Journal articles on the topic 'Fission Reactor Physic'
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1
Ripani, M. "Energy from nuclear fission." EPJ Web of Conferences 268 (2022): 00010. http://dx.doi.org/10.1051/epjconf/202226800010.
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The physics of nuclear fission will be briefly illustrated, from the basic mechanism behind this phenomenon to the relevant physical quantities like nuclear cross sections, neutron flux and reaction products, together with the accompanying phenomenon of neutron capture and its role in determining how the fuel transforms in a nuclear reactor. The basic concepts underlying the operation of different types of nuclear reactors will be illustrated, along with the concept of fuel cycle. The aspects of radioactive waste, fuel resources and safety will also be briefly illustrated.
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Ripani, M. "Energy from nuclear fission." EPJ Web of Conferences 246 (2020): 00010. http://dx.doi.org/10.1051/epjconf/202024600010.
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The physics of nuclear fission will be briefly illustrated, from the basic mechanism behind this phenomenon to the relevant physical quantities like nuclear cross sections, neutron flux and reaction products, together with the accompanying phenomenon of neutron capture and its role in determining how the fuel transforms in a nuclear reactor. The basic concepts underlying the operation of different types of nuclear reactors will be illustrated, along with the concept of fuel cycle. After touching on the aspect of safety, the current situation of nuclear power in the world, with its costs, its role in reducing carbon emissions, the available resources and finally the issues of waste management and accidents will be briefly illustrated.
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Verbeke, Jérôme M., Odile Petit, Abdelhazize Chebboubi, and Olivier Litaize. "Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® ." EPJ Web of Conferences 170 (2018): 01019. http://dx.doi.org/10.1051/epjconf/201817001019.
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Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using either FREYA or FIFRELIN, are compared to experimental results. For 240Pu(sf), the measured correlations were used to tune the model parameters.
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Burgio, N., M. Carta, V. Fabrizio, L. Falconi, A. Gandini, R. Gatto, V. Peluso, E. Santoro, and M. B. Sciarretta. "SUBCRITICALITY MONITORING IN FUSION-FISSION HYBRID REACTORS." Problems of Atomic Science and Technology, Ser. Thermonuclear Fusion 44, no. 2 (2021): 27–41. http://dx.doi.org/10.21517/0202-3822-2021-44-2-27-41.
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Mills, Robert W., David J. Mountford, Jonathon P. Coleman, Carl Metelko, Matthew Murdoch, and Yan-Jie Schnellbach. "Modelling of the anti-neutrino production and spectra from a Magnox reactor." EPJ Web of Conferences 170 (2018): 07008. http://dx.doi.org/10.1051/epjconf/201817007008.
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The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.
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Anastasiou, Maria. "6Li(n,t)α reaction event-identification for the 235U(n,f)/6Li(n,t) cross section ratio measurement in the NIFFTE fissionTPC." HNPS Advances in Nuclear Physics 28 (October 17, 2022): 30–35. http://dx.doi.org/10.12681/hnps.3567.
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While nuclear data play an important role in nuclear physics applications, it has become important to have a better understanding of the data and try to minimize the uncertainties. In particular, there is a need for precision neutron-induced fission cross section measurements on fissile nuclei. Neutron-induced fission cross sections are typically measured as ratios, with a well-known standard in the denominator. While the 235U(n,f) reaction is a well measured standard, some light particle reactions are also well-known and their use as reference can provide information to remove shared systematic uncertainties that are present in an actinide-only ratio. A recent measurement of the 235U(n,f) reaction using as a reference the standard 6Li(n,t) reaction, was conducted at the Los Alamos Neutron Science Center using the NIFFTE collaboration’s fission time projection chamber (fissionTPC). The fissionTPC is a 2×2π charged particle tracker designed for measuring neutron-induced fission. Detailed 3D track reconstruction of the reaction products enables evaluation of systematic effects and corresponding uncertainties which are less directly accessible by other measurement techniques. This work focuses on the analysis for the event identification of the 6Li(n,t)α reaction in the fissionTPC.
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Tonchev, Anton P., Jack A. Silano, Chris Hagmann, Roger Henderson, Mark A. Stoyer, Matthew Gooden, Todd Bredeweg, et al. "Toward short-lived and energy-dependent fission product yields from neutron-induced fission." EPJ Web of Conferences 239 (2020): 03001. http://dx.doi.org/10.1051/epjconf/202023903001.
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Fission product yields (FPYs) are an important source of information that are used for basic and applied physics. They are essential observables to address questions relevant to nucleosynthesis in the cosmos that created the elements from iron to uranium, for example, in energy generating processes from fission recycling in binary neutron star mergers; resolving the reactor neutrino anomaly; decay heat release in nuclear reactors; and many national security applications. While new applications will require accurate energy-dependent FPY data over a broad set of incident neutron energies, the current evaluated FPY data files contain only three energy points: thermal, fast, and 14-MeV incident energies. Recent measurements using mono-energetic and pulsed neutron beams at the Triangle Universities Nuclear Laboratory (TUNL) tandem accelerator and employing a dual fission ionization chambers setup have produced self-consistent, high-precision data critical for testing fission models for the neutron-induced fission of the major actinide nuclei. This paper will present new campaign just beginning utilizing a RApid Belt-driven Irradiated Target Transfer System (RABITTS) to measure shorter-lived fission products and the time dependence of fission yields, expanding the measurements from cumulative towards independent fission yields.
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Davies, Sebastian, Dzianis Litskevich, Bruno Merk, Andrew Levers, Paul Bryce, and Anna Detkina. "DYN3D and CTF Coupling within a Multiscale and Multiphysics Software Development (Part II)." Energies 15, no. 13 (July 1, 2022): 4843. http://dx.doi.org/10.3390/en15134843.
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Traditionally, the complex coupled physical phenomena in nuclear reactors has resulted in them being treated separately or, at most, simplistically coupled in between within nuclear codes. Currently, coupling software environments are allowing different types of coupling, modularizing the nuclear codes or multi-physics. Several multiscale and multi-physics software developments for LWR are incorporating these to deliver improved or full coupled reactor physics at the fuel pin level. An alternative multiscale and multi-physics nuclear software development between NURESIM and CASL is being created for the UK. The coupling between DYN3D nodal code and CTF subchannel code can be used to deliver improved coupled reactor physics at the fuel pin level. In the current journal article, the second part of the DYN3D and CTF coupling was carried out to analyse a parallel two-way coupling between these codes and, hence, the outer iterations necessary for convergence to deliver verified improved coupled reactor physics at the fuel pin level. This final verification shows that the DYN3D and CTF coupling delivers improved effective multiplication factors, fission, and feedback distributions due to the presence of crossflow and turbulent mixing.
