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1
Kataria, S. K., V. S. Ramamurthy, M. Blann, and T. T. Komoto. "Shell-dependent level densities in nuclear reaction codes." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 288, no. 2-3 (March 1990): 585–88. http://dx.doi.org/10.1016/0168-9002(90)90155-y.
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Denikin, Andrey, Alexander Karpov, Mikhail Naumenko, Vladimir Rachkov, Viacheslav Samarin, and Vycheslav Saiko. "Synergy of Nuclear Data and Nuclear Theory Online." EPJ Web of Conferences 239 (2020): 03021. http://dx.doi.org/10.1051/epjconf/202023903021.
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The paper describes the NRV web knowledge base on low-energy nuclear physics developed in the Joint Institute for Nuclear Research. The NRV knowledge base working through the Internet integrates a large amount of digitized experimental data on the properties of nuclei and nuclear reaction cross sections with a wide range of computational programs for modeling of nuclear properties and nuclear dynamics. Today, the NRV becomes a powerful instrument for nuclear physics research as well as for educational applications. Advantages of the functioning scheme of the knowledge base provide the synergy of coexistence of the experimental data and computational codes within one platform.
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Özdoğan, H., İsmail Hakki Sarpün, Mert Şekerci, and Abdullah Kaplan. "Production cross-section calculations of 111In via proton and alpha-induced nuclear reactions." Modern Physics Letters A 36, no. 08 (February 18, 2021): 2150051. http://dx.doi.org/10.1142/s0217732321500516.
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[Formula: see text], a known gamma emitter, is used for many medical purposes such as imaging of myocardial metastases. It can be produced by using different nuclear reactions. In this study, the reactions of [Formula: see text]Ag([Formula: see text]2n)[Formula: see text], [Formula: see text](p,[Formula: see text]n)[Formula: see text], [Formula: see text](p,[Formula: see text]2n)[Formula: see text], [Formula: see text](p,[Formula: see text]3n)[Formula: see text] and [Formula: see text](p,[Formula: see text]4n)[Formula: see text], which are the production routes of [Formula: see text], were investigated. Production cross-section calculations were performed by using equilibrium and pre-equilibrium models of TALYS 1.95 and EMPIRE 3.2 nuclear reaction codes. Hauser–Feshbach Model was appointed in both codes for calculations of equilibrium approximations. Exciton and Hybrid Monte Carlo Simulation (HMS) models were used in the EMPIRE 3.2, whereas Two-Component Exciton and Geometry Dependent Hybrid Model, which is implemented to TALYS code, has been used in the TALYS 1.95 for pre-equilibrium reactions. Also, a weighting matrix of the nuclear models was obtained by using statistical variance analysis. The optimum beam energy to obtain [Formula: see text] has been determined by using the results obtained from this weighting matrix.
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Hilaire, Stephane, Eric Bauge, Pierre Chau Huu-Tai, Marc Dupuis, Sophie Péru, Olivier Roig, Pascal Romain, and Stephane Goriely. "Potential sources of uncertainties in nuclear reaction modeling." EPJ Nuclear Sciences & Technologies 4 (2018): 16. http://dx.doi.org/10.1051/epjn/2018014.
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Nowadays, reliance on nuclear models to interpolate or extrapolate between experimental data points is very common, for nuclear data evaluation. It is also well known that the knowledge of nuclear reaction mechanisms is at best approximate, and that their modeling relies on many parameters which do not have a precise physical meaning outside of their specific implementations in nuclear model codes: they carry both specific physical information, and effective information that is related to the deficiencies of the model itself. Therefore, to improve the uncertainties associated with evaluated nuclear data, the models themselves must be refined so that their parameters can be rigorously derived from theory. Examples of such a process will be given for a wide sample of models like: detailed theory of compound nucleus decay through multiple nucleon or gamma emission, or refinements to the width fluctuation factor of the Hauser-Feshbach model. All these examples will illustrate the reduction in the effective components of nuclear model parameters, through the reduced dynamics of parameter adjustment needed to account for experimental data. The significant progress, recently achieved for the non-fission channels, also highlights the difficult path ahead to improve our quantitative understanding of fission in a similar way: by relying on microscopic theory.
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Sarpün, İsmail Hakki, Hasan Özdoğan, Kemal Taşdöven, Hüseyin Ali Yalim, and Abdullah Kaplan. "Theoretical photoneutron cross-section calculations on Osmium isotopes by Talys and Empire codes." Modern Physics Letters A 34, no. 26 (August 30, 2019): 1950210. http://dx.doi.org/10.1142/s0217732319502109.
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In this study, the level density parameter and the gamma ray strength function effects on photoneutron reaction cross-section calculations for Osmium isotopes were investigated by employing available level density models and gamma ray strength functions within Talys v1.8 and Empire v3.1 nuclear codes. A relative variance analysis was done to determine the best gamma ray strength function. Then, the effect of level density models for the photoneutron reactions was investigated by using the best gamma ray strength function. The results were compared with each other and also with the experimental data taken from the literature.
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Şekerci, Mert, Hasan Özdoğan, and Abdullah Kaplan. "Level density model effects on the production cross-section calculations of some medical isotopes via (α, xn) reactions where x = 1–3." Modern Physics Letters A 35, no. 24 (June 23, 2020): 2050202. http://dx.doi.org/10.1142/s0217732320502028.
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Level density models have an undeniable importance for a better perception on the nature of nuclear reactions, which influences our life via various ways. Many novel and advanced medical application use radioisotopes, which are produced with nuclear reactions. By considering the connection between the level density models and the importance of theoretical calculations for the production routes of medically important isotopes, this study is performed to investigate the level density model effects on the production cross-section calculations of [Formula: see text]Zn, [Formula: see text]Ga, [Formula: see text]Kr, [Formula: see text]Pd, [Formula: see text]In, [Formula: see text]I and [Formula: see text]At radioisotopes via some alpha particle induced and neutron emitting reactions. For theoretical calculations; frequently used computation tools, such as TALYS and EMPIRE codes, are applied. Obtained theoretical results are then compared with the experimental data, taken from Experimental Nuclear Reaction Data (EXFOR) library. For a better interpretation of the results, a mean weighted deviation calculation for each investigated reaction is performed in addition to a visual comparison of the graphical representations of the outcomes.
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Henning, Greg, Antoine Bacquias, Catalin Borcea, Mariam Boromiza, Roberto Capote, Philippe Dessagne, Jean-Claude Drohé, et al. "MEASUREMENT OF 182,184,186W (N, N’ γ) CROSS SECTIONS AND WHAT WE CAN LEARN FROM IT." EPJ Web of Conferences 247 (2021): 09003. http://dx.doi.org/10.1051/epjconf/202124709003.
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Today’s development of nuclear installations rely on numerical simulation for which the main input are evaluated nuclear data. Inelastic neutron scattering (n, xn) is a reaction of importance because it modifies the neutron population, the neutron energy distribution and may create new isotopes. The study of this reaction on tungsten isotopes is interesting because it is a common structural material. Additionally, tungsten isotopes are a good testing field for theories. The IPHC group started an experimental program with the GRAPhEME setup installed at the neutron beam facility GELINA to measure (n, xn γ) reaction cross sections using prompt gamma spectroscopy and neutron energy determination by time-of-flight. The obtained experimental data provide constraints on nuclear reaction mechanisms models for 182,184,186W. Indeed, to reproduce correctly the experimental (n, n’ γ) cross-sections, the reaction codes must include accurate models of the reaction mechanism, nuclear de-excitation process and use correct nuclear structure information.
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Sabra, M. S., Robert A. Weller, Marcus H. Mendenhall, Robert A. Reed, Michael A. Clemens, and A. F. Barghouty. "Validation of Nuclear Reaction Codes for Proton-Induced Radiation Effects: The Case for CEM03." IEEE Transactions on Nuclear Science 58, no. 6 (December 2011): 3134–38. http://dx.doi.org/10.1109/tns.2011.2169989.
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Korbut, Tamara, Maksim Kravchenko, Ivan Edchik, and Sergey Korneev. "Yalina-thermal facility neutron characteristic computational study 129I, 237Np and 243Am transmutation reaction rates calculations." EPJ Web of Conferences 239 (2020): 22013. http://dx.doi.org/10.1051/epjconf/202023922013.
