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Автор: Grafiati

Опубліковано: 30 травня 2022

Оновлено: 31 травня 2022

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Статті в журналах з теми "Subcrital nuclear reactor"

Vega-Carrillo, Hector Rene, V. P. Singh, Claudia Rafela Escobedo-Galván, Diego Medina Castro, Arturo Agustin Ortiz Hernandez, Teodoro Rivera-Montalvo, and Segundo Agustín Martínez-Ovalle. "Mini Subcritical Nuclear Reactor." Journal of Nuclear Physics, Material Sciences, Radiation and Applications 6, no. 2 (February 26, 2019): 170–77. http://dx.doi.org/10.15415/jnp.2019.6.02.170-177.

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Hector Rene Vega-Carrillo, V. P. Singh, Claudia Rafela Escobedo-Galván, Diego Medina Castro, Arturo Agustin Ortiz Hernandez, Teodoro Rivera-Montalvo, and Segundo Agustín Martínez-Ovalle. "Mini Subcritical Nuclear Reactor." Journal of Nuclear Physics, Material Sciences, Radiation and Applications 6, no. 2 (February 26, 2019): 179–85. http://dx.doi.org/10.15415/jnp.2019.62026.

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A mini subcritical nuclear reactor was designed using Monte Carlo methods. The reactor has light water as moderator, natural uranium as fuel, and a 239PuBe neutron source. In the design uranium fuel was modeled in an arrangement of concentric rings: 8.5, 14.5, 20.5 26.5, 32.5 cm-inner radius, 3 cm-thick, and 36 cm-high. Different models were made from a single ring of natural uranium to five rings. For each case, the neutron spectra, the neutron fluence distribution, the effective multiplication factor, the amplification factor, and the reactor power were estimated. The ambient dose equivalent rate outside the mini reactor was also estimated. The maximum value for the keff (0.78) was obtained when five rings of fuel were used; this value is close to 0.86 which belongs to a Nuclear Chicago subcritical reactor which requires almost twice the amount of uranium than the mini subcritical reactor.

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Zhaohui, WANG, REN Jie, WU Hongyi, QIAN Jing, HUANG Hanxiong, WANG Wenming, JIANG Wei, et al. "Measurement of Gamma-Ray from Inelastic Neutron Scattering on 56Fe." EPJ Web of Conferences 239 (2020): 01036. http://dx.doi.org/10.1051/epjconf/202023901036.

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In nuclear reactors, inelastic neutron scattering is a significant energy-loss mechanism which has deep impacts on designments of nuclear reactor and radiation shielding. Iron is an important material in reactor. However, for the existing nuclear data for iron, there exists an obvious divergence for the inelastic scattering cross sections and the related gamma production sections. Therefore the precise measurements are urgently needed for satisfying the demanding to design new nuclear reactors (fast reactors), Accelerator Driven Subcritical System (ADS), and other nuclear apparatus. In this paper, we report a new system with an array of HPGe detectors, electronics and acquisition system. Experiments had been carried out on three neutron facilities.

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Kostic, Ljiljana. "Reactivity determination in accelerator driven reactors using reactor noise analysis." Nuclear Technology and Radiation Protection 17, no. 1-2 (2002): 19–26. http://dx.doi.org/10.2298/ntrp0202019k.

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Feynman-alpha and Rossi-alpha methods are used in traditional nuclear reactors to determine the subcritical reactivity of a system. The methods are based on the measurement of the mean value, variance and the covariance of detector counts for different measurement times. Such methods attracted renewed attention recently with the advent of the so-called accelerator driven reactors (ADS) proposed some time ago. The ADS systems, intended to be used either in energy generation or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those traditionally used by radioactive sources. In such reactors the monitoring of the subcritical reactivity is very important, and a statistical method, such as the Feynman-alpha method, is capable of resolving this problem.

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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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Heidet, Florent, Nicholas R. Brown, and Malek Haj Tahar. "Accelerator–Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles." Reviews of Accelerator Science and Technology 08 (January 2015): 99–114. http://dx.doi.org/10.1142/s1793626815300066.

