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arXiv 2024年

作者： Otuka, Naohiko Tada, Kenichi Cabellos, Oscar Iwamoto, Osamu Nuclear Data Section Division of Physical and Chemical Sciences Department of Nuclear Sciences and Applications International Atomic Energy Agency Wagramerstraße 5 WienA-1400 Austria Research Group for Reactor Physics Thermal Hydraulics Technology Japan Atomic Energy Agency Tokai-mura Naka-gun Ibaraki319-1195 Japan Department of Energy Engineering Division of Nuclear Engineering Universidad Politécnica de Madrid Madrid28006 Spain Instituto de Fusión Nuclear - "Guillermo Velarde" Universidad Politécnica de Madrid Madrid28006 Spain Nuclear Data Center Japan Atomic Energy Agency Tokai-mura Naka-gun Ibaraki319-1195 Japan

The uranium-233 neutron capture cross section between 3 keV and 1 MeV was evaluated with the new alpha value recently measured at the Los Alamos National Laboratory LANCE facility and compiled in the EXFOR library. The obtained capture cross section is systematically lower than those in the latest versions of the major general purpose nuclear data libraries, and the reduction from the JENDL-5 library is close to 50% around 20 keV. The newly evaluated cross section was benchmarked against 166 criticality experiments chosen from the ICSBEP handbook by performing Monte Carlo neutron transport calculation with the JENDL-5 library, and slight reduction of the cumulative chi-square value was achieved by adoption of the newly evaluated capture cross section. © 2024, CC BY.

关键词： Neutrons

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Current nuclear data needs for applications

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Physical Review Research 2022年第2期4卷 021001-021001页

作者： Karolina Kolos Vladimir Sobes Ramona Vogt Catherine E. Romano Michael S. Smith Lee A. Bernstein David A. Brown Mary T. Burkey Yaron Danon Mohamed A. Elsawi Bethany L. Goldblum Lawrence H. Heilbronn Susan L. Hogle Jesson Hutchinson Ben Loer Elizabeth A. McCutchan Matthew R. Mumpower Ellen M. O'Brien Catherine Percher Patrick N. Peplowski Jennifer J. Ressler Nicolas Schunck Nicholas W. Thompson Andrew S. Voyles William Wieselquist Michael Zerkle Nuclear and Chemical Sciences Division Lawrence Livermore National Laboratory Livermore California 94550 USA Department of Nuclear Engineering University of Tennessee Knoxville Tennessee 37996 USA Department of Physics and Astronomy University of California Davis California 95516 USA IB3 Global Solutions Oak Ridge Tennessee 37830 USA Physics Division Oak Ridge National Laboratory Oak Ridge Tennessee 37831 USA University of California Department of Nuclear Engineering Berkeley California 94720 USA Nuclear Science Division Lawrence Berkeley National Laboratory Berkeley California 94720 USA National Nuclear Data Center Brookhaven National Laboratory Upton New York 11973 USA Design Physics Division Lawrence Livermore National Laboratory Livermore California 94550 USA Rensselaer Polytechnic Institute Department of Mechanical Aerospace and Nuclear Engineering 110 8th Street Troy New York 12180 USA Signature Science and Technology Division Pacific Northwest National Laboratory Richland Washington 99352 USA Xe-Mobile Division X-Energy LLC Rockville Maryland 20852 USA Radioisotope Science and Technology Division Oak Ridge National Laboratory Oak Ridge Tennessee 37831 USA NEN-2 Advanced Nuclear Technology Los Alamos National Laboratory Los Alamos New Mexico 87545 USA Theoretical Division Los Alamos National Laboratory Los Alamos New Mexico 87545 USA Chemistry Division Los Alamos National Laboratory Los Alamos New Mexico 87545 USA Nuclear Criticality Safety Division Lawrence Livermore National Laboratory Livermore California 94550 USA Johns Hopkins University Applied Physics Laboratory Laurel Maryland 20723 USA Nuclear Energy and Fuel Cycle Division Oak Ridge National Laboratory Oak Ridge Tennessee 37831 USA Reactor Technology Department Naval Nuclear Laboratory West Mifflin Pennsylvania 15122 USA

