Nuclear data are the essential data for nuclear physics scientific research, the development of nuclear energy as well as the applications of nuclear technology, and it is an important bridge ...connecting fundamental nuclear physics research with nuclear engineering and nuclear technology application. Nuclear data play important roles in national defense, nuclear security, national economic development and basic science research etc. A nuclear data research system which including nuclear data measurement, nuclear data evaluation, model calculation, nuclear data library development and nuclear data integral benchmarking, has been established in China since 1960’s. A lot of highreliability nuclear data, highlevel methodologies and libraries have been achieved with the effect of the national nuclear research network since China Nuclear Data Center established in 1975. All the output feed the needs of nuclear energy and nuclear technology applications, and nuclear data research also prompted the development of nuclear physics study in China. The nuclear data evaluation is an important part of nuclear data research, and it involves experimental data evaluation, nuclear model calculation, recommendation, complicated physics checking of complete set of data, evaluated data file integral verification and validation, etc. This paper briefly introduces the main research procedure of nuclear data evaluation, corresponding theoretical models and calculation codes, and the method of the establishment and validation testing for the latest version of Chinese Evaluated Nuclear Data Library (CENDL32). Some notable advances in CENDL32 are noted, such as new evaluation of the nn and np scattering cross section, model dependent covariance data for main reaction cross sections of some fission product nuclides and the updated evaluation of nuclear reaction data of several key nuclides, such as 7Li, 233,235U, 56Fe, 240Pu, etc. The evaluated nuclear data integral verification and validation is the essential links in the application of nuclear data to nuclear engineering. In order to verify the physical rationality, systematic comparisons between CENDL-3.2 and other major evaluated libraries (e.g. ENDF, JENDL, BROND, JEFF and TENDL) have been implemented using a suite of criticality benchmarks from the International Criticality Safety Benchmark Evaluation Project, and compared with available experimental data. This paper also briefly introduces relevant benchmarking testing and application results based on the CENDL32, and the χ2 value obtained implies that CENDL32 has a potential remarkable improvement of predictions for 235U and Pu systems. The CENDL32 library also has good performances on the reactor physics simulation of PWR, VENUS3 shielding benchmark by ARES transport code, HTR10 benchmark, etc. In general, integral validation performance of the CENDL32 library is improved relative to the previous CENDL31 library.
Accurate neutron cross section data are a vital input to the simulation of nuclear systems for a wide range of applications from energy production to national security. The evaluation of experimental ...data is a key step in producing accurate cross sections. There is a widely recognized lack of reproducibility in the evaluation process due to its artisanal nature and therefore there is a call for improvement within the nuclear data community. This can be realized by automating/standardizing viable parts of the process, namely, parameter estimation by fitting theoretical models to experimental data. There are numerous candidate methods to approach this type of problem, but many rely on large, labeled datasets that are not accessible to the nuclear data evaluator. For a reaction cross-section, there are usually just a handful of datasets, none of which can be considered labeled because evaluators never have access to the exact solution (cross section). Here this work leverages problem-specific physics, Monte Carlo sampling, and a general methodology for data synthesis to generate unlimited, labeled experimental cross-section data of high-utility. The synthesized data is said to be of high-utility because it is statistically similar to the observed data. Heuristic and, where applicable, rigorous statistical comparisons to observed data support this claim. The methodology is split into two generative models. The first generates a realization of an energy-differential cross section for a given isotope. The second takes the output from the first as a determined input and generates noisy experimental observables (radiation detector signals) from the determined cross section realization. The latter is the primary development of this article and is based/limited to transmission measurements at Rensselaer Polytechnic Institute (RPI). The former leverages an existing method for model parameter sampling in the resolved resonance region (RRR), thus limiting the current demonstration to the RRR of incident neutron energies. An open-source software is published alongside this article that executes the complete methodology to produce high-utility synthetic datasets. The goal of this work is to provide an approach and corresponding tool that will allow the evaluation community to begin exploring more data-driven, ML-based solutions to long-standing challenges in the field.
The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the ...release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment.
The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides
235,238U and
239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on
239Pu; and (9) A new decay data sublibrary.
Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications.
We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication M. B. Chadwick, P. Obložinský, M. Herman, N. M. Greene, R. D. McKnight, D. L. Smith, P. G. Young, R. E. MacFarlane, G. M. Hale, S. C. Frankle, A. C. Kahler, T. Kawano, R. C. Little, D. G. Madland, P. Moller, R. D. Mosteller, P. R. Page, P. Talou, H. Trellue, M. C. White, W. B. Wilson, R. Arcilla, C. L. Dunford, S. F. Mughabghab, B. Pritychenko, D. Rochman, A. A. Sonzogni, C. R. Lubitz, T. H. Trumbull, J. P. Weinman, D. A. Br, D. E. Cullen, D. P. Heinrichs, D. P. McNabb, H. Derrien, M. E. Dunn, N. M. Larson, L. C. Leal, A. D. Carlson, R. C. Block, J. B. Briggs, E. T. Cheng, H. C. Huria, M. L. Zerkle, K. S. Kozier, A. Courcelle, V. Pronyaev, and S. C. van der Marck, “ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,” Nuclear Data Sheets
107, 2931 (2006).