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This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of accelerator-driven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

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Seifritz, W. "Nuclear transmutation by flux compression." Kerntechnik 66, no. 5-6 (October 1, 2001): 225–28. http://dx.doi.org/10.1515/kern-2001-0093.

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Abstract A new idea for the transmutation of minor actinides, long (and even short) lived fission products is presented. It is based an the property of neutron flux compression in nuclear (fast and/or thermal) reactors possessing spatially non-stationary critical masses. An advantage factor for the burn-up fluence of the elements to be transmuted in the order of magnitude of 100 and more is obtainable compared with the classical way of transmutation. Three typical examples of such transmuters (a subcritical ringreactor with a rotating reflector, a sub-critical ring reactor with a rotating spallation source, the socalled “pulsed energy amplifier”, and a fast burn-wave reactor) are presented and analysed with regard to this purpose.

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Vega-Carrillo, Hector Rene, Isvi Ruben Esparza-Garcia, and Alvaro Sanchez. "Features of a subcritical nuclear reactor." Annals of Nuclear Energy 75 (January 2015): 101–6. http://dx.doi.org/10.1016/j.anucene.2014.08.006.

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Babenko, Vladimir, Volodymyr Pavlovych, and Volodymyr Gulik. "The pulsed subcritical amplifier of neutron flux driven by high-intensity neutron generator." Nuclear Technology and Radiation Protection 34, no. 1 (2019): 1–12. http://dx.doi.org/10.2298/ntrp180629018b.

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The subcritical reactor driven by external neutron source could apply as useful instrument for modern nuclear energy applications requiring high-level irradiation of different materials by the high-energy and high-intense neutron flux (e. g., nuclear waste transmutation, radiopharmaceutical production, etc.). The propagation of neutron pulses through the subcritical nuclear system was considered in the present paper. Simple homogeneous subcritical systems and a model of two-zone subcritical reactor were computationally investigated using Monte Carlo MCNP4c transport code. The propagation of one initial neutron pulse and series of one hundred neutron pulses through the presented subcritical nuclear models were simulated. In this study, the neutron multiplication factor, the neutron flux, the energy amplification factor, the total energy of neutrons in initial pulse, etc. were obtained and analyzed. The presented calculations have shown that the considered pulse subcritical systems can be successfully used as effective amplifiers of neutron flux from the initial source. The modeling results indicate that there is an achievement of a stable, high level of neutron flux caused by the accumulation of delayed neutrons from previous pulses in series of one hundred pulses for both homogeneous and heterogeneous systems.

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Tumanyan, A. R., and A. G. Khudaverdyan. "Subcritical accelerator-controlled medium-power nuclear reactor." Atomic Energy 79, no. 1 (July 1995): 486–87. http://dx.doi.org/10.1007/bf02406211.

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Більше джерел

Дисертації з теми "Subcrital nuclear reactor"

Онисимчук, Тетяна Михайлівна. "Трансмутація радіоактивних відходів з удосконаленням системи сповільнення швидких нейтронів". Master's thesis, Київ, 2018. https://ela.kpi.ua/handle/123456789/25799.

