ENTRY 22934 20071224 21912293400000001 SUBENT 22934001 20071224 21912293400100001 BIB 16 52 2293400100002 TITLE Pb-207(N,2NG)Pb-206 cross section measurement by in- 2293400100003 beam gamma-ray spectroscopy. 2293400100004 AUTHOR (P.Baumann, C.Borcea, E.Jericha, S.Jokic, M.Kerveno, 2293400100005 S.Lukic, L.C.Mihailescu, A.Pavlik, A.J.M.Plompen, 2293400100006 G.Rudolf) 2293400100007 INSTITUTE (2ZZZGEL) Borcea, Plompen, Mihailescu, and 2293400100008 experimental site 2293400100009 (2FR STR) Baumann, Kerveno, Rudolf 2293400100010 (3RUMBUC) Borcea, 2293400100011 (2AUSTHV) Jericha 2293400100012 (3YUGBKB) Lukic 2293400100013 (2AUSPVI) Pavlic 2293400100014 REFERENCE (C,2005NOTRED,,561,2005) Main reference, graph given 2293400100015 (J,NIM/A,531,375,2004) ENTRY 22870,L.C.Mihailescu+, 2293400100016 More details of experiments. 2293400100017 (C,2004SANTA,1,876,2004)Exp.details.1st author A.Pavlik2293400100018 FACILITY (LINAC,2ZZZGEL) The 150 MeV Geel electron linac, 2293400100019 GELINA. 2293400100020 INC-SOURCE (PHOTO) Rotary uranium target that is struck by an 2293400100021 electron beam with a mean energy of 105 2293400100022 MeV and an average current of 70 microA. 2293400100023 (BRST) Neutrons are generated by (gamma,xn) and 2293400100024 (gamma,f) reaction following bremsstrahlung. 2293400100025 INC-SPECT GELINA Pulsed white neutron source. 2293400100026 Neutron spectrum is combination of evaporation and 2293400100027 fission spectra with rapid intensity decrease above 2293400100028 2 MeV neutron energy. 2293400100029 SAMPLE The lead sample, 89.9 g, enriched to 92.4% in Pb-207, 2293400100030 was placed at 198.6 m from the neutron source. 2293400100031 ADD-RES (COMP) The TALYS and EMPIRE II code predictions. 2293400100032 METHOD (TOF) Algorithms for pulse shaping, pulse-height 2293400100033 (PSD) determination, and derivation of time-of-flight 2293400100034 (PHD) ("digital trigger") are implemented. 2293400100035 PART-DET 1(G) Prompt gamma rays emitted by sample. 2293400100036 DETECTOR 1(HPGE) 2 detectors, 70% and 100% rel.eff., among the 2293400100037 three used. 2293400100038 Positioned at 110 and 150 degrees of gamma-ray 2293400100039 emission angles. 2293400100040 (FISCH) Neutron flux was measured with a U-235 2293400100041 fission chamber. 2293400100042 MONITOR (92-U-235(N,F),,SIG) 2293400100043 ANALYSIS Gamma-ray spectra were extracted from raw data for 2293400100044 1.MeV-wide neutron energy groups up to 20 MeV neutron 2293400100045 energy. 2293400100046 ERR-ANALYS (DATA-ERR) The error is reported by authors on fig. 2293400100047 (ERR-DIG) Digitizing error of data. 2293400100048 (EN-ERR-DIG) Digitizing error of incident energy. 2293400100049 E-RSL - Energy resolution at 803.keV gamma-ray energy.2293400100050 HISTORY (20060529C) 2293400100051 (20061004U) Last corrections have been done. 2293400100052 (20071224A) Experimental details from 3rd reference 2293400100053 were added. M.M. 2293400100054 ENDBIB 52 0 2293400100055 NOCOMMON 0 0 2293400100056 ENDSUBENT 55 0 2293400199999 SUBENT 22934002 20071224 21912293400200001 BIB 3 3 2293400200002 REACTION (82-PB-207(N,2N)82-PB-206,PAR,SIG,G) 2293400200003 EN-SEC (E,G) 2+ -> 0+ gamma transition energy. 2293400200004 STATUS (CURVE) Fig.1 of main reference. 2293400200005 ENDBIB 3 0 2293400200006 COMMON 4 3 2293400200007 E EN-ERR-DIG ERR-DIG E-RSL 2293400200008 KEV MEV MB KEV 2293400200009 803. 0.12 8. 3.2 2293400200010 ENDCOMMON 3 0 2293400200011 DATA 3 31 2293400200012 EN DATA DATA-ERR 2293400200013 MEV MB MB 2293400200014 0.53 14. 2293400200015 0.95 12. 2293400200016 2.48 7. 11. 2293400200017 2.91 10. 2293400200018 4.03 11. 2293400200019 4.45 10. 2293400200020 4.96 8. 10. 2293400200021 5.56 11. 2293400200022 5.90 9. 2293400200023 6.59 7. 2293400200024 7.53 19. 2293400200025 8.10 98. 2293400200026 8.49 306. 2293400200027 8.93 596. 28. 2293400200028 9.38 766. 18. 2293400200029 9.85 953. 18. 2293400200030 10.33 1038. 20. 2293400200031 10.74 1140. 40. 2293400200032 11.41 1120. 40. 2293400200033 11.97 1060. 30. 2293400200034 12.38 1180. 30. 2293400200035 12.91 1070. 50. 2293400200036 13.29 1120. 30. 2293400200037 13.84 1180. 40. 2293400200038 14.29 1090. 40. 2293400200039 14.93 890. 20. 2293400200040 15.87 763. 28. 2293400200041 17.01 677. 18. 2293400200042 17.97 577. 23. 2293400200043 18.97 400. 23. 2293400200044 20.04 332. 25. 2293400200045 ENDDATA 33 0 2293400200046 ENDSUBENT 45 0 2293400299999 ENDENTRY 2 0 2293499999999