The TENDL library is now established as one of the major nuclear data libraries in the world, striving for completeness and quality of nuclear data files for all isotopes, evaluation methods, ...processing and applied performance. To reach this status, some basic principles have been applied which sets it apart from other libraries: reproducible dedicated evaluations when differential data are available, through determination of nuclear models implemented in TALYS and their parameters, completeness (with or without experimental data), format and processing standardization, automation of production and reproducibility. In this paper, we will outline how such an approach has become a reality, and recall some of the past successes since the first TENDL release in 2008. Next, we will demonstrate the performance of the latest TENDL releases for different application fields, as well as new approaches for uncertainty quantification based on Bayesian inference methods and possible differential and integral adjustments. Also, current limitations of the library performances due to modelling and needs for new and more precise experimental data will be outlined.
The evaluated 35Cl(n,p) cross sections in the ENDF/B-VIII.0 nuclear data library remain uncertain because of the lack of measured data in the 100–600 keV neutron energy range. In this energy region, ...the R-matrix analysis relies heavily on estimated (n,p) partial widths and level density information. This resulted in significant differences for the evaluated cross sections and predicted reactivity in nuclear reactor applications with respect to previous nuclear data libraries. This paper presents a methodology to quantify the cross sections for the 35Cl(n,p) reaction channel from the combination of measured total cross sections and model calculations.
With the increasing improvement of the core physics calculation method, the uncertainty introduced by the calculation software and model approximation in the process of nuclear reactor core physics ...design is becoming smaller and smaller, while the uncertainty of the calculation results caused by the uncertainty of the input parameters itself is becoming more and more important. As an important input parameter of core physics calculation, the uncertainty of nuclear data will be transmitted, which will affect the accuracy of core physics calculation results. Quantifying the uncertainty of the core physics calculation results caused by nuclear data is an important element in the nuclear reactor safety design process. On the other hand, according to the target accuracy requirements of the nuclear reactor design, the key nuclear data and the priority of the improvement when the physical parameters of the nuclear reactor core reach the target accuracy requirements are evaluated, which can provide reference standards for the improvement of nuclear data and guide the development of specific experimental measurements work.
•Development and validation of a nuclear data uncertainty calculation program.•Heat pipe cooled reactor sensitivity and uncertainty analysis.•Nuclear data target accuracy assessment.•Nuclear data improvement priority evaluation.
Measurement of the 145Sm half-life Kajan, I.; Pommé, S.; Pelczar, K. ...
Applied radiation and isotopes,
December 2021, 2021-12-00, 20211201, Letnik:
178
Journal Article
Recenzirano
Odprti dostop
The half-life of 145Sm has been measured by means of the reference source method with a HPGe detector. The long-lived radionuclide 44Ti was mixed into the source for reference. The time-dependency of ...the 145Sm/44Ti activity ratio was followed by assessing the count-rate ratio of their characteristic gamma-ray emissions at 61.2 keV (145Sm) and 67.9/78.3 keV (44Ti) in spectra recorded over periods of typically one day. In total, 220 measurements were performed over a period of 384 days or about one half-life period. The experiment and ensuing uncertainty budget are discussed in detail. Different error propagation is applied for random uncertainties, autocorrelated structures in the fit residuals, and potential systematic errors. The result for the 145Sm half-life, 345 (16) d, is compatible with the scarce literature values, however the experimental details of the old measurements were barely documented.
•Measured half-life of 145Sm by reference source method on a HPGe detector.•Used count ratio of 61.2 keV (145Sm) with reference peaks of 67.9 and 78.3 keV (44Ti).•Result T1/2(145Sm)=345 (16) d is consistent with recommended value of 340 (3) d.•Old experiments and data analysis are insufficiently documented.
Interest in safeguards verification measurements using passive thermal neutron counting to assay 235U content in large 30B UF6 canisters has grown in recent years. The prohibitively high cost and ...impracticality of using reference 30B calibration cylinders extensively will likely make accurate simulations of increasing interest. Accuracy of the simulated response will define the confidence in the predicted response and the extent to which simulations can reasonably be relied upon. With 234U driven 19F(α, n) reactions being the main neutron source in low enriched UF6 the uncertainties of the 19F(α, n) energy spectrum and the thick target yield of 234U in UF6 propagate into the uncertainty in the predicted response and represent a major influence of basic nuclear data. Here sensitivity of the simulated total (Singles) and coincidence (Doubles) count rates are assessed for the Passive Neutron Enrichment Meter using six potential 19F(α, n) neutron energy spectra over a range of enrichments and material distributions. The results indicate that variations in the Singles and Doubles due to simulated (α, n) neutron spectrum are less than 1.5% for this set of simulated neutron spectra, with dependence varying inversely with enrichment. Singles uncertainty is only slightly less than that of the thick target 19F(α, n) yield corresponding to the primary neutron source, whereas the 19F(α, n) yield dependence of the Doubles is reduced by the non-negligible 238U(SF) coincident neutron emissions. Based on available thick target 19F(α, n) yield estimates the uncertainty is on the order of 5%, establishing this as the main nuclear data limitation when simulating thermal neutron detectors response for 30B UF6 storage cylinders. Based on these findings, it appears that the measurement and evaluation of the thick target 19F(α, n) yield for uranium hexafluoride is due.
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