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Winterberg, F. "Mini Fission-Fusion-Fission Explosions (Mini-Nukes). A Third Way Towards the Controlled Release of Nuclear Energy by Fission and Fusion." Zeitschrift für Naturforschung A 59, no. 6 (June 1, 2004): 325–36. http://dx.doi.org/10.1515/zna-2004-0603.
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Chemically ignited nuclear microexplosions with a fissile core, a DT reflector and U238 (Th232) pusher, offer a promising alternative to magnetic and inertial confinement fusion, not only burning DT, but in addition U238 (or Th232), and not depending on a large expensive laser of electric pulse power supply. The prize to be paid is a gram size amount of fissile material for each microexplosion, but which can be recovered by breeding in U238.In such a “mini-nuke” the chemical high explosive implodes a spherical metallic shell onto a smaller shell, with the smaller shell upon impact becoming the source of intense black body radiation which vaporizes the ablator of a spherical U238 (Th232) pusher, with the pusher accelerated to a velocity of ∼200 km/s, sufficient to ignite the DT gas placed in between the pusher and fissile core, resulting in a fast fusion neutron supported fission reaction in the core and pusher. Estimates indicate that a few kg of high explosives are sufficient to ignite such a “mini-nuke”, with a gain of ∼103, releasing an energy equivalent to a few tons of TNT, still manageable for the microexplosion to be confined in a reactor vessel.A further reduction in the critical mass is possible by replacing the high explosive with fast moving solid projectiles. For light gas gun driven projectiles with a velocity of ∼ 10 km/s, the critical mass is estimated to be 0.25 g, and for magnetically accelerated 25 km/s projectiles it is as small as ∼ 0.05 g.With the much larger implosion velocities, reached by laser- or particle beam bombardment of the outer shell, the critical mass can still be much smaller with the fissile core serving as a fast ignitor.Increasing the implosion velocity decreases the overall radius of the fission-fusion assembly in inverse proportion to this velocity, for the 10 km/s light gas gun driven projectiles from 10 cm to 5 cm, for the 25 km/s magnetically projectiles down to 2 cm, and still more for higher implosion velocities.
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Li, Jia, and Shanliang Zheng. "Feasibility Study to Byproduce Medical Radioisotopes in a Fusion Reactor." Molecules 28, no. 5 (February 22, 2023): 2040. http://dx.doi.org/10.3390/molecules28052040.
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Currently, international nuclear fission reactors producing medical isotopes face the problem of shutdown and maintenance, decommissioning, or dismantling, while the production capacity of domestic research reactors for medical radioisotopes is inadequate, and the supply capacity for medical radioisotopes faces major challenges in the future. Fusion reactors are characterized by high neutron energy, high flux density, and the absence of highly radioactive fission fragments. Additionally, compared to fission reactors, the reactivity of the fusion reactor core is not significantly affected by the target material. By building a preliminary model of the China Fusion Engineering Test Reactor (CFETR), a Monte Carlo simulation was performed for particle transport between different target materials at a fusion power of 2 GW. The yields (specific activity) of six medical radioisotopes (14C, 89Sr, 32P, 64Cu, 67Cu, and 99Mo) with various irradiation positions, different target materials, and different irradiation times were studied, and compared with those of other high-flux engineering test reactors (HFETR) and the China Experimental Fast Reactor (CEFR). The results show that this approach not only provides competitive medical isotope yield, but also contributes to the performance of the fusion reactor itself, e.g., tritium self-sustainability and shielding performance.
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Wagemans, J., L. Borms, A. Kochetkov, A. Krása, C. Van Grieken, and G. Vittiglio. "Nuclear instrumentation in VENUS-F." EPJ Web of Conferences 170 (2018): 04027. http://dx.doi.org/10.1051/epjconf/201817004027.
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VENUS-F is a fast zero power reactor with 30 wt% U fuel and Pb/Bi as a coolant simulator. Depending on the experimental configuration, various neutron spectra (fast, epithermal, and thermal islands) are present. This paper gives a review of the nuclear instrumentation that is applied for reactor control and in a large variety of physics experiments. Activation foils and fission chambers are used to measure spatial neutron flux profiles, spectrum indices, reactivity effects (with positive period and compensation method or the MSM method) and kinetic parameters (with the Rossi-alpha method). Fission chamber calibrations are performed in the standard irradiation fields of the BR1 reactor (prompt fission neutron spectrum and Maxwellian thermal neutron spectrum).
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Chiba, Go, and Shunsuke Nihira. "Uncertainty quantification works relevant to fission yields and decay data." EPJ Nuclear Sciences & Technologies 4 (2018): 43. http://dx.doi.org/10.1051/epjn/2018022.
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In the present paper, firstly, we review our previous works on uncertainty quantification (UQ) of reactor physics parameters. This consists of (1) development of numerical tools based on the depletion perturbation theory (DPT), (2) linearity of reactor physics parameters to nuclear data, (3) UQ of decay heat and its reduction, and (4) correlation between decay heat and β-delayed neutrons emission. Secondly, we show results of extensive calculations about UQ on decay heat with several different numerical conditions by the DPT-based capability of a reactor physics code system CBZ.
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Dranga, Ruxandra, Laura Blomeley, and Rebecca Carrington. "AN MCNP PARAMETRIC STUDY OF GEORGE C. LAURENCE'S SUBCRITICAL PILE EXPERIMENT." AECL Nuclear Review 3, no. 2 (December 1, 2014): 91–99. http://dx.doi.org/10.12943/anr.2014.00037.
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In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon–uranium arrangement (or “pile”) was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada’s first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile’s neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada’s nuclear science and technology program.
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Pesic, Milan, Yury Titarenko, Viacheslav Batyaev, Kirill Pavlov, Alexey Titarenko, Valeriy Zhivun, Mikhail Igumnov, Viacheslav Konev, and Vladimir Legostaev. "Validation of minor actinides fission neutron cross-sections." Nuclear Technology and Radiation Protection 30, no. 1 (2015): 1–10. http://dx.doi.org/10.2298/ntrp1501001p.
