Development of a Unified Cross-section Set ADJ2000 based on Adjustment Technique for Fast Reactor Analysis

Itō, Makoto; Sugino, Kazuteru; Sato, Wakaei; Numata, Kazuyuki

doi:10.1080/00223131.2002.10875287

Cited by 25 publications

(17 citation statements)

References 1 publication

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“…This stems from the steadily growing need for nuclear data uncertainty information to be used in reactor core calculations (e.g., estimation of uncertainty in k eff , criticality safety studies, in the adjustment of nuclear data libraries [196,197]), in radiation shielding designs), and for various other applications, in order to assess safety, reliability, and cost effectiveness issues.…”

Section: H Covariancesmentioning

confidence: 99%

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

Chadwick

Obložinský

Herman

et al. 2006

Nuclear Data Sheets

1,930

908

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We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VII.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes. The new evaluations are based on both experimental data and nuclear reaction theory predictions.The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th, Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, 6 Li, 10 B, Au and for 235,238 U fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC); (3) Improved thermal neutron scattering; (4) An extensive set of neutron cross sections on fission products developed through a WPEC collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron-and proton-induced reactions up to an energy of 150 MeV; (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra; (9) New radioactive decay data; and (10) New methods developed to provide uncertainties and covariances, together with covariance evaluations for some sample cases.The paper provides an overview of this library, consisting of 14 sublibraries in the same, ENDF-6 format, as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The long-standing underprediction of low enriched U thermal assemblies is removed; (b) The 238 U, 208 Pb, and 9 Be reflector biases in fast systems are largely removed; (c) ENDF/B-VI.8 good agreement for simulations of highly enriched uranium assemblies is preserved; (d) The underprediction of fast criticality of 233,235 U and 239 Pu assemblies is removed; and (e) The intermediate spectrum critical assemblies are predicted more accurately.We anticipate that the new library will play an important role in nuclear technology applications, including transport simulations supporting national security, nonproliferation, advanced reactor and fuel cycle concepts, criticality safety, medicine, space applications, nuclear astrophysics, and nuclear physics facility design. The ENDF/B-VII.0 library is archived at the National Nuclear Data Center,

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Section: H Covariancesmentioning

confidence: 99%

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

Chadwick

Obložinský

Herman

et al. 2006

Nuclear Data Sheets

1,930

908

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“…Self-shielded cross sections were prepared for the homogeneous cell representations by using the adjusted cross-section library ADJ2000R [5,6]. This library incorporates integral experimental information about critical facilities and existing fast reactors.…”

Section: The Reference Fbr Corementioning

confidence: 99%

Correlations among FBR core characteristics for various fuel compositions

Maruyama

Ohki

Ōkubo

et al. 2012

Journal of Nuclear Science and Technology

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In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations.

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“…The detailed methodology of the cross section adjustment adopted here is reported in ref. [4]. The basic pre-adjustment library before applied here is JENDL-3.2 and its covariance data, but 237 Np data was adopted from JENDL-3.3, since there is no covariance data of 237 Np in JENDL-3.2.…”

Section: Improvement Of Prediction Accuracy For Np-loaded Cores By Usmentioning

confidence: 99%

Validation of Np-237 cross sections by analysis of BFS critical experiments with massively neptunium-loaded cores

Itō

Hazama

2007

Nd2007

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Abstract. Three series of BFS critical experiments loaded approximately with 10 kg of 237 Np-dioxide were analyzed to verify the performance of major libraries with the 237 Np nuclear data. The latest JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1 showed satisfactory results from the 237 Np data viewpoint; however some other major data such as 239 Pu ν and fission, 238 U µ and/or the sodium inelastic scattering cross sections of some libraries may need re-evaluation. Further, the effect of integral experimental information to obtain better predictions of reactor core parameters was demonstrated by the cross section adjustment method.

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Development of a Unified Cross-section Set ADJ2000 based on Adjustment Technique for Fast Reactor Analysis

Cited by 25 publications

References 1 publication

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

Correlations among FBR core characteristics for various fuel compositions

Validation of Np-237 cross sections by analysis of BFS critical experiments with massively neptunium-loaded cores

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