The Optimization Design of Lower Plenum and Distribution Plates in MSR

2013 ◽  
Author(s):  
Jianjun Zhou ◽  
Suizheng Qiu ◽  
Zhangpeng Guo ◽  
Guanghui Su
Keyword(s):  
Molten Salt ◽  
Optimization Design ◽  
Reactor Core ◽  
Fission Products ◽  
Physical Property ◽  
Maximum Temperature ◽  
Concept Design ◽  
Spent Fuel ◽  
Fuel Temperature ◽  
Cfd Method

Molten salt reactor was one of six Generation IV reactor types, which uses the liquid molten salt as the coolant and fuel solvent. In transmutation of actinides and long-lived fission products have marked advantages. As a liquid reactor the physical property and thermo-characteristic is different to solid fuel and water coolant reactors, which was influenced by many factors. MOSART was one of the advanced molten salt reactors concept design, which can burners TRU from LWR spent fuel. The reactor core does not contain graphite structure elements, so the flow pattern was potentially complex and may significantly affect the fuel temperature distributions. The optimizations of the salt flow may be needed, the present work designed three core models and three kinds of distribution plates to investigate the influence of lower plenum and distribution plates to thermohydraulics characteristics of the reactor core with CFD method use software FLUENT. Velocity field and maximum temperature of the core was simulated in each model at different mass flow rate.

Preliminary LOFC Analysis of a TMSR-SF1 Concept Design Using CFD Method

2017 ◽  
Author(s):  
Linsen Li ◽  
Kai Xie ◽  
Feng Shen ◽  
Qiming Li ◽  
Naxiu Wang
Keyword(s):  
Molten Salt ◽  
Cooling System ◽  
Reactor Core ◽  
Air Cooling ◽  
Concept Design ◽  
Preliminary Calculation ◽  
Calculation Results ◽  
Air Cooling System ◽  
Cfd Method ◽  
Residual Heat

The thorium molten salt experimental reactor with solid fuels (TMSR-SF1) was one of the conceptual designs developed by Shanghai Institute of Applied Physics, Chinese Academy of Sciences (SINAP, CAS). The fuel pebbles at the reactor core and the pool type structure of the vessel increase the complexity of thermal-hydraulic (T-H) analysis of the reactor. In order to analysis the T-H feature of TMSR-SF1 in case of the postulated Loss Of Forced Cooling (LOFC) accident, and investigate whether its external air cooling system with nature convection is capable of removing the residual heat, the Computational Fluid Dynamics (CFD) method was used to model the reactor and simulate the transient. In this research, an integrated pseudo 2-D thermal-hydraulic model of the core was developed and a simulation and analysis of the LOFC accident has been conducted. The preliminary calculation results using CFD method show that the external air cooling system has the capability of removing the residual heat. The calculation results also indicate that the peak temperatures of the fuel pebbles, key components and structures of TMSR-SF1 remain under the safety limits and the temperature of the molten salt remains below boiling point.

scholarly journals Nuclear policy-nuclear fuels for peaceful, safer co-generation and environmentally benign applications

2020 ◽  
Vol 9 (3) ◽  
pp. 724
Author(s):  
Syazwani Mohd Fadzil ◽  
Shafi Qureshi ◽  
Sekhar Basu ◽  
K. Kasturirangan ◽  
Anil Kakodkar ◽  
...  
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Reactor Core ◽  
Negative Temperature ◽  
Hydrogen Gas ◽  
Hydraulic Diameter ◽  
Nuclear Fuels ◽  
Spent Fuel ◽  
Fuel Temperature ◽  
Nuclear Policy

Here, safer nuclear fuels which can sustain in the high temperature and fluence environment of the reactor core are investigated to utilize nuclear energy peacefully. At Nuclear Fuel Complex in Hyderabad, nuclear fuels are being manufactured which are best suited for high temperature and fluence environment of the reactor core even in accidental scenarios. In this paper, nuclear fuels manufactured at NFC, Hyderabad are presented. The developed nuclear fuels have higher equivalent hydraulic diameter and breeding capability to produce U^233. Nuclear fuels having higher equivalent hydraulic diameter reduce the reactor core temperature substantially. These fuels have negative temperature coefficient of reactivity. Thus, in case of an accident, the fuel temperature never exceeds the safety limit. Therefore, the thermal heat available across the secondary of a heat exchanger can be utilized for different industrial processes. This allows the development of key technologies, such as safer co-generation of electricity and Hydrogen. The Three-Stage Indian Nuclear Power Program has been explained for nuclear disarmament. The product Hydrogen gas has been utilized in many ways for different applications. Moreover, the processing of iron ore with the energy obtained from the IHX secondary side, eliminates the burning of coals and CO2 emissions into the environment. Several radioisotopes have been developed for medical applications from spent fuel.  

