PhD Student, Nuclear Engineering, North Carolina State University
Nuclear power plants produce a lot of heat which is cooled using a coolant such as water and is then used to generate electricity. Majority of commercially operating reactors around the world are cooled by water which is the older generation (Generation II/III) of Nuclear Reactors. However, my research involves development of newest generation of nuclear reactors (Generation IV) cooled by liquid metal (Lead) instead of water cooled. Generation III water cooled reactors have an extensive plant design and are constricted to being built near a water body. However, these issues can be successfully tackled by using liquid lead as coolant. Liquid lead as reactor coolant has several advantages over water such as the compact reactor design, reduced capital cost, more economical and reliable. Lead-cooled reactors also supersede the current water-cooled reactors in safety aspect as lead has a much higher melting a boiling point, thus can operate at high temperatures without the risk of boiling which is considered a safety issue in nuclear reactors. In addition, construction site has no restriction based on the coolant. It can be constructed in the dry Sahara desert or the frozen Gobi dessert. Development of Lead-Cooled Fast Reactors is a unique and innovative step in the field of Nuclear Engineering. Westinghouse Electric Co. is developing an advanced lead-cooled fast reactor with the potential to achieve best in class safety and economics performances. Commercialization of the technology will require demonstration, as a preliminary step, in an appropriately scaled demonstration reactor that will ensure best scalability and flexibility towards the commercial reactor. My research includes performing a sensitivity study on the reactor size, fuel cost and cycle length. This is will help provide the base platform form towards development of a Lead-cooled fast reactor.
Abstract: Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U–10 wt% Zr and U–10 wt% Zr–5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions.
Pub.: 08 Jun '14, Pinned: 22 Aug '17
Abstract: Publication date: December 2016 Source:Annals of Nuclear Energy, Volume 98 Author(s): R.C. Lopez-Solis, J.L. François, G.E. Bastida-Ortiz, M. Becker, V.H. Sánchez-Espinoza Sodium cooled fast reactors are one of the systems considered as candidates in the Generation IV Forum. As most of Generation-IV systems are under development, efficient and reliable computational tools are needed to obtain accurate results in a reasonable computer time. In this work, the KArlsruhe Neutronic EXtended Tool (KANEXT), which is a modular code system for deterministic reactor calculations consisting of one kernel and several modules, is introduced and compared against the Monte Carlo SERPENT code, for fuel depletion calculations of a small core of a sodium fast reactor (SFR). The sodium fast reactor particularities, important for the core analysis, are described, and the model assumptions made for both codes are explained. Selected parameters, e.g. the effective neutron multiplication factor, the isotopes inventory and axial and radial power distributions against burnup are compared. The results of these comparisons show a very good agreement between the transport P3 solution of KANEXT and SERPENT.
Pub.: 03 Aug '16, Pinned: 22 Aug '17
Abstract: Publication date: 1 December 2016 Source:Nuclear Engineering and Design, Volume 309 Author(s): Sunghwan Yun, Sang Ji Kim, Jae-Yong Lim We study the characteristics of in-vessel shielding design for small-size pool-type burner and breeder sodium-cooled fast reactors (SFRs). The breeder SFR shows a better performance as regards axial neutron shielding due to the existence of the axial blanket. The diameter of the reactor vessel with respect to the no-shield case is increased by 17% for the breeder SFR, while the diameter of the reactor vessel is increased by 23% for the burner SFR because of the shielding assemblies. However, this disadvantage of the shielding characteristics of the burner SFR is noticeably mitigated upon adopting the cylindrical shielding design concept. Although there are several other parameters that influence the size of the reactor vessel, the configuration of the in-vessel shielding structure is one of the most important parameters that determine the final size of the reactor vessel for a pool-type SFR. Therefore, the cylindrical fixed shield design concept is more efficient than the design concept utilizing replaceable shield assemblies to prevent secondary sodium activation at the intermediate heat exchangers (IHXs) for small-size pool-type burner SFRs.
