPTED IN
ACCORDANCE WITH EXPERIMENTAL DATA COMPILED IN
REF.[2] FOR IR-191(N,2N)IR-190G+M1,
IR-191(N,2N)IR-190M2, IR-193(N,2N)IR-192G+M1.
PROBABILITY OF EXCITATION OF SECOND ISOMERIC
STATE OF IR-192 IN THE (N,2N) REACTION ON
IR-193 IS ADOPTED EXACTLY THE SAME AS FOR
IR-191.
MT = 17 (N,3N) REACTION CROSS SECTION FOR IR-191 IS
ADOPTED IN ACCORDANCE WITH EXPERIMENTAL DATA
CITED IN REF.[2]. RATIO OF (N,3N) TO (N,2N)
REACTION CROSS SECTIONS FOR IR-193 IS ADOPTED
TO BE EQUAL TO THOSE FOR IR-191.
MT = 51-58 EXCITATION CROSS SECTIONS FOR THE NEXT
LEVELS ARE GIVEN.
----------------------------------------
MT ! ISOT ! E - LEV ! J-PI ! T 1/2
----------------------------------------
51 193 73 KEV 1/2+
52 193 80 KEV 11/2- 10.6 DAY
53 191 81 KEV 1/2+
54 191 129 KEV 5/2+
55 193 139 KEV 5/2+
56 191 171 KEV 11/2+ 4.9 SEC
57 191 178 KEV 3/2+
58 193 180 KEV 3/2+
----------------------------------------
CROSS SECTION CALCULATED BY THE TNG CODE
REF.[4]. ABOVE 7 MEV ALL THIS CROSS SECTIONS
PUT EQUAL TO ZERO.
MT = 91 CONTINUUM INELASTIC CROSS SECTION CALCULATED
ON THE BASIS OF STATISTICAL MODEL. BELOW THE
THRESHOLD OF (N,3N) REACTION THIS CROSS
SECTION CALCULATED BY TNG CODE. ABOVE THIS
THRESHOLD CONTINUUM INELASTIC CROSS SECTION
IS THE DIFFERENCE BETWEEN SUM OF CROSS
SECTIONS MT = 4+16+17+102+103+107 CALCULATED
BY TNG CODE AND SUM ADOPTED CROSS SECTIONS OF
(N,2N), (N,3N), (N,GAMMA), (N,P) AND (N,ALFA)
MT = 102 RADIATIVE CROSS SECTION IN THE MF = 3 FILE
IN THE REGION 10-5 EV - 160 EV PUT EQUAL TO
ZERO (CROSS SECTIONS IN THIS INTERVAL ARE
DESCRIBED BY RESONANCE PARAMETERS); IN THE
REGION 160 EV - 300 EV CAPTURE CROSS SECTION
CONTAINS CONTRIBUTION OF THAT FOR IR-191
CALCULATED ON THE BASIS OF AVERAGED RESONANCE
PARAMETERS; IN THE INTERVAL 300 EV - 100 KEV
CROSS SECTION IN THIS SECTION INCLUDE
CONTRIBUTION OF BOTH ISOTOPES CALCULATED VIA
THE AVERAGE PARAMETERS FITTED TO EXPERIMENTAL
DATA. FROM 10 KEV TO 2 MEV EYE GUIDE CURVE
BASED ON THE EXPERIMENTAL DATA COMPILED IN
REF.[2]. ADOPTED HERE. ABOVE 2 MEV CROSS
SECTION PUT EQUAL TO THOSE OF AU-197
RENORMALIZED ON IRIDIUM EVALUATED CAPTURE
CROSS SECTION IN THE 1 - 2 MEV INTERVAL.
MT = 103 CROSS SECTION OF (N,P) REACTION IS
CALCULATED BY TNG CODE AND RENORMALIZED ON
QAIM ET AL. REF.[5] EXPERIMENTAL DATA AT 14.7
MEV.
MT = 107 CROSS SECTION OF (N,ALFA) REACTION IS
CALCULATED BY TNG CODE AND RENORMALIZED FOR
IR-191 ON THE COLEMAN ET AL. [6] EXPERIMENTAL
DATA AT 14.5 MEV AND FOR IR-193 ON KHURANA ET
AL. DATA [7] AT 14.0 MEV.
MF = 4 ANGULAR DISTRIBUTIONS:
MT = 2 ELASTIC SCATTERING ANGULAR DISTRIBUTION
ACCEPTED AS THOSE FOR GOLD (ENDF/B-5 DATA).
MT = 16, 17, 51-56, 91 (N,2N), (N,3N) AND INELASTIC
SCATTERING NEUTRON ANGULAR DISTRIBUTIONS
ACCEPTED AS ISOTROPIC IN THE LAB SYSTEM.