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Verification of neutron fission cross-sections of minor actinides from some recently available evaluated nuclear data libraries was carried out by comparison of the reaction rates calculated by the MCNP6.1 computer code to the experimental values. The experimental samples, containing thin layers of 235U, 237Np, 238,239,240,241Pu, 242mAm, 243Cm, 245Cm, and 247Cm, deposited on metal support and foils of 235U (pseudo-alloy 27Al + 235U), 238U, natIn, 64Zn, 27Al, and multi-component sample alloy 27Al + 55Mn + natCu + natLu + 197Au, were irradiated in the channels of the tank containing fluorine salts 0.52NaF + 0.48ZrF4, labelled as the Micromodel Salt Blanket, inserted in the lattice centre of the MAKET heavy water critical assembly at the Institute for Theoretical and Experimental Physics, Moscow. This paper is a continuation of earlier initiated scientific-research activities carried out for validation of the evaluated fission cross-sections of actinides that were supposed to be used for the quality examination of the fuel design of the accelerator driven systems or fast reactors, and consequently, determination of transmutation rates of actinides, and therefore, determination of operation parameters of these reactor facilities. These scientific-research activities were carried out within a frame of scientific projects supported by the International Science and Technology Center and the International Atomic Energy Agency co-ordinated research activities, from 1999 to 2010. Obtained results confirm that further research is needed in evaluations in order to establish better neutron cross-section data for the minor actinides and selected nuclides which could be used in the accelerator driven systems or fast reactors.
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Jagannathan, V., Usha Pal, R. Karthikeyan, Devesh Raj, Argala Srivastava, and Suhail Ahmad Khan. "Reactor physics ideas to design novel reactors with faster fissile growth." Energy Conversion and Management 49, no. 8 (August 2008): 2032–46. http://dx.doi.org/10.1016/j.enconman.2008.02.019.
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Pavlovych, Volodymyr, Volodymyr Khotyayintsev, and Olena Khotyayintseva. "The physical basis of the fission wave reactor." Nuclear Technology and Radiation Protection 23, no. 2 (2008): 3–15. http://dx.doi.org/10.2298/ntrp0802003p.
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The main idea of slow nuclear fission wave reactor is discussed and short review of the existing works is also presented. The aim of this paper is to clarify the physics of processes, which define the stationary wave of nuclear burning, and to develop the approaches determining the wave parameters. It is shown that the diffusion equation for fluence can be used to describe the stationary and non-stationary processes in the nuclear fission wave. Two conditions of stationary wave existence are first formulated in the paper. The rule of determination of wave velocity as the eigenvalue of boundary problem is also formulated.
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Gill, Daniel F. "INLINE THERMAL AND XENON FEEDBACK ITERATIONS IN MONTE CARLO REACTOR CALCULATIONS." EPJ Web of Conferences 247 (2021): 04019. http://dx.doi.org/10.1051/epjconf/202124704019.
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In this work, we describe a method for converging nonlinear feedback during the convergence of the neutron fission source in a Monte Carlo reactor simulation. This approach involves updating feedback physics during discard batches in the Monte Carlo simulation rather than fully (or partially) converging the neutronics prior to the nonlinear update. This approach is demonstrated for a single PWR pin with thermal feedback and with both thermal and xenon feedback. Converging these feedbacks inline with the fission source is shown to have the benefit of reducing numerical instability by effectively underrelaxing the tallied quantities in the Monte Carlo simulation and improving computational performance by converging feedback within (or near to) the number of discard batches required to converge the fission source even without any feedback.
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Bhattacharya, P., A. Sen, T. K. Ghosh, N. Majumdar, and S. Mukhopadhyay. "Development of Micro-Pattern Gaseous Detectors for Nuclear Reaction Studies." Journal of Physics: Conference Series 2349, no. 1 (September 1, 2022): 012017. http://dx.doi.org/10.1088/1742-6596/2349/1/012017.
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One of the frontiers of today’s nuclear physics research is the synthesis of Super Heavy Elements (SHE). Fusion-fission dynamics, namely the competition between quasi fission and fusion is one of the key challenges to optimize the SHE. To have an insight into the dynamics, one requires the study of fission fragment mass and angular distribution near barrier energies for heavy-ion induced fission reactions. Recent successful installation of linear accelerators in India offers a unique opportunity to study the dynamics of nuclear reactions and formation process of SHE. For the effective utilization of these current, as well as upcoming facilities, development of novel detectors to study reaction dynamics, formation process of SHE with heavier projectiles and higher beam energies is needed. Gaseous detectors have undergone a rapid improvement in terms of spatial, temporal and energy resolution, rate capability, radiation hardness, ion feedback etc., ushering in a new genre of micro-structured devices based on semi-conductor technology, commonly known as Micro-Pattern Gaseous Detectors (MPGDs). Although many of the MPGD structures were primarily developed for high-rate tracking of charged particles in high energy physics experiments, stability of operation, simplicity of construction and relatively low cost make these detectors suitable for other applications, such as low-energy nuclear physics experiments. The present activities encompass a detailed evaluation of the operational conditions of Micromesh-Multi Wire and THGEM-Multi Wire hybrid detector operated in low-pressure isobutane gas with a view to optimizing their use in the detection of charged particles and fission fragments.
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Shaginyan, Ruben, Valery Kolesov, and Evgeny Ivanov. "A MESOSCOPIC OXIDE FUEL CLUSTERING AND ITS GLOBAL PERFORMANCE." EPJ Web of Conferences 247 (2021): 15012. http://dx.doi.org/10.1051/epjconf/202124715012.
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Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppler broadening, moderation ratio, and, sometimes, neutron gas temperature etc.) and on variations of thermal-physics parameters (temperature distributions, fuel elongation and moderator density). Usually, in a rough reactor analysis one ignores the very details of temperature distributions largely staying in a frame of so-called adiabatic assumptions (when temperature and density distribution are changing in sync keeping given spatial shapes). In majority of practical applications the radially distributed temperature fields are represented as monotonically smeared ones as if fissile and other materials are homogeneously mixed. Moreover, no one measurement technique allows counting precise correlation between reactivity feedback and in-pellet temperature and materials space-time distributions. However, if fuel is made of Mixed Oxide Plutonium-Uranium compound the behavior of Light Water Reactor would be impacted by an appearance of Pu-rich agglomerates that could be large enough to change physical processes. In such case the fuel reacts on power and temperature variations no more as a homogeneous but a heterogeneous media (on a mesoscopic scale, of course). It leads to changes in a fission product distributions, a fission gas release and, even, to an appearance of multiple components in a Fuel Temperature Coefficient and in a Power Reactivity feedback. These components would depend non-linearly on power, power rate and on some details of a heat transfer. This paper is the only first step of a broad research program where we are estimating the relevant phenomena just by an order of magnitude.