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Дисертацію присвячено впровадженню системи сповільнення швидких нейтронів в підкритичних ядерних установках, керованих зовнішнім джерелом, виконаної із вторинно переробленого поліетилентерефталату. У дисертації визначено показники безпеки функціонування підкритичних систем з пластиковим сповільнювачем, доведено можливість застосування вторинного полімеру в якості сповільнювача швидких нейронів та встановлено залежність коефіцієнту пропускання іонізуючого випромінювання від циклу переробки матеріалу захисного шару. Визначено, що економічна ефективність використання вторинної сировини з врахуванням збільшення кількості полімеру в 1,3 рази в порівнянні з еталонним поліетиленом, окреслюється економією капіталовкладень у розмірі 1 197 683 грн на момент проведення розрахунку. Отримано залежністькоефіцієнту пропускання іонізуючого випромінювання від товщини зовнішнього шару сповільнювача після циклічних етапів переробки, яка описується поліномом 3-го порядку. Виявлено, що допустима кількість циклів переробки для поліетилентерефталату становить три, при якій товщина сповільнювача становитиме 750 мм. При подальших циклах переробки використання вторинної сировини економічно недоцільне. Розроблений стартап-проект реалізації технології на вітчизняному ринку прогнозує отримання разового прибутку у розмірі близько 800 тис. грн протягом 1 року від дати отримання сертифікату відповідності.
The dissertation is devoted to the introduction of a fast neutron slowdown system in subcritical nuclear installations controlled by an external source made from recycled polyethylene terephthalate. The dissertation defines the performance indicators of subcritical systems with plastic moderator, proved the possibility of using the secondary polymer as a moderator of fast neurons, and the dependence of the ionizing radiation transmittance coefficient on the cycle of processing the material of the protective layer has been established. It is determined that the economic efficiency of the use of secondary raw materials, taking into account the increase in the amount of polymer in 1,3 times compared with the reference polyethylene, is defined by the savings of investments in the amount of 1 197 683 UAH at the time of calculation. The dependence of transmittance coefficient of ionizing radiation on the thickness of the outer layer of the retarder after the cyclic stages of processing, which is described by the polynomial of the 3rd order, is obtained.It was found that the permissible number of recycling cycles for polyethylene terephthalate is three, in which the thickness of the moderator will be 750 mm. In subsequent cycles of recycling, the use of secondary raw materials will be economically impractical. The developed start-up project of technology implementation on the domestic market predicts a one-time profit of about 800 thousand UAH for 1 year from the date of receipt of the certificate of conformity.
Диссертация посвящена внедрению системы замедления быстрых нейтронов в подкритических ядерных установках, управляемых внешним источником, выполненной из вторично переработанного полиэтилентерефталата. В диссертации определены показатели безопасности функционирования подкритических систем с пластиковым замедлителем, доказана возможность применения вторичного полимера в качестве замедлителя быстрых нейронов и установлена зависимость коэффициента пропускания ионизирующего излучения от цикла переработки материала защитного слоя. Определено, что экономическая эффективность использования вторичного сырья с учетом увеличения количества полимера в 1,3 раза в сравнении с эталонным полиэтиленом, определяется экономией капиталовложений в размере 1 197 683 грн на момент проведения расчета. Получена зависимость коэффициента пропускания ионизирующего излучения от толщины внешнего слоя замедлителя после циклических этапов переработки, которая описывается полиномом 3-го порядка. Выявлено, что допустимое количество циклов переработки для полиэтилентерефталата составляет три, при котором толщина замедлителя составит 750 мм. При последующих циклах переработки использование вторичного сырья экономически нецелесообразно. Разработанстартап-проект реализации технологии на отечественном рынке прогнозирует получение разового дохода в размере около 800 тыс. грн в течение 1 года от даты получения сертификата соответствия.

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Sommer, Christopher Michael. "Subcritical transmutation of spent nuclear fuel." Diss., Georgia Institute of Technology, 2011. http://hdl.handle.net/1853/41205.

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A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner Reactor (SABR). SABR is a fusion-fission hybrid reactor that combines the leading sodium cooled fast reactor technology with the leading tokamak plasma technology based on ITER physics. Two general fuel cycles were considered for the SABR system. The first fuel cycle is one in which all of the transuranics from light water reactors are burned in SABR. The second fuel cycle is a minor actinide burning fuel cycle in which all of the minor actinides and some of the plutonium produced in light water reactors are burned in SABR, with the excess plutonium being set aside for starting up fast reactors in the future. The minor actinide burning fuel cycle is being considered in European Scenario Studies. The fuel cycles were evaluated on the basis of TRU/MA transmutation rate, power profile, accumulated radiation damage, and decay heat to the repository. Each of the fuel cycles are compared against each other, and the minor actinide burning fuel cycles are compared against the EFIT transmutation system, and a low conversion ratio fast reactor.