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Present work describes Monte-Carlo calculations of the neutron field and minor actinide transmutation reaction rates within the Yalina-Thermal sub-critical assembly of the Joint Institute for Power and Nuclear Research – Sosny of the National Academy of Sciences of Belarus. The computer model of the facility was prepared for the corresponding calculations via MCU-PD and MCNP Monte-Carlo codes. The model neutron characteristics estimations were performed as well as the nuclear safety analysis. The up-to-date ENDF B/VIII, JEFF 3.3 and JENDL 4.0 nuclear data libraries were used during research.
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Voinov, A. V., S. M. Grimes, C. R. Brune, A. Bürger, A. Görgen, M. Guttormsen, A. C. Larsen, T. N. Massey, and S. Siem. "Level Density Inputs in Nuclear Reaction Codes and the Role of the Spin Cutoff Parameter." Nuclear Data Sheets 119 (May 2014): 255–57. http://dx.doi.org/10.1016/j.nds.2014.08.070.
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Bird, Robert F., Patrick Gillies, Michael R. Bareford, Andy Herdman, and Stephen Jarvis. "Performance Optimisation of Inertial Confinement Fusion Codes using Mini-applications." International Journal of High Performance Computing Applications 32, no. 4 (November 2, 2016): 570–81. http://dx.doi.org/10.1177/1094342016670225.
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Despite the recent successes of nuclear energy researchers, the scientific community still remains some distance from being able to create controlled, self-sustaining fusion reactions. Inertial Confinement Fusion (ICF) techniques represent one possible option to surpass this barrier, with scientific simulation playing a leading role in guiding and supporting their development. The simulation of such techniques allows for safe and efficient investigation of laser design and pulse shaping, as well as providing insight into the reaction as a whole. The research presented here focuses on the simulation code EPOCH, a fully relativistic particle-in-cell plasma physics code concerned with faithfully recreating laser-plasma interactions at scale. A significant challenge in developing large codes like EPOCH is maintaining effective scientific delivery on successive generations of high-performance computing architecture. To support this process, we adopt the use of mini-applications – small code proxies that encapsulate important computational properties of their larger parent counterparts. Through the development of a mini-application for EPOCH (called miniEPOCH), we investigate a variety of the performance features exhibited in EPOCH, expose opportunities for optimisation and increased scientific capability, and offer our conclusions to guide future changes to similar ICF codes.
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Enferadi, Milad, Mahdi Sadeghi, and Fatemeh Bolourinovin. "Accelerator production and nuclear aspects of 88Y: An efficient radiotracer." Nuclear Technology and Radiation Protection 26, no. 3 (2011): 201–8. http://dx.doi.org/10.2298/ntrp1103201e.
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Yttrium-88 (T1/2 = 106.6 d, Ib+ = 0.2% and IEC = 99.8%) is used in mixed gamma efficiency calibration standards and also as a substitute for 90Y to quantify the biodistribution of Y-pharmaceuticals in animals. Yttrium-88 and strontium-85 as gamma emitting radiotracers were produced via natSrCO3(p, xn) and natRbCl(p, xn) nuclear processes at AMIRS, with 18 and 15 MeV protons, respectively, at a current of 20 ?A for 10 hours. The deposition of natSrCO3 on the Cu backing was carried out by means of the sedimentation method. Yt- trium-88 was separated in a 92 ? 5% radiochemical yield using the precipitation technique by precipitating natSr as SrSO4 (strontium-85 as a tracer) while radioyttrium passed through the filter paper. Also, a theoretical study of the nuclear reaction cross-sections for proton and deuteron induced reactions on natSr and natRb for the production of 88Y and 85Sr, respectively, was performed using the EMPIRE (version 3.1 Rivoli), TALYS-1.26 codes and the TENDL-2010 database.
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Yiğit, M., and M. E. Korkmaz. "On the behavior of cross-sections of charged particle-induced reactions of 181Ta target." Modern Physics Letters A 33, no. 26 (August 24, 2018): 1850155. http://dx.doi.org/10.1142/s0217732318501559.
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The production of nuclear energy from fusion reaction with no CO2 emission is one of the most attractive sources for the future. The unprecedented physical and mechanical properties for structural materials are very important in the nuclear fusion reactor design. So, tantalum material is a valuable candidate for plasma-facing materials in the fusion devices. In this paper, nuclear excitation functions for the [Formula: see text], [Formula: see text], [Formula: see text], [Formula: see text], [Formula: see text] and [Formula: see text] reactions are obtained using the nuclear codes TALYS 1.8 and ALICE/ASH. The contribution of pre-equilibrium and equilibrium processes in these reactions is investigated. In the calculations, the Weisskopf–Ewing and Hauser–Feshbach formalisms for the equilibrium particle emission, and the two-component exciton, hybrid and geometry-dependent hybrid formalisms for the pre-equilibrium particle emission are used. Hence, the cross-sections calculated are compared with the measured values. It is observed that the cross-section results of the geometry-dependent hybrid model match fairly well with the experimental measurements.
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Kerveno, Maëlle, Marc Dupuis, Catalin Borcea, Marian Boromiza, Roberto Capote, Philippe Dessagne, Greg Henning, et al. "What can we learn from (n,xnγ) cross sections about reaction mechanism and nuclear structure?" EPJ Web of Conferences 239 (2020): 01023. http://dx.doi.org/10.1051/epjconf/202023901023.
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Inelastic (n,n') cross section is a key quantity to accurately simulate reactor cores, and its precision was shown to need significant improvements. To bypass the experimental difficulties to detect neutrons from (n,xn) reaction and to discriminate inelastically scattered neutrons from those following the fission process in case of fissile targets, an indirect but yet powerful method is used: the prompt γ-ray spectroscopy. Along this line, our collaboration has developed the GRAPhEME setup, optimized for actinides, at the GELINA facility to measure partial (n,xn γ) cross sections, from which the total (n,xn) cross section can be inferred. (n,xn γ) experiments with actinides are still particularly challenging, as their structure presents a high level density at low energy, and the competing neutron-induced fission reaction contaminates the γ-energy distribution. New precise measurements of the partial (n,xn γ) cross sections provide a stringent test to theoretical model and offer a way to improve them. This is a path to a better determination of the total inelastic scattering cross sections. In this contribution we discuss modeling aspects of the 238U and 182W (n,n' γ) reactions, also measured with GRAPhEME, using the three codes TALYS, EMPIRE and CoH. We will highlight the needed/expected improvements on reaction modeling and nuclear structure input.
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Enferadi, Milad, and Mahdi Sadeghi. "122Sb - a potential radiotracer: Evaluation of cyclotron production via novel routes." Nuclear Technology and Radiation Protection 26, no. 1 (2011): 58–63. http://dx.doi.org/10.2298/ntrp1101058s.
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Antimony-122, having a half-life of 2.723 d and I?-=97.59%, is an important radiotracer in studies of environmental contamination and food crops. For the work discussed in this paper, the production of 122Sb was done via the natSn(p, xn)122Sb nuclear reaction. Radiochemical separation was performed by silica gel column chromatography and liquid-liquid extraction methods. Excitation functions for the 122Sb radionuclide, via 122Sn(p, n)122Sb, natSn(p, xn)122Sb, 122Sn(d, 2n)122Sb, natSn(d, xn)122Sb, 124Sn(p, 3n)122Sb and 124Sn(d, 4n)122Sb reactions, were calculated by ALICE/91, ALICE/ASH, and TALYS-1.2 codes and compared with existing data.
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Watanabe, Yukinobu, Hiroki Sadamatsu, Shouhei Araki, Keita Nakano, Shoichiro Kawase, Tadahiro Kin, Yosuke Iwamoto, et al. "Study of the Li(d, xn) reaction for the development of accelerator-based neutron sources." EPJ Web of Conferences 239 (2020): 20012. http://dx.doi.org/10.1051/epjconf/202023920012.
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Double-differential neutron production cross sections (DDXs) for deuteron-induced reactions on Li at 200 MeV were measured for emission angles ranging from 0◦ to 25◦ in steps of 5◦ by means of a time of flight (TOF) method with EJ301 liquid organic scintillators at the Research Center for Nuclear Physics (RCNP), Osaka University. The measured DDXs were compared to theoretical model calculations with the DEURACS and PHITS codes and TENDL-2017 nuclear data. It was found that the DEURACS calculation is in better agreement with the measured DDXs than the PHITS calculation, while TENDL-2017 fails to reproduce both the spectral shape and magnitude of the measured DDXs for all angles.
