Journal articles: 'Subcrital nuclear reactor'

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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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Shelaev, L. A., A. M. Baldin, A. I. Malakhov, and E. –J Langrock. "An accelerator–driven nuclear reactor." Kerntechnik 66, no. 5-6 (October 1, 2001): 246–53. http://dx.doi.org/10.1515/kern-2001-0099.

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Abstract The basic parameters of the accelerator facility for a proton beam of 100 mA and 1 GeV using a superconducting separate orbit cyclotron, are given. By means of such an accelerator system one can activate a subcritical 3 GW thermal power nuclear reactor of the well known RMBK–type 1000. The system is stable against any “Chernobyl–type” power excursion. When this accelerator system will be coupled to existing RMBK–type 1000 power stations, the resulting electricity production would become substantially more safe. Associated problems and design parameters are discussed.

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Holbert, Keith E. "A Study of the Minimum Thermal Power of a Nuclear Reactor." Journal of Nuclear Engineering 2, no. 4 (October 20, 2021): 412–21. http://dx.doi.org/10.3390/jne2040031.

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The minimum mass for a critical reactor is well studied whereas the minimum heat production from a nuclear reactor has received little attention. The thermal power of a (sub)critical reactor originates from fission as well as radioactive decay. Fission includes neutron-induced and spontaneous fission. For an idealized critical core, we find that the minimum theoretical power is ER/Λ, whereas for a subcritical reactor comprising fissionable material undergoing spontaneous fission, the minimum power is dictated by subcritical multiplication. Interestingly, radioisotopic heat generation exceeds the minimum theoretical fission power for most of the fissile materials examined in this study.

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Luo, Run, Shripad T. Revankar, and Fuyu Zhao. "Comparative Safety Analysis of Accelerator Driven Subcritical Systems and Critical Nuclear Energy Systems." Applied Sciences 11, no. 17 (September 3, 2021): 8179. http://dx.doi.org/10.3390/app11178179.

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The accelerator driven subcritical system (ADS) has been chosen as one of the best candidates for Generation IV nuclear energy systems which could not only produce clean energy but also incinerate nuclear waste. The transient characteristics and operation principles of ADS are significantly different from those of the critical nuclear energy system (CNES). In this work, the safety characteristics of ADS are analyzed and compared with CNES by a developed neutronics and thermal-hydraulics coupled code named ARTAP. Three typical accidents are carried out in both ADS and CNES, including reactivity insertion, loss of flow, and loss of heat sink. The comparison results show that the power and the temperatures of fuel, cladding, and coolant of the CNES reactor are much higher than those of the ADS reactor during the reactivity insertion accident, which means ADS has a better safety advantage than CNES. However, due to the subcriticality of the ADS core and its low sensitivity to negative reactivity feedback, the simulation results indicate that the inherent safety characteristics of CNES are better than those of ADS under loss of flow accident, and the protection system of ADS would be quickly activated to achieve an emergency shutdown after the accident occurs. For the loss of heat sink, it is found that the peak temperatures of the cladding in the ADS and CNES reactors are lower than the safety limit, which imply these two reactors have good safety performance against loss of heat sink accidents.

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Thompson, Nicholas, Jesson Hutchinson, Rian Bahran, David Hayes, William Myers, Jennifer Arthur, John Bounds, et al. "National Criticality Experiments Research Center (NCERC) - capabilities and recent measurements." EPJ Web of Conferences 239 (2020): 18003. http://dx.doi.org/10.1051/epjconf/202023918003.

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The National Criticality Experiments Research Center (NCERC) located at the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS) and operated by Los Alamos National Laboratory (LANL) is home to four critical assemblies which are used to support of range of missions, including nuclear criticality safety and nuclear nonproliferation. Additionally, subcritical systems can also be assembled at NCERC. NCERC is providing critical and subcritical experiments valuable to the nuclear data community and experiments performed at NCERC are often published as benchmarks in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. This manuscript will give a broad overview of recent experiments performed at NCERC, upcoming experiments, and why integral measurements are important and useful to the nuclear data community. The four critical assemblies are GODIVA IV, FLATTOP, COMET, and PLANET. GODIVA IV is a cylindrical metal fast burst reactor, the fourth in the GODIVA series that dates back to the 1950’s. FLATTOP is an highly enriched uranium (HEU) or Pu core reflected by natural uranium. COMET and PLANET are vertical lift assemblies, where one half of the reactor can be lifted to the upper half of the reactor to create a critical system. Some recent experiments include various critical intermediate energy assemblies with lead, and subcritical measurements of plutonium reflected by copper, tungsten, and nickel. Work is also underway to make a better measurement of the critical mass of neptunium, using a neptunium sphere surrounded by nickel shells. Additionally, measurements will be performed next year with HEU shells from Rocky Flats. These HEU shells will be stacked together to make larger systems, allowing for a large range of criticality (from subcritical to delayed critical). Other upcoming measurements include an HEU critical assembly sensitive to intermediate energy neutrons.