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Bopp, Andrew T. "The calculation of fuel bowing reactivity coefficients in a subcritical advanced burner reactor." Thesis, Georgia Institute of Technology, 2013. http://hdl.handle.net/1853/50295.

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The United States' fleet of Light Water Reactors (LWRs) produces a large amount of spent fuel each year; all of which is presently intended to be stored in a fuel repository for disposal. As these LWRs continue to operate and more are built to match the increasing demand for electricity, the required capacity for these repositories grows. Georgia Tech's Subcritical Advanced Burner Reactor (SABR) has been designed to reduce the capacity requirements for these repositories and thereby help close the back end of the nuclear fuel cycle by burning the long-lived transuranics in spent nuclear fuel. SABR's design is based heavily off of the Integral Fast Reactor (IFR). It is important to understand whether the SABR design retains the passive safety characteristics of the IFR. A full safety analysis of SABR's transient response to various possible accident scenarios needs to be performed to determine this. However, before this safety analysis can be performed, it is imperative to model all components of the reactivity feedback mechanism in SABR. The purpose of this work is to develop a calculational model for the fuel bowing reactivity coefficients that can be used in SABR's future safety analysis. This thesis discusses background on fuel bowing and other reactivity coefficients, the history of the IFR, the design of SABR, describes the method that was developed for calculating fuel bowing reactivity coefficients and its validation, and presents an example of a fuel bowing reactivity calculation for SABR.

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Persson, Carl-Magnus. "Reactivity Assessment in Subcritical Systems." Licentiate thesis, Stockholm : Fysiska institutionen, Kungliga Tekniska högskolan, 2007. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-4363.

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Zino, John Frederick. "Analysis of subcritical experiments using fresh and spent research reactor fuel assemblies." Diss., Georgia Institute of Technology, 1999. http://hdl.handle.net/1853/17507.

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Adams, William Mark 1961. "APPLICATION OF THE VARIANCE-TO-MEAN RATIO METHOD FOR DETERMINING NEUTRON MULTIPLICATION PARAMETERS OF CRITICAL AND SUBCRITICAL REACTORS (REACTOR NOISE, FEYNMAN-ALPHA)." Thesis, The University of Arizona, 1985. http://hdl.handle.net/10150/275438.

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Sumner, Tyler Scott. "A safety and dynamics analysis of the subcritical advanced burner reactor: SABR." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2008. http://hdl.handle.net/1853/24636.

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Lee, David. "Neutron production with thorium fuel in accelerator driven subcritical reactors." Thesis, University of Huddersfield, 2018. http://eprints.hud.ac.uk/id/eprint/34579/.

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ADSRs (Accelerator-Driven Subcritical Reactors) incorporate a spallation technique which is an efficient way to produce high neutron flux by externally supplying neutrons into the reactor. In this thesis, a review of spallation is given explaining the spallation reaction process, describing spallation reactions in dense metal and analysing the resulting neutron energy spectra. The thesis also discusses current spallation sources around the world. Studies involving proton-induced neutron production in spallation target are demonstrated. Spallation reactions on a lead target have been simulated using a Monte Carlo transport code called GEANT4, and the benchmarking of these simulations against experimental neutron spectra produced from a thick lead target bombarded with 0.5 and 1.5 GeV protons is discussed. This is followed by discussion of the angular distribution of neutrons of different energies in order to understand the emission of neutrons from a spallation target. Lead and Lead Bismuth Eutectic (LBE) are both widely utilised to produce neutrons; this is due to the fact that lead is a high Z, heavy metal element which is relatively cheap to use. This thesis provides a comparison between lead and LBE in terms of their effect on neutron energy spectra at various projecting angles. Given the confidence in the GEANT4 simulation provided by the benchmarking studies, the thesis goes on to discuss neutron production and behaviour in the environment of thorium-fuelled ADSRs. With a spallation target composed of LBE, the design of the MYRRHA reactor developed by SCKCEN has been configured to explore neutron production created from each layer filled with thorium fuel. This is then followed by a focus on neutrons escaping from the LBE-composed reflector, and the thesis provides an analysis of the effect of several different materials used for inner and outer shielding in the reactor core. By using the latest nuclear data library and numerical techniques provided in the GEANT4 code, the author has been able to simulate the actual usage of thorium in an ADSR reactor set-up which has not yet been fully demonstrated. The thesis concludes with the idea that thorium could well be utilised as an actual fuel source in an ADSR for the purpose of transmutation, and possibly for energy production as well.