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Jiang, Yifeng, Benoit Geslot, Vincent Lamirand, and Pierre Leconte. "Review of kinetic modulation experiments in low power nuclear reactors." EPJ Nuclear Sciences & Technologies 6 (2020): 55. http://dx.doi.org/10.1051/epjn/2020017.
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The safety improvement of nuclear reactors requires continuous efforts in understanding the fundamental physical quantities related to the fission process. In neutronic models, the reactor dynamics is covered by the kinetic parameters to characterize the temporal behavior of the neutron population subject to perturbations. The reactor transfer function is a frequency domain analogy of this temporal description. It can be measured experimentally through transfer function analysis via noise analysis or kinetic modulation, for the study of reactor stability and kinetic parameters. This paper summarizes the experimental measurements of reactor transfer function through kinetic modulation. Extensive work have been conducted experimentally, starting from the beginning of reactor physics research. An overview is given regarding various experimental designs and conducted analyses. The concepts of the modulation system are also discussed. The current work is limited to online contents and internal archives of CEA Cadarache due to difficulties in accessing references traced back to 1950s.
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Chapman, Pete, Jonathan Mueller, Jason Newby, and John Mattingly. "Exploiting fission chain reaction dynamics to image fissile materials." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 935 (August 2019): 198–206. http://dx.doi.org/10.1016/j.nima.2019.05.001.
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Jaroszewicz, Janusz, Zuzanna Marcinkowska, and Krzysztof Pytel. "Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis." Nukleonika 59, no. 2 (July 8, 2014): 43–52. http://dx.doi.org/10.2478/nuka-2014-0009.
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Abstract The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.
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Artun, Ozan. "Estimation of fission yield and cross-section for thorium-based molten salt reactor and accelerator driven system." Modern Physics Letters A 36, no. 21 (July 2, 2021): 2150147. http://dx.doi.org/10.1142/s0217732321501479.
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Thorium is the primary candidate fuel in Thorium-based Molten Salt Reactor (TMSR) and Accelerator Driven System (ADS). However, to utilize Thorium in the reactor and accelerator systems, nuclear data in the literature are not sufficient for neutron-induced fission reaction process. Therefore, we estimated neutron fission cross-section [Formula: see text] (mb), independent [Formula: see text] and primary fission fragment yield [Formula: see text], mass-dependent average neutron multiplicity [Formula: see text], energy-dependent average neutron multiplicity [Formula: see text] and neutron emission multiplicity distribution [Formula: see text] of [Formula: see text] for different neutron incident energy values, 1 MeV, 1.5 MeV, 3 MeV, 5.5 MeV, 7.7 MeV, 10 MeV, 14.8 MeV and 20 MeV, where the energy region is important for the reactor and accelerator systems used in the industry. The obtained results are compared with experimental data obtained from the literature and are discussed for each energy value.
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Orsitto, F. P., M. Angelone, and M. Tardocchi. "DIAGNOSTICS AND CONTROL OF FUSION-FISSION HYBRID TOKAMAKBASED REACTORS: THE TECHNOLOGY FOR MEASUREMENT SYSTEMS." Problems of Atomic Science and Technology, Ser. Thermonuclear Fusion 44, no. 2 (2021): 78–85. http://dx.doi.org/10.21517/0202-3822-2021-44-2-78-85.
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Li, Jinfeng. "Monte Carlo Investigation of the UK’s First EPR Nuclear Reactor Startup Core Using Serpent." Energies 13, no. 19 (October 4, 2020): 5168. http://dx.doi.org/10.3390/en13195168.
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Computationally modelling a nuclear reactor startup core for a benchmark against the existing models is highly desirable for an independent assessment informing nuclear engineers and energy policymakers. For the first time, this work presents a startup core model of the UK’s first Evolutionary Pressurised Water Reactor (EPR) based on Monte Carlo simulations of particle collisions using Serpent 2, a state-of-the-art continuous-energy Monte Carlo reactor physics burnup code. Coupling between neutronics and thermal-hydraulic conditions with the fuel depletion is incorporated into the multi-dimensional branches, obtaining the thermal flux and fission reaction rate (power) distributions radially and axially from the three dimensional (3D) single assembly level to a 3D full core. Shannon entropy is quantified to characterise the convergence behaviour of the fission source distribution, with 3 billion neutron histories tracked by parallel computing. Source biasing is applied for the variance reduction. Benchmarking the proposed Monte Carlo 3D full-core model against the traditional deterministic transport computation suite used by the UK Office for Nuclear Regulation (ONR), a reasonably good agreement within statistics is demonstrated for the safety-related reactivity coefficients, which creates trust in the EPR safety report and informs the decision-making by energy regulatory bodies and global partners.
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Andriamirado, M., A. B. Balantekin, H. R. Band, C. D. Bass, D. E. Bergeron, N. S. Bowden, C. D. Bryan, et al. "PROSPECT-II physics opportunities." Journal of Physics G: Nuclear and Particle Physics 49, no. 7 (June 6, 2022): 070501. http://dx.doi.org/10.1088/1361-6471/ac48a4.
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Abstract The precision reactor oscillation and spectrum experiment, PROSPECT, has made world-leading measurements of reactor antineutrinos at short baselines. In its first phase, conducted at the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory, PROSPECT produced some of the strongest limits on eV-scale sterile neutrinos, made a precision measurement of the reactor antineutrino spectrum from 235U, and demonstrated the observation of reactor antineutrinos in an aboveground detector with good energy resolution and well-controlled backgrounds. The PROSPECT collaboration is now preparing an upgraded detector, PROSPECT-II, to probe yet unexplored parameter space for sterile neutrinos and contribute to a full resolution of the reactor antineutrino anomaly, a longstanding puzzle in neutrino physics. By pressing forward on the world’s most precise measurement of the 235U antineutrino spectrum and measuring the absolute flux of antineutrinos from 235U, PROSPECT-II will sharpen a tool with potential value for basic neutrino science, nuclear data validation, and nuclear security applications. Following a two-year deployment at HFIR, an additional PROSPECT-II deployment at a low enriched uranium reactor could make complementary measurements of the neutrino yield from other fission isotopes. PROSPECT-II provides a unique opportunity to continue the study of reactor antineutrinos at short baselines, taking advantage of demonstrated elements of the original PROSPECT design and close access to a highly enriched uranium reactor core.