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Kovbasenko, Yu, and Yevgen Bilodid. "Analysis of criticality of melt during severe accidents in reactor vessel." Nuclear and Radiation Safety, no. 2(78) (June 7, 2018): 3–10. http://dx.doi.org/10.32918/nrs.2018.2(78).01.

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The article investigates the possibility of a self-sustaining chain nuclear fission reaction during the development of a severe accident in the core at nuclear power plants with reactors WWER-1000 of Ukraine. Some models for calculating a criticality at different stages of the severe accident in the reactor VVER-1000 vessel were developed and calculations of multiplication properties of fuel containing masses were performed. The severe accident in the VVER-1000 core approximately divided into seven major stages: the intact reactor core, beginning of cladding damage (swelling), cladding melting and flowing down to the support grid, melting of constructional materials, homogenization of the materials at the bottom of the reactor vessel, stratification of corium at the bottom of the reactor vessel, the exit of the corium from the reactor shaft. It was shown that at the beginning of an accident, if fuel rods geometry is maintained, criticality might appear even if the emergency protection rods is triggered. With further development of the accident, the melt of fuel and structural materials will be deeply subcritical if water cannot penetrate into the pores or voids of the melt. In the case of the formation of pores or voids in the melt and the ingress of water into them, a recriticality may arise. A compensating measure is the addition of a boric acid solution to a cooling water with a certain concentration. According to the results of the computation analysis, a reactor core loaded with TVSA fuel (Russian production) requires a higher concentration of boric acid in water to compensate the multiplication properties of the fuel system in emergency situations compared to the core loaded with TVS-WR fuel (manufactured by Westinghouse), i.e. TVS-WR fuel is safer from the criticality point of view.

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Alekseev, P. N., V. V. Ignat'ev, O. E. Kolyaskin, V. I. Mostovoi, L. I. Men'shikov, V. N. Prusakov, N. N. Ponomarev-Stepnoi, et al. "Cascade subcritical enhanced-safety reactor." Atomic Energy 79, no. 5 (November 1995): 733–42. http://dx.doi.org/10.1007/bf02416363.

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Medina-Castro, Diego, Pablo L. Hernández-Adame, Consuelo Letechipía de León, Laszlo Sajo-Bohus, and Hector Rene Vega-Carrillo. "Designing a heterogeneous subcritical nuclear reactor with thorium-based fuel." Annals of Nuclear Energy 96 (October 2016): 455–58. http://dx.doi.org/10.1016/j.anucene.2016.07.002.

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Polozov, S. M., and A. D. Fertman. "High-energy proton beam accelerators for subcritical nuclear reactors." Atomic Energy 113, no. 3 (January 2013): 192–200. http://dx.doi.org/10.1007/s10512-012-9616-4.

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Bess, John Darrell, Tatiana Ivanova, and J. Blair Briggs. "Contributions to integral nuclear data in ICSBEP and IRPhEP since ND2016." EPJ Web of Conferences 239 (2020): 18007. http://dx.doi.org/10.1051/epjconf/202023918007.

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The contributions to the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) was last presented to the international nuclear data community at ND2016. Since ND2016, integral benchmark data that are available for nuclear data testing has continued to increase. The 2018 edition of the International Handbook of Evaluated Criti-cality Safety Benchmark Experiments (ICSBEP Handbook) now contains 574 evaluations with benchmark specifications for 4,916 critical, near-critical, or subcritical configurations, 45 criticality alarm placement/shielding configuration with multiple dose points apiece, and 215 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. The 2018 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) contains data from 159 different experimental series that were performed at 54 different nuclear facilities. Currently 156 of the 159 evaluations are published as approved benchmarks with the remaining three evaluations published as drafts. Measurements found in the IRPhEP Handbook include criticality, buckling and extrapolation length, spectral characteristics, reactivity effects, reactivity coefficients, kinetics, reaction-rate distributions, power distributions, isotopic compositions, and/or other miscellaneous types of measurements for various types of reactor systems. Additional benchmark evaluations will be included in the 2019 editions of these handbooks. These handbooks continue to represent the standard for neutronics benchmark experiment evaluation.

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Nailatussaadah, N., and I. Irsyad. "Neutronic Analysis of The SMART Modular Reactor Fuel Using SRAC 2006." Computational And Experimental Research In Materials And Renewable Energy 4, no. 2 (November 24, 2021): 60. http://dx.doi.org/10.19184/cerimre.v4i2.28370.