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Berglöf, Carl. "On measurement and monitoring of reactivity in subcritical reactor systems." Doctoral thesis, KTH, Reaktorfysik, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-12483.

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Accelerator-driven systems have been proposed for incineration of transuranic elements from spent nuclear fuel. For safe operation of such facilities, a robust method for reactivity monitoring is required. Experience has shown that the performance of reactivity measurement methods in terms of accuracy and applicability is highly system dependent. Further investigations are needed to increase the knowledge data bank before applying the methods to an industrial facility and to achieve license to operate such a facility. In this thesis, two systems have been subject to investigation of various reactivity measurement methods. Conditions for successful utilization of the methods are presented, based on the experimental experience. In contrast to previous studies in this field, the reactivity has not only been determined, but also monitored based on the so called beam trip methodology which is applicable also to non-zero power systems. The results of this work constitute a part of the knowledge base for the definition of a validated online reactivity monitoring methodology for facilities currently being under development in Europe (XT-ADS and EFIT).
QC 20100621

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David, Sylvain. "Capacités des réacteurs hybrides au plomb pour la production d'énergie et l'incinération avec multirecyclage des combustibles : évolution des paramètres physiques : radiotoxicités induites." Université Joseph Fourier (Grenoble), 1999. http://www.theses.fr/1999GRE10042.

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Le concept des reacteurs sous-critiques pilotes par accelerateur (reacteurs hybrides), relance au debut des annees 90 par c. Rubbia et c. D. Bowman, permet d'ouvrir des voies nouvelles quant a la gestion des dechets nucleaires, qu'il s'agisse des dechets actuels ou de la production d'energie plus propre basee sur l'utilisation du cycle du thorium. Dans un premier temps, ce travail concerne l'etude des caracteristiques de la multiplication neutronique dans un cur de reacteur sous-critique et met en evidence les differences fondamentales existant avec les systemes critiques et les avantages qui en decoulent. Cette etude est liee aux series de mesures realisees a cadarache (experiences muse) dont les premiers resultats sont presentes. La sous-criticite d'un reacteur hybride rend ce systeme tres flexible, et permet d'envisager differentes utilisations, comme la production d'energie ou l'incineration de dechets. La deuxieme partie de ce travail concerne donc l'etude par simulation monte carlo, des capacites des systemes hybrides a spectre rapide et refroidis au plomb, a produire de l'energie en utilisant differents cycles de combustible (uranium et thorium), tout en assurant la regeneration de la matiere fissile et le maintien de la reactivite sans intervention exterieure. Nous envisageons differents types de multirecyclage du combustible. Les resultats permettent de quantifier les avantages lies a l'utilisation de la filiere thorium, en terme de reduction de radiotoxicite notamment. L'etude des phases transitoires necessaires pour aboutir a cette filiere a partir des combustibles actuels (plutonium issu des reacteurs thermiques et uranium enrichi) propose une gestion efficace des actinides produits par les reacteurs actuels en tant que matiere fissile d'appoint. Enfin, l'incineration des actinides en fin de cycle (scenario d'arret de la filiere) est envisagee et permet de decrire les avantages des systemes hybrides refroidis au plomb pour la reduction de la radiotoxicite d'un inventaire en fin de cycle.

