Status of ENDF/B-VII.1 library

Transcription

1 Status of ENDF/B-VII.1 library Report to WPEC, May 12-13, 2011, NEA Headquarters, Paris, France M. Herman for CSEWG National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY , USA The US nuclear data community, coordinated by CSEWG, is preparing release of the ENDF/B- VII.1 library. This new version will address deficiencies identified in ENDF/B-VII.0, include improved neutron evaluations for several materials, provide covariances for 185 materials, revise fission product yields and contain essential update of the decay data sub-library. This report summarizes current work, describes status of the new files and outlines further developments. I. INTRODUCTION The ENDF/B-VII.0 library, made available in December 2006, has been the first major release of the US nuclear reaction data library in 16 years. It was extensively documented in the special issue of Nuclear Data Sheets, which since then, has ben cited more than 330 times. Intensive validation proved generally good performance of the library but a number of deficiencies were discerned. The sodium void worths are under-predicted and exhibit changes with core loading. Deficiencies in 2 H cross sections impact upon calculated D 2 O moderated HEU benchmark eigenvalues and ZED-2 coolant void reactivity coefficients. Mosteller noted significant (about 1%) decrease in the calculated k eff for HEU-D 2 O benchmarks attributed to elastic scattering angular distribution changes below 3 MeV. Several minor actinides, carried over from the previous versions of the library, are often crude since role of these materials in applications used to be marginal. Minor actinides gained importance only with the advent of new generation of nuclear reactors. It was recognized that that nuclear data structural materials: Ti, V, Mn, Cr, Ni, W in ENDF/B-VII.0 need upgrading, due to poor performance in some critical assemblies, such as ANL ZPR. Dean noted that thermal capture for certain fission products are deficient and need to be revisited. A study of fission product yields for plutonium identified that the England and Rider (ENDF/B- VI) data for the important fission product 99Mo are too low by 4%. It turned out that theory based delayed neutrons, which replaced ENDF/B-VI data, were problematic and we decided to revert to previous estimates. In 2008 CSEWG decided to undertake multilaboratory effort leading to VII.1 version of the library to be released in December Highlights of the new library include (i) a consistent set of covariance data targeting explicit needs of the Advanced Fuel Cycle Initiative, (ii) new R-matrix based evaluations for several light nuclei, (iii) evaluations for reactions on structural materials in both the fast neutron region and the resonance region, (iv) improvements of resonances regions and thermal cross sections for certain fission products and neutron absorber materials (Cd, Gd), (v) improvements in minor actinide evaluations for isotopes of U, Np, Pu, and Am, (vi) adoption of JENDL4.0 evaluations for the Cm, Bk, Cf, Es, Fm, and some other minor actinides, (vii) fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239 Pu, and (viii) a new Decay Data sub-library. We note that no significant changes are expected for the major actinides 235,8 U and 239 Pu, except reverting delayed neutron data to ENDF/B-VI.8 and adding covariances. Current update concerns only the most important neutron sub-library plus fission yields, and decay data leaving the remaining 11 sub-libraries unchanged. II. STATUS OF ENDF/B-VII.1 Three preliminary versions of the library have been released for testing. The most recent beta2 was made available on February 22, At the time this report is being written the beta3 version is being assembled and verified at BNL. Its release is expected by the middle of May While most of the new cross section evaluations were already available in beta2 the forthcoming version will contain a full set of covariances expected in ENDF/B-VII.1. Since January 2010 the ENDF/B library is maintained under the NNDC GForge server ( The previous ENDF/A library containing newly submitted files has been replaced by the Subversion system, which keeps track of all new submittals and modifications. The GForge system stores also all beta releases of the future library, evaluation and release reports, CSEWG minutes, and other auxiliary files. It manages also the ENDF mailing list.