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Neutronic analysis of The SMART modular reactor fuel using SRAC 2006 has been carried out. Electrical energy is important today because the need is increasing along with the increase in human population, advanced technology and the economy. On the other hand, there are demands from the community for the clean, efficient and consistent energy. This is the reason why nuclear power plants are considered as one of the candidates for electrical energy suppliers in Indonesia in particular. This study evaluates a SMART reactor with Gadolinium as the burnable absorber material. The two kinds of fuel assembly were analyzed using the SRAC 2006 code system with the JENDL 4.0 as nuclear data library. This study aims to observe the neutronic characteristics of the fuel assembly designs according to the reference used. The results of the study show that of all types of fuel assemblies used can reach criticality at the beginning of the operating cycle and last up to 3 till 5 years when it finally reaches subcritical condition. Another parameter observed is the conversion ratio value, which from this study is in accordance with the characteristics of the conversion ratio for thermal reactors.

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Arslan, Alper Buğra, İlayda Yilmaz, Gizem Bakir, and Hüseyin Yapici. "Transmutations of Long-Lived and Medium-Lived Fission Products Extracted from CANDU and PWR Spent Fuels in an Accelerator-Driven System." Science and Technology of Nuclear Installations 2019 (October 20, 2019): 1–13. http://dx.doi.org/10.1155/2019/4930274.

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This study presents the time-dependent analyses of transmutations of long-lived fission products (LLFPs) and medium-lived fission products (MLFPs) occurring in thermal reactors in a conceptual helium gas-cooled accelerator-driven system (ADS). In accordance with this purpose, the CANDU-37 and PWR 15 × 15 spent fuels are separately considered. The ADS consists of LBE-spallation neutron target, subcritical fuel zone, and graphite reflector zone. While the considered ADS is fueled with the spent nuclear fuels extracted from each thermal reactor without the use of additional fuel, fission products extracted from same thermal reactor are also placed into transmutation zone in graphite reflector zone. The LLFP transmutation performance of the modified ADS is analyzed by considering three different spent fuels extracted from the thermal reactors. Spent fuels are extracted from CANDU-37 in case A, from PWR-15 × 15 in case B, and from CANDU-37 fueled with mixture of PWR 15 × 15 spent fuel and 46% ThO2 in case C. The LBE target is bombard with protons of 1000 MeV. The proton beam power is assumed as 20 MW, which corresponds to 1.24828·1017 protons per second. MCNPX 2.7 and CINDER 90 computer codes are used for the time-dependent burn calculations. The ADS is operated under subcritical mode until the value of keff increases to 0.984, and the maximum operation times are obtained as 3400, 3270, and 5040 days according to the spent fuel cases of A, B, and C, respectively. The calculations bring out that in the modified ADS, LLFPs and MLFPs, which are extracted from thermal reactors, can be transformed to stable isotopes in significant amounts along with energy production.

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KIM, Jaewan, Sang-In BAK, Tae-Sun PARK, and Seung-Woo HONG. "How to Burn Nuclear Wastes ? The Accelerator?driven Subcritical Thorium Reactor." Physics and High Technology 24, no. 3 (March 31, 2015): 36. http://dx.doi.org/10.3938/phit.24.014.

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Korbut, T. N., M. V. Bobkova, E. A. Rudak, and I. A. Zubets. "Neutron and Neutron-Breeding Medium Interaction Process Description Within the Physical Birth-and-Death Model." Nonlinear Phenomena in Complex Systems 23, no. 4 (December 4, 2020): 428–34. http://dx.doi.org/10.33581/1561-4085-2020-23-4-428-434.

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Analytic methods for nuclear reactor physics problems are still of current interest. They allow the physical interpretation to be obtained for studied processes within simple mathematical apparatus. This work proposes a new approach of neutron and neutronbreeding medium interaction process description based on birth-and-death model. This approach was named the physical birth-and-death model. The equations for the main kinetic characteristic are presented and reactivity values are estimated for two subcritical nuclear assemblies: MASURCA and KUCA.

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Shen, Yaosong. "Burning high-level TRU waste in fusion fission reactors." International Journal of Modern Physics: Conference Series 44 (January 2016): 1660227. http://dx.doi.org/10.1142/s2010194516602271.

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Recently, the concept of actinide burning instead of a once-through fuel cycle for disposing spent nuclear fuel seems to get much more attention. A new method of burning high-level transuranic (TRU) waste combined with Thorium–Uranium (Th–U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper. The thorium-based TRU fuel burns all of the long-lived actinides via a hard neutron spectrum while outputting power. A one-dimensional model of the reactor concept was built by means of the ONESN_BURN code with new data libraries. The numerical results included actinide radioactivity, biological hazard potential, and much higher burnup rate of high-level transuranic waste. The comparison of the fusion–fission reactor with the thermal reactor shows that the harder neutron spectrum is more efficient than the soft. The Th–U cycle produces less TRU, less radiotoxicity and fewer long-lived actinides. The Th–U cycle provides breeding of [Formula: see text]U with a long operation time (>20 years), hence significantly reducing the reactivity swing while improving safety and burnup.

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Pesic, Milan. "A new approach on modeling of the b-viii, the ultimate achievement of the second “Uranverain”." Nuclear Technology and Radiation Protection 33, no. 1 (2018): 1–23. http://dx.doi.org/10.2298/ntrp1801001p.

