Caldeira_2010_Annals-of-Nuclear-Energy

Technical noteOn processing material23290Thfrom ENDF/B-VII.0with the NJOY systemAlexandre D.Caldeira *Instituto de Estudos Avançados (IEAv),12231-970São Josédos Campos,SP,Brazila r t i c l e i n f o Article history:Received 24November 2009Accepted 5May 2010Available online 11June 2010Keywords:ENDF/B-VII.0NJOY ThoriumEnergy distribution Fission reactiona b s t r a c tSome problems encountered when processing material 23290Thfrom ENDF/B-VII.0with the NJOY 99.259system are reported.Ó2010Elsevier Ltd.All rights reserved.1.IntroductionAn Evaluated Nuclear Data File (ENDF)includes targets or mate-rials for neutron,proton,deuteron and tritium reactions,thermal neutron scattering,photonuclear reactions and others and a new version becomes available from time to time.The ENDF/B-VI.8li-brary was issued in October 2001and has been frozen since then.This library has been superseded by ENDF/B-VII.0.The ENDF/B-VII.0library has been released on December 15,2006after inten-sive evaluation and testing of preliminary versions were carried out in various laboratories (National Nuclear Data Center,2009).Among the activities conduced by the Nuclear Data Subdivision of the Institute for Advanced Studies (IEAv),one of them is a con-sultant’s service on evaluated nuclear data libraries for Brazilian institutions.So,every time a new version of an evaluated nuclear data library is released,at least a consistency check is performed.In this superficial analysis for a material,basically the computer file structure is checked.Another activity is the processing of se-lected materials with the objective of producing pointwise and multigroup nuclear data.Along these lines,material 23290Th is espe-cially interesting for Brazil mainly due to its great mineral reserve and possible utilization in (Chadwick et al.,2006):the design of various reactor systems;innovative fuel cycle concepts;and accel-erator-driven systems for the transmutation of nuclear waste.As in the ENDF/B-VII.0validation process all transport calcula-tions were performed using stochastic methods (Chadwick et al.,2006;Van der Marck,2006),where pointwise nuclear data were utilized,in this work the generation of multigroup nuclear datafor material 23290Thfor applications in deterministic methods isaddressed.2.Recovering and analyzing material23290Thfrom ENDF/B-VII.0First of all,evaluations for material 23290Th,MAT number 9040,from ENDF/B-VII.0were taken from the National Nuclear Data Cen-ter (/)and the International Atomic Energy Agency Nuclear Data Services (/)sites.The evaluations were compared by contents and no differ-ences were found,as expected.In ENDF/B,MF and MT numbers specify file and reaction types,respectively.While cross sections are represented in file MF =3,angular and/or energy distributions are represented in files MF =4,5and 6,as described in Table 1.All angular and/or energy distributions of emitted particles available in ENDF/B-VII.0for material 23290Th are presented in Table 2.3.Processing material23290Thwith the NJOY systemNowadays,considering the criteria of versatility and complete-ness (Caldeira and Carlson,2004),the NJOY nuclear data process-ing system (Nuclear Energy Agency,2009)is the best choice for spectrum averaged cross section calculations.Among the NJOY in-put data,MFD numbers specify vector or matrix reactions,as dis-played in Table 3,and MTD numbers are related with reactions and some of them have special meanings.The input instructions for producing multigroup nuclear data include the generation of a fission spectrum,through the short-cut fission spectrum option,selecting MFD =5and MTD =18.0306-4549/$-see front matter Ó2010Elsevier Ltd.All rights reserved.doi:10.1016/j.anucene.2010.05.005*Tel.:+551239475495;fax:+551239441177.E-mail address:alexdc@ieav.cta.brTaking a look at the angular and/or energy distributions for23290Th,presented in Table2,it is noted the energy distributionfile forfis-sion reaction,MF=5and MT=18,is not found.In other words,if the short-cutfission spectrum is requested an error in subroutine FINDF of NJOY system occurs.At the beginning of a processing of every vector or matrix reac-tion,in GROUPR module,a message such as‘‘group constants at t=temp for mf mfd and mt mtd mtname”is printed out in the out-putfile,where:temp is the desired temperature;mfd is a vector or matrix