2 A. Covariances The ENDF/B-VII.1beta3 release will contain 185 materials with covariances in the neutron sublibrary. These constitute the vast majority of covariances in the future VII.1 and represent one of the major improvements. Covariances for 110 materials explicitly address needs of the AFCI community. Specifically, they are intended to facilitate data adjustment for fast reactors. The current library of covariances builds on three previous projects, with the third one (COMMARA-2.0) being actual source of the 110 core covariance files. BOLNA A pilot library created by five laboratories (BOLNA = Brookhaven-Oak Ridge-Los Alamos- NRG Petten-Argonne) within WPEC Subgroup 26. The library contains covariances for 52 materials, put together on a short timetable. Low-Fidelity A library created in by four US National Laboratories under a DOE project for testing nuclear criticality safety methods. A full set of materials included in ENDF/B-VII.0 using simple, yet not unreasonable, methods. The emphasize was on completeness rather than quality, hence the name low-fidelity. COMMARA-2.0 A library produced through a BNL- LANL collaboration during for 110 materials of importance for the Advanced Fuel Cycle Initiative. In addition, ENDF/B-VII.1 will contain covariances already present in VII.0, new covariances produced by ORNL and LANL which do not make part of COMMARA-2.0 and a consistent set of covariances in the adopted JENDL-4.0 files for minor actinides. Last but not least, standards covariances have been incorporated in the respective VII.1 files, although in a few limited energy regions in 235,238 U and 239 Pu central values of cross sections differ from those recommended for standards. In these cases standards covariances will be properly scaled to account for the difference. The detailed list of materials with covariances is given in the file ENDF-B-VII.1beta3-covar.txt at B. Light nuclei As is true for all the light-element evaluations done at Los Alamos, these are based on R-matrix analyses of experimental data for the neutron plus target compound system, using the LANL Energy Dependent Analysis code, EDA. The multichannel R-matrix formalism has been implemented in its full generality, and without any approximations, in the EDA code. It accommodates any number of two-body channels having particles with arbitrary spins, masses, and charges. The formulation is relativistic, so that even zero-mass particles, such as photons, are treated correctly. Experimental data can be modified by the use of adjustable normalizations and energy shifts, and the calculations can fold in the effects of beam energy resolution/spread. In ENDF/B-VII.1 new analyses have been performed for neutrons plus 4 He, 6 Li, 9 Be, and 16 O. The 9 Be evaluation includes new data from RPI. In the following we focus on 16 O because of it s paramount importance for various applications. The ENDF/B-VII.0 evaluation, carried over from ENDF/B- VI.8, combined analyses at Los Alamos above 3.4 MeV and work at the KAPL naval reactor Lab below this energy. It was, however, performing well in integral validation. For the VII.1 we have developed a new, more consistent, R-matrix evaluation that better represents the available cross section data. The new evaluation of the n+ 16 O cross sections is based on an R-matrix analysis of reactions in the 17 O system at energies up to 7 MeV. Several new measurements were added to the data set for α + 13 C reactions. This resulted in a change that essentially undid the previous revision that had lowered the evaluated (n,α) cross section by 32% at energies below about 9 MeV. Validation performed so far indicates that the new file performs at least as good as the previous one, while the cross sections are in better agreement with the differential measurements. C. Halogens and alkali metals These new evaluations include resonance ranges in 19 F, 35,37 Cl, 39,41 K evaluated by ORNL and full new evaluation of 23 Na, which is still under development. The ORNL evaluations of neutron cross sections in the resolved resonance region were done with the multilevel Reich-Moore R-matrix formalism of the code SAMMY and incorporate recent high-resolution capture and transmission measurements. The resonance parameters for 19 F, 35 C were the first to use extended Reich-Moore formalism coded in the ENDF representation with the LRF=7 option. This novelty allows to describe open inelastic channels with the resonance parameters. For example, in the energy range up to 1 MeV 19 F has two inelastic channels starting at the energies kev with spin 1/2- and kev with spin 5/2+. These two inelastic channels were included in the SAMMY evaluation. Comparison of the thermal elastic, capture and total cross sections at room temperature with the data listed in Atlas of Neutron Resonances is generally very favorable. Only the case of thermal scattering cross-section for 41 K the Atlas value is almost 3 times smaller than that of the present evaluation. The reasons for the discrepancies are not known. The BNL evaluation of 23 Na is still in progress but it is worth mentioning because of the novelty in the evaluation approach. The methodology, applied in collaboration with INL, is called consistent adjustment or assimilation and involves use of both differential and in- 2