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Vityuk, V. A. "DESIGN-BASIS JUSTIFICATION FOR IMPLEMENTING TARGETED ENERGY RELEASE IN TEST OBJECTS OF THE IMPULSE GRAPHITE REACTOR." Eurasian Physical Technical Journal 17, no. 2 (December 24, 2020): 87–95. http://dx.doi.org/10.31489/2020no2/87-95.
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The article presents the approaches and technical solutions applied to assure preset axial and radial distribution of energy release in simulative fuel rods and fuel assemblies in the tests at the impulse graphite reactor. It considers the procedure for the design-basis justification of solutions that provide a given volume distribution of energy release in a test unit. The considerations are based on the example of heterogeneous fuel assemblies with the altitude separation of enriched fuel into two zones by a depleted uranium layer used to reproduce fissile material. The implementation of the procedure and making appropriate technical solutions made it possible to provide a targeted profile of the axial and radial distribution of energy release in a simulative fuel assembly at the design stage of an irradiation device. By the result of study, it is demonstrated that uniform radial energy release and targeted average energy release in upper and lower fission zones of experimental fuel assembly could be obtained at the level of 90.6 W/g (UO2) and 74 W/g (UO2), respectively. The measures include profiling of fuel pellets enrichment in fuel rods, using the pellets with an absorber at the zone ends, and a certain altitudinal positioning of the irradiation device in the reactor.
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Vogt, R., J. Randrup, N. Vassh, T. Sprouse, and R. Surman. "Employing FREYA for fission product yield evaluations." EPJ Web of Conferences 242 (2020): 03002. http://dx.doi.org/10.1051/epjconf/202024203002.
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The fast event-by-event fission code FREYA (Fission Reaction Event Yield Algorithm) generates large samples of complete fission events while employing only a few physics-based parameters. Not only is FREYA fast, it is also flexible, able to employ a variety of input formats to test the implications of various fission yield evaluations on neutron and photon observables. We describe how FREYA was applied to the neutron-rich nuclei needed for r-process nucleosynthesis calculations as an example of this flexibility. Finally, we discuss how we plan to make use of this flexibility to extend FREYA to calculations of cumulative fission product yields to aid evaluations of these yields in the future.
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Suryono, T. J., Sudarno, S. Santoso, and R. Maerani. "Modelling of FPGA-based Reactor Protection Systems of an Experimental Power Reactor." Journal of Physics: Conference Series 2048, no. 1 (October 1, 2021): 012038. http://dx.doi.org/10.1088/1742-6596/2048/1/012038.
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Abstract The reactor protection system of nuclear power plants including an experimental power reactor which will be built by Indonesia is a safety system that actuates the control rods to be inserted in the reactor core to absorb the neutron to stop the fission reaction and then shut down the reactor (reactor trip). The reactor protection system (RPS) is actuated when the level of signals from the sensors of important components in the reactors deviates from the setpoint determined in the bi-stable processor of the RPS. RPS for the experimental power reactor has 3 redundant channels for reliability and to minimize fake signals from the sensors due to electrical noise. It can be done by selecting the channels in local coincidence logic in the RPS by voting 2 of 3 channels which are eligible to generate actuation signals to trip the reactor. Recently, the RPSs are based on the programmable logic controller (PLC). However, now the trend changes to FPGA-based RPS because of its simplicity and reliability. This paper investigates the model of the FPGA-based RPS for an experimental power reactor and the functionality of each component of the model. The results show that the model can represent the functionality of RPS for the experimental power reactor.
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Li Maosheng, 李茂生, 师学明 Shi Xueming, 刘荣 Liu Rong, 鹿心鑫 Lu Xinxin, 朱通华 Zhu Tonghua, 王新华 Wang Xinhua, 余泳 Yu Yong, et al. "Progress in physics design of fusion-fission hybrid energy reactor." High Power Laser and Particle Beams 26, no. 10 (2014): 100203. http://dx.doi.org/10.3788/hplpb20142610.100203.
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Li, Xian. "The entropy of the Au + Au system in low energy nuclear reaction." Modern Physics Letters A 33, no. 33 (October 29, 2018): 1850191. http://dx.doi.org/10.1142/s0217732318501912.
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Considering the boundary effect, we adopt the Tsallis entropy and compute the time evolution of the entropy in low energy reaction system for the first time, dealing with the initial, compression, expansion and fission stages with a consistent method. We find that it rises up in the compression period and reduces slightly after separation, because of the energy exchange between the collective motion and the internal excitation. The research shows that the entropy curves in binary, sequential or direct ternary fissions are distinct, which declares that the entropy may be a novel criterion to study the breakup mechanism of heavy nuclear reactions in low energy.
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Pritychenko, B., O. Schwerer, J. Totans, and O. Gritzay. "Present Status of Neutron-, Photo-induced and Spontaneous Fission Yields Experimental Data." EPJ Web of Conferences 242 (2020): 02001. http://dx.doi.org/10.1051/epjconf/202024202001.
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Nuclear reaction data collection, evaluation and dissemination have been pioneered at the Brookhaven National Laboratory since the early 50s. These activities gained popularity worldwide, and around 1970 the experimental nuclear reaction data interchange or exchange format (EXFOR) was established. The original EXFOR compilation scope consisted only of neutron reactions and spontaneous fission data, while many other nuclear data sets were ignored. Due to the high cost of new experiments, it is very important to find and recover the previously disregarded data using scientific publications, data evaluations and nuclear databases comparisons. Fission yields play a very important role in applied and fundamental physics, and such data are essential in many applications. The comparative analysis of Nuclear Science References (NSR) and Experimental Nuclear Reaction (EXFOR) databases shows a large number of unaccounted experiments and provides a guide for the recovery of fission cross sections, yields and covariance data sets. The dedicated fission yields data compilation effort is currently underway in the Nuclear Reaction Data Centers (NRDC) network, and includes identification, compilation, storage and Web dissemination of the recovered data sets.
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Sauer, G. "An attempt to describe the swelling of U3Si 2 fuel by a simple formula." Kerntechnik 66, no. 1-2 (January 1, 2001): 62–64. http://dx.doi.org/10.1515/kern-2001-0016.
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Abstract Fuel swelling is a key parameter to be studied in the process of qualifying a fuel for use in reactors. Generally, test irradiations are performed to quantify the swelling. However, all reactor conditions can often not be simulated in tests. This situation occurs, for example, when a fuel shall be used in a new reactor where the fission rate is higher than in existing reactors available for the qualification tests. The swelling under the conditions of the new reactor must be predicted based on the data found under these test conditions. For this prediction a formula is required. Such a formula is derived for one of the most popular fuels for high flux reactors: U3Si2.
