94-Pu-241 JAEA+ EVAL-JAN10 O.Iwamoto,N.Otuka,S.Chiba,+ DIST-MAY10 20100325 ----JENDL-4.0 MATERIAL 9443 -----INCIDENT NEUTRON DATA ------ENDF-6 FORMAT History 06-09 Numbers of neutrons per fission were revised. 07-05 Theoretical calculation was made with CCONE code. 07-11 Fission cross section was revised with simultaneous evaluation. 07-12 Fission cross section was revised with a result of new simultaneous evaluation. 08-01 Fission cross section was revised. 08-02 Fission cross section and nu-p were revised. CCONE calculation was made with revised parameters. Data were compiled as JENDL/AC-2008/1/ 09-03 (1,452) and (1,455) were revised. 09-08 (MF1,MT458) was evaluated. 09-09 fission and total sigs and URP were revised. 09-10 fission cross section was revised. 09-12 New theoretical calculation was made with CCONE code. Nu-d was revised. 10-01 Data of prompt gamma rays due to fission were given. 10-03 Covariance data were given. MF= 1 General information MT=452 Number of Neutrons per fission Sum of MT's = 455 and 456. MT=455 Delayed neutron data Determined from nu-d of the following three nuclides and partial fission cross sections calculated with CCONE code/2/. Pu-242 = 0.0160 measured by Benedetti et al./3/ Pu-241 = 0.0064 data measured by Benedetti et al./3/ reduced by a factor of 0.7 Pu-240 = 0.0046 data for Pu-239, reduced by a factor of 0.7 Decay constants were adopted from Ref./4/. MT=456 Number of prompt neutrons per fission Experimental data of Conde et al./5/, Frehaut et al./6/ and D'yachenko et al./7/ were reproduced with two straight lines. A thermal value of 2.929 evaluated by Holden and Zucker/8/ was also considered. They were re-normalized to the nu-p of Cf-252 spontaneous fission of 3.756/9/. MT=458 Components of energy release due to fission Total energy and prompt energy were calculated from mass balance using JENDL-4 fission yields data and mass excess evaluation. Mass excess values were from Audi's 2009 evaluation/10/. Delayed energy values were calculated from the energy release for infinite irradiation using JENDL FP Decay Data File 2000 and JENDL-4 yields data. For delayed neutron energy, as the JENDL FP Decay Data File 2000/11/ does not include average neutron energy values, the average values were calculated using the formula shown in the report by T.R. England/12/. The fractions of prompt energy were calculated using the fractions of Sher's evaluation/13/ when they were provided. When the fractions were not given by Sher, averaged fractions were used. MF= 2 Resonance parameters MT=151 Resolved resonance parameters (RM: below 300 eV) Revised Derrien's evaluation/14/ below 20 eV was adopted. Numerical data were taken from ENDF/B-VII.0/15/. See Apendix A-1. Unresolved resonance parameters (300 eV - 30 keV) Parameters were determined with ASREP code/16/ to reproduce capture cross section measured by Weston and Todd/17/, fission cross section measured by Weston and Todd/17/, Migneco et al./18/ and Gerasimov et al./19/, and total cross section calculated with CCONE code. The parameters are used only for self-shielding calculations. Thermal cross sections and resonance integrals (at 300K) ------------------------------------------------------- 0.0253 eV reson. integ.