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De Saint Jean, Cyrille, Pierre Tamagno, Pascal Archier, and Gilles Noguere. "CONRAD – a code for nuclear data modeling and evaluation." EPJ Nuclear Sciences & Technologies 7 (2021): 10. http://dx.doi.org/10.1051/epjn/2021011.
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The CONRAD code is an object-oriented software tool developed at CEA since 2005. It aims at providing nuclear reaction model calculations, data assimilation procedures based on Bayesian inference and a proper framework to treat all uncertainties involved in the nuclear data evaluation process: experimental uncertainties (statistical and systematic) as well as model parameter uncertainties. This paper will present the status of CONRAD-V1 developments concerning the theoretical and evaluation aspects. Each development is illustrated with examples and calculations were validated by comparison with existing codes (SAMMY, REFIT, ECIS, TALYS) or by comparison with experiment. At the end of this paper, a general perspective for CONRAD (concerning the evaluation and theoretical modules) and actual developments will be presented.
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Chasapoglou, Sotirios, A. Tsantiri, A. Kalamara, M. Kokkoris, V. Michalopoulou, A. Stamatopoulos, R. Vlastou, A. Lagoyannis, Z. Eleme, and N. Patronis. "Study of the 232Th(n,f) Cross section at NCSR ‘Demokritos’ using Micromegas Detectors." HNPS Proceedings 27 (April 17, 2020): 106. http://dx.doi.org/10.12681/hnps.2994.
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The accurate knowledge of neutron-induced fission cross sections in actinides, is of great importance when it comes to the design of fast nuclear reactors, as well as accelerator driven systems. Specifically for the 232Th(n,f) case, the existing experimental datasets are quite discrepant in both the low and high energy MeV regions, thus leading to poor evaluations, a fact that in turn implies the need for more accurate measurements.In the present work, the total cross section of the 232Th(n,f) reaction has been measured relative to the 235U(n,f) and 238U(n,f) ones, at incident energies of 7.2, 8.4, 9.9 MeV and 14.8, 16.5, 17.8 MeV utilizing the 2H(d,n) and 3H(d,n) reactions respectively, which generally yield quasi-monoenergetic neutron beams. The experiments were performed at the 5.5 MV Tandem accelerator laboratory of N.C.S.R. “Demokritos”, using a Micromegas detector assembly and an ultra thin ThO2 target, especially prepared for fission measurements at n_ToF, CERN during its first phase of operations, using the painting technique. The masses of all actinide samples were determined via α-spectroscopy. The produced fission yields along with the results obtained from activation foils were studied in parallel, using both the NeusDesc [1] and MCNP5 [2] codes, taking into consideration competing nuclear reactions (e.g. deuteron break up), along with neutron elastic and inelastic scattering with the beam line, detector housing and experimental hall materials. Since the 232Th(n,f) reaction has a relatively low energy threshold and can thus be affected by parasitic neutrons originating from a variety of sources, the thorough characterization of the neutron flux impinging on the targets is a prerequisite for accurate cross-section measurements, especially in the absence of time-of-flight capabilities. Additional Monte-Carlo simulations were also performed coupling both GEF [3] and FLUKA [4] codes for the determination of the detection efficiency.
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Raj, Prasoon, Ulrich Fischer, Axel Klix, and JET Contributors. "Comparative survey of evaluated nuclear data libraries for fusion-relevant neutron activation spectrometry." EPJ Web of Conferences 239 (2020): 21003. http://dx.doi.org/10.1051/epjconf/202023921003.
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The neutron flux-spectrum in a fusion device is frequently determined with activation foils and adjustment of a guess-spectrum in unfolding codes. Spectral-adjustment being a rather complex and uncertain procedure, we are carefully streamlining and evaluating it for upcoming experiments. Input nuclear cross-section data holds a vital position in this. This paper presents a survey of common dosimetry reactions and available data files relevant for fusion applications. While the IRDFF v1.05 library is the recommended source, many reactions of our interest are found missing in this. We investigated other standard sources: ENDF/B-VIII.0, EAF-2010, TENDL-2017, JENDL-4.0 etc. And, we analysed two experiments to ascertain the sensitivity of the spectral adjustment to the choice of nuclear data. One was performed with D-D (approx. 2.5 MeV peak) neutrons at the Joint European Torus (JET) machine and another with a white neutron field (approx. 33 MeV endpoint energy) at Nuclear Physics Institute (NPI) of Řež. Choice of cross-section source has affected the integral fluxes (<5%), reaction rates (<10%), total fluxes in some sensitive energy-regions (>20%) and individual group fluxes (<30%). Based on this experience, essential qualitative conclusions are made to improve the fusion activation-spectrometry.
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Hedden, O. F. "Status of Code Cases Related to Application of Risk-Informed Inservice Inspection in ASME Code Section XI." Journal of Pressure Vessel Technology 120, no. 4 (November 1, 1998): 438–40. http://dx.doi.org/10.1115/1.2842356.
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ASME Code Section XI Cases N-577 and N-578, for application of risk-informed technology to examination of piping systems in nuclear power plants, are proceeding, with review and acceptance by ASME Board on Nuclear Codes and Standards and by the U.S. Nuclear Regulatory Commission remaining before implementation. Sources of support for a favorable reaction by NRC will be reviewed, starting with developmental research sponsored by NRC in the late 1980s. Extensive discussion in the engineering community as exemplified by forums presented by ASME PVP in 1994 and 1995 will be cited. Recent academic and NRC managerial support for risk-informed performance-based regulation will also be cited. The expressed need for risk neutrality will then be addressed.
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Werneth, Charles M., K. M. Maung, M. D. Vera, and L. W. Townsend. "Optical potential for light nuclei and momentum-space eikonal phase function." Canadian Journal of Physics 96, no. 6 (June 2018): 642–49. http://dx.doi.org/10.1139/cjp-2017-0219.
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The space radiation environment comprises all of the nuclei in the periodic table with energies that extend from a fraction of an MeV/n to TeV/n. The vast range of projectile–target and energy combinations necessitates highly efficient and accurate cross section codes for use in radiation transport codes. As particles in the space radiation environment impinge on shielding materials, nuclear reactions, such as nuclear fragmentation, occur. One way of estimating nuclear fragmentation cross sections is to use an abrasion–ablation model, which describes how nucleons are dislodged from the nuclei as a result of nuclear collisions and the mechanism by which excited pre-fragments decay via particle emission to more stable states. The well-known partial wave solution method cannot be used directly for the computation of abrasion cross sections. Instead, abrasion cross sections may be computed by slightly altering the Eikonal solution method, which is a high energy (small scattering angle) approximation that depends on the nucleus–nucleus optical potential. The aim of the current work is to present two efficient methods for the computation of the Eikonal phase shift function. Analytic formulas of the optical potential are presented in the position-space representation for nuclei that are well-represented by harmonic-well nuclear matter densities (A < 20), which reduces the Eikonal phase factor to an integration over a single dimension. Next, the Eikonal phase function is presented in the momentum-space representation, which is particularly useful when the Fourier transform of the position-space optical potential is known. These new methods increase the computational efficiency by three orders of magnitude and allow for rapid prediction of elastic differential, total, elastic, and reaction cross sections in the Eikonal approximation.
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Khaliel, A., T. J. Mertzimekis, A. Psaltis, I. Psyrra, A. Kanellakopoulos, V. Lagaki, V. Foteinou, M. Axiotis, and S. Harrisopulos. "Experimental Investigation of radiative proton-capture reactions relevant to Nucleosynthesis." HNPS Proceedings 24 (April 1, 2019): 168. http://dx.doi.org/10.12681/hnps.1861.