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German Nazi state conducted researches in nuclear technologies as an attempt to achieve various military goals. As the result of these researches, German scientists developed different, advanced nuclear technologies in years before and during World War II. In an attempt to develop the ?Uranmaschinen?, in which controlled release of high energy in fission process can be achieved, various approaches were examined, theoretically and experimentally. These studies were conducted under support of the German Nazi state and were known as the First and Second ?Uranverain? (Uranium Society/Club). Versions of the ?Uranmaschinen? were based, mainly, on natural uranium fuel and moderators of heavy water, regular water or paraffin. The latest known fission device was the subcritical nuclear fission reactor B-VIII, re-built in village Haigerloch, Bavaria, Southern Germany, in first months of 1945. It was a tank type device with natural uranium metal fuel and heavy water moderator, reflected by graphite. Radiation shielding of the device was achieved, primarily, by surrounding the reactor tank by regular water. The whole device construction was assembled inside a concrete hole in the floor of an underground cave, ex beer cellar. A recent neutronics study of this reactor was done, assuming fuel rods with lumped parameters approximation, by Italian Bologna University LIN (Laboratorio Ingegneria Nucleare) research group in 2009. This paper is a new approach to the neutronics study of the B-VIII reactor with an attempt to model real fuel-moderator geometry. This study points out many approximations and simplifications, made during the B-VIII material composition and geometry modeling, due to missing data. The paper investigates the influence to criticality of numerous uncertainties in the material compositions, mass densities and geometry of the facility. The Monte Carlo MCNP6.1 code with the latest ACE type neutron nuclear cross section data is used for that purpose. Additionally, an attempt of estimation of the uncertainty of the experimental result of the neutron multiplication was given. Differences in the calculated values of the neutron multiplication and the experimental one are investigated and tried to explain. These analyses show that the B-VIII was a subcritical device, as it was shown by the experimental results of the German scientists achieved in March-April 1945 in Haigerloch. This article has been corrected. Link to the correction <a href="http://dx.doi.org/10.2298/NTRP1802230E">10.2298/NTRP1802230E</a>

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Bowen, Douglas G., and Travis M. Greene. "VERIFICATION OF SUBCRITICAL LIMITS IN ANSI/ANS-8.1-2014." EPJ Web of Conferences 247 (2021): 17005. http://dx.doi.org/10.1051/epjconf/202124717005.

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The ANSI/ANS-8.1 standard, “Safety Standard for Operations with Fissionable Materials Outside Reactors,” has been available since 1964 as ASA N6.1-1964. In 1969, this standard was revised as ANSI N16.1-1969, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.” This version of the standard includes a variety of subcritical limits (SCLs) for uniform aqueous solutions and metals containing fissile nuclides for 233U, 235U, and 239Pu. Furthermore, SCLs are also included for uranium-water lattices. In the 1983 version of ANSI/ANS-8.1 (a revision of ANSI N16.1-1975), the suite of SCLs in the standard expanded to include 235U enrichment limits for homogeneous uranium-water mixtures and dry/damp oxides, uniform aqueous solutions of low-enriched uranium, and uniform aqueous mixtures of Pu(NO3)4 containing 240Pu, in addition to the SCLs included in ANSI N16.1-1969. The SCLs have changed little in subsequent revisions (ANSI/ANS-8.1-1998 and ANSI/ANS-8.1-2014). The ANSI/ANS-8.1-2014 standard is currently being revised to include new SCLs (uranium metal and compounds with enrichments up to 20 wt. % 235U) and possible updates to the current SCLs already in the standard, although these SCLs will not be available to the nuclear criticality safety community for a number of years. The bases for these SCLs were documented in journal articles such as Nuclear Science and Engineering, and the American Nuclear Society’s meeting transactions; however, the bases were ambiguous enough that sites and regulators in the United States are reluctant to endorse them for safety purposes. The purpose of this paper is to present the results of a comparison study for the SCLs in the ANSI/ANS-8.1-2014 standard using modern codes and cross sections (SCALE/ENDF/B-VIII) to provide some assurance about their quality (bias and bias uncertainty) for use in nuclear criticality safety applications.

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Lafuente, A., and M. Piera. "Nuclear fission sustainability with subcritical reactors driven by external neutron sources." Annals of Nuclear Energy 38, no. 4 (April 2011): 910–15. http://dx.doi.org/10.1016/j.anucene.2010.11.024.

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King, Jeffrey C., and Mohamed S. El-Genk. "Submersion-Subcritical Safe Space (S4) reactor." Nuclear Engineering and Design 236, no. 17 (September 2006): 1759–77. http://dx.doi.org/10.1016/j.nucengdes.2005.12.010.

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Bznuni, S. A., V. M. Zhamkochyan, and A. G. Khudaverdyan. "Parameters of two-reactor subcritical accelerator-controlled systems." Atomic Energy 88, no. 4 (April 2000): 331–35. http://dx.doi.org/10.1007/bf02673622.