3 tegral data to constrain nuclear reaction model parameters. Thus the adjustment to the integral data occurs on the level of physical model parameters rather than on the level of multi-group cross sections, removing application dependance of the classical adjustment procedure. The consistent adjustment requires that the evaluation be fully determined by a set of model parameters and related nuclear reaction models and needs clean, well defined integral experiments, which are predominantly sensitive to the considered material. D. Structural materials New evaluations have been prepared for the following structural materials: Ti, and V (LANL), 55 Mn and W (IAEA), 50,52,53,54 Cr and 58,60 Ni (ORNL), 181 Ta, 185,187 Re (LLNL). It is interesting to note that important modification to the 48 Ti was elastic scattering angular distributions at low energies. In ENDF/B- VII.0 the elastic scattering angular distributions are calculated with the optical model and the Hauser-Feshbach statistical model. We found that the calculated P1 (the L = 1 component of the Legendre expansion coefficients) for the differential elastic scattering tends to overestimate reflection of neutrons in the critical assemblies with a Ti reflector, and adjustment of the optical potential parameters does not solve this problem. This was finally by replacing the elastic scattering angular distributions of 48 Ti up to 4 MeV by those in ENDF/B-VI in which the angular distributions were evaluated based experimental data. The larger P1 values in ENDF/B- VII.1 in the fast energy range give more neutron scattering in the forward angles, which results in less reflection of neutrons inside the reflector. New evaluations for tungsten isotopes are the high points of the VII.1 library. Notoriously poor performance of tungstens in VII.0 has been dramatically improved thanks to the consistent modeling in the fast neutron region with EMPIRE code. Here, the essential factors were the optical model potentials developed specifically for the current evaluations. This new class of dispersive optical model potentials serve as backbone of undertaken nuclear reaction modeling. Selected benchmarks covering both criticality and fusion domains were calculated for the assembled evaluated data file; benchmark results were analyzed to further constrain the theoretical modeling and model parameters. Then, the whole process was repeated starting from a new theoretical calculation and concluding with a new least- squares fitting that produces a new evaluated data file. Iterations were carried on until benchmarks performance does not improve anymore. Described feedback to the theoretical modeling based on benchmarks results is considered to be an extremely important step in improving evaluated data file. A big advantage of going from benchmarks results back to modeling stage is that internal consistency and general applicability of derived evaluated files are preserved. Such process involves a tight interaction of the modelers and benchmark evaluators and requires a truly interdisciplinary evaluation team. E. Fission products and absorbers Modified were: 95 Mo, 99 Tc, 103 Rh, 109 Ag, 133 Cs, 143,145 Nd, 147,149,152 Sm, 153 Eu, 62 Ni, 90,91 Zr, 113 Cd, and 157 Gd. Improvements to the fission products and absorbers regarded mostly adjustment of the thermal cross sections and in some cases of the full resonance region. These modifications were performed by Said Mughabghab (BNL) partly in response to the criticism by Dean who reported poor performance of VII.0 in his reactivity worths measurements. New analysis, also taking into account new experimental data when available, brought VII.1 significantly closer to the Dean s results. In addition, there were dramatic changes in thermal cross sections for 90 Zr and 91 Zr, for 113 Cd and 157 Gd suggested by the new experiments. In summary: Except for 95 Mo, the major significant discrepancies of the reactivity-worth results of Dean are removed in the ENDF/B-VII.1 evaluations. The 62 Ni capture cross section was reevaluated so that the computed ENDF/B-VII.1 30-keV Maxwellian capture cross section agreed with direct measurements. The thermal capture cross sections of 90 Zr and 91 Zr were reevaluated to account for a new measurement and those of 113 Cd and 157 Gd modified to reflect recent differential and integral measurements. F. Actinides While major actinides were not modified, except reverting to ENDF/B- VI.8 delayed neutron parameters, minor actinides 237 Np, 233, 236, 237, 238, 239U, 238,239,240 Pu, and 241 Am were either reevaluated or improved. The 237 Np evaluation has been updated using a recent evaluation of Maslov et al. Their evaluation concentrated on the production of the short-lived 236 Np isomer and replaces the (n,2n) and (n,3n) cross sections. Considered for adoption is Phil Youngs latest evaluation for 233 U that reevaluates the inelastic scattering cross sections following feedback from Trkov. We have made two modifications to the 236 U evaluation: in the energy region near 100 kev the fission cross section has been modified slightly to provide a smoother match to the unresolved resonance region; and radiative capture has been increased by about 10% for the energy region above 100 kev. The capture modification was motivated mainly by MCNP simulations of 236 U(n,γ) 3