Pub.: 02 Oct '16, Pinned: 22 Aug '17
Abstract: Publication date: April 2017 Source:Annals of Nuclear Energy, Volume 102 Author(s): Reuven Rachamin, Sören Kliem A comparative study has been performed to evaluate the prediction capability of the DYN3D-Serpent code system for sodium fast reactor (SFR) cores. In this study, the calculation system was tested against the BFS-73-1 and BFS-62-3A experiments conducted at the Russian Institute of Physics and Power Engineering (IPPE). These experiments were designed for full-scale modeling of SFR cores, and for validation of codes and nuclear data. The study was performed in two parts. The first part is aimed at developing and validating a 3D full-core heterogeneous model of each of the experiments using the Serpent Monte-Carlo code (MC) code. This part meant as a first step towards the use of the Serpent MC code as a tool for preparation of homogenized group constants, and as a reference solution for code-to-code comparison with the DYN3D code. The second part is devoted to the homogenized group constants generation procedure and the DYN3D steady-state calculations. This paper covers the first part of the study. The experiments were simulated using the Serpent MC code, and the basic neutronic characteristics were evaluated and compared against experimental values. The calculated results agreed well with the measured values on most of the neutronic characteristics. It suggests that the Serpent MC code can be used for the preparation of homogenized group constants and as a reference solution for code-to-code verification with the DYN3D code.
Pub.: 05 Jan '17, Pinned: 22 Aug '17
Abstract: Publication date: March 2017 Source:Nuclear Engineering and Design, Volume 313 Author(s): Lokesh Verma, Anil Kumar Sharma, K. Velusamy Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also established that a single plate core catcher can safely accommodate decay heat arising due to ∼70% of the core debris by establishing natural circulation in the lower sodium pool. The influence of heat removal rate by natural circulation on availability of the number of decay heat exchangers (DHX) dipped in the upper pool is also analyzed. It is seen that the temperatures in the inner vessel, source plate and the maximum debris temperature do not increase significantly even when the DHXs are deployed 5h after the accident, demonstrating the benefit of large thermal inertia of the pool.
Pub.: 12 Jan '17, Pinned: 22 Aug '17
Abstract: An open fuel pin failure is a breach in the fuel pin cladding that allows direct contact between the primary coolant and the nuclear fuel. In this paper we focus on the sodium-fuel interactions in a Sodium cooled Fast neutrons Reactor (SFR), reviewing the main aspects of the fuel pin failure evolution with an emphasis on the Reaction between the Oxide fuel and the Sodium (ROS). This reaction leads to the formation of an uranoplutonate phase with approximately half the density of the initial oxide. In turn this can cause significant fuel swelling and the potential further degradation of the fuel pin. The maximal fuel swelling due to the formation of the uranoplutonate can be estimated for non-irradiated fuel based on the physico-chemical properties of the pellets, as further described in this paper.
Pub.: 01 Dec '16, Pinned: 22 Aug '17
Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.
Pub.: 25 Oct '16, Pinned: 22 Aug '17
Abstract: Publication date: November 2016 Source:Progress in Nuclear Energy, Volume 93 Author(s): H. György, Sz. Czifrus The main goal of this paper is to show how thorium, as an alternative nuclear fuel, could be applied as fuel in a Generation IV reactor. The paper focuses on the multiplication factor, the produced 233U and delayed neutron fraction in infinite lattice models. For the investigations, simplified models of a fuel assembly of five design types of the six reactor concepts were elaborated. The MSR reactor type is out of scope of this paper due to the fact that it is designed for the utilization of thorium. Although the fissile isotope content was not increased to compensate the thorium caused multiplication factor decrease, the burnup calculations suggest that the designs of ESFR (European Sodium-Cooled Fast Reactor) and ELSY (European Lead-cooled System) are the most promising types according to the trend of the multiplication factor changes and the amount of produced fissionable 233U.
Pub.: 21 Sep '16, Pinned: 22 Aug '17
Abstract: Publication date: November 2016 Source:Nuclear Engineering and Design, Volume 308 Author(s): Hyeong-Yeon Lee The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.
Pub.: 08 Sep '16, Pinned: 22 Aug '17
Abstract: A concept is proposed for a fast reactor which has a potential for overcoming the safety and fuel-cycle related fears of large-scale development of nuclear power. This proposal is based on ideas advanced in 1958 at the Second Geneva Conference on World Use of Atomic Energy and is aimed at the implementation of the innovation potential of nuclear power for solving the most complex safety and fuel supply problems of reactor technology.
Pub.: 23 Nov '12, Pinned: 30 Jun '17
Abstract: The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.