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Nifenecker, H., S. David, J. M. Loiseaux, and O. Meplan. "Basics of accelerator driven subcritical reactors." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 463, no. 3 (May 2001): 428–67. http://dx.doi.org/10.1016/s0168-9002(01)00160-7.

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31

Zhang, Tao, Zhiyuan Xing, Ling Zhang, and Yuncan Zhang. "Recycling of resin cured IIR-based ground bladder rubber with the assistance of subcritical fluids." Journal of Elastomers & Plastics 50, no. 8 (February 22, 2018): 677–96. http://dx.doi.org/10.1177/0095244318757836.

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In this work, the devulcanization reaction of isobutylene–isoprene rubber (IIR)-based ground bladder rubber (GBR) in GBR/ethylene–propylene–diene monomer (EPDM) blend was investigated through a co-rotating twin-screw extruder. The influences of subcritical fluids (blank sample, water, ethanol, and n-propanol) and temperatures (160°C, 180°C, and 200°C) were investigated. The results confirmed the effectiveness of subcritical fluids in decreasing the gel content. Moreover, gel permeation chromatography analysis demonstrates that the devulcanizing processes with subcritical fluids are more homogeneous, making the molecular weight of sol detached from the devulcanized blend more uniform. Proton nuclear magnetic resonance spectra confirm that the reactivity of devulcanization of subcritical ethanol was the best. The optimal extrusion temperature for devulcanization is 180°C, at which the mechanical properties of the revulcanized IIR/(devulcanized ground bladder rubber [DGBR]/EPDM) blends achieve the best state. When promoting agent alkylphenol polysulfide (450) works with the assistance of subcritical ethanol at the best reaction condition (180°C, 2.0 MPa, and 500 r min−1), the tensile strength and elongation at break of the revulcanizate retain 94.7% and 110.2% of vulcanized IIR (15.91 MPa, 483.62%), respectively.

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Stacey, W. M., Z. Abbasi, C. J. Boyd, A. H. Bridges, E. A. Burgett, M. W. Cymbor, S. W. Fowler, et al. "A Subcritical, Helium-Cooled Fast Reactor for the Transmutation of Spent Nuclear Fuel." Nuclear Technology 156, no. 1 (October 2006): 99–123. http://dx.doi.org/10.13182/nt06-a3777.

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Iwamoto, Hiroki, Alexey Stakovskiy, Luca Fiorito, and Gert Van den Eynde. "Sensitivity and uncertainty analysis of βeff for MYRRHA using a Monte Carlo technique." EPJ Nuclear Sciences & Technologies 4 (2018): 42. http://dx.doi.org/10.1051/epjn/2018023.

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This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction βeff for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo N-Particle transport code MCNP. The βeff sensitivities are calculated by the modified k-ratio method proposed by Chiba. Comparing the βeff sensitivities obtained with different scaling factors a introduced by Chiba shows that a value of a = 20 is the most suitable for the uncertainty quantification of βeff. Using the calculated βeff sensitivities and the JENDL-4.0u covariance data, the βeff uncertainties for the critical and subcritical cores are determined to be 2.2 ± 0.2% and 2.0 ± 0.2%, respectively, which are dominated by delayed neutron yield of 239Pu and 238U.

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34

Blanovsky, Anatoly. "ICONE11-36579 PROSPECTS FOR NUCLEAR WASTE TRANSMUTATION USING HIGH FLUX SUBCRITICAL REACTORS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003 (2003): 204. http://dx.doi.org/10.1299/jsmeicone.2003.204.

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35

Dyachenko, Petr P., Anatoly V. Zrodnikov, Oleg F. Kukharchuk, and Alexey A. Suvorov. "Problem of nuclear-laser power engineering and methods of their solution." Nuclear Energy and Technology 5, no. 3 (September 25, 2019): 257–63. http://dx.doi.org/10.3897/nucet.5.46381.

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The concept of a high power reactor-laser system based on a nuclear pumped optical quantum amplifier (OKUYaN) was formulated at IPPE in the mid-1980-ies. The idea amounted to the use of wide-aperture OKUYaN as an amplifier within the already well-known “master laser – two-pass amplifier with phase conjugation” scheme. The structure of such an amplifier includes a system of two neutron-coupled units – an ignition reactor (RB) and a nuclear pumped laser amplifier (LB). The ignition unit is a compact multi-core pulsed fast neutron reactor. The laser amplifier unit operates on thermal neutrons and, with regard to the neutronics, it is a subcritical booster zone of the ignition reactor unit. Unique reactor-laser complex incorporating demonstration sample of a pulsed reactor-laser system based on OKUYaN (test facility “Stand B”) having no analogues anywhere in the world, was developed and put into operation at IPPE in 1999 for the purpose of substantiation of basic principles of the OKUYaN concept and demonstration of the possibility of its practical implementation, as well as verification of calculation codes and development of relevant equipment elements. Problems overcome in the development and construction of “Stand B” test facility, the choice and justification of the neutronics and laser characteristics of the OKUYaN demonstration sample are discussed in the present paper. Provided are the results of a detailed computational-experimental study of the demonstration sample characteristics, the data from systems studies of direct conversion of nuclear fission energy into laser radiation energy in complex reactor-laser devices and the results of examination of prospects for the development of nuclear-laser power engineering.