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One of the primary objectives of the field of Nuclear Astrophysics is the study of the elemental and isotopic abundances in our solar system. Although a lot of progress has been made regarding a large number of nuclides, there is still a number of neutron-deficient nuclei, ie the p nuclei, which cannot be created via the s and r processes. These processes are responsible for the production of the bulk of heavy nuclides. The pre-explosive or explosive phases of massive stars are considered potential loci for p nuclides production via various combinations of photodisintegrations and nucleon captures, along with β+ decays and electron captures. For the study of the vast network of nuclear reactions (over 20'000) that are responsible for observed isotopic abundances, the statistical model of Hauser-Feshbach is employed. The model requires the knowledge of nuclear reaction cross sections, quantities that can be measured in the laboratory. In this work, we report on recent experimental attempts to measure such cross sections in radiative proton-capture reactions involving 107,109Ag near the astrophysically relevant energy window. Measurements have been performed at the Tandem Accelerator Laboratory of the N.S.C.R. “Demokritos”. The results are compared to various theoretical models, using the TALYS and EMPIRE codes, in an attempt to provide experimental input to astrophysical models.
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Al-Adili, A., A. Solders, and V. Rakopoulos. "Employing TALYS to deduce angular momentum rootmean-square values, Jrms, in fission fragments." EPJ Web of Conferences 239 (2020): 03019. http://dx.doi.org/10.1051/epjconf/202023903019.
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Fission fragments exhibit large angular momenta J, which constitutes a challenge for fission models to fully explain. Systematic measurements of isomeric yield ratios (IYR) are needed for basic nuclear reaction physics and nuclear applications, especially as a function of mass number and excitation energy. One goal is to improve the current understanding of the angular momentum generation and sharing in the fission process. To do so, one needs to improve the modeling of nuclear de-excitation. In this work, we have used the TALYS nuclear-reaction code to relax excited fission fragments and to extract root-mean-square (rms) values of initial spin distributions, after comparison with experimentally determined IYRs. The method was assessed by a comparative study on 252Cf(sf) and 235U(nth,f). The results show a consistent performance of TALYS, both in comparison to reported literature values and to other fission codes. A few discrepant Jrms values were also found. The discrepant literature values could need a second consideration as they could possibly be caused by outdated models. Our TALYS method will be refined to better comply with contemporary sophisticated models and to reexamine older deduced values in literature.
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Yokoyama, Atsushi, Shogo Katsura, and Akira Sugawara. "Biochemical Analysis of Histone Succinylation." Biochemistry Research International 2017 (2017): 1–7. http://dx.doi.org/10.1155/2017/8529404.
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Posttranslational modification (PTM) of proteins is used to regulate protein activity and stability. Histone PTMs are regarded as some of the most important, as they can directly regulate gene expression through chromatin reorganization. Recently, histone proteins were found to undergo succinylation, adding to other well-known PTMs such as acetylation, methylation, and phosphorylation. However, there is little information regarding the enzyme which catalyzes histone lysine succinylation. In fact, it is unclear whether this reaction is enzymatic. In this study, we tested histone succinylation activity in vitro using cell nuclear extracts of HepG2 cells. Although whole nuclear extracts did not show histone succinylation activity, we found that an SP 1.0 M KCl fraction of nuclear extracts indeed had such activity. These data offer the first direct evidence that histone succinylation is an enzymatic PTM as are other histone codes in the nucleus.
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Bahadir, Tamer. "BEAVRS BENCHMARK EVALUATION WITH CASMO5 AND SIMULATE5." EPJ Web of Conferences 247 (2021): 06022. http://dx.doi.org/10.1051/epjconf/202124706022.
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The MIT BEAVRS benchmark problem, which was primarily setup for the verification and validation of high-fidelity tools that have coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion models, has also found extensive usage in the reactor physics community for validating core analysis tools. The primary purpose of this paper is to provide an accurate, comprehensive evaluation of the BEAVRS benchmark with CASMO5 and SIMULATE5 codes. The CMS5 calculated results for low-power physics tests (hot zero power critical boron, control rod worth and isothermal temperature coefficients) and full power operation (boron let-down and flux map reaction rate distributions) are compared to plant measured data provided in the benchmark specification. The CMS5 model, using ENDF/BVII.1 nuclear data library, predicts HZP critical boron concentration for all-rods-out conditions within 10 ppm for Cycle-1, and 25 ppm in Cycle-2; the control rod worth is predicted with a difference of 0.7% ± 3.8%, where the maximum difference is less than 10%. For the core follow calculations at the hot full power condition, the average difference in predicting the critical boron concentration is less than 20 ppm. In addition, the radial and nodal reaction rate distributions are predicted with a mean difference of about 1.6% and 3.8%, respectively. The CMS5 calculations are repeated using the most recent ENDF/B-VIII.0 library. No significant difference is observed in predicting measured plant parameters with different nuclear data libraries. Additionally, the impact of various modeling options, which are typically employed with nodal diffusion codes, on the predictions of important core parameters are presented as part of the benchmark evaluation.
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Fleming, M., J.-C. David, J. L. Rodríguez-Sánchez, L. Fiorito, M. Gilbert, and T. Stainer. "The High-Energy Intra-Nuclear Cascade Liège-based Residual (HEIR) nuclear data library." EPJ Web of Conferences 239 (2020): 20001. http://dx.doi.org/10.1051/epjconf/202023920001.
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It is standard practice for nuclear data files to include tabulated data for distinct reaction channels for incident energies up to 20-30 MeV. Above these energies, the assumptions implicit in the definition of individual channels break down and event generators are typically used within codes that simulate nuclear observables in applications. These offer robust simulation of the physics but increase the computational burden. So-called ‘high-energy’ nuclear data files have been produced, but the well-known libraries are more than a decade old and rely upon models developed many years before their release. This presentation describes a modern library with a high level of production automation that offers regular updates as the models it is based upon are improved. The most recent versions of the intra-nuclear cascade and de-excitation models available within Geant4 were used to generate tabulated data of residual nuclide production. For the first released library, the INCL++5.3 and ABLA version within Geant4 v10.3 were used to calculate over 1012 incident protons over 2095 target isotopes with incident energies up to 1 GeV. These were collated into tabulated data in the international-standard ENDF-6 format. The resulting files were provided as group-wise files and were distributed as HEIR-0.1 with the FISPACT-II version 4.0 release. A second library, HEIR-0.2, has been generated using the new INCL++6.0 and C++ translation of the ABLA07 model available within Geant4 v10.4. Simulations were performed using incident protons, neutrons, deuterons and π±. An improved agreement is observed in the comparison to experimental data not only between the two versions, but against the other well-known high-energy nuclear data files and models available within Geant4. This benchmark includes mass and isotopic distributions, as well as incident-energy dependent cumulative and independent cross sections from the EXFOR database.
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Milocco, A., and A. Trkov. "Modelling of the Production of Source Neutrons from Low-Voltage Accelerated Deuterons on Titanium-Tritium Targets." Science and Technology of Nuclear Installations 2008 (2008): 1–7. http://dx.doi.org/10.1155/2008/340282.
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Fast quasi-monoenergetic neutrons can be produced by accelerating charged deuterons on tritium solid targets. Benchmark experiments were performed in many laboratories with intense D-T neutron sources. The aim is to validate the computational models and nuclear data for fusion applications. The detailed information on the neutron source term is highly important for the benchmark analyses. At present, the MCNP family of codes cannot explicitly model the D-T reaction for Deuterons in the KeV energy range. The physics for the D-T neutron production was modelled at ENEA (Italy) in the SOURCE and SRCDX subroutines to compile with the MCNP source code. Some improvements to the original subroutines were introduced. The differential cross-sections for the D-T reaction from the ENDF/B-VII library were built into the code. The relativistic approach was implemented for neutron kinematics. The new D-T neutron source model was applied to the MCNP5 simulation of the tungsten integral experiment performed at the OKTAVIAN facility. The uncertainty associated with the realistic D-T reactions was separated from the total uncertainty of the source term. The outcome of the benchmark analysis was an improvement in the quality of the computational model of the experiment.
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Vinayak, A., M. M. Hosamani, P. N. Patil, and N. M. Badiger. "Determination of single neutron spectroscopic factor of doubly shell closed, neutron shell closed and neutron-rich nuclei through (d,p) reaction." International Journal of Modern Physics E 29, no. 06 (June 2020): 2050030. http://dx.doi.org/10.1142/s0218301320500305.