Pub.: 15 Aug '09, Pinned: 30 Jun '17
Abstract: Radiation damage in structural materials for fast reactors is caused by the action of many different mechanisms, depending on the irradiation conditions, the composition and state of the material, and external factors. This damage affects considerably the physicomechanical and operating characteristics of the material and, thus, the economy of fast reactors. Experimental data and theoretical models of radiation damage make it possible to predict the basic factors which limit the efficiency of the structural material for high burnup values and also single out the basic problems in improving the core elements and materials. This applies in the first place to problems of accommodation of structural material swelling and problems of securing sufficiently high mechanical characteristics for large fluence values. Experimental data on swelling and long-term mechanical characteristics have a rather large scatter. It is necessary to understand the causes of this scatter in order to eliminate indeterminacies in design calculations and determine the conditions ensuring the greatest resistance of materials to radiation damage.The lower values of swelling and plasticity loss encourage optimism and show the potentialities of the materials already in use. However, it must be considered that most reactor data pertain to doses of 50–70 d/α, while it is necessary to know the behavior of the material characteristics for doses of up to 100–120 d/α. Therefore, in substantiating the choice of materials for fuel-element jackets and assembly casings for projected high-power reactors, it is necessary to obtain the characteristics of materials with different compositions for such doses under conditions close to the actual operating conditions.
Pub.: 01 Jul '77, Pinned: 30 Jun '17
Abstract: The principal results from studies of the state of fuel elements with nitride fuel for subsequent work on use in fast reactors are reviewed. It is shown that fuel elements with nitride fuel and helium filler can secure serviceability to burnup, determined by the onset of kernel–cladding interaction. Nitride fuel in fuel elements with a liquid-metal contact layer is a promising variant for attaining deep burnup.
Pub.: 02 Oct '12, Pinned: 30 Jun '17
Abstract: Astatine isotopes can be produced in liquid lead-bismuth eutectic targets through proton-induced double charge exchange reactions on bismuth or in secondary helium-induced interactions. Models implemented into the most common high-energy transport codes generally have difficulties to correctly estimate their production yields as was shown recently by the ISOLDE Collaboration, which measured release rates from a lead-bismuth target irradiated by 1.4 and 1 GeV protons. In this paper, we first study the capability of the new version of the Liège intranuclear cascade model, INCL4.6, coupled to the deexcitation code ABLA07 to predict the different elementary reactions involved in the production of such isotopes through a detailed comparison of the model with the available experimental data from the literature. Although a few remaining deficiencies are identified, very satisfactory results are found, thanks in particular to improvements brought recently on the treatment of low-energy helium-induced reactions. The implementation of the models into MCNPX allows identifying the respective contributions of the different possible reaction channels in the ISOLDE case. Finally, the full simulation of the ISOLDE experiment is performed, taking into account the likely rather long diffusion time from the target, and compared with the measured diffusion rates for the different astatine isotopes, at the two studied energies, 1.4 and 1 GeV. The shape of the isotopic distribution is perfectly reproduced as well as the absolute release rates, assuming in the calculation a diffusion time between 5 and 10hours. This work finally shows that our model, thanks to the attention paid to the emission of high-energy clusters and to low-energy cluster induced reactions, can be safely used within MCNPX to predict isotopes with a charge larger than that of the target by two units in spallation targets, and, probably, more generally to isotopes created in secondary reactions induced by composite particles.
Pub.: 04 Mar '13, Pinned: 30 Jun '17
Abstract: The growing number of countries wishing to use nuclear energy, and the expansion in the geography of NPPs entails the risk of nuclear weapons proliferation, given that political leaders in some countries may want to purchase or develop sensitive nuclear technologies. A certain risk of proliferation through nuclear power technologies and materials cannot be excluded altogether. In the nuclear fuel cycle there are large inventories of nuclear materials, including fissile materials, (many hundreds and thousands of tons). The problem of spent nuclear fuel with plutonium in it, especially for novice countries and countries with small nuclear power program, also increases the risk of proliferation, including the growing risk of actions on the part of subnational or terrorist organizations because of the proliferation of nuclear technologies and materials as respective protection measures are insufficient in these countries.
Pub.: 28 Feb '16, Pinned: 30 Jun '17
Abstract: In summary, we have proposed a new method for producing89Sr for medical purposes from natural yttrium according to the reaction (n, p) in fast-neutron reactors. Investigations confirm the computational parameters of the production: from 2 to 15 mCi89Sr per gram of the starting yttrium. We have shown that90Sr can be extracted and the final product with the required radionuclide purity can be obtained. Commercial production of89Sr in BR-10 and BOR-60 has now started.
Pub.: 01 May '97, Pinned: 30 Jun '17
Abstract: The stages of the development of fast reactors in the world are analyzed. It is shown that substantial progress has been made in the development and operation of sodium-cooled fast reactors and accident-free operation of the main liquid-metal equipment, equal to the performance of general industrial equipment. Ways to make nuclear-power-plant units of this type competitive are discussed.The status of the work on fast reactors with other coolants – gas, steam, and heavy metals – is briefly reviewed. The main problems which must be solved to implement these directions of the development of fast reactors are indicated.