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36

Bakir, Gizem, and Huseyin Yapici. "Analysis of fuel rejuvenation times in a fusion breeder reactor fuelled with a mixture of uranium-thorium oxides for the CANDU reactor." Nuclear Technology and Radiation Protection 32, no. 3 (2017): 193–203. http://dx.doi.org/10.2298/ntrp1703193b.

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This study presents the determination of fuel rejuvenation times in a D-T fusion breeder reactor fuelled with a mixture of natUO2 and ThO2 for multi-reuse of nuclear fuels in CANDU-37 reactors. To determine the effect of thorium on the fuel enrichment and rejuvenation times, neutronic analyses are performed by increasing the percentage of ThO2 in the fuel mixture from 10 to 35. The time-dependent neutronic calculations are carried out in three stages. In the first stage, which is the fuel enrichment or rejuvenation process in the fusion breeder reactor, the subcritical calculations of the fusion breeder reactor fuelled with the fuel mixtures are performed by using the MCNPX 2.7/CINDER under a fusion neutron wall loading of 1 MWm-2, corresponding to neutron flux of 4.444?1013 cm-2s-1 (energy of every fusion neutron is 14.1 MeV). In the second stage, which is the thermal reactor analysis, the fuel rods enriched at the end of the first stage are placed in the CANDU-37 reactor, and the critical calculations of this reactor are performed by using MCNPX 2.7 and MONTEBURNS codes separately. The numerical results show that the neutronic values obtained from both codes are very near each other. The third stage is the two-year cooling process of CANDU spent fuels. The values obtained by numerical calculations show that this fusion breeder reactor is self-sufficient in terms of tritium and has a high performance in terms of energy multiplication as well as fuel rejuvenation and thorium utilization.

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Edchik, I. A., T. N. Korbut, A. V. Kuzmin, S. E. Mazanik, V. P. Togushov, and M. O. Kravchenko. "Experimental methods for determining the effective neutron multiplication factor of the “Yalina-Thermal” subcritical assembly." Proceedings of the National Academy of Sciences of Belarus, Physical-Technical Series 65, no. 2 (July 7, 2020): 235–42. http://dx.doi.org/10.29235/1561-8358-2020-65-2-235-242.

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To study the kinetics of subcritical systems and determine the optimal conditions for the transmutation of longlived radioactive waste in the neutron spectrum of ADS-systems the “Yalina” research nuclear facility was created at Joint Institute for Power and Nuclear Research – Sosny (Minsk, Belarus). The main safety indicator of a subcritical system (active zone reactivity) was measured for a “Yalina-Thermal” assembly via three independent methods: inverse multiplication, probabilistic and impulse ones. For the inverse multiplication method, the neutron flux density was monitored during assembly loading. For a fuel load of 285 EK-10 rods the neutron multiplication was M = 22.3±0.6, and the effective neutron multiplication coefficient was keff = 0.9551± 0.0016. The probabilistic method (Feynman-alpha method), based on measuring fluctuations in the neutron density level within a system with a fission chain reaction, gave the ratio of the variance to the average counting rate value D/n = 1.779±0.005, which corresponds to keff = 0.9597 ±0.0003. The pulse method is aimed at studying the neutron flux behavior of after the neutron pulse injection into the breeding system. Measurements were held with the same setup, used in the Feynman-alpha method. The measured decay constant of instantaneous neutrons is α = –670±0.7 1/s, which corresponds to keff = 0.9560±0.0001. The effective multiplication factor keff of the subcritical assembly “Yalina-Thermal”, obtained via three different independent methods, is around average value of keff = 0.9569 ± 0.0018. The methods considered can be used for subcritical level monitoring for ADS-systems and research nuclear facilities.

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38

Degtyarev, A. M., A. K. Kalugin, O. E. Kolyaskin, A. A. Myasnikov, L. I. Ponomarev, F. I. Karmanov, M. B. Seregin, and S. F. Sidorkin. "Cascade subcritical liquid-salt reactor for burning transplutonium actinides." Atomic Energy 101, no. 2 (August 2006): 569–77. http://dx.doi.org/10.1007/s10512-006-0132-2.

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39

Kuzina, Ju, M. Arnoldov, Yu Orlov, and A. Sorokin. "THERMOPHYSICAL INVESTIGATIONS: FROM THE FIRST TO STAND LARGE-SCALE NUCLEAR ENERGY." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 2 (June 26, 2021): 236–55. http://dx.doi.org/10.55176/2414-1038-2021-2-236-255.