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The spectroscopic factor (SF) of doubly-magic nuclei, neutron shell closed and neutron-rich nuclei has been determined through ([Formula: see text], [Formula: see text]) reaction in the projectile energy range from 3 to 26[Formula: see text]MeV. The theoretical angular differential cross-sections of ([Formula: see text], [Formula: see text] reactions in scattering center-of-mass angles from [Formula: see text] to [Formula: see text] have been calculated using FRESCO and NRV-DWUCK5 codes. By comparing the theoretical angular differential cross-sections with available experimental angular differential cross-sections, the values of SF have been determined. The exponential increase of SF as a function of neutron separation energy normalized by spin of the recoil nuclei has been shown for the first time for doubly-magic nuclei. The similar type of trend has also been observed for neutron-rich as well as neutron shell closed nuclei as a function of neutron separation energy normalized by asymmetric factor of recoil nucleus. More experimental data are required to verify the trend predicted by this investigation.
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Özdoğan, Hasan, Yiğit Ali Üncü, Mert Şekerci, and Abdullah Kaplan. "A study on the estimations of (n, t) reaction cross-sections at 14.5 MeV by using artificial neural network." Modern Physics Letters A 36, no. 23 (July 30, 2021): 2150168. http://dx.doi.org/10.1142/s0217732321501686.
Full textAbstract:
In this paper, calculations of the [Formula: see text] reaction cross-sections at 14.5 MeV have been presented by utilizing artificial neural network algorithms (ANNs). The systematics are based on the account for the non-equilibrium reaction mechanism and the corresponding analytical formulas of the pre-equilibrium exciton model. Experimental results, obtained from the EXFOR database, have been used to train the ANN with the Levenberg–Marquardt (LM) algorithm which is a feed-forward algorithm and is considered one of the well-known and most effective methods in neural networks. The Regression [Formula: see text] values for the ANN estimation have been determined as 0.9998, 0.9927 and 0.9895 for training, testing and for all process. The [Formula: see text] reaction cross-sections have been reproduced with the TALYS 1.95 and the EMPIRE 3.2 codes. In summary, it has been demonstrated that the ANN algorithms can be used to calculate the [Formula: see text] reaction cross-section with the semi-empirical systematics.
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Kodeli, Ivan A., Maurizio Angelone, and Davide Flamini. "NUCLEAR DATA SENSITIVITY/UNCERTAINTY PRE-ANALYSIS OF FNG WCLL FUSION BENCHMARK." EPJ Web of Conferences 247 (2021): 15004. http://dx.doi.org/10.1051/epjconf/202124715004.
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To assure tritium self-sufficiency in future fusion reactors such as DEMO the accuracy of TRP calculations has to be demonstrated within the design uncertainties. A new neutronics experiment representing a mock-up of the Water Cooled Lithium Lead (WCLL) Test Blanket Module (TBM) is under preparation at the Frascati neutron generator (FNG) with the objective to provide an experimental validation of accuracy of nuclear data and neutron transport codes for the tritium production rate (TPR) calculations. The mock-up will consist of LiPb bricks, EUROFER plates and Perspex substituting water. The mock-up will be irradiated by 14 MeV neutrons at the FNG facility, and the TPR and detector reaction rates will be measured using Li2CO3 pellets and activation foils placed at different positions up to about 55 cm inside the mock-up. Computational pre-analyses for the design of the WCLL neutronics experiment using the SUSD3D sensitivity/uncertainty (S/U) code system is described and compared with the results of some similar FNG experiments performed in the past, in particular the FNG HCPB Tritium Breeder Module Mock-up (2005) and FNG-HCLL Tritium Breeder Module Mock-up (2009). The objective of the pre-analysis is to provide the calculated nuclear responses including the uncertainties due to the uncertainties in nuclear data and thus contributes to the optimisation of the design of the experimental set-up.
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Simmonds, M. J., J. H. Yu, Y. Q. Wang, M. J. Baldwin, R. P. Doerner, and G. R. Tynan. "Expanding the capability of reaction-diffusion codes using pseudo traps and temperature partitioning: Applied to hydrogen uptake and release from tungsten." Journal of Nuclear Materials 508 (September 2018): 472–80. http://dx.doi.org/10.1016/j.jnucmat.2018.05.080.
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Isazadeh, Farzad, and akbar abdi saray. "CALCULATION OF CROSS SECTION AND PRODUCTION YIELD OF RADIOPHARMACEUTICAL PRASEODYMIUM-139 THROUGH 140CE(P, 2N)139PR REACTION USING GEANT4 AND TALYS NUCLEAR CODES." Studies in Medical Sciences 31, no. 9 (November 1, 2020): 680–89. http://dx.doi.org/10.29252/umj.31.9.680.
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Yuferov, Anatoliy G. "Infological models of the ENDF-format nuclear data." Nuclear Energy and Technology 5, no. 1 (March 20, 2019): 53–59. http://dx.doi.org/10.3897/nucet.5.33984.
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Issues involved in the infologic modeling of the ENDF-format nuclear data libraries for the purpose of converting ENDF files into a relational database have been considered. The transfer to a relational format will make it possible to use standard readily available tools for nuclear data processing which simplify the conversion and operation of this data array. Infological models have been described using formulas of the “Entity (List of Attributes)” type. The proposed infological formulas are based on the physical nature of data and theoretical relations. This eliminates the need for a special notation to be introduced to describe the structure and the content of data, which, in turn, facilitates the use of relational formats in codes and solution of nuclear data evaluation problems. The concept of nuclear informatics has been formulated based on relational DBMS technologies as one of the tools for solving the “big data” problem in modern science and technology. The organizational and technological grounds for the transfer of ENDF libraries to a relational format are presented. Requirements to the nuclear data presentation formats supported by relational DBMS are listed. Peculiarities of the infological model construction, conditioned by the hierarchical nature of nuclear data, are identified. The sequence for the ENDF metadata saving is presented, which can be useful for the verification and validation (testing of the structural and syntactical validity and operability) of both source data and the procedures for the conversion to a relational format. Formulas of infological models are presented for the cross sections file, the secondary neutron energy distributions file, and the nuclear reaction product energy-angle distributions file. A complete array of infological models for ENDF libraries and the generation modules of respective relational tables are available on a public website.
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Wang, Jong Rong, Hao Tzu Lin, Yung Shin Tseng, and Chun Kuan Shih. "Application of TRACE and CFD in the Spent Fuel Pool of Chinshan Nuclear Power Plant." Applied Mechanics and Materials 145 (December 2011): 78–82. http://dx.doi.org/10.4028/www.scientific.net/amm.145.78.
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In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. After Fukushima NPP event, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of the spent fuel pool for Chinshan NPP which also assumed the cooling system of the spent fuel pool failed. The geometry of the Chinshan NPP spent fuel pool is 12.17 m × 7.87 m × 11.61 m and the initial condition is 60 ¢J / 1.013 × 105 Pa. In general, the NPP safety analysis is performed by the thermal hydraulic codes. The advanced thermal hydraulic code named TRACE for the NPP safety analysis is developing by U.S. NRC. Therefore, the safety analysis of the spent fuel pool for Chinshan NPP is performed by TRACE. Besides, this safety analysis is also performed by CFD. The analysis result of TRACE and CFD are similar. The results show that the uncovered of the fuels occur in 2.7 days and the metal-water reaction of the fuels occur in 3.5 days after the cooling system failed.
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Solis, Augusto Hernandez, Ivan Merino Rodriguez, Luca Fiorito, and Gert Van den Eynde. "A NOVEL COMPUTATIONAL PLATFORM FOR THE PROPAGATION OF NUCLEAR DATA UNCERTAINTIES THROUGH THE FUEL CYCLE CODE ANICCA." EPJ Web of Conferences 247 (2021): 13007. http://dx.doi.org/10.1051/epjconf/202124713007.