MF = 5 ENERGY DISTRIBUTIONS:
MT = 16, 17, 91 CONTINUOUS ENERGY DISTRIBUTIONS
(N,2N), (N,3N) AND INELASTIC SCATTERING
NEUTRONS ARE CALCULATED BY THE NEVA CODE
[8] IN THE FRAME OF EVAPORATION MODEL.
EVAPORATION TEMPERATURES ESTIMATED
ACCORDING TO GILBERT-CAMERON PRESCRIPTION.
COMPETITION FROM THE RADIATIVE CHANNEL NOT
TAKEN INTO ACCOUNT. PRECOMPOUND EMISSION
TAKEN INTO ACCOUNT APPROXIMATELY:
ABOVE 6 MEV EVAPORATION SPECTRUM OF THE FIRST
NEUTRONS IS CONSTANT.
MF = 12 MULTIPLICITIES OF RADIATIVE TRANSITIONS:
MT = 16 MULTIPLICITIES OF PHOTONS EMITTED IN (N,2N)
REACTION CALCULATED BY TNG CODE. THEIR
SPECTRA ACCEPTED TO BE CONTINUOUS.
MT = 17 MULTIPLICITY OF PHOTONS EMITTED IN (N,3N)
REACTION ACCEPTED SUCH THAT BEING
MULTIPLIED BY AVERAGE PHOTON ENERGY (SEE MF =
15) TO FULFILL TO ENERGY BALANCE WITH THE
TAKING INTO ACCOUNT OF NEUTRON SPECTRA,
LISTED IN THE MF = 5 FILE.
MT = 51, 53, 54, 55, 57, 58 FOR EACH OF THIS
REACTIONS LINEAR PHOTON SPECTRUM ARE GIVEN.
PHOTON LINES DETERMINED OR BY TRANSITIONS ON
THE UNDERLYING LEVELS OR IN THE CASE OF
INTERNAL CONVERSION BY CORRESPONDING
ROENTGEN RAYS. RADIATIVE TRANSITION SCHEMES
TAKEN FROM REF.[16]. OPTION L0=1 IS USED.
MT = 91 FOR CONTINUOUS INELASTIC SCATTERING THE
PHOTON SPECTRUM CONTAINS DISCRETE LINES
DETERMINED IN THE MF = 12 FILE (IT CAUSED BY
TRANSITIONS BETWEEN LOW ENERGY LEVELS
POPULATED BY THE TRANSITIONS FROM THE
CONTINUUM) AND CONTINUOUS SPECTRUM
(DETERMINED IN THE MF = 15 FILE). IN THE
DISCRETE LINES OF SPECTRUM ROENTGEN RAYS
CONNECTED WITH THE CONVERSION ELECTRON
EMISSION IS TAKEN INTO ACCOUNT.
MT = 102 FOR RADIATIVE CAPTURE IN THE MF = 12 FILE
TWO GROUP OF PHOTONS ARE DETERMINED. FIRST
GROUP IS THE GROUP OF LOW ENERGY (E<0.7 MEV)
PHOTONS WITH ENERGIES AND RELATIVE
INTENSITIES MEASURED BY KRUGER ET AL. [14].
SECOND GROUP IS GROUP OF PRIMARY PHOTONS
(E>4.5 MEV) WHICH ENERGIES AND RELATIVE
INTENSITIES MEASURED IN THE SAME WORK. IT IS
SUPPOSED THAT OTHER PRIMARY PHOTONS NOT
EMITTED. THUS SUM OF RELATIVE INTENSITIES
OF MEASURED HARD PHOTONS IS NORMALIZED TO
UNITY. IN THE INTERVAL FROM 0.7 TO 4.5 MEV
PHOTON SPECTRUM ARE ACCEPTED TO BE CONSTANT.
YIELDS OF THIS CONTINUOUS SPECTRUM ARE FITTED
IN ORDER TO FULFILL ENERGY BALANCE FOR
THERMAL NEUTRON CAPTURE. YIELDS OF LOW ENERGY
PHOTONS ARE NORMALIZED SO THAT SUM OF YIELDS
OF PHOTONS WITH THE ENERGIES FROM 0.4 TO 7
MEV WOULD BE EQUAL TO YIELD PHOTONS OF
CONTINUOUS SPECTRUM IN THE INTERVAL FROM 0.7
TO 1 MEV. ENERGY BALANCE FOR FAST NEUTRON
CAPTURE CONSERVED BY INCREASING OF ENERGY OF
PRIMARY PHOTONS. ENERGY DEPENDENCES OF
MULTIPLICITIES ARE REPRESENTED BY TWO STEPS:
IN THE REGION FROM 10-5 EV TO 0.46416 EV
GAMMA SPECTRA OF IR-192 AND IR-194 AVERAGED
WITH THE WEIGHTS OF THERMAL CAPTURE CROSS
SECTIONS OF IR-191 AND IR-193. IN THE REGION
FROM 0.46416 EV TO 20 MEV GAMMA SPECTRA OF
IR-192 AND IR-194 AVERAGED WITH THE WEIGHTS
OF RESONANCE INTEGRALS OF IR-191 AND IR-193.
