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[en] Highlights: • The fuel pellet with a core-shell structure was designed by an in-situ process. • The relationship between the shell thickness and mixing parameters were studied. • An empirical formula for the shell thickness and mixing parameters was established. - Abstract: Using the segregation phenomenon of different materials in the mixing process, the dispersion nuclear fuel pellet with a core-shell structure was firstly designed and fabricated by an in-situ preparation process. This core-shell structure could prevent the metal matrix from reacting with water to generate hydrogen, which improves the safety property of the pellet. Furthermore, it provides a perspective on integrated manufacture of pellets and claddings by one-step preparation. Effects of mixing process parameters on the shell thickness were investigated in detail. The results indicate that: a three-quarter cycle sinusoid could be used to describe the change in shell thickness with rotational speed; and there is a negative linear relationship between shell thickness and volume percentage of ZrO2 microspheres; the shell thickness increases as the particle size difference of ZrO2 microspheres/glass beads increases. What’s more, an empirical formula describing the relationship between the shell thickness and these parameters has been established.
[en] Highlights: • Assembly-level analyses of a long-life marine SMR loaded with ATF is performed. • The new assembly is designed and its burn-up reaches to 95 MWD/KgU. • The new design is proved to be safe with negative FTC and MTC. • Accident Tolerant Fuel/Cladding is implemented and it performs well. - Abstract: A call for flexible power generation for a wide range of applications is gradually increasing, making Small Modular Reactor (SMR) a heat topic, especially for marine power. Unlike normal reactors, marine reactors demand a long lifetime up to 15 years without refueling, a high power-density and also a higher reliability and security. The standard PWR cannot achieve this goal. Here, a new 13 × 13 assembly loaded with Accident Tolerant Fuels (ATFs): U3Si2 – FeCrAl system, is proposed and analyzed on an assembly-level. U3Si2 is one kind of ATFs with high heavy metal density and thermal conductivity. It has a lower parasitic absorption. FeCrAl has a great radiation resistance, corrosion resistance and thermodynamic properties although with serious neutron penalties. It is currently the fastest researched accident tolerant cladding material. The fuel enrichment of this assembly is raised to 13%, prolonging the burnup to up to 95,000 MWD/tU, which means that the reactor can survive for more than 15 years at its rated power density. Using 15 B4C control rods can help achieve the beginning-of-life cool shutdown (K < 0.95). The power of the assembly is well distributed during the whole burnup. In consider of the inherent safety of reactor operation, the reactivity coefficients, especially Fuel Temperature Coefficient (FTC) and Moderator Temperature Coefficient (MTC) are computed and proved to be negative values, forming a negative feedback effect. The preliminary design is well-pleasing and it may be a better choice for future marine SMRs.
[en] Highlights: • Development of a new integral-type rack design for spent nuclear fuel pools. • Boron plates are replaced by racks made of (Gd 0.7 at% + Eu) stainless steel. • Criticality analysis is conducted with SCALE 6.2 for region I and II. • New racks have potential for increased storage capacity or reduced burnup limits. • Setting objectives for feasibility/manufacturability studies of gadolinium steels. - Abstract: This study presents the development of a new integral-type rack design, characterized by the use of gadolinium (Gd)-containing structure materials that enhances the capacity of spent fuel (SF) pool storage by exploiting the high neutron absorption capability of Gd. Appropriate types and contents of Gd-based neutron-absorbing materials are selected for the new design through parametric studies. For high-reactivity fuels (region I of SF storage pools), neutron-absorbing material composed of Gd 0.7 at% with Eu 2.73 at% is found to be an optimal neutron absorber whereas for low-reactivity fuels (region II of SF storage pools), a composition of Gd 0.7 at% with Eu 8.38 at% is found to be optimal. A criticality safety analysis shows that the newly designed racks are more subcritical than conventional racks for both regions I and II. The additional reactivity margin yielded by the new integral-type design can be used to reduce the pitch of the rack while maintaining equivalent subcriticality compared to conventional rack design. This study demonstrates the potential of Gd-based neutron absorbers in structure materials for increasing the total amount of fuel assemblies that can be stored in a SF storage pool.