(*) (barns) (barns) ------------------------------------------------------- total 1386.67 elastic 11.26 fission 1012.34 567 capture 363.06 180 ------------------------------------------------------- (*) In the energy range from 0.5 eV to 10 MeV. MF= 3 Neutron cross sections Cross sections above the resolved resonance region except for the elastic scattering (MT=2), fission (MT=18, 19, 20, 21, 38) and capture cross sections were calculated with CCONE code/2/. The model parameters were determined by considering integral data as well as measured capture cross sections data and fission cross section of JENDL-3.3. MT= 1 Total cross section The cross section was calculated with CC OMP of Soukhovitskii et al./20/. MT=2 Elastic scattering cross section From 300 eV to 30 keV, CCONE calculation was adopted. Above 30 keV, it was calculated as (total)-(non-elastic scattering cross sections). MT=18 Fission cross section From 300 eV to 10 keV, the data of Weston and Todd/17/, Migneco et al./18/ and Gerasimov et al. /19/ were taken into consideration. Above 10 keV, experimental data measured after 1960 were analyzed by simultaneous fitting of U-233, U-235, U-238, Pu-239, Pu-240 and Pu-241 fission cross sections and their ratios by the SOK code /21/. -------------------------------------------------------------- Cross section -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- 30548.003 1.48E+7 N.A.Khan+ /22/ 10636.002 5.00E+3 - 6.50E+4 G.W.Carlson+ /23/ 20570.004 1.18E+6 - 2.63E+6 I.Szabo+ /24/ 20567.004 3.50E+4 - 9.70E+5 I.Szabo+ /24/ 20484.002 5.00E+3 - 2.98E+4 J.Blons+ /25/ 40636.007 1.46E+7 M.I.Kazarinova+ /26/ -------------------------------------------------------------- Ratio to U-235(n,f) cross section -------------------------------------------------------------- EXFOR Energy range (eV) Authors Reference -------------------------------------------------------------- 40474.003 2.40E+4 - 7.40E+6 B.I.Fursov+ /27/ 40474.003 1.27E+5 - 7.00E+6 B.I.Fursov+ /27/ 10563.002 5.00E+3 - 2.96E+7 J.W.Behrens+ /28/ 20364.002 1.37E+4 - 1.13E+6 F.Kaeppeler+ /29/ -------------------------------------------------------------- Cross section was slightly modified in the energy region from 1 to 4 MeV, and from 7 to 8 MeV for JENDL-4. MT=19, 20, 21, 38 Multi-chance fission cross sections Calculated with CCONE code, and renormalized to the total fission cross section (MT=18). MT=102 Capture cross section From 300 eV to 30 keV, the cross section was calculated from the unresolved resonance parameters mentioned above. CCONE calculation was adopted in the energy range above 50 keV. The experimental data of Weston and Todd/17/ were used to determine the parameters in the CCONE calculation. MF= 4 Angular distributions of secondary neutrons MT=2 Elastic scattering Calculated with CCONE code. MT=18 Fission Isotropic distributions were assumed in the laboratory system. MF= 5 Energy distributions of secondary neutrons MT=18 Fission spectra Calculated with CCONE code. MT=455 Delayed neutron spectra (Same as JENDL-3.3) Results of summation calculation made by Brady and England/4/ were adopted. MF= 6 Energy-angle distributions Calculated with CCONE code/2/. Distributions from fission (MT=18) are not included. MF=12 Photon production multiplicities MT=18 Fission Calculated from the total energy released by the prompt gamma-rays due to fission given in MF=1/MT=458 and the average energy of gamma-rays. MF=14 Photon angular distributions MT=18 Fission Isotoropic distributions were assumed. MF=15 Continuous photon energy spectra MT=18 Fission Experimental data measured by Verbinski et al./30/ for Pu-239 thermal fission were adopted. MF=31 Covariances of average number of neutrons per fission MT=452 Number of neutrons per fission Combination of covariances for MT=455 and MT=456. MT=455 Number of delayed neutrons per fission Uncertainty of 5% /3/ was assumed in the energy region below 5 MeV and 15% above 5 MeV. MT=456 Number of prompt neutrons per fission Covariance matrix was obtained by fitting to the experimental data of nu-p. MF=33 Covariances of neutron cross sections Covariances