MT = 103, 107 PHOTON MULTIPLICITIES FOR (N,P) AND
(N,ALFA) REACTIONS CALCULATED BY TNG CODE.
THEIR SPECTRA ACCEPTED TO BE CONTINUOUS.
MF = 14 PHOTON ANGULAR DISTRIBUTIONS:
MT = 16, 17, 51, 53-55, 57, 58, 91, 102 ALL ANGULAR
DISTRIBUTIONS ACCEPTED TO BE ISOTOPIC IN THE
LAB SYSTEM.
MF = 15 CONTINUOUS PHOTON SPECTRA.
MT = 16 CONTINUOUS PHOTON SPECTRA FOR (N,2N)
REACTION CALCULATED BY TNG CODE.
MT = 17 CONTINUOUS PHOTON SPECTRA FOR (N,3N)
REACTION ESTIMATED FROM THOSE FOR (N,2N)
REACTION BY RENORMALIZATION OF ENERGY SCALE
IN
EIN+Q3N-3*EOUT3N
---------------- TIMES
EIN+Q2N-2*EOUT2N
MT = 91 CONTINUOUS PHOTON SPECTRUM FOR INELASTIC
SCATTERING CALCULATED BY TNG CODE WITH
INTRODUCING OF THE NEXT CORRECTIONS: LOW
ENERGY PHOTONS DESCRIBED BY DISCRETE SPECTRUM
ARE EXCLUDED, PHOTON YIELD IN THE 0.2 - 0.4
MEV INTERVAL REDUCED ON 25% FOR TAKING INTO
ACCOUNT OF INTERNAL CONVERSION.
MT = 102 CONTINUOUS SPECTRUM OF RADIATIVE CAPTURE
DESCRIBES ONLY CASCADE PHOTONS (SEE COMMENTS
TO MF = 12, MT = 102). IT IS NOT DEPEND FROM
THE ENERGY OF CAPTURED NEUTRONS.
MT = 103, 107 CONTINUOUS PHOTON SPECTRA (N,P) AND
(N,ALFA) REACTIONS CALCULATED BY TNG CODE.
REFERENCES
1. NEUTRON CROSS SECTIONS. V1. NEUTRON RESONANCE
PARAMETERS AND THERMAL CROSS SECTIONS. PART B.
S.F.MUGHABGHAB, M.DIVADEENAM, N.E.HOLDEN. NNDC BNL,
ACADEMIC PRESS 1981.
2. NEUTRON CROSS SECTIONS. V2. NEUTRON CROSS SECTIONS
CURVES. V.MCLANE, C.L.DUNFORD, P.F.ROSE. NNDC BNL,
ACADEMIC PRESS 1988.
3. MANTUROV G.N. ET AL. VANT "YADENYE CONSTANTY" ISSUE
1(50),P.50. MOSCOW 1983.
4. SHIBATA K., FU C.Y. ORNL/TM-10093, 1986.
5. QAIM ET AL. NUCL.PHYS/A.V.283, P.269, 1977.
6. COLEMAN ET AL. PROC.PHYS.SOC. K.73,P.218, 1959.
7. KHURANA ET AL. PROC. OF LOW ENERGY NUCLEAR PHYS. SYMP.
WALTAIR, P.297,1960.
8. NIKOLAEV M.N. AND GILFANOVA O. NEVA - CODE FOR
CALCULATION OF SPECTRA CONSEQUENTLY EVAPORATED
NEUTRONS. FEI, 1988. UNPUBLISHED.
9. BROADHEAD ET AL. INT.J. OF APPLIED RADIATION AND
ISOTOPES. V.18, P.279, 1967.
10. BORNEMISZA ET AL. ATOMIC KOZLEMENYEK, V.10(2),P.112.
DEBREZEN, 1968.
11. ANDERS ET AL. PROC. OF INT. CONF. NUCL. DATA FOR
SCIENCE AND TECHN. ANTWERP, 1982, P.859.
12. BAYHURST ET AL. PHYS.REV./C. V.12, P.451, 1975.
13. SIDDAPPA ET AL. PROC. OF NUCL. AND SOLID STATE PHYS.
SIMP. MADARAI. 1970, V.2, P.29.
14. KRUGER ET AL. NUCL. PHYS/A, V.169, P.363, 1971.
************************ C O N T E N T S ************************
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