Pub.: 01 May '04, Pinned: 30 Jun '17
Abstract: The present status of research on dense nitride fuel for liquid-metal cooled fast reactors is examined. It is noted that within the framework of project Proryv nitride fuel is to be used in reactors with sodium (BN-1200) as well as lead coolant (BREST). Carbothermal synthesis of nitride from oxide powders has been chosen to fabricate the experimental fuel elements. Several experimental assemblies with mixed uranium-plutonium nitride fuel have been fabricated for tests in BOR-60 and BN-600. It is shown that the existing global experience in studying nitride fuel is insufficient for validating such fuel for use in BN-1200 and BREST. A complex program for computational-experimental validation of fuel elements is being implemented for validating the serviceability of fuel elements.
Pub.: 30 Nov '14, Pinned: 30 Jun '17
Abstract: Substantiation is given for the development of nuclear power based on inherently-safe fast reactors with a mononitride core. Fundamental studies and design work on the development of such reactors with lead (BREST-OD-300), lead–bismuth (SVBR-75/100), and sodium coolant (BN-800) are being performed. The development of nuclear power in our country is based on organizing a closed fuel cycle. The results of experimental investigations of the properties of mononitride fuel are correlated. Mononitride fuel meets all requirements for fast-reactor fuel.
Pub.: 01 Sep '03, Pinned: 30 Jun '17
Abstract: Currently, forty-two thallium, forty-two lead, forty-one bismuth, and forty-two polonium isotopes have so far been observed; the discovery of these isotopes is discussed. For each isotope a brief summary of the first refereed publication, including the production and identification method, is presented.
Pub.: 21 Jan '12, Pinned: 30 Jun '17
Abstract: The density of molten Pb–Bi eutectic is measured by the pycnometer method in a temperature range of 410–726 K. The confidence of error of measurement, made up by the systematic and random components, does not exceed 0.1%. The experimental results are compared with the experimental data on the density of a liquid lead–bismuth alloy of eutectic composition available in the literature.
Pub.: 01 Mar '03, Pinned: 30 Jun '17
Abstract: The heat transfer characteristics between liquid lead bismuth eutectic (LBE) and helium are of great significance for the two-loop cooling system based on accelerator-driven system (ADS). This paper presents an experimental study on resistance characteristic and heat transfer performance in LBE-helium experimental loop of ADS. Pressure drops in the LBE loop and the main heat transfer (MHX), the coupled heat transfer characteristics between LBE and helium are investigated experimentally. The temperature of LBE has a significant effect on the LBE thermo-physical properties, and is therefore considered in the prediction of pressure drops. The results show that the overall heat transfer coefficient increases with the increasing helium flow rate, and the decreasing inlet temperature of helium. Increasing the LBE Reynolds number and LBE inlet temperature promotes the heat transfer performance of MHX and thus the overall heat transfer coefficient. The experimental results give insight into the flow and heat transfer properties in LBE-helium heat exchanger and are helpful for the optimization of an ADS system design.
Pub.: 10 May '16, Pinned: 30 Jun '17
Abstract: Molecular dynamics method was applied to simulate the physical properties of liquid metals: lead, bismuth and a binary alloy-- lead-bismuth eutectic (LBE). The embedded atom method (EAM), an empirical model rooted in density-functional theory, was used to represent the many-body interaction within the liquid metals. The atomic-scale interactions, structure and thermal physical properties of lead, bismuth and LBE were obtained through the simulation, and then compared to the available experimental results. The theoretical results of the physical properties calculated through the MD simulations are in good agreements with the available experimental data.
Pub.: 30 Aug '16, Pinned: 30 Jun '17
Abstract: Significant structural steels for nuclear power engineering are chromium-nickel austenitic stainless steels. The presented paper evaluates the kinetics of the fatigue crack growth of AISI 304L and AISI 316L stainless steels in air and in corrosive environments of 3.5% aqueous NaCl solution after the application of solution annealing, stabilizing annealing, and sensitization annealing. Comparisons were made between the fatigue crack growth rate after each heat treatment regime, and a comparison between the fatigue crack growth rate in both types of steels was made. For individual heat treatment regimes, the possibility of the development of intergranular corrosion was also considered. Evaluations resulted in very favourable corrosion fatigue characteristics of the 316L steel. After application of solution and stabilizing annealing at a comparable ∆K level, the fatigue crack growth rate was about one half compared to 304L steel. After sensitization annealing of 316L steel, compared to stabilizing annealing, the increase of crack growth rate during corrosion fatigue was slightly higher. The obtained results complement the existing standardized data on unconventional characteristics of 304L and 316L austenitic stainless steels.
