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9.323900+4 2.369990+2 0 1 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 0 197 1
93-Np-239 ORNL EVAL-DEC88 R. Q. WRIGHT
DIST-MAY05 REV1-MAY05 20050504
----JEFF-31 MATERIAL 9352
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
***************************** JEFF-3.1 *************************
** **
** Original data taken from: JEFF-3.0 **
** **
******************************************************************
***************************** JEFF-3.0 ***********************
DATA TAKEN FROM :- ENDF/B-VI.4 (DIST-FEB90)
JENDL-3.2 (MF1MT452,455,456)
Same LNU=2 representation for all neutron multiplicities
******************************************************************
ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC
CONVERTED FROM JENDL-2 EVALUATION, MAT 2932 (SEE BELOW)
93-NP-239 KYUSHU U.+ EVAL-MAR76 Y.KANDA,JENDL-CG
*****************************************************************
ORNL, REVISED MARCH 2, 1987 R. Q. WRIGHT
THE JENDL2 NP-239 EVALUATION, MAT 2932, HAS BEEN REVISED BELOW
4.0 EV. IN THIS ENERGY RANGE THE CAPTURE CROSS SECTION, MF=3,
MT=102, IS GIVEN BY:
SIGC(E) = 77.0*SQRT(E0)/SQRT(E)
WHERE E IS THE ENERGY IN EV AND E0 = 0.0253 EV.
THIS CHANGE WAS MADE IN ORDER FOR THE THERMAL CAPTURE CROSS
SECTION TO BE IN AGREEMENT WITH THE VALUE GIVEN IN REFERENCE 1.
THE TOTAL CROSS SECTION WAS MODIFIED TO BE IN AGREEMENT WITH
THE SUM OF THE ELASTIC AND THE REVISED CAPTURE CROSS SECTIONS.
IN ADDITION, THE TOTAL CROSS SECTIONS AT 6.2533 AND 7.5000 MEV
WERE INCREASED BY ABOUT 0.1% SO THAT THEY WOULD BE IN AGREEMENT
WITH THE SUM OF THE PARTIAL CROSS SECTIONS AT THESE ENERGIES.
THE FORMAT WAS CHANGED FROM ENDF/B-IV TO ENDF/B-V.
REFERENCE 1. S. F. MUGHABGHAB, "NEUTRON CROSS SECTIONS: VOL 1,
NEUTRON RESONANCE PARAMETERS AND THERMAL CROSS SECTIONS,
PART B: Z=61-100" (1984), ACADEMIC PRESS.
*****************************************************************
HISTORY
76-03 THE EVALUATION FOR JENDL-1 WAS PERFORMED BY KANDA (KYUSHU
UNIV.) AND JENDL-1 COMPILATION GROUP. DETAILS ARE GIVEN
IN REF. /1/.
83-03 JENDL-1 DATA WERE ADOPTED FOR JENDL-2 AND EXTENDED TO 20
MEV. MF=5 WAS REVISED.
84-01 COMMENT DATA WERE ADDED.
MF=1 GENERAL INFORMATION
MT=451 DESCRIPTIVE DATA AND DICTIONARY
MT=452 NUMBER OF NEUTRONS PER FISSION
TAKEN FROM THE NP-237 DATA OF ENDF/B-IV.
MF=2 RESONANCE PARAMETERS
MT=151 NO RESONANCE PARAMETERS WERE GIVEN.
2200-M/SEC CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS.
2200 M/SEC RES. INTEG.
ELASTIC 10.50 B -
CAPTURE 37.00 B 445. B
FISSION 0.0 B 7.06 B
TOTAL 47.50 B -
MF=3 NEUTRON CROSS SECTIONS
BELOW 4.0 EV.
MT=1 TOTAL
SUM OF PARTIAL CROSS SECTIONS.
MT=2 ELASTIC SCATTERING
THE CONSTANT CROSS SECTION OF 10.5 BARNS WAS ASSUMED FROM
SIG=4*3.14*(0.147*A**(1/3))**2.
MT=18 FISSION
ASSUMED TO BE ZERO BARNS.
MT=102 CAPTURE
THE FORM OF 1/V WAS ASSUMED. THE 2200-M/SEC CROSS SECTION
WAS ADOPTED FROM THE EXPERIMENTAL DATA BY STOUGHTON AND
HALPERIN /2/.
ABOVE 4.0 EV.
MT=1 TOTAL
CALCULATED WITH OPTICAL AND STATISTICAL MODEL CODE CASTHY
/3/. OPTICAL POTENTIAL PARAMETERS WERE OBTAINED BY OHTA AND
MIYAMOTO /4/ BY USING THE TOTAL CROSS SECTION OF PU-239.
V = 45.87-0.2*EN, WI= 0.06, WS= 14.1, VSO= 7.3 (MEV)
R = 1.27 , RI= 1.27, RS=1.302, RSO= 1.27(FM)
A0= 0.652 , AI=0.315, AS= 0.98, ASO=0.652(FM)
MT=2 ELASTIC SCATTERING
CALCULATED WITH CASTHY /3/.
MT=4,51-58,91 INELASTIC SCATTERING
CALCULATED WITH CASTHY /3/. THE LEVEL SCHEME WAS ADOPTED
FROM NUCL. DATA SHEETS VOL.6.