Accurate nuclear data provide an essential foundation for advances in a wide range of fields, including nuclear energy, nuclear safety and security, safeguards, nuclear medicine, and planetary and space exploration. In these and other critical domains, outdated, imprecise, and incomplete nuclear data can hinder progress, limit precision, and compromise safety. Similar nuclear data needs are shared by many applications, thus prioritizing these needs is especially important and urgently needed. Many levels of analysis are required to prepare nuclear measurements for employment in end-user applications. Because research expertise is typically limited to one level, collaboration across organizations and international borders is essential. This perspective piece provides the latest advances in nuclear data for applications and describes an outlook for both near- and long-term progress in the field.

关键词： nuclear engineering nuclear reactors Quantum computation Monte Carlo methods nuclear data analysis & compilation

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Development of a fast reactor for minor actinides transmutation - Improvement of prediction accuracy for ma-related integral parameters based on cross-section adjustment technique

Development of a fast reactor for minor actinides transmutat...

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Physics of reactors 2016: Unifying Theory and Experiments in the 21st Century, PHYSOR 2016

作者： Yokoyama, Kenji Maruyama, Shuhei Numata, Kazuyuki Ishikawa, Makoto Takeda, Toshikazu Nuclear Data and Reactor Engineering Division Nuclear Science and Engineering Center Japan Atomic Energy Agency Ibaraki Japan Research Institute of Nuclear Engineering University of Fukui Fukui Japan Fast Reactor Computational Engineering Department Advanced Fast Reactor Cycle System Research and Development Center Japan Atomic Energy Agency Ibaraki Japan

ISBN: (纸本)9781510825734

In order to improve the prediction accuracy of the minor actinide transmutation in fast reactors, the cross-section adjustment technique was applied. In the present study, we utilized the MA-related integral experimental data that were extensively collected and evaluated in detail as a part of the ongoing project "Study on Minor Actinides Transmutation Using Monju data." This is one of the national projects entitled "technology development for the environmental burden reduction," which has been launched by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in Japan. Those data include neutronics parameters, such as the criticality and the sodium void reactivity measured in massively Np-loaded cores, MA fission reaction rate ratios measured in fast critical assemblies, MA samples irradiated in fast reactors, and neutronics parameters measured in an 241Am-accumulated fast breeder reactor. These MA-related integral experimental data were applied to the cross-section adjustment with the existing general, i.e., non-MA-related experimental data. For the criticality and sodium void reactivity, significant improvement of prediction accuracy is seen both for the cores with MA and without MA. As for the fission rate ratios, the cross-section adjustment result suggests the large reduction of 244Cm fission cross section.

关键词： Fast reactors

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Development and verification of three-dimensional hex-z burnup sensitivity solver based on generalized perturbaion theory

Development and verification of three-dimensional hex-z burn...

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2014 International Conference on Physics of reactors, PHYSOR 2014

作者： Yokoyama, K. Research Group for Reactor Physics Division of Nuclear Data and Reactor Enginnering Nuclear Science and Engineering Directorate Japan Atomic Energy Agency Tokai-mura Ibaraki319-1195 Japan

A burnup sensitivity analysis solver based on generalized perturbation theory has been de-veloped in a multi-purpose analysis framework MARBLE. The new solver has capability to cal-culate sensitivity coefficients of transmuted nuclear fuel composition after burnup, i.e. atomic number density, with respect to nuclear data. Based on the diffusion theory, the new solver calculates the burnup sensitivity coefficients not only in 2-dimensional R-Z model but also in 3-dimensional Hexagonal-Z model, that is not supported in a existing code system in JAEA. The new solver has been verified by comparing with results of so-called direct calculation. Accord-ing to the generalized perturbation theory, it is possible to calculate a breakdown of the burnup sensitivity coefficient by using five terms which appear in the formulation. In the present paper, a new numerical verification method named component-wise direct calculation is proposed to check the results of each term. A numerical experiment of the new verification method was per-formed, and it shows that the new verification method is valid. In addition, a sample application result is shown in order to evaluate calculation model effects due to difference between 2-and 3-dimensional models. The sample application result suggests that 3-dimensional calculation model is necessary for some heterogeneous core configuration. © 2014 Westinghouse Electric Company LLC. All Rights Reserved.