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This paper presents the first results of a computational platform dedicated to the propagation of nuclear data covariances, all the way to fuel cycle scenario observables. Such platform, based on in-house codes developed at SCK•CEN in Belgium, both for the creation of the many-randomized nuclear data libraries based on ENDF format and for fuel cycle scenario-studies (known as SANDY and ANICCA, respectively), was employed for the uncertainty assessment of the time-dependent inventory computed from a mono-recycling of Plutonium scenario based on a PWR fleet. An essential part of the procedure that deals with the creation of input data libraries to ANICCA, has been carried out this time by the SERPENT2 code. Due to the fact that its neutron transport and depletion parallelized calculation in 72 cores for up to 1640 days and 60 MWd/kg-HM takes almost one hour, it is feasible to finish a total of 100 ANICCA runs based on randomized input libraries created from ENDF/B-VII.1 neutron-reaction covariances in about one week. Therefore, it is consider that the computation of the output population statistics can be inferred from 100 observables representing time-dependent mass inventories. To mention a few results from the aforementioned NEA/OECD benchmark scenario, it was found out that the relative standard deviation of the accumulated plutonium in the final disposal after 120 years was of 7%, while for curium it corresponded to 8%. Thus, sources of uncertainty arising from neutron-reaction covariances do have an impact in the final quantitative analysis of the fuel cycle output uncertainties.
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Azizakram, Hamid, Mahdi Sadeghi, Parviz Ashtari, and Farhad Zolfagharpour. "A Monte Carlo approach to calculate the production prerequisites of 124I radioisotope towards the activity estimation." Nuclear Technology and Radiation Protection 33, no. 1 (2018): 68–74. http://dx.doi.org/10.2298/ntrp1801068a.
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The Monte Carlo simulation code MCNPX has been used to simulate the production of 124I by 124,125Te(p, xn) and 123,124Te(d,xn) reactions to form high activity 124I. For this reason, the TALYS-1.8 and ALICE/ASH codes were used to calculate the reaction cross-section. The optimal energy range of projectile is selected for this production by identifying the maximum cross-section and the minimum impurity due to other emission channels. Target geometry is designed by SRIM code based on stopping power calculations with identical dimensions as the experimental data. The thick target yield of reactions is predicted because of the excitation functions and stopping power. All of the prerequisites obtained from the above interfaces are adjusted in MCNPX code and the production process is simulated according to benchmark experiments. Thereafter, the energy distribution of proton in targets, the amount of residual nuclei during irradiation, were calculated. The results are in good agreement with the reported data, thus confirming the usefulness and accuracy of MCNPX as a tool for the optimization of other radionuclides production. Based on the results, the 124Te(p,n)124I process seems to be the most likely candidate to produce the 124I in low-energy cyclotrons.
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Ruan, Zhenglin, and Haibing Guo. "A HIGH-FIDELITY SIMULATION OF THE C5G7 BENCHMARK BY USING THE PARALLEL ENTER CODE." EPJ Web of Conferences 247 (2021): 06023. http://dx.doi.org/10.1051/epjconf/202124706023.
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In simulation of advanced nuclear reactors, requirements like high precision, high efficiency and convenient to multi-physics coupling are putting forward. The deterministic transport method has the advantage of high efficiency, capable of obtaining detailed flux distribution and efficient in multi-physics coupling, but its accuracy is limited by the homogenized reaction cross-section data and core modelling exactness. The traditional two-steps homogenization strategy may introduce substantial deviation during the assembly calculation. It is possible to conduct a whole core deterministic transport simulation pin-by-pin to achieve higher accuracy, which eliminates the assembly homogenization process. The C5G7 benchmarks were proposed to test the ability of a modern deterministic transport code in analyzing whole core reactor problems without spatial homogenization. Different deterministic code that developed by different methods were applied to the benchmark simulation and some of them solved the benchmark accurately. However, there still exist some drawbacks in the given calculation processes which carried out by some other deterministic transport codes and we could find that the fuel pin cell in the assembly were not exactly geometrically modelled owing to the limit of the code. Consequently, the calculation precision could be improved by utilizing a high-fidelity geometry modelling. In this paper, the C5G7 benchmarks with different control rod position and different configuration were calculated by the finite element SN neutron transport code ENTER [1], and the results were presented after massively parallel computation on TIANHE-II supercomputer. By introducing a large scale high-fidelity unstructured meshes, high fidelity distributions of power and neutron flux were gained and compared with the results from other codes, excellent consistency were observed. To sum up, the ENTER code can meet those new requirements in simulation of advanced nuclear reactors and more works and researches will be implemented for a further improvement.
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Kushnir, Doron, and Boaz Katz. "An accurate and efficient numerical calculation of detonation waves in multidimensional supernova simulations using a burning limiter and adaptive quasi-statistical equilibrium." Monthly Notices of the Royal Astronomical Society 493, no. 4 (February 28, 2020): 5413–33. http://dx.doi.org/10.1093/mnras/staa594.
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ABSTRACT Resolving the small length-scale of thermonuclear detonation waves (TNDWs) in supernovae is currently not possible in multidimensional full-star simulations. Additionally, multidimensional simulations usually use small, oversimplistic reaction networks and adopt an ad hoc transition criterion to nuclear statistical equilibrium (NSE). The errors due to the applied approximations are not well understood. We present here a new accurate and efficient numerical scheme that accelerates the calculations by orders of magnitudes and allows the structure of TNDWs to be resolved. The numerical scheme has two important ingredients: (1) a burning limiter that broadens the width of the TNDW while accurately preserving its internal structure, and (2) an adaptive separation of isotopes into groups that are in nuclear statistical quasi-equilibrium, which resolves the time-consuming burning calculation of reactions that are nearly balanced out. Burning is calculated in situ employing the required large networks without the use of post-processing or pre-describing the conditions behind the TNDW. In particular, the approach to and deviation from NSE are calculated self-consistently. The scheme can be easily implemented in multidimensional codes. We test our scheme against accurate solutions of the structure of TNDWs and against homogeneous expansion from NSE. We show that with resolutions that are typical for multidimensional full-star simulations, we reproduce the accurate thermodynamic trajectory (density, temperature, etc.) to an accuracy that is better than a per cent for the resolved scales (where the burning limiter is not applied), while keeping the error for unresolved scales (broadened by the burning limiter) within a few per cent.
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Privas, Edwin, Cyrille De Saint Jean, and Gilles Noguere. "On the use of the BMC to resolve Bayesian inference with nuisance parameters." EPJ Nuclear Sciences & Technologies 4 (2018): 36. http://dx.doi.org/10.1051/epjn/2018042.
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Nuclear data are widely used in many research fields. In particular, neutron-induced reaction cross sections play a major role in safety and criticality assessment of nuclear technology for existing power reactors and future nuclear systems as in Generation IV. Because both stochastic and deterministic codes are becoming very efficient and accurate with limited bias, nuclear data remain the main uncertainty sources. A worldwide effort is done to make improvement on nuclear data knowledge thanks to new experiments and new adjustment methods in the evaluation processes. This paper gives an overview of the evaluation processes used for nuclear data at CEA. After giving Bayesian inference and associated methods used in the CONRAD code [P. Archier et al., Nucl. Data Sheets 118, 488 (2014)], a focus on systematic uncertainties will be given. This last can be deal by using marginalization methods during the analysis of differential measurements as well as integral experiments. They have to be taken into account properly in order to give well-estimated uncertainties on adjusted model parameters or multigroup cross sections. In order to give a reference method, a new stochastic approach is presented, enabling marginalization of nuisance parameters (background, normalization...). It can be seen as a validation tool, but also as a general framework that can be used with any given distribution. An analytic example based on a fictitious experiment is presented to show the good ad-equations between the stochastic and deterministic methods. Advantages of such stochastic method are meanwhile moderated by the time required, limiting it's application for large evaluation cases. Faster calculation can be foreseen with nuclear model implemented in the CONRAD code or using bias technique. The paper ends with perspectives about new problematic and time optimization.
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Fazaeli, Yousef, Mohammadreza Aboudzadeh, Khosro Aardaneh, Tayyeb Kakavand, Fatemeh Bayat, and Kamran Yousefi. "A new approach to targetry and cyclotron production of 45Ti by proton irradiation of 45Sc." Nuclear Technology and Radiation Protection 29, no. 1 (2014): 28–33. http://dx.doi.org/10.2298/ntrp1401028f.
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Titanium-45 with a half-life of 3.09 hours decays by emission of positrons (85%) and the electron capture process (15%). These properties make this radionuclide useful in the diagnosis of tumors by positron emission tomography. In this study, after having considered the excitation functions for the 45Sc(p, n)45Ti reaction using TALYS and ALICE/ASH codes and after the comparison with other experimental data, 45Ti was produced by dint of the pressing method and a newly designed and manufactured shuttle and capsule, resulting in an experimental yield of 403.3 MBq/mAh. Essential target thickness and physical yield were calculated. The scandium oxide target was irradiated at a 20 mA current and a 21 MeV proton beam energy for 1 hour.