Pub.: 16 Dec '16, Pinned: 29 Jun '17
Abstract: In the development process for the BREST-OD-300 lead-cooled fast reactor, 10Kh15N9S3B1 and 10Kh9NSMFB steels were chosen for the main equipment, the corrosion resistance and mechanical properties were investigated, and the effect of the lead coolant on the short-time mechanical properties, long-time strength, creep, low-cycle fatigue and crack kinetics under low-cycle loading were studied. 16Kh12MVSFBR steel was chosen for fuel-element cladding. The physical-mechanical, corrosion in liquid-lead flow, radiation, technological and other properties were studied. Complex-alloyed (Cr, Mo, W, Si, Al, N) steel or bimetallic (trimetallic) materials were studied as promising cladding materials.
Pub.: 07 Mar '13, Pinned: 29 Jun '17
Abstract: In nuclear systems, operators have to carry out corrective actions when abnormal situations occur. However, operators might make mistakes under pressure. In order to avoid serious consequences of the human errors, a new method for operators support based on intelligent dynamic interlock was proposed. The new method based on full digital instrumentation and control system, contains real-time alarm analysis process, decision support process and automatic safety interlock process. Once abnormal conditions occur, necessary safety interlock parameter based on analysis of real-time alarm and decision support process can be loaded into human-machine interfaces and controllers automatically, and avoid human errors effectively. Furthermore, the new method can make recommendations for further use and development of this technique in nuclear power plant or fusion research reactor.
Pub.: 16 Sep '16, Pinned: 29 Jun '17
Abstract: The most important candidate for 4th generation nuclear system is lead or lead alloy cooled fast reactor. In the design phase of a lead cooled reactor, it is a top priority to finish the analysis work for the core. The detailed sub-channel analysis code KMC-Sub (Keda multi-physics and multi-scale coupling platform) has been developed to analyze steady state thermal hydraulic issues of SNCLFR-100 (Small Modular Lead-cooled Natural Circulation Fast Reactor, which was designed by University of Science and Technology of China). The model used in this code had taken the influence of cross flow into account, both forced and natural circulation can be simulated, to assess the development status of KMC-sub, experimental data from ORNL 19 pin tests (sodium cooled) and CAS 61 rods test (lead bismuth eutectic cooled) are compared to results from the code. The author found that in most flow rate and power density regime, the results coincides well with the tests data. After the V&V work, the code was used to analyze the flow and temperature distribution in important assemblies of SNCLFR-100 core, which showed that the design is reasonable and feasible.
Pub.: 14 Feb '17, Pinned: 29 Jun '17
Abstract: Two variants of the arrangement of a fast reactor cooled by a eutectic alloy of lead and bismuth are studied. The first one is obtained by solving the problem of minimizing the void coefficient with drying of the central zone of the reactor. The second one is obtained by increasing the power by a factor of 1.5. Both problems contain a constraint for the reliability functionals, characterizing the nominal operating regime of the reactor, and the safety functionals, characterizing the intrinsic self-shielding from ATWS-type accidents.A fast reactor cooled with the eutectic alloy Pb–Bi possesses a high potential from the standpoint of increasing the power of the power-generating unit while maintaining safety at an acceptable level. 1 figure, 2 tables, 3 references.
Pub.: 01 Mar '01, Pinned: 29 Jun '17
Abstract: SNCLFR-100, a 100 MWth lead-cooled small modular reactor with a passive cooling feature to both normal and abnormal operations, was proposed by University of Science and Technology of China (USTC). The reactor is well suited as a remote power source because of its compact size, as well as because it has a refueling interval of 10 years without assembly reconfiguration. The reactor is a typical pool-type fast reactor with an array of heterogeneous square fuel assemblies loaded with MOX fuels. In this paper, the overall design and neutronics features were illustrated and evaluated. The steady state thermal-hydraulic performance, mass flow distribution characteristics and sub-channel T/H features were analyzed and discussed. Two major accident scenarios including unprotected overpower transient (UTOP) and unprotected loss of heat sink transient (ULOHS) were selected for a first evaluation of its dynamic behavior. The results show that the safety criteria are satisfied and reactor is tolerant to the UTOP and ULOHS transients. This implies that the conceptual design of SNCLFR-100 is acceptable and the reactor has excellent inherent safety characteristics.
Pub.: 26 Feb '16, Pinned: 29 Jun '17
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