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Книги з теми "Subcrital nuclear reactor"

Italy) International Workshop on Fusion Neutrons and Subcritical Nuclear Fission (2011 Varenna. Fusion for neutrons and subcritical nuclear fission: Proceedings of the international conference : Varenna, Italy, 12-15 September 2011. Edited by Källne Jan. Melville, N.Y: American Institute of Physics, 2012.

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International Atomic Energy Agency Specialists' Meeting on Subcritical Crack Growth (2nd (1985 Sendai, Japan). Proceedings of the second International Atomic Energy Agency Specialists' Meeting on Subcritical Crack Growth: Held at Sendai, Japan, May 15-17, 1985. Washington, D.C: The Office, 1986.

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David, S., H. Nifenecker, and O. Meplan. Accelerator Driven Subcritical Reactors. Taylor & Francis Group, 2003.

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David, S., H. Nifenecker, and O. Meplan. Accelerator Driven Subcritical Reactors (Fundamental and Applied Nuclear Physics Series.). Taylor & Francis, 2003.

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H, Cullen W., International Atomic Energy Agency, T͡Sentralʹnyĭ nauchno-issledovatelʹskiĭ institut tekhnologii mashinostroenii͡a (Soviet Union), U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research., Materials Engineering Associates, and Argonne National Laboratory, eds. Proceedings of the Third International Atomic Energy Agency Specialists' Meeting on Subcritical Crack Growth: Held at Moscow, USSR, May 14-17, 1990. Washington, D.C: The Commission, 1990.

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Proceedings of the Third International Atomic Energy Agency Specialists' Meeting on Subcritical Crack Growth: Held at Moscow, USSR, May 14-17, 1990. Washington, D.C: The Commission, 1990.

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Частини книг з теми "Subcrital nuclear reactor"

Verma, Vinod Kumar, and Karel Katovsky. "Special Hybrid Systems and Molten-Salt Reactors." In Spent Nuclear Fuel and Accelerator-Driven Subcritical Systems, 21–30. Singapore: Springer Singapore, 2018. http://dx.doi.org/10.1007/978-981-10-7503-2_2.

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Yamanaka, Masao. "Effective Delayed Neutron Fraction." In Accelerator-Driven System at Kyoto University Critical Assembly, 83–123. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_4.

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AbstractIn kinetic analyses on ADS, although adjoint flux distribution is defined under the existence of an external neutron source, an issue of the proper determination of the weighting function still remains in the definition to obtain the kinetics parameters in the fixed-source calculations. Here, an alternative methodology is proposed with the combined use of the k-ratio method and the reaction rates obtained by the fixed-source calculations, when the subcriticality level or the spectrum of the external neutron source is varied. In ADS experiments, the measurement of βeff is expected to provide complementary verification of the calculation and reliability of nuclear data. Then, the formulation of the Rossi-α method in the pulsed-neutron source has been already available for application to the subcriticality measurement in the pulsed-neutron source (PNS) experiments. Accordingly, the methodology is applied uniquely to deduce the βeff value with the pulsed-neutron source (spallation neutrons), with the combined use of the results of experiments and calculations. Using parameters α and ρ$, the values of βeff/Λ are deduced at near-critical configurations through experimental analyses. To estimate the numerical precision of Λ, the value of βeff/Λ is used as an index of Λ evaluation that is defined by a ratio of Λ values in the super-critical and subcritical states.

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Kerlin, Thomas W., and Belle R. Upadhyaya. "Subcritical operation." In Dynamics and Control of Nuclear Reactors, 53–56. Elsevier, 2019. http://dx.doi.org/10.1016/b978-0-12-815261-4.00005-6.

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"Deep underground disposal of nuclear waste." In Accelerator Driven Subcritical Reactors. Taylor & Francis, 2003. http://dx.doi.org/10.1201/9781420034738.axi.