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Nicol, D. G., P. C. Malte, A. J. Hamer, R. J. Roby, and R. C. Steele. "Development of a Five-Step Global Methane Oxidation-NO Formation Mechanism for Lean-Premixed Gas Turbine Combustion." Journal of Engineering for Gas Turbines and Power 121, no. 2 (April 1, 1999): 272–80. http://dx.doi.org/10.1115/1.2817117.
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It is known that many of the previously published global methane oxidation mechanisms used in conjunction with computational fluid dynamics (CFD) codes do not accurately predict CH4 and CO concentrations under typical lean-premixed combustion turbine operating conditions. In an effort to improve the accuracy of the global oxidation mechanism under these conditions, an optimization method for selectively adjusting the reaction rate parameters of the global mechanisms (e.g., pre-exponential factor, activation temperature, and species concentration exponents) using chemical reactor modeling is developed herein. Traditional global mechanisms involve only hydrocarbon oxidation; that is, they do not allow for the prediction of NO directly from the kinetic mechanism. In this work, a two-step global mechanism for NO formation is proposed to be used in combination with a three-step oxidation mechanism. The resulting five-step global mechanism can be used with CFD codes to predict CO, CO2, and NO emission directly. Results of the global mechanism optimization method are shown for a pressure of 1 atmosphere and for pressures of interest for gas turbine engines. CFD results showing predicted CO and NO emissions using the five-step global mechanism developed for elevated pressures are presented and compared to measured data.
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Zhan, Dekui, Xinhai Zhao, Shaoxiong Xia, Peng Chen, and Huandong Chen. "Numerical Simulation and Validation for Early Core Degradation Phase under Severe Accidents." Science and Technology of Nuclear Installations 2020 (August 3, 2020): 1–12. http://dx.doi.org/10.1155/2020/6798738.
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Early core degradation determines the amount of hydrogen generated by cladding oxidation as well as the temperature, the mass, and the composition of corium that further relocates into the lower head of reactor pressure vessel (RPV), which is essential for the effectiveness analysis of in-vessel retention (IVR) and hydrogen recombiners. In this paper, the mechanisms of controlling phenomena in the early phase of core degradation are analysed at first. Then, numerical models adopted to calculate (1) core heating up, (2) cladding oxidation, (3) dissolution between molten zirconium and fuel pellets, and (4) formation of a molten pool in the core active section are presented. Compared with integral codes for severe accident analysis (such as MAAP and MELCOR), the models in this paper are established at the fuel pin level and the calculation is performed in 3D, which can capture the detail local phenomena during the core degradation and eliminate the average effect due to equivalent rings used in integral codes. In addition, most of the control equations in this paper are calculated by implicit schemes, which can improve the accuracy and stability of the calculation. In the simulation, the calculation oxidation is calculated by using the oxygen diffusion model, while the dissolution is calculated with Kim, Hayward, Hofmann, and IBRAE models to perform uncertainty analysis. For the validation, the cladding oxidation model is verified by Olander theoretical cases in the conditions of both steam-rich and steam-starved. The dissolution models are validated by the RIAR experiment. The code is overall verified by Phebus FPT0 on the integral phase of core early degradation. According to the simulation results, it can be inferred that the dissolution reaction between the molten zirconium and fuel pellets is the main reason for the melting of UO2 at low temperature. In the case of starved steam, part of the fuel pellets can melt down even at 2248 K and relocate to the bottom of the core, which is much lower than the melting point of UO2 (3113 K).
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Faure, Quentin, and Tatjana Jevremovic. "Molecular dynamics and reaction kinetics analyses of gamma radiation impact on concrete hydration." Nuclear Technology and Radiation Protection 35, no. 1 (2020): 1–15. http://dx.doi.org/10.2298/ntrp2001001f.
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MOPAC and LAMMPS molecular dynamics codes and reaction kinetics code based on multi-ionic continuum-based model are used to analyze the impact of gamma radiation on concrete hydration. The experimental studies showed that while cured with the low gamma dose concrete shows a statistically significant increase in its strength compared to conventionally cured concrete. The potential reason is the interactions of gamma rays with water causing concrete faster hydration. The question then to ask is would the higher gamma dose enhance the concrete curing further producing its higher strength. This paper provides in-depth numerical analyses of the high-dose gamma radiation effect on concrete based on molecular dynamics and reaction kinetics models. Under these conditions, it is assumed that gamma radiation interacting with water within the concrete induces water radiolysis. These numerical simulations show that the reactivity is generally increased in the presence of electrophiles. However, the early hydration models of tricalcium silicate (alite) and dicalcium silicate (belite) with H+, OH-, and H3O+ show that the hydration process is slowed down leading to a lower concrete strength. Additionally, the reaction kinetics model used to estimate the effect of [OH-] on tricalcium silicate hydration shows that an increase or decrease of [OH-] during tricalcium silicate hydration can respectively slow down or enhance its rate of hydration. The dose necessary to produce the water radiolysis resulting in varying [OH-] during tricalcium silicate hydration is required to be extremely high and therefore, will damage the concrete structure itself. This leads to the conclusion that increasing the gamma dose to concrete above that used in the experimental studies in order to induce water radiolysis will not improve concrete strength, therefore water radiolysis is not the required condition for improving concrete strength when cured under gamma radiation.
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Zizin, M. N., V. F. Boyarinov, V. A. Nevinitsa, P. A. Fomichenko, Yu N. Volkov, and A. E. Kruglikov. "Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility." Kerntechnik 85, no. 1 (December 1, 2020): 4–8. http://dx.doi.org/10.1515/kern-2020-850103.
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Abstract Coupled neutronic and thermal hydraulic calculation codes are verified for calculating the design of modern and prospective types of nuclear reactors. This verification is done by comparing experimental and calculated results for stationary and transient conditions. This paper presents ShIPR (Shell of Intelligent Package for Reactor) Integrated Development Environment with automatic generation of head programs based on the chain of computational modules. The aim of this study is to find the reason of a discrepancy in the modelling of sub-critical states that was found in previous work. The comparison of ShIPR stationary module with Monte-Carlo code (MCU) and experimental results on ASTRA HTGR critical facility was presented in the paper. To compare the detector readings and MCU calculation with the ShIPR module, the interpolation cross-section procedure was performed. This procedure allows simulating 235U fission reaction rates (detector readings) in the complicate, annular core, using the micro-cross sections prepared by the cell-lattice code. We found that the calculation accuracy of stationary ShIPR module is on an acceptable level but the macro constants for control rods need to be prepared independently, given with surroundings.
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Ferrari, A., and P. R. Sala. "Nuclear Reactions in Monte Carlo Codes." Radiation Protection Dosimetry 99, no. 1 (June 1, 2002): 29–38. http://dx.doi.org/10.1093/oxfordjournals.rpd.a006788.
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Rizk, N. K., and H. C. Mongia. "Three-Dimensional Gas Turbine Combustor Emissions Modeling." Journal of Engineering for Gas Turbines and Power 115, no. 3 (July 1, 1993): 603–11. http://dx.doi.org/10.1115/1.2906749.
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An emission model that combines the analytical capabilities of three-dimensional combustor performance codes with mathematical expressions based on detailed chemical kinetic scheme is formulated. The expressions provide the trends of formation and/or the consumption of Nox, CO, and UHC in various regions of the combustor utilizing the details of the flow and combustion characteristics given by the three-dimensional analysis. By this means, the optimization of the combustor design to minimize pollutant formation and maintain satisfactory stability and performance could be achieved. The developed model was used to calculate the emissions produced by several engine combustors that varied significantly in design and concept, and operated on both conventional and high-density fuels. The calculated emissions agreed well with the measurements. The model also provided insight into the regions in the combustor where excessive emissions were formed, and helped to understand the influence of the combustor details and air admissions arrangement on reaction rates and pollutant concentrations.
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Kleynhans, G. F., and D. W. Childs. "The Acoustic Influence of Cell Depth on the Rotordynamic Characteristics of Smooth-Rotor/Honeycomb-Stator Annular Gas Seals." Journal of Engineering for Gas Turbines and Power 119, no. 4 (October 1, 1997): 949–56. http://dx.doi.org/10.1115/1.2817079.