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The article presents the main research results of thermal physicists of the IPPE from its inception to the present time. Research results in the areas of heat and mass transfer and hydrodynamics of coolants (liquid metals, water), physical chemistry and technology of liquid metal coolants for nuclear power plants for various purposes (nuclear power plants, nuclear submarines, space nuclear power plants), development codes, innovative projects, non-nuclear technologies for the use of liquid metals, heat pipes, analysis and generalization of thermophysical data are considered in the article. As a result of a large complex of experimental and computational studies, the fundamental physicochemical and thermohydraulic regularities of the coolant - impurities - structural materials - protective gas have been studied, scientific foundations have been created for the use of liquid metal coolants in nuclear power. Studies have been carried out to substantiate the technical and economic characteristics of nuclear fuel for operating, under construction and future NPPs of VVER RP, design solutions for passive safety, technical solutions and hydrogen safety devices, heat removal from the reactor through a steam generator and PHRS in case of beyond design basis accidents. As well as design solutions and safety for NPP designs with BN-1200 reactor with sodium coolant, BREST-OD-300 reactor with lead coolant, SVBR-100 reactor with lead-bismuth alloy, MBIR research reactor. The results of these studies made it possible, together with institutes and design organizations, to scientifically substantiate thermal-hydraulic parameters and highly efficient technological processes, develop and practically implement devices and systems that ensure the successful operation of fundamentally new nuclear power plants cooled by water and liquid metals, with original scientific and technical solutions that had no analogue in world practice. R&D works were carried out to substantiate the innovative project VVER with supercritical pressure, the concept of an electro-nuclear subcritical blanket based on the modular principle of constructing an core with liquid-salt melts of fissile materials, studies of thermal hydraulics, mass transfer of high-temperature sodium and the development of a combined coolant purification system to justify the BN-HT reactor with temperature sodium ~900 °C for hydrogen production. The directions of investigations at the present stage are discussed.

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40

Difilippo, F. C. "Stochastic Processes in a Subcritical Nuclear Reactor in the Presence of a Fission Source." Nuclear Science and Engineering 90, no. 1 (May 1985): 13–18. http://dx.doi.org/10.13182/nse85-a17426.

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41

Gonnelli, E., S. M. Lee, L. N. Pinto, H. R. Landim, R. Diniz, R. Jerez, and A. dos Santos. "An alternative experimental approach for subcritical configurations of the IPEN/MB-01 nuclear reactor." Journal of Physics: Conference Series 630 (July 15, 2015): 012007. http://dx.doi.org/10.1088/1742-6596/630/1/012007.

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42

Gnanapragasam, N., D. Ryland, and S. Suppiah. "Hydrogen Co-Production From Subcritical Water-Cooled Nuclear Power Plants In Canada." AECL Nuclear Review 2, no. 1 (June 1, 2013): 49–60. http://dx.doi.org/10.12943/anr.2013.00006.

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Subcritical water-cooled nuclear reactors (Sub-WCR) operate in several countries including Canada providing electricity to the civilian population. The high-temperature-steam-electrolysis process (HTSEP) is a feasible and laboratory-demonstrated large-scale hydrogen-production process. The thermal and electrical integration of the HTSEP with Sub-WCR-based nuclear-power plants (NPPs) is compared for best integration point, HTSEP operating condition and hydrogen production rate based on thermal energy efficiency. Analysis on integrated thermal efficiency suggests that the Sub-WCR NPP is ideal for hydrogen co-production with a combined efficiency of 36%. HTSEP operation analysis suggests that higher product hydrogen pressure reduces hydrogen and integrated efficiencies. The best integration point for the HTSEP with Sub-WCR NPP is upstream of the high-pressure turbine.

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43

Arkani, Mohammad. "Diagnostic methods applied to Esfahan light water subcritical reactor (ELWSCR)." Nuclear Engineering and Technology 53, no. 7 (July 2021): 2133–50. http://dx.doi.org/10.1016/j.net.2021.01.023.

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44

Frieß, Friederike, Wolfgang Liebert, and Nikolaus Müllner. "Assessment of Partitioning and Transmutation of High-Level Waste and Hypothetical Implementation Scenarios in Germany." Safety of Nuclear Waste Disposal 1 (November 10, 2021): 261–62. http://dx.doi.org/10.5194/sand-1-261-2021.

