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Fission Product Data for Nuclear Technology Applications P. B. Hemmig Introduction .The ENDFIB-IV Fission Product Data .file (1,2) is the basis of many nuclear technology applications in the U.S. This.file is the culmination of a two year task force effort by the Cross Section Evaluation Working Group and con-. tains approximately 300,000 data entries on 824 fission products. Cross sections are provided for 181 of the major nuclides; decay parameters (half 'lives, B and y energies, branching ratios, etc.) are provided for 712 nuclides; and 13,000 fission yield parameters are provided for the six major fissionable nuclei. ENDF/B-IV data is routinely processed for use with the RIBD-II, (3) CINDER-7,(4) and ORIGEN(5) codes. These codes have been checked out and agree exactly for several "burst!' and long exposure calculations when the same input data base is used. Decay Heat Analysis A recent paper by R. Schenter and F. Schmittroth (6) shows summation results using the RIBD-II computer code for thermal neutron fission of U-235 with "infinite" exposure for the cooling time interval 10 to 10 3 seconds. Also shown is the evaluation of Perry, et al ~(7) of integral data and the .evaluation of Shure(*) which is the basis of the ANS 5.1 decay heat standard. (9) Figure 2 shows the same functions as those given in Figure 1 for coding intervals of lo3 to lo6 seconds.. When the curves of Figure 1 were split into~B and y components, agreement with Perry's evaluations was not as good for the components as for the total, but the agreement was still within the error bars assigned by Perry. The 1arges.t difference between ORIGEN calculations and Perry's curve was found by Kee, et al (10) to be 12% <l set following discharge~and <4% at 10 sec. The results of ORIGEN decay heat calculations for CRBR core assemblies were reported by Morrison, et al. (11) It was noted that for discharge times of 30-720 days, 17 nuclides contributed more than 90% of the decay heat. 95200001 L .~.,,,.., ,~~^-l--‘--.a-.r~-~~ -2- Nuclides from only 3 mass chains (95, 106 and 144) contributed 76.8% at 720 days. Thus only a few isotope yields, decay constants and Q values were found to be crucial for fuel storage calculations. Gamma Spectra Calculations Gamma spectra above Be-9 and H-2 photoneutron thresholds were calculated by Stramatelatos.and England (12) for cooling time up to 1000 hours. After a cooling time of 10 ,hours, the spectra hardens with time and the total y emission rate remains nearly constant up to 200 hours. This behavior is important to sourceless startups in LWR reactors or reactors containing large D20 or Be moderators. Prompt and Delayed Neutrons Calculations of prompt and delayed neutrons per fission were reported by Schenter and England. (13) Delayed neutron summations were found to be very sensitive to direct fission yield data. The neutron emission probabilities were measured values or calculated using the approximate formula P, = .97 (~,q-B~)l'*~ for 57 precursors. (14) The values of 9 were calculated using Terrell's summation method. (15) The summations in Table 1 are compared with RNDF/B-IV evaluations of integral measurements by Cox, et al. (16) Table 1(13) i indicates good agreement for thermal fission in U-235 but poor agreement for the higher energy fission yields. It is anticipated that forthcoming improve- ments in the fast neutron yield data for the major delayed neutron precursors will improve this agreement, although 5-10% discrepancies are still noted in integral yield measurements for isotopes such as U-238. Integral Measurements There are several recent efforts in the U.S. to provide improved integral -measurements of decay heat.(17y18) Results are also expected to be reported during the next several months from experiments underway at ORNL, I&L and the University of Illinois. - . 1 . : . / . Table I -. ..