Transport of High-Temperature Molten Salt Slurry for Pyro-Reprocessing

2009 ◽  
Author(s):  
Takatoshi Hijikata ◽  
Tadafumi Koyama
Keyword(s):  
Molten Salt ◽  
Fuel Cycle ◽  
Fission Products ◽  
Spent Fuel ◽  
Supply Tank ◽  
Transport Studies ◽  
Advanced Fuel ◽  
Inner Diameter ◽  
Porous Anode ◽  
Transport Tube

Pyro-reprocessing is one of the most promising technologies for advanced fuel cycle with favorable economic potential and intrinsic proliferation resistance. The development of transport technology for molten salt is a key issue in the industrialization of pyro-reprocessing. As for pure molten LiCl-KCl eutectic salt at approximately 773 K, we have already reported the successful results of transport using gravity and a centrifugal pump. However, molten salt in an electrorefiner mixes with insoluble fines when spent fuel is dissolved in porous anode basket. The insoluble consists of noble metal fission products, such as Pd, Ru, Mo, and Zr. There have been very few transport studies of a molten salt slurry (metal fines - molten salt mixture). Hence, transport experiments on a molten salt slurry were carried out to investigate the behavior of the slurry in a tube. The apparatus used in the transport experiments on a molten salt slurry consisted of a supply tank, a 10° inclined transport tube (10 mm inner diameter), a valve, a filter, and a recovery tank. Stainless steel (SS) fines with diameters from 53 to 415 μm were used. To disperse these fines homogenously, the molten salt and fines were stirred in the supply tank by an impeller at speeds from 1200 to 2100 rpm. The molten salt slurry containing 0.2 to 0.4 vol.% SS fines was transported from the supply tank to the recovery tank through the transportation tube. In the recovery tank, the fines were separated from the molten salt by the filter to measure the transport behavior of molten salt and SS fines. When the velocity of the slurry was 0.02 m/s, only 1% of the fines were transported to the recovery tank. On the other hand, most of the fines were transported when the velocity of the slurry was more than 0.6 m/s. Consequently, the molten salt slurry can be transported when the velocity is more than 0.6 m/s.

Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Yuiko Motome ◽  
Yoshiya Akiyama ◽  
Hiroyuki Murao
Keyword(s):  
Radiation Effects ◽  
Research Reactor ◽  
Reactor Core ◽  
Fission Products ◽  
Internal Exposure ◽  
Spent Fuel ◽  
Loss Of Function ◽  
Safety Research ◽  
Spent Fuel Storage ◽  
Fuel Storage

Abstract The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics—annular core pulse reactor (TRIGA-ACPR) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident (RIA) conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. The graded approach is introduced in the regulation. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. There is fuel in the reactor core, fresh fuel storage, and spent fuel storage. As the effects from reactor core, we evaluate the external exposure to radiation and exposure from the release of fission products assuming that loss of function to shut down the reactor, break of cladding tubes, loss of reactor pool water, and collapse of the reactor building. As the effects from fresh fuel storage, we evaluate the internal exposure assuming that the fresh fuel particles released into the air because of breaking into pieces. In addition, we evaluate the critical safety assuming that the critical safety shapes of the fresh fuel storage are lost. As the effects from spent fuel storage, we evaluate the critical safety assuming that the critical safety shapes of the spent fuel storage are lost. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions caused by the tornado, earthquake, and the accompanying tsunami.

scholarly journals An Eulerian Single-Phase Transport Model for Solid Fission Products in the Molten Salt Fast Reactor: Development of an Analytical Solution for Verification Purposes

2021 ◽  
Vol 9 ◽  
Author(s):  
Andrea Di Ronco ◽  
Stefano Lorenzi ◽  
Francesca Giacobbo ◽  
Antonio Cammi
Keyword(s):  
Boundary Conditions ◽  
Molten Salt ◽  
Fast Reactor ◽  
Liquid Fuel ◽  
Mixed Type ◽  
Single Phase ◽  
Transport Model ◽  
Reactor Core ◽  
Fission Products ◽  
Phase Transport