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Choi, Hangbok, Robert W. Schleicher, and Puja Gupta. "A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle." Science and Technology of Nuclear Installations 2013 (2013): 1–10. http://dx.doi.org/10.1155/2013/618707.
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In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2), which is a compact gas-cooled fast reactor (GFR). The EM2augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.
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Hill, I. "IDENTIFICATION OF REACTOR PHYSICS BENCHMARKS FOR NUCLEAR DATA TESTING: TOOLS AND EXAMPLES." EPJ Web of Conferences 247 (2021): 10028. http://dx.doi.org/10.1051/epjconf/202124710028.
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Measurements of reactor physics quantities aimed at identifying the reactivity worth of materials, spectral ratios of cross-sections, and reactivity coefficients have ensured reactor physics codes can accurately predict nuclear reactor systems. These measurements were critical in the absence of sufficiently accurate differential data, and underpinned the need for experiments through the 50s, 60s, 70s and 80s. Data from experimental campaigns were routinely incorporated into nuclear data libraries either through changes to general nuclear data libraries, or more commonly in the local libraries generated by a particular institution or consortium interested in accurately predicting a specific nuclear system (e.g. fast reactors) or parameters (e.g. fission gas release, yields). Over the last three decades, the model has changed. In tandem access to computing power and monte carlo codes rose dramatically. The monte carlo codes were well suited to computing k-eff, and owing to the availability of high quality criticality benchmarks and these benchmarks were increasing used to test the nuclear data. Meanwhile, there was a decline in the production of local libraries as new nuclear systems were not being built, and the existing systems were considered adequately predicted. The cost-to-benefit ratio of validating new libraries relative to their improved prediction capability was less attractive. These trends have continued. It is widely acknowledged that the checking of new nuclear data libraries is highly skewed towards testing against criticality benchmarks, ignoring many of the high quality reactor physics benchmarks during the testing and production of general-purpose nuclear data libraries. However, continued increases in computing power, methodology (GPT), and additional availability reactor physics experiments from sources such as the International Handbook of Evaluated Reactor Physics Experiments should result in better testing of new libraries and ensured applicability to a wide variety of nuclear systems. It often has not. Leveraging the wealth of historical reactor physics measurements represents perhaps the simplest way to improve the quality of nuclear data libraries in the coming decade. Resources at the Nuclear Energy Agency can be utilized to assist in interrogating available identify benchmarks in the reactor physics experiments handbook, and expediting their use in verification and validation. Additionally, high quality experimental campaigns that should be examined in validation will be highlighted to illustrate potential improvements in the verification and validation process.
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Pal, Usha, and V. Jagannathan. "Physics principles to achieve comparable fission power from fertile and fissile rods of the conceptual ATBR/FTBR reactors." Annals of Nuclear Energy 35, no. 9 (September 2008): 1636–41. http://dx.doi.org/10.1016/j.anucene.2008.02.015.
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Petti, D., D. Crawford, and N. Chauvin. "Fuels for Advanced Nuclear Energy Systems." MRS Bulletin 34, no. 1 (January 2009): 40–45. http://dx.doi.org/10.1557/mrs2009.11.
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AbstractFuels for advanced nuclear reactors differ from conventional light water reactor fuels and also vary widely because of the specific architectures and intended missions of the reactor systems proposed to deploy them. Functional requirements of all fuel designs for advanced nuclear energy systems include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintenance of a geometry that can be cooled. In all cases, anticipated fuel performance is the limiting factor in reactor system design, and cumulative effects of increased utilization and increased exposure to inservice environments degrade fuel performance. In this article, the current status of each fuel system is reviewed, and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed.
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Kulikov, Gennady G., Anatoly N. Shmelev, Vladimir A. Apse, and Evgeny G. Kulikov. "Safety features of fast reactor with heavy atomic weight weakly neutron absorbing reflector." Nuclear Energy and Technology 6, no. 1 (March 11, 2020): 15–21. http://dx.doi.org/10.3897/nucet.6.50867.
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The purpose of the present study is the justification of the possibility of improving fast reactor safety by surrounding reactor cores with reflectors made of material with special neutron physics properties. Such properties of 208Pb lead isotope as heavy atomic weight, small neutron absorption cross section, and high inelastic scattering threshold result in certain peculiarities in neutron kinetics of the fast reactor equipped with 208Pb reflector, which can significantly enhance reactor safety. The reflector will also make possible generation of additional delayed neutrons characterized by the “dead” time. This will improve the resistibility of the fission chain reaction to stepwise reactivity excursions and exclude prompt supercriticality. Let us note that generation of additional delayed neutrons can be shaped by reactor designers. The relevance of the study amounts to the fact that generation of additional delayed neutrons in the reflector will make it possible mitigating the consequences of a reactivity accident even if the introduced reactivity exceeds the effective fraction of delayed neutrons. At the same time, the role of the fraction of delayed neutrons as the maximum permissible reactivity for reactor safety is depreciated. Scientific originality of the study pertains to the fact that the problem of yield of additional neutrons with properties close to normal delayed neutrons, has not been posed before. The authors suggest a new method for enhancing safety of fast reactors by increasing the fraction of delayed neutrons due to the time delay of prompt neutrons during their transfer in the reflector. In order to benefit from the expected advantages, the following combination is acceptable: lead enriched by 208Pb is used as a neutron reflector while natural lead or other material (sodium, etc.) is used as a coolant in the reactor core.
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PAKVASA, SANDIP. "NEUTRINO PHYSICS AND GEOPHYSICS WITH A DEEP OCEAN ANTINEUTRINO OBSERVATORY." Modern Physics Letters A 22, no. 25n28 (September 14, 2007): 1887–92. http://dx.doi.org/10.1142/s0217732307025108.
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This paper presents the science potential of a deep ocean antineutrino observatory being developed at Hawaii. The observatory design allows for relocation from one site to another. Positioning the observatory 60 km distant from a nuclear reactor complex enables precision measurement of neutrino mixing parameters, leading to a determination of neutrino mass hierarchy and θ13. At a mid-Pacific location the observatory measures the flux and ratio of uranium and thorium neutrinos from earth's mantle and performs a sensitive search for a hypothetical natural fission reactor in earth's core. A subsequent deployment at another mid-ocean location would test lateral heterogeneity of uranium and thorium in earth's mantle.