4 in fast critical assemblies that previously underpredicted the measured LANL data by about 10%. The ENDF/B-VII.0 evaluation for 237 U had various limitations, originating largely from the fact that very little measured data exist for this unstable nucleus and nuclear reaction modeling has a limited predictive power for reactions that involve fission. Presently, model calculations have an improved predictive power once they involve some calibration to measured data on nearby uranium isotopes. These results were employed to derive new recommendations for 237 U in fast neutron region. These data were smoothly matched onto the JENDL3.3 resolved and unresolved resonance evaluation below 100 kev (we modified the unresolved resonance representation slightly to facilitate this). In performing our new evaluation for 239U, we changed the resonance region and all of the cross section data. Our new (n,f) cross section evaluation combines the reanalyses of Younes and Britt with the new surrogate reaction experiment performed by Burke et al. The resonance range was generated with the EMPIRE code within a picket fence model with realistic resonance spacings. A new evaluation was performed for neutron-induced reactions on Pu-238 in the fast neutron region. The evaluation is based on model calculations, as well as analysis of experimental data. The ECIS94 code was used to perform coupled-channels optical model calculations, and obtain total, shape and reaction cross sections, as well as all discrete elastic and inelastic cross sections and angular distributions. Neutron transmission coefficients used for statistical Hauser-Feshbach calculations were also inferred from the coupled-channels results. The optical model potential developed recently by Soukhovitskii et al. for even-even plutonium isotopes was used in this work. The GNASH and COMNUC codes, which implement the Hauser-Feshbach equations, width fluctuation corrections as well as pre-equilibrium components, were used to compute (n, xn) reaction cross sections. The COH code was used for computing the neutron radiative capture cross section. The GLUCS statistical analysis code was used to analyze experimental data sets, and in particular, infer the fission cross section as well as prompt fission neutron multiplicity. The JENDL-4.0 evaluation [160] was also used to complement the present work in certain areas. The average prompt fission neutron spectrum (PFNS) and multiplicity (PFNM) were evaluated using Los Alamos (Madland-Nix) model calculations. The systematics developed by Tudora and Vladuca for the model input parameters were used as prior parameters in our analysis. The evaluation of neutron-induced reactions on Pu- 240 follows the exact same methodology used for Pu-238. The ECIS, COH, GNASH, COMNUC, GLUCS, PFNS codes were also used here. Our americium 241 Am evaluation for ENDF/B-VII.1 builds on the VII.0 work, and makes some modest changes for fission and capture. We performed a new SOK code statistical analysis of measured data to obtain a new evaluated fission cross section down to 150 ev. The sub-threshold fission cross sections are given in File 3. G. Fission yields For ENDF/B-VII.1 a new fission product yield (FPY) evaluation for n+239pu was developed by Chadwick et al., updating the previous evaluation for ENDF/B- VI by England and Rider. The evaluation was not changed for incident neutrons at thermal energies as we find that the original evaluation is reliable here; but significant changes were made for incident neutrons with energies corresponding to fission spectrum, and 14 MeV neutrons. The fission spectrum evaluation work is described in detail in a previous issue of Nuclear Data Sheets. H. Decay data The decay data sub-library is based on the Evaluated Nuclear Structure Data File (ENSDF), following a translation into the ENDF-6 format. On average, data in ENSDF are updated every 7 years, which basically means that an important amount of the data in VII.0 was reevaluated. The VII.1 version of the sub-library includes fixes to errors in VII.0, such as the positron intensity, which was left as zero in the VII.0 version. Additionally, the atomic radiation part of the library was modified. Atomic radiation, X-rays and Auger electrons, are produced from the filling of atomic vacancies produced in electron capture and electron conversion. In the VII.1 version, the atomic data from the Evaluated Atomic Data Library (EADL) developed by LLNL was used. All the K-L, K-M and K-N as well as the L?, L? and L? X-rays are included. In addition, the KLL, KLX, KXY, LLX, LMM, LMX, LXY, MMX, and MXY average Auger electrons are also listed. The electron conversion to atomic sub-shells was calculated with the code BRICC. Finally, the average electromagnetic energy (EEM) and average light particle energy (ELP) values from the recently published TAGS data for 102,104,105,106,107 Tc, 101 Nb and 102 Mo, were also included in the VII.1 release. The inclusion of these data has improved the calculation of decay heat for n+ 239 Pu at cooling times seconds. III. RELEASE SCHEDULE The next preliminary release of the library (beta3) is currently being assembled and should be made available by middle of May This release will differ from the beta2 mostly in the AFCI and standards covariances that will be included. Subsequent beta4 version might be released between mini-csewg meeting in June and full CSEWG meeting in November This one will mostly contain formal fixes and eventual modifications 4

5 resulting from the validation exercise. A very few, previously agreed upon and well validated, new evaluations might be included (e.g., 23 Na) at this stage. Final release of the library is scheduled for December It will be accompanied by the dedicated issue of Nuclear Data Sheets with extended documentation of the library. Contained will be Big Paper II, about 100 pages, describing cross section part of the neutron sub-library and several papers dedicated to the covariances prepared by the participating Labs (BNL, LANL, ORNL). In addition, general overview of the covariances in ENDF/B-VII.1 will be given in a paper by D. Smith. 5