[en] Highlights: • Description of safety potentialities of the submerged SMR concept. • Preliminary design and basic safety strategy of IRIS-160, an integral PW-SMR designed to operate in a submerged containment. • Numerical investigation of the long-term decay heat removal strategy in a submerged SMR with Relap5-Mod3.3. • Discussion of limitations and validation issues of the results. - Abstract: Following the Fukushima-Daiichi nuclear accident in March 2011, innovations are needed to improve the reliability of new generation nuclear power plants toward scenarios where electrical power and ultimate heat sink are lost. This paper describes an integral design and basic passive safety strategy of a Small Modular Reactor (SMR) submerged in the sea or in an artificial lake, then performing a preliminary analysis of the long-term decay heat removal. The analysis considers a pressurized reactor placed in a horizontal cylindrical hull, which is surrounded by the external water. The simulated system is based on the Flexblue concept, developed by French company DCNS (now Naval Group). The object of the investigation is the natural circulation in the submerged containment, which is the key component for the long-term cooling. Following a rupture in the primary circuit, decay heat must be removed according to a fully passive safety strategy for an indefinitely long period. The purpose of this work is to study the effectiveness of a sump natural circulation flow to cool the fuel rods, up to several days after the scram. Decay heat generates steam in the core, which is released in the containment and condensed on the metal surface, transferring the heat to the exterior. Relap5-Mod3.3 has been employed to simulate the accident scenario. Results show the consistency of the safety principles and stimulate experimental investigations. However, the sensitivity analysis identifies the nodalization of the reactor containment as a modeling and numerical issue, deserving further analyses.
[en] Highlights: • Potential advantages of 99Mo production in an Aqueous Homogeneous Reactor (AHR) are discussed. • An AHR conceptual design using low-enriched uranium for the production of 99Mo is studied. • Aspects related to the neutronic behavior of the AHR were studied using MCNPX code. • The reactivity feedback introduced by the production of radiolytic gas bubbles was analyzed. • The feasibility of 99Mo production in an Aqueous Homogeneous Reactor is discussed. - Abstract: Nowadays, 99mTc is the most common radioisotope used in nuclear medicine, with up to 30–40 million procedures worldwide every year. Furthermore, medical diagnostic imaging techniques using 99mTc represent approximately 80% of all nuclear medicine procedures. Currently, 99mTc is almost exclusively produced from the beta-decay of its 66-h parent 99Mo. The 99Mo production in an Aqueous Homogeneous Reactor (AHR) is potentially advantageous because of its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This paper studies an AHR conceptual design using low-enriched uranium for the production of 99Mo. Aspects related to the neutronic behavior such as critical height, medical isotopes production, uranium consumption, plutonium production and the reactivity feedback introduced in the solution by the volumetric expansion of the fuel solution were evaluated using the computational code MCNPX version 2.6e. In addition, important reactor kinetic parameters such as the effective delayed neutron fraction, βeff, and mean neutron generation time, Λ, were calculated. A benchmarking exercise was solved using available results of critical experiments performed at the Russian Research Center “Kurchatov Institute”. The neutronic calculations demonstrated that the reactor is able to produce 99 six-day Ci of 99Mo in operation cycles of five days. The reactivity feedbacks introduced by the volumetric expansion of the fuel solution is at about −1460 pcm, which represents approximately the 48% of the planned initial reactivity reserve in the core. The calculated effective delayed neutron fraction and the average mean neutron generation time were 785 pcm and 134.08 µs, respectively.
[en] Highlights: • An open specification for system code modeling was developed for the MSRE based on the literature. • An example solution using the NEAMS tool SAM was developed and compared to RELAP5-3D results. • Sensitivity studies to the molten salt equations of state were performed with and without actinides in the salt. - Abstract: System analysis codes have a long history of providing best-estimate and conservative safety analysis for both light water and advanced reactor technologies, including molten salt reactors. As interest continues to expand with advanced reactor concepts, system analysis codes will need revisions to accommodate the behavior of these technologies. Legacy system analysis codes will need to be updated to the latest numerical techniques to shorten execution time and increase the accuracy of results. One example of a modern system analysis code that already encompasses these characteristics is the System Analysis Module (SAM). One key objective of this paper was to review available information for system code modeling of the Molten Salt Reactor Experiment (MSRE) from sources in the open literature and collect the information from these open sources in one place for the first time. This supports the potential objective of developing an open specification for system code analysis for MSRE steady state and transients with and without reactor kinetics. Data from actual MSRE tests will serve as the basis for code-to-code comparison exercises, including the MSRE zero power physics tests, the fuel pump start-up and coast down tests, and the natural circulation transient. The objective is to produce a code-to-code benchmark with a standardized set of comparison problems, recognizing the limitations of the original data. To demonstrate an initial application of this objective and the usefulness of compiling this open data, two Molten Salt Reactor Experiment (MSRE)-related models were developed to evaluate SAM for liquid fueled molten salt reactors. One model was the SAM MSRE hydraulic mockup, which provided experimental data for pressure drop measurements. The second model was the complete MSRE primary loop. The MSRE primary loop model incorporated a fluoride salt fuel/coolant with heat transfer in both the core and heat exchanger. For both the hydraulic mockup and MSRE primary loop models, a holistic 1-D system description was built using open documentation, an open description that can be readily modified and applied for any system analysis code. SAM results for the pressure drop of the hydraulic mockup model were within 6% with measurements. Coolant temperatures for the primary loop model matched the expected axial change in temperature from historical calculations. Using alternative coolant properties obtained from the literature, corresponding to salts with different actinide contents, returned similar trends in core temperature profiles. A thermal hydraulic demonstration of a loss-of-flow transient showed the importance of coupling SAM thermal hydraulic analysis to neutronics. This coupling is essential for simulating MSR transients with system analysis codes.