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Pioro, I. L., and C. O. Zvorykin. "Thermophysical properties of fluids at subcritical and critical/supercritical conditions." In Handbook of Generation IV Nuclear Reactors, 771–94. Elsevier, 2016. http://dx.doi.org/10.1016/b978-0-08-100149-3.15003-7.

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Тези доповідей конференцій з теми "Subcrital nuclear reactor"

Yao, Chengzhi, Yuerong Fan, Jinshan Zhang, Haifen Han, Dayong Yi, and Zhanli Zhang. "Structure Design of Jordan Subcritical Reactor." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15971.

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The subcritical reactor is a kind of small nuclear facility, which may realize the chain reactor reaction through introducing an external neutron source to meet with the production, research, teaching, testing and training purposes. Meanwhile, the subcritical reactor will not reach the critical status, which has advantages of simple structure, safety, reliability. Jordan subcritical reactor is a light water moderated reactor, it employs the Pu-Be as external neutron source, and the neutron multiplication factor in the range of 0.94～0.95. Jordan subcritical reactor mainly consists of fuel elements, reactor core and reactor core vessel assembly, operation platform, water loop system, neutron source driving system, which can be used for the purposes of teaching, training, testing and research. This paper reviews the international application history of subcritical reactor and its status of research and development, describes the design purpose and requirement of Jordan subcritical reactor. The detailed structure design of Jordan subcritical reactor is illustrated. Furthermore, the structure design characteristics and some difficulties of Jordan subcritical reactor are mentioned in detail, such as the choose of structure materials, lay out of Jordan subcritical reactor, control of machining precision and sealing of pipes. Then, the possible solutions of these problems are presented. Now, the manufacture and installation of Jordan subcritical reactor has been completed, which fulfill the anticipated design requirement.

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Saltanov, Eugene, Romson Monichan, Elina Tchernyavskaya, and Igor Pioro. "Steam-Reheat Option for SCWRs." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-76061.

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Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30 – 35% to about 45 – 48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency a nuclear steam reheat has to be introduced inside a reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, USA and some other countries have developed and implemented the nuclear steam reheat at subcritical-pressure in experimental reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW nuclear-reactor concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the nuclear steam reheat at subcritical pressures in pressure-tube reactors is easier task than that in pressure-vessel reactors. Some design features related to the nuclear steam reheat are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors with the nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.

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Saltanov, Eugene, Wargha Peiman, Amjad Farah, Krysten King, Maria Naidin, and Igor Pioro. "Steam-Reheat Options for Pressure-Tube SCWRs." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29972.

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Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 40-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30–35% to about 45–48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency Nuclear Steam Reheat (NSR) has to be introduced inside a reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat at subcritical-pressure experimental boiling reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the nuclear steam reheat at subcritical pressures in pressure-tube reactors is easier task than that in pressure-vessel reactors. Some design features related to the NSR are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors with the nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.

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Srinivasan, R., and FTR Team. "Indian fusion test reactor." In FUSION FOR NEUTRONS AND SUBCRITICAL NUCLEAR FISSION: Proceedings of the International Conference. AIP, 2012. http://dx.doi.org/10.1063/1.4706850.

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Wang, Feng, Xue Qin, Zilong An, and Bo Cui. "Physics Analysis of the Accelerator Driven Subcritical Reactor Core." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15846.

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The reactor core of an accelerator driven sub-critical system has been physically analyzed by the MCNP code. Neutron flux density of different area within the reactor has been calculated, and the influence on its distribution has also been analyzed. Results show that there exists higher fast neutron flux variation at different element layer in fast region, and relatively lower thermal neutron flux variation at different element layer in thermal region. The calculated neutron flux meets the general design requirements in the reflector and shielding layer. Neutron multiplication factor is remarkable in the fast neutron spectrum area, and it realizes the energy amplification in the thermal spectrum area. The statistical particle number of code can influence the accuracy of the calculation and variation of the core design parameters can change the neutron flux distribution in the reactor core.