NO. ENERGY(MEV) SPIN-PARITY
G.S. 0.0 5/2 +
1 0.03114 7/2 +
2 0.07112 9/2 +
3 0.07467 5/2 -
4 0.11766 11/2 +
5 0.1230 7/2 -
6 0.17305 9/2 -
7 0.2414 11/2 -
8 0.320 13/2 -
LEVELS ABOVE 430 KEV WERE ASSUMED TO OVERLAPPING. IN THE
CALCULATION THE CAPTURE, FISSION, (N,2N) AND (N,3N) CROSS
SECTIONS WERE CONSIDERED AS COMPETING PROCESSES.
MT=16,17 (N,2N) AND (N,3N)
CALCULATED WITH PEARLSTEIN'S METHOD /5/.
MT=18 FISSION
ESTIMATED FROM THE NP-237 FISSION CROSS SECTION BY NORMALIZ-
ING WITH NEUTRON SEPARATION ENERGIES.
MT=102 CAPTURE
BELOW 100 KEV, THE CROSS SECTION WAS CALCULATED FROM
SIG = 435 / SQRT(EN) BARNS.
ABOVE 100 KEV, THA SHAPE OF THE EXPERIMENTAL DATA FOR NP-237
BY NAGLE ET AL. /6/ WAS ADOPTED AND NORMALIZED TO 1.4 BARNS
AT 100 KEV.
MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS
MT=2 CALCULATED WITH CASTHY CODE /3/.
MT=51-58 ISOTROPIC IN THE CENTER-OF-MASS SYSTEM.
MT=16,17,18,91 ISOTROPIC IN THE LABORATORY SYSTEM.
MF=5 ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS
MT=16,17,91 EVAPORATION SPECTRUM.
MT=18 MAXWELLIAN FISSION SPECTRUM ESTIMATED FROM
Z**2/A SYSTEMATICS /7/.
REFERENCES
1) IGARASI S. ET AL.: JAERI 1261 (1979).
2) STOUGHTON R.W. AND HALPERIN J.: NUCL. SCI. ENG., 6, 100
(1959).
3) IGARASI S.: J. NUCL. SCI. TECHNOL., 12, 67 (1975).
4) OHTA M. AND MIYAMOTO K.: J. NUCL. SCI. TECHNOL., 10, 583
(1973).
5) PEARLSTEIN S.: NUCL. SCI. ENG., 23, 238 (1965).
6) NAGEL R.J. ET AL.: 1971 KNOXVILLE CONF., 259 (1971).
7) SMITH A.B. ET AL.: ANL/NDM-50 (1979).
******************************************************************
Relevant comments taken from JENDL-3.2
--------------------------------------
HISTORY
76-03 THE EVALUATION FOR JENDL-1 WAS PERFORMED BY KANDA (KYUSHU
UNIV.) AND JENDL-1 COMPILATION GROUP. DETAILS ARE GIVEN
IN REF. /1/.
83-03 JENDL-1 DATA WERE ADOPTED FOR JENDL-2 AND EXTENDED TO 20
MEV. MF=5 WAS REVISED.
87-07 DATA FORMAT WAS CONVERTED INTO ENDF-5 FORMAT AND ADOPTED
TO JENDL-3.
94-06 JENDL-3.2.
NU-P, NU-D AND NU-TOTAL WERE MODIFIED.
COMPILED BY T.NAKAGAWA (NDC/JAERI)
***** MODIFIED PARTS FOR JENDL-3.2 ********************
(1,452), (1,455), (1,456)
***********************************************************
MF=1 GENERAL INFORMATION
MT=451 DESCRIPTIVE DATA AND DICTIONARY
MT=452 NUMBER OF NEUTRONS PER FISSION
SUM OF NU-P NAD NU-D.
MT=455 DELAYED NEUTRONS PER FISSION
AVERAGE VALUES OF SYSTEMATICS BY TUTTLE/2/, BENEDETTI ET
AL./3/ AND WALDO ET AL./4/ DECAY CONSTANTS WERE ASSUMED TO
BE THE SAME AS THOSE OF NP-237 EVALUATED BY BRADY AND
ENGLAND/5/.
MT=456 PROMPT NEUTRONS PER FISSION
BASED ON SYSTEMATICS BY MANERO AND KONSHIN/6/, AND BY
HOWERTON/7/.
REFERENCES
1) IGARASI S. ET AL.: JAERI 1261 (1979).
2) TUTTLE R.J.: INDC(NDS)-107/G+SPECIAL, P.29 (1979),
3) BENEDETTI G. ET AL.: NUCL. SCI. ENG., 80, 379 (1982).
4) WALDO R. ET AL.: PHYS. REV., C23, 1113 (1981).
5) BRADY M.C. AND ENGLAND T.R.: NUCL. SCI. ENG., 103, 129 (1989).
6) MANERO F. AND KONSHIN V.A.: AT. ENERGY REV.,10, 637 (1972).
7) HOWERTON R.J.: NUCL. SCI. ENG.,62, 438 (1977).
8) STOUGHTON R.W. AND HALPERIN J.: NUCL. SCI. ENG., 6, 100
(1959).
9) IGARASI S. AND FUKAHORI T.: JAERI 1321 (1991).
10) OHTA M. AND MIYAMOTO K.: J. NUCL. SCI. TECHNOL., 10, 583
(1973).
11) PEARLSTEIN S.: NUCL. SCI. ENG., 23, 238 (1965).
12) NAGEL R.J. ET AL.: 1971 KNOXVILLE CONF., 259 (1971).
13) SMITH A.B. ET AL.: ANL/NDM-50 (1979).
******************************************************************
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