[en] Highlights: • Integral effect test to inspect the performance of passive emergency core cooling system. • The various thermal hydraulic phenomena occurring inside SIT were investigated. • The theoretical condition for injection of Hybrid SIT coolant was derived. • The appropriate operating condition of auto depressurization system was suggested. - Abstract: This paper reports an experimental research on the performance of the passive emergency core cooling system (PECCS) using an integral effect test facility in Korea Atomic Energy Research Institute (KAERI). The PECCS consists of two hybrid safety injection tanks (H-SIT), two medium pressure safety injection tanks (MP-SITs) and an automatic depressurization system consisting of four stages. In this study, an integral effect test (IET) was conducted on a loss of coolant accident (LOCA), which requires the operation of the primary makeup system. The H-SIT (SITs #1 and #3) injection starts successfully when the primary side pressure falls to 10.0 MPa, and the MP-SIT (SITs #2 and #4) injection starts when the pressure falls further to 4.21 MPa. The coolant injection of H-SIT occurs intermittently at certain points of time, rather than continuously. It can be seen that the flow generation of the H-SIT near the pressure plateau occurs exactly when the secondary side is depressurized.
[en] Highlights: • Pressure wave propagation following LOCA in piping systems is evaluated. • Laplace Transform Finite Volume, LTFV, is implemented. • The comparing results with experimental data show reliable accuracy. • The results can be extended to predict depressurization of nuclear reactors. - Abstract: The detailed dynamic modeling and the simulation of the rapid depressurization of PWR following leak or loss of coolant accident, LOCA, is a key element of the safety analysis in the nuclear power plants. Early in a LOCA, the blowdown at the break point causes the propagation of an acoustic wave through the primary circuit. The local pressure gaps due to the depressurization wave propagation may lead to the component recoils and the internal structure movements. In this paper to obtain the pressure depressurization rate behavior in hypothetical LOCA, a thermal hydraulic test loop, THTL facility, has been designed and constructed. The pressure change data are provided to model the trend of the pressure decreasing along the water hammer oscillation peaks that are produced as soon as the accident is initiated. For determination of the intensity of the water hammer pressure waves due to LOCA, the governing equation is solved using Laplace Transform Finite Volume, LTFV, method and the boundary conditions are set on using the obtained experimental data from the THTL facility. To evaluate the method, the analysis of the water hammer data which is obtained experimentally from THTL facility are compared with simulation results of LTFV technique which reveals the reliability of the method.
[en] Highlights: • Wilks’ method for setting tolerance limits is derived and verified. • Higher order Wilks analysis increases the accuracy and precision of the predicted tolerance. • In most practical applications, higher order analysis is unnecessary. • Wilks’ method is applied to the Dittus-Boelter equation. - Abstract: Wilks’ non-parametric method for setting tolerance limits using order statistics has recently become popular in the nuclear industry. The method allows analysts to predict a desired tolerance limit with some confidence that the estimate is conservative. The method is popular because it is simple and fits well into established regulatory frameworks. A critical analysis of the underlying statistics is presented in this work, including a derivation, analytical and statistical verification, and a broad discussion. Possible impacts of the underlying assumptions for application to computational tools are discussed. An in-depth discussion of the order statistic rank used in Wilks’ formula is provided, including when it might be necessary to use a higher rank estimate.
[en] The coasting behavior of the reactor coolant pump is one of the important indicators for the safe operation of the nuclear power system. In order to ensure the optimal coasting behavior of the reactor coolant pump under the power cutoff accident condition, this manuscript takes the guide vanes of AP1000 reactor coolant pump as the subject of this study and the main factors parameters of the vanes as optimization variables, and aims to reduce the flow loss and uniform the velocity distribution in the flow channel. And based on the orthogonal experimental design and grey relational analysis method, this manuscript analyzes, optimizes and verifies the correlation degree of the main factors parameters of the guide vanes of the reactor coolant pump. The results show that three factors of the vane wrap angle φ, the guide impeller clearance Rt and the outlet blade angle α4, have a great influence on the efficiency and head. The correlation degree of each factor is ranked as α4, φ, Rt, α3, δ, b4. The optimized guide vane evidently improves the efficiency and head of the pump in small flow conditions. On the basis of the unchanged or improved performance parameters corresponding to the design conditions, the high efficient zone shift to small flow, the coasting time extends, and the coasting behavior is optimized.