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Mao, Lisheng, Minghuang Wang, Xuewei Fu, Jieqiong Jiang, and Yican Wu. "Preliminary Fuel Cycle Analysis of a Fusion-Driven Subcritical Reactor." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15588.

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The levelized cost of electricity (LCOE) has been performed to compare two fuel cycle scenarios: a once-through cycle (LWR OT) and a fusion Cdriven reactor, namely FDS-SFB, recycling employing PUREX (Purex-SFB). In order to estimate the LCOEs, the mass flows based on an equilibrium mode were analysed. The sensitivity of the results to variations in key parameters was also performed. A simple dynamic model was also constructed to consider other important factors that characterize a fuel cycle, e.g. resource utilization, environmental effects. The results of economics are as flows: LWR OT 29mills/KWh, Purex-SFB 48.19mills/KWh. It was found that the capital cost accounts for the largest proportion of the LCOEs. The fuel cycle cost analysis indicates that the FDS-SFB fuel cycle cost will be competitive with the once-through fuel cycle. Also, sensitivity analysis indicates that fuel cycle cost of LWR would be higher than that of the LWR-SFB fuel cycle with the uranium Price rising. Dynamic model analysis indicates that Purex-SFB could reduce the amounts of MA and the amounts of natural uranium considerably compared with LWR OT.

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Yang, Xie, Lei Shi, and Ding She. "Neutronics Analyses of Small Compact Prismatic Nuclear Reactors for the Preliminary Design." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66290.

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Due to the advantages of small volume, light weight and long-time running, nuclear reactor can provide an idea energy source for submarines, ships and even space crafts. In this paper, two small compact prismatic nuclear reactors with different core block material are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with high enriched uranium nitride fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa respectively. High temperature helium flowing out of the reactor can be used as the working medium for Closed Brayton Cycle (CBC) power conversion to generate at least 1 MW electricity due to the high efficiency of CBC. Neutronics analyses of reactors for the preliminary design in this paper are performed using Reactor Monte-Carlo (RMC) code developed by Tsinghua University. Both the two reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, which have at least $1 reactivity shutdown margin, and remains at least $1 subcritical in the submersion accident as well as one control drum failed accident. Finally, the optimization design is determined after comparing the U-235 mass and the total reactor mass of the above two prismatic reactors.

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Moiseenko, V. E., S. V. Chernitskiy, O. Ågren, and K. Noack. "A fuel for sub-critical fast reactor." In FUSION FOR NEUTRONS AND SUBCRITICAL NUCLEAR FISSION: Proceedings of the International Conference. AIP, 2012. http://dx.doi.org/10.1063/1.4706860.

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Orsitto, Francesco Paolo. "Diagnostics for hybrid reactors." In FUSION FOR NEUTRONS AND SUBCRITICAL NUCLEAR FISSION: Proceedings of the International Conference. AIP, 2012. http://dx.doi.org/10.1063/1.4706882.

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Wu, Yican, Yunqing Bai, Yong Song, Qunying Huang, Zhumin Zhao, Gang Song, Liqin Hu, and Jieqiong Jiang. "Design and R&D Progress of China Lead-Based Reactor." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-31136.

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Liquid lead or lead-based alloy is a potential candidate coolant for fast reactors and ADS subcritical reactors because of its many unique nuclear, thermal-physical and chemical attributes. Chinese Academy of Sciences (CAS) had launched an engineering project to develop ADS system and lead-based reactors. Series CLEAR reactor conceptual designs were finished, and the preliminary engineering design for the China Lead-based research reactor (CLEAR-I) was underway. The key components prototypes for engineering validation including the control rod drive system, refueling system, fuel assembly have been constructed, the validation experiment are carrying out. KYLIN series PbBi experimental loops has already been built to perform structure material corrosion experiment, thermal-hydraulics experiment and safety experiment for CLEAR series reactors. The highly intensified neutron generator HINEG will be constructed to take the benchmark experiment of neutronics simulation codes. In this paper, the design and R&D progress are presented.

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