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A two-control-volume model is employed for honeycomb-stator/smooth-rotor seals, with a conventional control-volume used for the throughflow and a “capacitance-accumulator” model for the honeycomb cells. The control volume for the honeycomb cells is shown to cause a dramatic reduction in the effective acoustic velocity of the main flow, dropping the lowest acoustic frequency into the frequency range of interest for rotordynamics. In these circumstances, the impedance functions for the seals cannot be modeled with conventional (frequency-independent) stiffness, damping, and mass coefficients. More general transform functions are required to account for the reaction forces, and the transfer functions calculated here are a lead-lag term for the direct force function and a lag term for the cross-coupled function. Experimental measurements verify the magnitude and phase trends of the proposed transfer functions. These first-order functions are simple, compared to transfer functions for magnetic bearings or foundations. For synchronous response due to imbalance, they can be approximated by running-speed-dependent stiffness and damping coefficients in conventional rotordynamics codes. Correct predictions for stability and transient response will require more general algorithms, presumably using a state-space format.
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Кураченко, Ю., Yu Kurachenko, Е. Онищук, E. Onischuk, Е. Матусевич, E. Matusevich, В. Коробейников, and V. Korobeynikov. "High-Intensity Bremsstrahlung of Electron Accelerator in Photoneutron and Radioisotopes Production for Medicine." Medical Radiology and radiation safety 64, no. 5 (October 21, 2019): 42–47. http://dx.doi.org/10.12737/1024-6177-2019-64-5-42-47.
Full textAbstract:
Purpose: To study the binary possibility of using the available linear electron accelerators for the neutron therapy and radioisotopes production. For both applications, calculations were performed and the results were normalized to the characteristics of the Mevex accelerator (average electron current 4 mA at a monoenergetic electron beam 35 MeV). It turns out that the production of both photoneutrons and radioisotopes is effective when using bremsstrahlung radiation generated in the giant dipole resonance of a heavy metal target.
Material and methods: The unifying problem for both applications is the task of target cooling: at beam power ~ 140 kW, half of it or more is deposited directly in the target. Therefore the liquid heavy metal was selected as a target, in order to conjoin high thermohydraulics quality with maximal productivity both bremsstrahlung radiation and photoneutrons. The targets were optimized using precise codes for radiation transport and thermohydraulics problems. The optimization was also carried out for the installations as a whole: 1) for the composition of the material and configuration of the photoneutron extraction unit for neutron capture therapy (NCT) and 2) for the scheme of bremsstrahlung generation for radioisotopes production.
Results: The photoneutron block provides an acceptable beam quality for NCT with a high neutron flux density at the output ~2·1010 cm–2s–1, which is an order of magnitude higher than the values at the output of the reactor beams that worked in the past and are currently being designed for neutron capture therapy. As for radioisotopes production, using optimal reaction channel (γ, n) 43 radioisotopes in 5 groups were received. For example, by the Mo100(γ,n)99Mo reaction the precursor 99Mo of main diagnostic nuclide 99mTc with specific activity ~6 Ci/g and total activity of the target 1.8 kCi could be produced after 1 day irradiation exposure.
Conclusion: The proposed schemes of neutron and bremsstrahlung generation and extraction have a number of obvious advantages over traditional techniques: a) the applying of the electron accelerators for neutron production is much safer and cheaper than to use conventional reactor beams; b) accelerator with the target, the beam output unit with the necessary equipment and tooling can be placed on the territory of the clinic without any problems; c) the proposed target for NCT is liquid gallium, which also serves as a coolant; it is an “environmentally friendly” material, its activation is rather low and rapidly (in ~4 days) falls to the background level.
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Кураченко, Ю., Yu Kurachenko, Е. Онищук, E. Onischuk, Е. Матусевич, E. Matusevich, В. Коробейников, and V. Korobeynikov. "High-Intensity Bremsstrahlung of Electron Accelerator in Photoneutron and Radioisotopes Production for Medicine." Medical Radiology and radiation safety 64, no. 5 (October 21, 2019): 48–53. http://dx.doi.org/10.12737/1024-6177-2019-64-5-48-53.
Full textAbstract:
Purpose: To study the binary possibility of using the available linear electron accelerators for the neutron therapy and radioisotopes production. For both applications, calculations were performed and the results were normalized to the characteristics of the Mevex accelerator (average electron current 4 mA at a monoenergetic electron beam 35 MeV). It turns out that the production of both photoneutrons and radioisotopes is effective when using bremsstrahlung radiation generated in the giant dipole resonance of a heavy metal target.
Material and methods: The unifying problem for both applications is the task of target cooling: at beam power ~ 140 kW, half of it or more is deposited directly in the target. Therefore the liquid heavy metal was selected as a target, in order to conjoin high thermohydraulics quality with maximal productivity both bremsstrahlung radiation and photoneutrons. The targets were optimized using precise codes for radiation transport and thermohydraulics problems. The optimization was also carried out for the installations as a whole: 1) for the composition of the material and configuration of the photoneutron extraction unit for neutron capture therapy (NCT) and 2) for the scheme of bremsstrahlung generation for radioisotopes production.
Results: The photoneutron block provides an acceptable beam quality for NCT with a high neutron flux density at the output ~2·1010 cm–2s–1, which is an order of magnitude higher than the values at the output of the reactor beams that worked in the past and are currently being designed for neutron capture therapy. As for radioisotopes production, using optimal reaction channel (γ, n) 43 radioisotopes in 5 groups were received. For example, by the Mo100(γ,n)99Mo reaction the precursor 99Mo of main diagnostic nuclide 99mTc with specific activity ~6 Ci/g and total activity of the target 1.8 kCi could be produced after 1 day irradiation exposure.
Conclusion: The proposed schemes of neutron and bremsstrahlung generation and extraction have a number of obvious advantages over traditional techniques: a) the applying of the electron accelerators for neutron production is much safer and cheaper than to use conventional reactor beams; b) accelerator with the target, the beam output unit with the necessary equipment and tooling can be placed on the territory of the clinic without any problems; c) the proposed target for NCT is liquid gallium, which also serves as a coolant; it is an “environmentally friendly” material, its activation is rather low and rapidly (in ~4 days) falls to the background level.
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Pasaribu, Rosenti, Kusminarto Kusminarto, and Yohannes Sardjono. "Clinical trial design of Boron Neutron Capture Therapy on breast cancer using D-D coaxial compact neutron generator as neutron source and Monte Carlo N-Particle simulation method." Indonesian Journal of Physics and Nuclear Applications 1, no. 1 (February 28, 2016): 34. http://dx.doi.org/10.24246/ijpna.v1i1.34-43.
Full textAbstract:
<span>A clinical trial simulation of Boron Neutron Capture Therapy (BNCT) for breast cancer was conducted at National Nuclear Energy Agency Yogyakarta, Indonesia. This was motivated by high rate of breast cancer in the world, especially in Indonesia. BNCT is a type of therapy by nuclear reaction </span><sup>10</sup><span>B(n,α)</span><sup>7</sup><span>Li that produces kinetic energy totaling 2.79 MeV. High Linear Energy Transfer (LET) radiation of α-particle and recoil </span><sup>7</sup><span>Li would locally deposit their energy in a range of 5-9 μm, which corresponds to the human cell diameter. Fast neutron coming out of Compact Neutron Generator (CNG) was moderated using Fe and MgF</span><sub>2</sub><span> material. A collimator, along with breast cancer and the corresponding organ at risk were designed compatible to Monte Carlo N-Particle X (MCNPX). The radiation were simulated by the MCNPX software and the physical quantities were counted by tally MCNPX codes. The highest neutron thermal flux was found at a depth of 1.4 cm on fat tissue. En face and upward intersection radiation techniques were adopted for the breast cancer radiation. The average dose rate of radiation used on breast cancer was 1.72×10</span><sup>-5 </sup><span>Gy/s for the en face method and 8.98×10</span><sup>-6 </sup><span>Gy/s for the upward intersection method. Dose 50±3 Gy was given into cancer cell, (4.18±0.06) ×10</span><sup>-2</sup><span> Gy into heart and (8.16±0.06) ×10</span><sup>-2</sup><span>Gy into lung for 806.34 hours irradiation.</span>