*. : . *.. .I '. -.' :I 'ENDF/B-IV Calculated and'lkaluated . Prdmpt and Dclaycd Neutron Comparisons . '. Delayed KeutronsflOO Fissions Pronot Neutrons per Fiss~ion, Fission. ! .\ - Nuclidc Calculated' ' Evaluated* Calculated' Eval,~ncrc'-~ i35U(T) .1.604 ' i.67iO.07 w . 2.41 2.40 235U(F) . 1.483 1.67t0.07 2.38 2.53-2.65 '235U(lE) 1.095 0.9O"O.l . 3.63 4.38-4.X 238U(F) 2.934 4.6020.25 2.70 2.43-2.53 -. 238U(HE) 1.953 2.6,?0.2, 4.02 4.43-4.55 239PU(T) 0.520 0.645'0.04 2.92 2.87 239PU(F) 0.508 0.645kO.04. 2.77 3.01-3.15 .1.047 1.57+0.15 3.00 2.92 0.821' 0.7405.04 - .:s..2.47 . 3.933 5.27Si.4 2.39 i" . 1. These calculation; we&e made using branchin? probabTliZes,~fron tSe FP fiic and flssiox yield values given with the fissionable nucil.Ze. 29 EXIF/B-Iv evaluation discussed in "Delayed Yeutron Data -- Rev&:: and haluation," s.. A. Cor, A.x~,mX-5 , April 1974 (uncertainties are 2ls6 .taken from this report). _. 3* Thermal (T) values are at 0.0253 eV. Fast (F) are given at l:b k-id 2.0 IleV. High Energy ('ZZ:) are given.at 14 and 15 XeV (linear-linear interpolation applies). : ‘0 - . : ‘_ ,I ‘_ :..: ” . ‘. : -. _’ . - . I .:. -’ . *’ ‘_ .‘,,‘.. : ._ . a .:‘ . * ; . .Z’... ’ “F \ ,. :. .* .oo I < ‘\ \ .\ -’ ENOF/B-IVL7SI \, ---- SHUREtGLI ‘\ ‘- -* StfUREc20 PERCENT -\, Q PERRY ET F)Ll731 . 6-00 1 cl’ COOLZNG TlfYE .T ‘(SEC1 FIG. 1. Total integral afterheat for 235U thermal , "infinite" exposure. I . ._ 4.50 \ - ENDF/8-IV~‘?S~ 4.00 1,. * ---0 SHUREtGL) 3.50 ‘\\ \; k’ \\ ‘\ ‘, ” ‘: 3.00 2.50 . : 2.00 1.50 O3 1.0’ lo5 1 oe COOLING TIME (SEC) : FIGI 2. ,Total integral afterheat for 235U thermal , “infinite” exposure. References 1. Summary of ENDF/B Fission Product Data, BNL (to be pubiished), 1975. 2. C. Reich and R. He&r, Radioactive-Nuclide Decay Data in Science and Technology, Washington, D.C. Conference, 3175. (see also ANCR-1157, 8174) 3. A User's Manual for Computer Code RIBD-II a Fission Product Inventory Code, HEDL-TM%75-26. 4. T. England, R. Wilcsynski, and N. Whittemore, CINDER-7, an Interim User's Report,. LA-5885-MS. .5. M. Bell, ORIGEN, The ORIK Isotope Generation and Depletion Code, ORNL-4628, Oak Ridge National Laboratory, May 1973. 6. R. Schenter and F. Schmittroth, Radioactive Decay Heat Analyses, to be published in Proceedings of the Conference on Neutron Cross Section Technology, Washington, D.C., March 1975. 7. A. Perry, F. Maienschein and 0. Vondy, Fission-Product Afterheat - A Review of Experiments Pertinent to the Thermal-Neutron Fission of u-235, ORNL-TM-4197, Oak Ridge National Laboratory, October 1973. 8. K. Shure, Fission-Product Decay Energy, Bettis Technical Review, WAPD-BT-24, December 1961. 9. American Nuclear Society, Proposed ANS Standard Decay Energy Release Rates Following Shutdown of Uranium Fueled Thermal Reactors, Draft Standard ANS 5.1, approved by Subcommittee ANS-5 of the ANS Standard Committee, October 1971, revised October 1973. b 10. C. Kee, C. Weisbin and R. Schenter, Processing and Testing ENDF/B-IV Fission Product and Transmutation Data, Transactions of American 0 Nuclear Society 19,398, 1974. 11. G. Morrison, et al, Decay Heat Analysis for an LMFBR Fuel Assembly Using ENDFIB-IV Data, Washington, D.C. Conference, 3/75. 12. M. Stametelatos and T. England, Fission Product Gamma-Ray and Photoneutron Spectra, Washington, D.C. Conference, 3/75. 13. R. Schenter and T. R. England, Nuclear Data for Calculating Radioactivity Effects, Transactions of American Nuclear Society (to be published), June 1975. .-. 14. L. Tomlinson, Atomic and N&clear Data Tables 12, 179-194 (1973). 15. J. Terrell, Neutron Yields from Individual Fission Fragments, Physical Review 127, 880, 1962. 95200006 . , . -2- 16. S. A. Cox, ANL/NDM-5, April 1974. 17. B. Gunst, et al, Decay Heating Measurements and Calculations for Irradiated 2351J, ='U, LJyPu and zJJTh, WAPD-m-1183, 7174. 18. B. Gunst, et al, Measurements and Calculations of Heavy Isotopes in Irradiated Fuels and of L'hJ Fission-Product Poisoning, WAPD-m-1182, 7174. (see also NS&E 56, 214, 1975) . - .
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