关键词： Marble

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Accelerator-based nuclear data measurements for nuclear science and technology 12

Accelerator-based nuclear data measurements for nuclear scie...

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12th International Topical Meeting on nuclear Applications of Accelerators 2015, AccApp 2015

作者： Plompen, A.J.M. Chadwick, M.B. Haight, R.C. Chiaveri, E. Danon, Y. Kimura, A. Ledoux, X. Ruan, X. European Commission Joint Research Center Inst. for Reference Materials and Measurements Retieseweg 111 Geel2440 Belgium Los Alamos National Laboratory Los AlamosNM87545 United States CERN Geneva 23CH-1211 Switzerland Gaerttner LINAC Center Rensselaer Polytechnic Institute NES Bldg 1-9 TroyNY12180 United States Nuclear Data and Reactor Engineering Division Nuclear Science and Engineering Center Japanese Atomic Energy Agency Ibaraki Tokai-mura319-1195 Japan Grand Accélérateur National d'Ions Lourds Bd. Henri Becquerel BP 55027 Caen Cedex 0514076 France Chinese Nuclear Data Center China Institute of Atomic Energy P.O. Box 275 Beijing102413 China

ISBN: (纸本)9781510842205

Accelerator-development and new experimental techniques at accelerator-based neutron sources are at the heart of improvements of nuclear data for science and technology. An important driver is the new interest for safety estimates based on full uncertainty propagation as opposed to the exclusive use of safety margins based on expert judgment. Large scale modeling facilitated by the massive growth of computing power emphasize the importance of nuclear data uncertainties which are often the main contributor to performance uncertainty. Advanced reactor development with its emphasis on resource economy, minimization of high-level waste and improved safety is another important driver. Accelerator driven systems may prove essential new systems for closing the fuel cycle. Global security, emergency preparedness and safeguards change the focus towards fission and neutron- and gamma-induced yields that help forensics and detecting illegal activities The challenges for nuclear data lie in the systematic provision of uncertainties and covariances, the substantial improvement of well-known key quantities such as those addressed in the CIELO initiative (Collaborative International Evaluated Library Organisation) and addressing some of the elusive nuclear data for minor actinides which are essential in addressing the high-level waste issues. In the field of fission measurements spectacular new techniques provide a wealth of data that opening new avenues. An overview of recent developments is presented from leading facilities around the world with a short outlook of where these developments may take us. © 12th International Topical Meeting on nuclear Applications of Accelerators 2015, AccApp 2015. All rights reserved.

关键词： Neutron sources

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Evaluation of direct disposal of spent fuel in existing dual-purpose canisters

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Radwaste Solutions 2014年第1期21卷 26-39页

作者： Hardin, E.L. Clayton, D.J. Howard, R.L. Clarity, J. Scaglione, J.M. Carter, J.T. Nutt, W.M. Clark, R.W. Advanced Nuclear Fuel Cycle Technologies Advanced Nuclear Energy Programs Sandia National Laboratories United States Used Fuel Systems Group Reactor and Nuclear Systems Division Oak Ridge National Laboratory United States Nuclear Data and Criticality Safety Group Reactor and Nuclear Systems Division Oak Ridge National Laboratory United States Savannah River National Laboratory United States Nuclear Engineering Division Argonne National Laboratory United States Department of Energy United States

The US nuclear power industry is accumulating spent nuclear fuel (SNF) in dry storage at the rate of approximately 2,000 metric tons (t) per year. Dry storage sites are associated with both operating and decommissioned power plants. The concept of using a common canister design for the storage, transport, and disposal of SNF originated in the 1990s as dry-storage systems were being deployed by the US utility industry. The potential advantages of standardized canisters were recognized, giving rise to multi-purpose canister (MPG) concepts developed for the US. Canister capacity and fuel burnup assumptions bracket the DPCs currently in storage, and also address the projected larger canisters and higher SNF burnup. The 150-year time limit on storage and disposal operations avoids more protracted, longer-term commitment to waste management activities.