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Vogt, R., J. Randrup, P. Talou, J. T. Van Dyke, and L. A. Bernstein. "Parameter Optimization and Sensitivity Studies of Spontaneous Fission with FREYA." EPJ Web of Conferences 239 (2020): 05003. http://dx.doi.org/10.1051/epjconf/202023905003.
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For many years, the state of the art for simulating fission in transport codes amounted to sampling from average distributions. However, such "average" fission models have limited capabilities. Energy is not explicitly conserved and no correlations are available because all particles are emitted independently. However, in a true fission event, the emitted particles are correlated. Recently, Monte Carlo codes generating complete fission events have been developed, thus allowing the use of event-by-event analysis techniques. Such techniques are particularly useful because the complete kinematic information is available for the fission products and the emitted neutrons and photons. It is therefore possible to extract any desired observables, including correlations. The fast event-by-event fission code FREYA (Fission Reaction Event Yield Algorithm) generates large samples of complete fission events, employing only a few physics-based parameters. A recent optimization of these parameters for the isotopes in FREYA that undergo spontaneous fission is described and results are presented. The sensitivity of neutron observables in FREYA to the input yield functions is also discussed and the correlation between the average neutron multiplicity and fragment total kinetic energy is quantified.
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Julien-Laferrière, S., L. Thombansen, G. Kessedjian, A. Chebboubi, O. Serot, C. Sage, O. Méplan, et al. "Status of fission fragment observables measured with the LOHENGRIN spectrometer." EPJ Web of Conferences 211 (2019): 04004. http://dx.doi.org/10.1051/epjconf/201921104004.
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Nuclear fission yields are key parameters to evaluate reactor physics observables, such as fuel inventory, decay heat, spent fuel radiotoxicity, criticality but also for understanding the fission process. Despite a significant effort allocated to measure fission yields during the last decades, the recent evaluated libraries still need improvements in particular in the description of the uncertainties with the associated correlations. Additional kinds of measurements provide complementary information in order to test the models used in the nuclear data evaluation. Moreover, some discrepancies between these libraries must be explained. A common effort by the CEA, the LPSC and the ILL aims at tackling these issues by providing precise measurement of isotopic and isobaric fission yields with the related variance-covariance matrices. Nevertheless, the experimental program represents itself a large range of observables requested by the evaluations: isotopic yields, nuclear charge polarization, odd-even effect, isomeric ratio and their dependency with fission fragment kinetic energy as a probe of the nuclear de-excitation path in the (E*, Jπ) representation. Measurements for thermal neutron induced fission of 241Pu have been carried out at the Institut Laue Langevin using the LOHENGRIN mass spectrometer. Experimental program, observables reachable, results and comparison to model calculations are shown.
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Raflis, H., M. Ilham, Z. Su’ud, A. Waris, and D. Irwanto. "Core Configuration Analysis for Modular Gas-cooled Fast Reactor (GFR) using OpenMC." Journal of Physics: Conference Series 2072, no. 1 (November 1, 2021): 012007. http://dx.doi.org/10.1088/1742-6596/2072/1/012007.
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Abstract The core configuration analysis of modular Gas-cooled Fast Reactor (GFR) has been done to understand GFR performance. The modular GFR used a fast neutron spectrum and high-temperature helium gas, providing higher thermal efficiency than the other generation IV reactor candidates. In this paper, the variation of core configuration and dimension for core design has been applied in radial, axial, and radial-axial directions. The Monte Carlo method, named OpenMC code, has been used for the criticality and isotope evaluation of design core GFR. The OpenMC code provides the probabilistic solution to solve the neutron transport equation in a 3D model and non-homogenous physical volumes using Evaluated Nuclear Data File (ENDF/B-VII.b5) and continuous energy spectrum. The neutronics parameters characterized are the value of keff, fission rate and neutron flux distribution, and fissile material evolution to know of GFR core design’s performance. The analysis showed that the core configuration in radial direction gives a good understanding of the feasibility of GFR core design.
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Govers, Kevin, Lesley Adriaensen, Andrew Dobney, Mireille Gysemans, Christelle Cachoir, and Marc Verwerft. "Evaluation of the irradiation-averaged fission yield for burnup determination in spent fuel assays." EPJ Nuclear Sciences & Technologies 8 (2022): 18. http://dx.doi.org/10.1051/epjn/2022018.
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In order to derive the burnup of spent nuclear fuel from the concentration of selected fission products (typically the Nd isotopes and 137Cs), their irradiation-averaged fission yields need to be known with sufficient accuracy, as they evolve with the changes in the actinide vector over the irradiation history. To obtain irradiation-averaged values, radiochemists often resort to robust generic methods – i.e., based on simple mathematical relations – that weight the fission yields according to the actinides contributing to fission, without performing core physics calculations. In order to assess the performance of those generic methods, a database of about 30 000 spent nuclear fuel inventories has been constructed from neutron transport and depletion simulations, covering a representative range of fuel enrichment, burnup, assembly designs and reactor types. When testing several existing methods for effective fission yield calculation, some inaccuracies were identified, originating from improper one-group cross-section parameters that do not accurately reflect resonance and self-shielding effects, and too crude approximations in the estimation of the actinide concentration evolution. Revised effective fission and absorption cross-section parameters are then proposed here, as a first improvement to the earlier burnup determination methods. As a second step, a novel method is proposed that reduces the error on their radiation-averaged fission yield values, and hence on burnup, while retaining a straightforward calculation scheme.
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Portinari, D., A. Cammi, S. Lorenzi, M. Aufiero, Y. Calzavara, and A. Bidaud. "VOID COEFFICIENT SENSITIVITY ANALYSIS FOR THE TRIGA MARK II REACTOR AT L.E.N.A. (UNIPV)." EPJ Web of Conferences 247 (2021): 15005. http://dx.doi.org/10.1051/epjconf/202124715005.
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Sensitivity analysis studies the effect of a change in a given parameter to a response function of the system under investigation. In reactor physics, this usually translates into the study of how cross sections and fission spectrum modifications affect the value of the multiplication factor, the delayed neutron fraction or the void coefficient for example. Generalized Perturbation Theory provides a useful tool for the assessment of adjoint weighed functions such as keff and void coefficient sensitivities. In this work, the capability of SERPENT code to perform sensitivity calculation based on GPT is used to study the TRIGA Mark II research reactor installed at L.E.N.A. of University of Pavia. A general sensitivity analysis to the most important reactor’s cross sections has been performed in order to highlight the biggest reactivity contributions. Two numerically challenging tasks related to GPT calculation have been performed thanks to the relatively quick Monte Carlo approach allowed by this reactor: investigating the linearity of the reactivity injection caused by the flooding of the central channel, and calculating the fuel void coefficient sensitivity to the coolant density.