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Abstract. In the context of the search for a deep geological repository for high-level radioactive waste from nuclear energy a preliminary waste treatment is repeatedly called into play by partitioning and transmutation (P&T). Proponents of this approach promise that with P&T, the requirements for and the risks posed by a – then still necessary – repository could be significantly reduced. However, such technological promises have to be prospectively, promptly and publicly reasonably verifiable. Partitioning is reprocessing in which, in addition to separating uranium and plutonium from the fission products, other material streams (for example, the minor actinides) are extracted. In transmutation, radionuclides – especially through nuclear fission – are converted into other nuclides. Thus, conversion of the parent nuclides into nuclides with shorter half-lives, lower radiotoxicity, or into stable nuclides could be achieved. For the assessment of P&T, essential aspects are the current degree of maturity of necessary technologies, the requirements for research and development, technological development risks, the basic feasibility and objective, risks of a hypothetical operation of corresponding plants and the possible effects on nuclear waste disposal. More specifically, on the technological side, it is all about development periods, technical security requirements and licensability, proliferation risks and implementation periods. The presentation of the results of some hypothetical P&T scenarios is intended to help to assess the impacts on radioactive waste present in Germany, necessary facilities and operating periods. Thus, pyro-chemical and hydrochemical separation processes, special transuranic fuels based on mixed oxides (MOX) or uranium-free fuel types and critical fast reactors, subcritical (accelerator-driven) reactors, as well as molten salt reactors, are considered. One difficulty is that the multiple recycling of the transuranics changes the fuel composition. Detailed statements about these changes are only possible with complex simulation calculations and their influence on safe reactor operation. So far, this has not happened on an international scale. In the modelling presented here, an attempt was made to represent the restrictions that the reactor design has on the fuel composition more precisely, at least insofar as the element composition of the fuel remains the same for the duration of the scenario. Conclusions presented from the analysis of the hypothetical scenarios affect, among other things, necessary operating periods and the number of plants and changes achieved in the stock of both transuranics and fission products.

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Sumner, T. S., W. M. Stacey, and S. M. Ghiaasiaan. "Dynamic Safety Analysis of the SABR Subcritical Transmutation Reactor Concept." Nuclear Technology 171, no. 2 (August 2010): 123–35. http://dx.doi.org/10.13182/nt10-a10777.

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46

Sommer, C. M., W. M. Stacey, and B. Petrovic. "Fuel Cycle Analysis of the SABR Subcritical Transmutation Reactor Concept." Nuclear Technology 172, no. 1 (October 2010): 48–59. http://dx.doi.org/10.13182/nt10-a10881.

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47

Tran Minh, Tien, and Dung Tran Quoc. "Calculation of the Neutron Parameters for Accelerator-Driven Subcritical Reactors." Science and Technology of Nuclear Installations 2021 (December 20, 2021): 1–6. http://dx.doi.org/10.1155/2021/5284580.

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In this paper, the accelerator-driven subcritical reactor (ADSR) is simulated based on structure of the TRIGA-Mark II reactor. A proton beam is accelerated and interacts on the lead target. Two cases of using lead are considered here: firstly, solid lead is referred to as spallation neutron target and water as the coolant; secondly, molten lead is considered both as a target and as a coolant. The proton beam in the energy range from 115 MeV to 2000 MeV interacts with the lead to create neutrons. The neutron parameters as neutron yield Yn/p, neutron multiplication factor k, the radial and axial distributions of the neutron flux in the core have been calculated by using MCNPX program. The results show that the neutron yield increases as the energies of the proton beam increases. When using the lead target, the differences between the neutron yield are from 4.2% to 14.2% depending on the energies of the proton beam. The proportion of uranium in the mixtures should be around 24% to produce an effective neutron multiplier factor greater than 0.9. The neutron fluxes are much higher than the same calculations for the TRIGA-Mark II reactor model using tungsten target and light water coolant.

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Wan, J. S., R. Brandt, A. N. Sosnin, and M. I. Krivopustov. "Subcritical nuclear systems and their stability against changes in the geometrical set-up." Kerntechnik 66, no. 1-2 (January 1, 2001): 54–58. http://dx.doi.org/10.1515/kern-2001-0014.

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Abstract A specific subcritical reactor system has been suggested and the stability of its parameters against small changes in the geometrical set-up has been estimated using the model calculation code DCM/CEM from Dubna. The dimensions of the system are 200 cm in diameter and 170 cm in length and it supplies a thermal power of 900 MW with keff =0.943, using a 20 mA proton beam of an energy of 1 GeV. Considering the thermal → electricity power and electricity → beam power conversion efficiencies, the electric power amplification is about 8. Energy deposition and neutron energy distribution in the fission core are also studied. Some properties, such as the heat production per unit volume, are rather similar to modern fast breeders. The neutron multiplication factor keff is very sensitive to small changes in the geometrical set-up within the fission core.

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Gandini, A. "On the multiplication factor and reactivity definitions for subcritical reactor systems." Annals of Nuclear Energy 29, no. 6 (April 2002): 645–57. http://dx.doi.org/10.1016/s0306-4549(01)00073-1.

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50

Degtyarev, A. M., A. A. Myasnikov, O. N. Saltykova, T. E. Trofimova, O. A. Seryanina, and S. F. Sidorkin. "Statistical control k eff for subcritical fused-salt transplutonium burner reactor." Atomic Energy 113, no. 4 (February 2013): 227–35. http://dx.doi.org/10.1007/s10512-013-9622-1.

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Journal articles on the topic 'Subcrital nuclear reactor'

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