were given to all the cross sections by using KALMAN code/31/ and the covariances of model parameters used in the theoretical calculations. For the following cross sections, covariances were determined by different methods. MT=1, 2 Total and elastic scattering In the resolved resonance region, uncertainty of 5% was added to the contributions from resonance parameter uncertainties. Above 300 eV, covariances for CCONE calculation were adopted. MT=18 Fission cross section Error of 2 % was assumed in the resolved resonance region up to 300 eV. In the energy range from 300 eV to 10 keV, the fission cross section was determined from the experimetal data of Gerassimov et al./19/ Error of 5% was assumed in the region up to 9 keV. Above 9 keV, covariance matrix was obtained by simultaneous evaluation among the fission cross sections of U-233, U-235, U-238, Pu-239, Pu-240 and Pu-241(See MF=3, MT=18, and /1/). Since the variances are very small, they were adopted by multiplying a factor of 2. MT=102 Capture cross section Error of 10 % was assumed in the resolved resonance region. In the energy region from 300 eV to 2 keV, error of 15% was assumed, and from 2 to 30 keV, error of 10%, by considering dispersion of average cross sections of Weston and Todd/17/. Above 30 keV, Covariance matrix was obtained with CCONE and KALMAN codes/31/. MF=34 Covariances for Angular Distributions MT=2 Elastic scattering Covariances were given only to P1 components. MF=35 Covariances for Energy Distributions MT=18 Fission spectra Estimated with CCONE and KALMAN codes. ***************************************************************** Calculation with CCONE code ***************************************************************** Models and parameters used in the CCONE/2/ calculation 1) Coupled channel optical model Levels in the rotational band were included. Optical model potential and coupled levels are shown in Table 1. 2) Two-component exciton model/32/ * Global parametrization of Koning-Duijvestijn/33/ was used. * Gamma emission channel/34/ was added to simulate direct and semi-direct capture reaction. 3) Hauser-Feshbach statistical model * Moldauer width fluctuation correction/35/ was included. * Neutron, gamma and fission decay channel were included. * Transmission coefficients of neutrons were taken from coupled channel calculation in Table 1. * The level scheme of the target is shown in Table 2. * Level density formula of constant temperature and Fermi-gas model were used with shell energy correction and collective enhancement factor. Parameters are shown in Table 3. * Fission channel: Double humped fission barriers were assumed. Fission barrier penetrabilities were calculated with Hill-Wheler formula/36/. Fission barrier parameters were shown in Table 4. Transition state model was used and continuum levels are assumed above the saddles. The level density parameters for inner and outer saddles are shown in Tables 5 and 6, respectively. * Gamma-ray strength function of Kopecky et al/37/,/38/ was used. The prameters are shown in Table 7. ------------------------------------------------------------------ Tables ------------------------------------------------------------------ Table 1. Coupled channel calculation -------------------------------------------------- * rigid rotor model was applied * coupled levels = 0,1,2,3,9 (see Table 2) * optical potential parameters /20/ Volume: V_0 = 49.682 MeV lambda_HF = 0.01004 1/MeV C_viso = 15.9 MeV A_v = 12.04 MeV B_v = 81.36 MeV E_a = 385 MeV r_v = 1.2568 fm a_v = 0.633 fm Surface: W_0 = 17.45 MeV B_s = 11.19 MeV C_s = 0.01361 1/MeV C_wiso = 23.5 MeV r_s = 1.1803 fm a_s = 0.601 fm Spin-orbit: V_so = 5.75 MeV lambda_so = 0.005 1/MeV W_so = -3.1 MeV B_so = 160 MeV r_so = 1.1214 fm a_so = 0.59 fm Coulomb: C_coul = 1.3 r_c = 1.2452 fm a_c = 0.545 fm Deformation: beta_2 = 0.24731 beta_4 = 0.05852 beta_6 = -0.02486 * Calculated strength function S0= 1.19e-4 S1= 2.62e-4 R'= 9.47 fm (En=1 keV) -------------------------------------------------- Table 2. Level Scheme of Pu-241 ------------------- No. Ex(MeV) J PI ------------------- 0 0.00000 5/2 + * 1 0.04197 7/2 + * 2 0.09578 9/2 + * 3 0.16131 11/2 + * 4 0.16168 1/2 + 5 0.17094 3/2 + 6 0.17505 7/2 + 7 0.22299 5/2 + 8 0.23194 9/2 + 9 0.23500 13/2 + * 10 0.24489 7/2 + 11 0.30117 11/2 + 12 0.33700 1/2 + 13 0.33714 9/2 + 14 0.37300 11/2 + 15 0.37600 1/2 - 16 0.38500 13/2 + 17 0.40445 9/2 - 18 0.40890 7/2 - 19 0.44600 11/2 - 20 0.47300 1/2 - 21 0.49500 7/2 + 22 0.50300 13/2 + 23 0.51881 5/2 - 24 0.53420 7/2 + 25 0.56142 7/2 - 26 0.57000 15/2 - 27 0.61484 9/2 - 28 0.64500 13/2 - 29 0.68100 1/2 - 30 0.75517 1/2 + 31 0.76927 1/2 - 32 0.77915 3/2 - 33 0.78415 3/2 + 34 0.80044 3/2 + 35 0.80048 5/2 + 36 0.81095 5/2 - 37 0.83159 5/2 + 38 0.83330 7/2 - 39 0.83484 3/2 + 40 0.84196 1/2 - ------------------- *) Coupled levels in CC calculation Table 3. Level density parameters -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Pu-242 18.5806 1.5428 2.4520 0.3710 0.0263 3.6230 Pu-241 19.1301 0.7730 2.1853 0.3372 -0.4308 2.4675 Pu-240 18.5012 1.5492 2.1440 0.3871 -0.0917 3.7899 Pu-239 18.4349 0.7762 1.8503 0.3560 -0.5001 2.5655 Pu-238 18.3685 1.5557 1.9652 0.3804 0.0287 3.6608 -------------------------------------------------------- Table 4. Fission barrier parameters ---------------------------------------- Nuclide V_A hw_A V_B hw_B MeV MeV MeV MeV ---------------------------------------- Pu-242 6.172 0.998 4.674 0.600 Pu-241 5.795 0.580 5.487 0.520 Pu-240 6.250 1.040 4.920 0.600 Pu-239 6.050 0.700 5.700 0.600 Pu-238 6.000 1.040 4.800 0.600 ---------------------------------------- Table 5. Level density above inner saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Pu-242 20.4971 1.7999 2.6000 0.3433 -0.8376 3.9999 Pu-241 20.4242 0.9018 2.6000 0.3647 -2.0579 3.4018 Pu-240 20.3513 1.8074 2.6000 0.3300 -0.6156 3.8074 Pu-239 20.2784 0.9056 2.6000 0.3523 -1.8394 3.2056 Pu-238 20.2054 1.8150 2.6000 0.3313 -0.6081 3.8150 -------------------------------------------------------- Table 6. Level density above outer saddle -------------------------------------------------------- Nuclide a* Pair Eshell T E0 Ematch 1/MeV MeV MeV MeV MeV MeV -------------------------------------------------------- Pu-242 20.4971 1.7999 0.5000 0.3933 -0.2466 4.1999 Pu-241 20.4242 0.9018 0.4600 0.3804 -0.9744 3.1018 Pu-240 20.5363 1.8074 0.4200 0.3796 -0.0661 4.0074 Pu-239 20.2784 0.9056 0.3800 0.3901 -1.0534 3.2056 Pu-238 20.2054 1.8150 0.3400 0.3694 0.1086 3.8150 -------------------------------------------------------- Table 7. Gamma-ray strength function for Pu-242 -------------------------------------------------------- K0 = 2.481 E0 = 4.500 (MeV) * E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 300.00 (mb) ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb) * M1: ER = 6.58 (MeV) EG = 4.00 (MeV) SIG = 3.87 (mb) * E2: ER = 10.11 (MeV) EG = 3.21 (MeV) SIG = 6.78 (mb) -------------------------------------------------------- References 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009). 2) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007). 3) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982). 4) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989). 5) H.Conde et al.: J. Nucl. Energy, 22, 53 (1968). 6) J.Frehaut et al.: CEA-R-4626 (1974). 