information,as presented in Table3;mtd is the reaction type;and mtname is a name for the reaction type.If one takes a careful look at the output of GROUPR module,it is noted the ab-sence of that message for the following MF/MT reactions:4/600; 4/800;6/102;6/649;and6/849,whose correspondent meaningsare presented in Table2.The main interest in material23290Th is for nuclear reactor calculations and,for this purpose,all these reac-tions are not so important.On the other hand,in medical applica-tions or material radiation damage analysis certainly these reactions will be necessary.It looks like the version99.259of the NJOY system is not prepared for processing thesefile/reaction combinations.4.Final commentsAn error was observed in the multigroup nuclear data genera-tion for material23290Th from ENDF/B-VII.0with version99.259of the GROUPR module of the NJOY system when the short-cutfission spectrum option is requested.This problem occurs because the en-ergy distributionfile for thefission reaction was not found in the evaluation.It was also noted the absence of a special message printed out in the GROUPR outputfile for some reactions.It is conceivable that this NJOY version is not prepared for processing them.Finally,it is recommended that the energy distributionfile for thefission reaction of this material be included in the next ENDF/B version.AcknowledgementsThe author wishes to thank Luiz H.Claro for providing the exe-cutablefile of NJOY system version99.259and Roberto D.M.Garcia for clarifying discussions and suggestions.ReferencesCaldeira, A.D.,Carlson, B.V.,puter codes for spectrum average cross section calculations.Annals of Nuclear Energy31(9),1069–1072. Chadwick,M.B.,Oblozinsky,P.,Herman,M.,Greene,N.M.,McKnight,R.D.,Smith,D.L.,Young,P.G.,MacFarlane,R.E.,Hale,G.M.,Frankle,S.C.,Kahler,A.C.,Kawano,T.,Little,R.C.,Madland,D.G.,Moller,P.,Mosteller,R.D.,Page,P.R.,Talou,P., Trellue,H.,White,M.C.,Wilson,W.B.,Arcilla,R.,Dunford,C.L.,Mughabghab,S.F., Pritychenko, B.,Rochman, D.,Sonzogni, A.A.,Lubitz, C.R.,Trumbull,T.H., Weinman,J.P.,Brown,D.A.,Cullen,D.E.,Heinrichs,D.P.,McNabb,D.P.,Derrien,H.,Dunn,M.E.,Larson,N.M.,Leal,L.C.,Carlson,A.D.,Block,R.C.,Briggs,J.B.,Cheng,E.T.,Huria,H.C.,Zerkle,M.L.,Kozier,K.S.,Courcelle,A.,Pronyaev,V.,Van der Marck,S.C.,2006.ENDF/B-VII.0:next generation evaluated nuclear data library for nuclear science and technology.Nuclear Data Sheets107(12),2931–3059.International Atomic Energy Agency Nuclear Data Services.</>or<http://www-nds.ipen.br/>.National Nuclear Data Center,Brookhaven National Laboratory,(a)<http:// /csewg/>and(b)</exfor/ endf00.jsp>.Nuclear Energy Agency.<http://www.nea.fr>.Van der Marck,S.C.,2006.Benchmarking ENDF/B-VII.0.Nuclear Data Sheets107(12),3061–3117.Table2ENDF/B-VII.0angular and/or energy distributions for material23290Th.MF/MT Description4/2Angular distribution for elastic scattering4/51to89Angular distributions for inelastic scattering from1st to39th excited state4/600Angular distribution for proton production4/800Angular distribution for alpha production5/455Energy distribution for average number of delayed neutrons released perfission6/5Energy-angle distribution for the sum of all reactions not given explicitly in another MT number6/16Energy–angle distribution for production of two neutrons6/17Energy–angle distribution for production of three neutrons6/18Energy–angle distribution forfission6/22Energy–angle distribution for production of a neutron and an alpha particle6/24Energy–angle distribution for production of two neutrons and an alpha particle6/28Energy–angle distribution for production of a neutron and a proton 6/41Energy–angle distribution for production of two neutrons and a proton6/91Energy–angle distribution for inelastic scattering in the continuum 6/102Energy–angle distribution for radiative capture6/649Energy–angle distribution for proton production in the continuum 6/849Energy–angle distribution for alpha production in the continuum Table3Some MFD definitions.MFD Meaning3Cross section or yield vector5Fission chi by short-cut method6Neutron–neutron matrix(MF=4and5) 8Neutron–neutron matrix(MF=6)Table1Some MF definitions.MF Meaning3Reaction cross sections4Angular distributions of emitted particles5Energy distributions(spectra)of emitted particles6Energy-angle distributions of emitted particlesA.D.Caldeira/Annals of Nuclear Energy37(2010)1420–14211421。