关键词： nuclear industry

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Role of multiparticle-multihole states of O18,19 in O18(n,γ)O19 reactions at keV energy

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Physical Review C 2008年第5期77卷 051303(R)-051303(R)页

作者： T. Ohsaki M. Igashira Y. Nagai M. Segawa K. Muto Research Laboratory for Nuclear Reactors Tokyo Institute of Technology 2-12-1 O-okayama Meguro Toyo 152-8550 Japan Research Center for Nuclear Physics Osaka University Ibaraki Osaka 667-0047 Japan Division of Nuclear Data and Reactor Engineering Japan Atomic Energy Agency Tokai Ibaraki 319-1195 Japan Quantum Beam Science Directorate Japan Atomic Energy Agency Tokai Ibaraki 319-1195 Japan Department of Physics Tokyo Institute of Technology 2-12-1 O-okayama Meguro Tokyo 152-8551 Japan

We have determined for the first time the partial-capture cross sections for discrete γ rays from neutron capture by O18 to the 5/2+ground, 1/2+second, and 3/2+ first excited states in O19 at 10⩽En⩽80 keV. The first two cross sections agree with the calculated ones based on a nonresonant direct p-wave capture by O18 within a factor of two. On the other hand, the cross section to the 3/2+ state with a three-particle (d5/2)3configuration was found to be about 180 times larger than the calculated one. We discussed a possible origin of the large discrepancy by referring to recent results of O18(n,γ)O19 at thermal energy.

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Role of multiparticle-multihole states in O18,19 in thermal neutron capture of O18

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Physical Review C 2007年第5期76卷 051301(R)-051301(R)页

作者： Y. Nagai M. Segawa T. Ohsaki H. Matsue K. Muto Research Center for Nuclear Physics Osaka University. Ibaraki Osaka 567-0047 Japan Division of Nuclear Data and Reactor Engineering Japan Atomic Energy Agency Tokai Ibaraki 319-1195 Japan Quantum Beam Science Directorate Japan Atomic Energy Agency Tokai Ibaraki 319-1195 Japan Research Laboratory for Nuclear Reactors Tokyo Institute of Technology 2-12-1 O-okayama Meguro Tokyo 152-8550 Japan Department of Physics Tokyo Institute of Technology 2-12-1 O-okayama Meguro Tokyo 152-8551 Japan

The primary E1γ-ray transition strength from the thermal neutron capture of O18 to the 3/2− state, 18 keV below the neutron threshold of O19, was found to be about 2×105 stronger than that to the 1/2− excited state, 729 keV below the threshold. In addition, the E1 strength from the 3/2− state leading to the 3/2+ state with three (d5/2)3 neutrons was more than six times stronger compared to that to the 5/2+ ground state with a single (d5/2) neutron. These anomalous γ-ray decay patterns give the first clear experimental evidence for the predicted property of the 3/2− state as mainly having a one-hole-four-particle configuration. Unique features of the O18 ground state with a two-hole-four-particle configuration and of the 3/2+ state combined with a discrete prompt γ-ray spectroscopy following the thermal neutron capture of O18 enabled us to exclusively identify the configuration of the 3/2− state.

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Interpretation of integral test results of fp cross sections in jendl-3.2 by analyzing the stek experiments

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Journal of nuclear Science and Technology 2002年 39卷 982-985页

作者： Kawai, Masayoshi Watanabe, Takashi Zukeran, Atsushi Matsunobu, Hiroyuki Chiba, Satoshi Nakagawa, Tsuneo Nakajima, Yutaka Sugi, Teruo Dietze, Klaus High Energy Accelerator Research Organization Tsukuba 305-0801 Japan Kawasaki Heavy Industry Co. Ltd. Tokyo 136-8588 Japan Power and Industrial System R and D Division Hitachi 319-1221 Hitachi Ltd. 7-2-1 Omika Japan Data Engineering Inc. Fujisawa 251-0038 5-13-21 Matsu-ga-oka Japan Japan Atomic Energy Research Institute Ibaraki 319-1195 Japan Research Organization of Information for Science and Technology Ibaraki-ken 319-1106 Japan Reactor Physics Research Group Japan Nuclear Cycle Development Institute Ibaraki-ken Japan