7) N.P.D'yachenko et al.: Sov. At. Energy, 36, 406 (1974). 8) N.E.Holden, M.S.Zucker: Nucl. Sci. Eng., 98, 174 (1988). 9) A.S.Vorobyev et al.: 2004 Santa Fe, Vol.1, p.613 (2004). 10) G.Audi: Private communication (April 2009). 11) J.Katakura et al.: JAERI 1343 (2001). 12) T.R.England et al.: LA-11151-MS (1988). 13) R.Sher, C.Beck: EPRI NP-1771 (1981). 14) H.Derrien et al.: Nucl. Sci. Eng., 150, 109 (2005). 15) M.B.Chadwick et al.: Nucl. Data Sheets, 107, 2931 (2006). 16) Y.Kikuchi et al.: JAERI-Data/Code 99-025 (1999) in Japanese. 17) L.W.Weston, J.H.Todd: Nucl. Sci. Eng., 65, 454 (1978). 18) E.Migneco et al.: 1970 Helsinki, p.437 (1970). 19) V.F.Gerasimov et al.: JINR-E3-97-213, p.348 (1997). 20) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005). 21) T.Kawano et al.: JAERI-Research 2000-004 (2000). 22) N.A.Khan et al.: Nucl. Instrum. Methods, 173, 137 (1980). 23) G.W.Carlson et al.: Nucl. Sci. Eng., 63, 149 (1977). 24) I.Szabo et al.: 1976 ANL, p.208 (1976). 25) J.Blons et al.: 1971 Knoxville, Vol.2, p.836 (1971). 26) M.I.Kazarinova et al.: Sov. J. At. Energy, 8, 125 (1961). 27) B.I.Fursov et al.: At. Energy, 44, 236 (1978). 28) J.W.Behrens et al.: Nucl. Sci. Eng., 68, 128 (1978). 29) F.Kaeppeler et al.: Nucl. Sci. Eng., 51, 124 (1973). 30) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973). 31) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in Japanese. 32) C.Kalbach: Phys. Rev. C33, 818 (1986). 33) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004). 34) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985). 35) P.A.Moldauer: Nucl. Phys. A344, 185 (1980). 36) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953). 37) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990). 38) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990). Appendix A-1: Resonance parameters ---------------------------------------------------------------- Reevaluation of rhe resonance parameters in the energy range up to 20 eV - Colaboration ORNL(USA)-CEN/CAD(FRANCE) H. Derrien and L.C Leal, ORNL A. Courcelle and A. Santamarina, CEN/CAD (submitted to NSE, 2003) Cross section integrals in the energy range 0.0021 eV to 20.0 eV ------------------------------------------------------------- Energy Range Capture b-eV Fission b-eV eV 1993 Present 1993 Present ------------------------------------------------------------- 0.0021-0.020 12.25 12.51 2.1% 31.06 31.43 1.2% 0.0200-0.030 3.67 3.72 1.4% 10.24 10.36 1.2% 0.0300-0.100 15.28 15.59 2.0% 49.02 49.14 0.2% 0.1000-0.500 110.58 117.40 6.2% 262.76 263.84 0.4% 0.5000-1.000 5.90 6.03 2.2% 17.93 17.89 -0.2% 1.0000-3.000 7.30 7.16 -2.0% 54.88 52.89 -3.8% 3.0000-20.00 1213. 1234. 1.7% 3039. 3026. -0.4% ------------------------------------------------------------- The cross sections at 0.0253 eV are the following: Total 1386.5 b Fission 1012.2 b Capture 363.0 b very close to the current standard. The fit of the experimental data (Young total, Wagemans fission, Weston capture and Weston fission ) were performed by assuming that young total and Wagemans fission were OK with the standard and that Weston original data(from the EXFOR file)needed a renor- zation to agree with the standard. A consistent normalization was obtained by a SAMMY fit leading to a decrease of 4.5% in the original Weston fission and a decrease of 3.0% in the original Weston capture.The differences between the 1988, 1993 and the present results are consistent with the experimental errors given by Weston for the measured fission (2-3%) and capture (10-15%). The default of normalization of the Weston fission,due to severe experimental errors in the thermal range, was responsible for the large uncertainty in the capture results.