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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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Liu, Chang-Qi, Dong-Ying Huo, Chao Han, Kang Wu, Xing-Yu Liu, Xu Yang, Xiao-Hou Bai, et al. "Monte-Carlo study of pre-neutron emission mass and energy for neutron-induced <sup>232</sup>Th fission." Acta Physica Sinica 71, no. 1 (2022): 012501. http://dx.doi.org/10.7498/aps.71.20211333.
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The development of fourth-generation reactors and advanced nuclear energy systems require high-precision, multi-nuclide, and wide-energy-area neutron nuclear data. However, the current nuclear energy-related nuclear fission data in the China Nuclear Data Evaluation Library (CENDL library) are incomplete and cannot meet the current need. It is extremely important to establish the reliable calculation methods and tools for the neutron nuclear data. Based on the Monte-Carlo method, a model for calculating the pre-neutron fission fragment is established in this work. The mass and kinetic energy distribution of <sup>232</sup>Th(n,f) reaction at the medium- and low- incident neutron energy are studied. The calculations of the mass distribution with the different values of incident energy are compared with the experimental results. The maximum deviation of this work from the experimental data is ~1%, which is advantageous compared with the GEF and TALYS code (maximum deviation from the experimental value is ~2%). The calculation of the pre-neutron fission fragment kinetic energy also shows good agreement with experimental result. The results indicate that this model can well describe and predict the characteristics of pre-neutron fission fragment for <sup>232</sup>Th(n,f) reaction at the medium- and low- incident neutron energy. It also provides a new idea for calculating the neutron-induced actinide fission reactions.
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Oprea, C., M. A. Ahmad, J. H. Baker, and A. I. Oprea. "Mathematical Modeling of Neutron Induced Fission of 237Np Nucleus." Ukrainian Journal of Physics 67, no. 1 (February 10, 2022): 11. http://dx.doi.org/10.15407/ujpe67.1.11.
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Recent progress of applied and fundamental researches in nuclear physics necessitates new neutron sources with highly improved intensity. For a few years at JINR (Dubna) the development of new neutron facilities that will replace the IBR-2 neutron pulsed research reactor, which will finish its activities in 2032, is carried on. Some projects use the fission process induced by neutrons in neptunium-based fuels. In the present research, we will study the neutron-induced fission of 237Np nucleus. The cross-section, mass distribution, yields of isotopes of interest, average number of emitted prompt neutrons, neutron fission spectra, and other parameters are obtained. The mathematical modeling is done partially by using the theoretical models implemented in Talys software (TALYS-1.2) and by computer codes realized by the authors. The presented results are compared with the available data and are of interest in the JINR projects for the design of new neutron facilities destined for researches.
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Golay, M. W. "Advanced Fission Power Reactors." Annual Review of Nuclear and Particle Science 43, no. 1 (December 1993): 297–332. http://dx.doi.org/10.1146/annurev.ns.43.120193.001501.
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Bostelmann, Friederike, Germina Ilas, and William A. Wieselquist. "Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0." Journal of Nuclear Engineering 2, no. 4 (September 30, 2021): 345–67. http://dx.doi.org/10.3390/jne2040028.
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The EBR-II benchmark, which was recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, served as a basis for assessing the performance of the SCALE code system for fast reactor analyses. A reference SCALE model was developed based on the benchmark specifications. Great agreement was observed between the eigenvalue calculated with this SCALE model and the benchmark eigenvalue. To identify potential gaps and uncertainties of nuclear data for the simulation of various quantities of interest in fast spectrum systems, sensitivity and uncertainty analyses were performed for the eigenvalue, reactivity effects, and the radial power profile of EBR-II using the two most recent ENDF/B nuclear data library releases. While the nominal results are consistent between the calculations with the different libraries, the uncertainties due to nuclear data vary significantly. The major driver of observed uncertainties is the uncertainty of the 235U (n,γ) reaction. Since the uncertainty of this reaction is significantly reduced in the ENDF/B-VIII.0 library compared to ENDF/B-VII.1, the obtained output uncertainties tend to be smaller in ENDF/B-VIII.0 calculations, although the decrease is partially compensated by increased uncertainties in 235U fission and ν¯.
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Abarca, A., M. Avramova, and K. Ivanov. "IMPLEMENTATION AND VALIDATION OF A FISSION GAS RELEASE MODEL FOR CTFFUEL USING THE NEA/OECD IFPE DATABASE." EPJ Web of Conferences 247 (2021): 10018. http://dx.doi.org/10.1051/epjconf/202124710018.
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The nuclear reactors themselves are complex systems whose responses are driven by interactions between different physics phenomena within the reactor core. Traditionally, the different physics phenomena have been analyzed separately and its interaction considered via boundary conditions or closure models. However, in parallel with the development of computational technology, multi-physics coupled simulations are being used to obtain accurate predictions thanks to the consideration of the feedback effects on the fly (on-line). In the nuclear systems the fuel temperature is an important feedback parameter used to obtain the nuclear cross sections at given conditions by the neutron kinetics codes. An accurate prediction of temperature profile within the fuel rod involve several physics such as neutron kinetics, mechanics, material behavior and properties, heat transfer, thermal-hydraulics, and even chemistry. The pellet to clad gap conductance is possibly the most important source of uncertainty in the solution of conductivity equation in the fuel rod and the fuel temperature prediction. The gap conductance depends on two effects: the pellet to gap distance and the conductivity of the gas species that fill the gap. In this research work, the authors are focused on improving of the prediction of the gap gas conductivity in CTFFuel by implementing a fission gas release model in the code. The objective of this contribution is the implementation of a transient fission gas release model in CTFFuel and its validation using the experimental data available in the OECD/NEA International Fuel Performance Experiments (IFPE) database. CTFFuel is an isolated fuel heat transfer capability within the framework of CTF code, the state-of-the-art version of the Coolant Boiling in Rod Arrays Code – Two-Fluid (COBRA-TF) sub-channel thermal-hydraulic code. The code is being jointly developed by North Carolina State University (NCSU) and Oak Ridge National Laboratory (ORNL) within the US Department of Energy (DOE) Consortium for Advanced Simulation of LWRs (CASL).