The integral test of the fission product cross sections in JENDL-3.2 was made as an activity of the Subgroup 10 in the Working Party for Evaluation Cooperation in NEANSC by analyzing the sample reactivity worths measured at the STEK cores. Analyses were separately made by a JNDC team at JAERI and *** at JNC by using the 70-group cross sections. Neutron and adjoint fluxes used in the analysis by JNDC were converted from the 25-group ones which were adjusted to reproduce the reactivity worth for standard samples by Janssen et al. At ECN, while those used for analysis by *** were calculated with the reactor core design code system of JNC. The results of both analyses were very similar to each other for strongly absorbing nuclides and in good agreement with the experimental values except for nuclides heavier than A=130 for which the reactivity worth was underestimated by about 10%. However, there were some discrepancies for weakly absorbing nuclides: The analysis by *** gave a better agreement with the experimental values than those by JNDC. In order to interpret the discrepancies among both calculations and the experimental values, the neutron and adjoint fluxes used in the JNDC calculations were compared with those calculated by ***. Additionally, flux calculations were also made with diffusion and Monte Carlo codes. For weakly absorbing nuclides, a detailed comparison of sample reactivity worths as well as adjoint spectra was made. © 2014 Atomic Energy Society of Japan. All Rights Reserved.

关键词： Adjoint flux Fission product Flux Integral test JENDL-3.2 Reaction rate Sample worth STEK experiments

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A COMPUTATIONAL STUDY OF HY AND HSLA STEEL PERFORMANCE IN A BULKHEAD TEST PANEL GEOMETRY - REPLY

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NAVAL ENGINEERS JOURNAL 1991年第4期103卷 131-133页

作者： DEGIORGI, V MATIC, P Virginia Gensheimer DeGiorgiis a member of the Mechanics of Materials Branch at the Naval Research Laboratory in Washington D.C. She received a B.S. and MEng in civil engineering from the University of Louisville. She received her Ph.D. in engineering mechanics from Southern Methodist University. Prior to joining NRL she was associated with the Nuclear Components Division and Breeder Reactor Component Project of Westinghouse Electric Corporation in Tampa and Pensacola Florida. Her research interests include nonlinear computational modeling large strain response and fracture of materials material-structural response to combined loading environmental effects and structural integrity evaluation. Peter Maticis a member of the Mechanics of Materials Branch at the Naval Research Laboratory in Washington D.C. He received a B.S. in mechanical engineering from the Illinois Institute of Technology and a Ph.D. in applied mechanics from Lehigh University. Prior to joining NRL he was associated with the Electric Boat Division of General Dynamics Corporation in Groton Conn. His research interests include material constitutive behavior its measurement using computational techniques in conjunction with experimental data the mechanics of fracture structural integrity and the use of complex systems theory to analyze and model material deformation. In 1989 he received an NRL Alan Berman Publication Award and the ASNE “Jimmie” Hamilton Award.

As nonlinear computational methods become integrated into the structural design and analysis process, the effects of large deformation material response on large deflection structural response can, in principle, be predicted. The accuracy of structural predictions ultimately depends on an accurate assessment of the material responses which are encountered in the structure. This is particularly true if information on locally severe plastic deformation or fracture initiation and flaw tolerance is desired. Previously determined parameters for the elastic-plastic response of HY-80, HY-100, HSLA-80 and HSLA-100 steels, accurate through large strains terminating at the point of fracture, were applied in finite element simulations to predict the performance of a bulkhead test panel geometry. Significant differences in panel performance under hydrostatic loading conditions were predicted for the materials considered. The plastic deflections were found to be strongly influenced by the relative local or diffuse nature of intense plastic deformation across the panel. These results suggest that the comparative nonlinearity of these materials was as important as the strength and ductility in affecting performance. Further analyses, aimed at assessing flaw tolerance, were considered by conducting small scale analyses of local details where severe deformation conditions were predicted by the larger scale panel model.

关键词： Steel

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