Nuclear reactor modeling has been shifting, over the last decades, towards full-core multiphysics analysis due to the ever-increasing safety requirements and complexity of the designs of innovative systems. This is particularly true for liquid-fuel reactor concepts such as the Molten Salt Fast Reactor (MSFR), given their strong intrinsic coupling between thermal-hydraulics, neutronics and fuel chemistry. In the MSFR, fission products (FPs) are originated within the liquid fuel and are carried by the fuel flow all over the reactor core and through pumping and heat exchange systems. Some of FP species, in the form of solid precipitates, can represent a major design and safety challenge, e.g., due to deposition on solid boundaries, and their distribution in the core is relevant to the design and safety analysis of the reactor. In this regard it is essential, both for the design and the safety assessment of the reactor, the capability to model the transport of solid FPs and their deposition to the boundary (e.g., wall or heat exchanger structures). To this aim, in this study, models of transport of solid FPs in the MSFR are developed and verified. An Eulerian single-phase transport model is developed and integrated in a consolidated multiphysics model of the MSFR based on the open-source CFD library OpenFOAM. In particular, general mixed-type deposition boundary conditions are considered, to possibly describe different kinds of particle-wall interaction mechanisms. For verification purposes, analytical solutions for simple case studies are derived ad hoc based on the extension of the classic Graetz problem to linear decay, distributed source terms and mixed-type boundary conditions. The results show excellent agreement between the two models, and highlight the effects of decay and deposition phenomena of various intensity. The resulting approach constitutes a computationally efficient tool to extend the capabilities of CFD-based multiphysics MSFR calculations towards the simulation of solid fission products transport.

scholarly journals SIMULATION OF BREED AND BURN FUEL CYCLE OPERATION OF MOLTEN SALT REACTOR IN BATCH-WISE REFUELING MODE

2021 ◽  
Vol 247 ◽  
pp. 13003
Author(s):  
Valeria Raffuzzi ◽  
Jiri Krepel
Keyword(s):  
Molten Salt ◽  
Fuel Cycle ◽  
Fission Products ◽  
Molten Salt Reactor ◽  
Core Size ◽  
Spent Fuel ◽  
Fluid Core ◽  
Active Core ◽  
Breed And Burn ◽  
Gen Iv

The Molten Salt Reactor (MSR) is one of the most revolutionary Gen-IV reactors and it can be operated, especially with chloride salts, in the so-called breed and burn fuel cycle. In this type of fuel cycle the fissile isotopes from spent fuel do not need to be reprocessed, because the excess bred fuel covers the losses. The liquid phase of the MSR fuel assures its instant homogenization, and the reactor can be operated with batch-wise refueling thus reaching an equilibrium state. At the same time, the active core of the chloride fast MSR needs to be bulky to limit neutron leakage. In this study, the code Serpent 2 was coupled to the Python script BBP to simulate batch-wise operation of the breed and burn MSR fuel cycle. The script, previously developed for solid assemblies shuffling, was modified to simulate fuel homogenization after fertile material addition. Several fuel salts and fission products removal strategies were simulated and their impact was analyzed. Similarly, the influence of blanket volume was assessed in a two-fluid core layout. The results showed that the reactivity initially grows during the irradiation period and later decreases. The blanket has a large impact on the performance and it can be used to further increase the fuel burnup or to shrink the active core size. The breed and burn fuel cycle in MSR can reach high fuel utilization without fuel reprocessing and a multi-fluid layout can help to decrease the core size.

scholarly journals Evaluation Of Radionuclides Release Estimation Of Power Reactor Using Scdap / Relap

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Jupiter Sitorus Pane
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Fuel Temperature ◽  
Primary Coolant ◽  
Calculation Results ◽  
Computer Codes ◽  
Radionuclide Release ◽  
Cladding Material

<p>Incident of radiation release to the environment is important event in reactor safety analysis. Numerous studies have been conducted using various computer codes, including SCDAP/RELAP, to calculate radionuclide releases into the reactor coolant during severe accident. This paper contains description of calculation results of radionuclide release from reactor core to primary coolant system in a1000 MW PWR reactor with the aim to study behavior of radionuclide releases during severe accident. The calculations using SCDAP/RELAP was done by assuming that there has been a station black out which ends up with some vapor released into the containment. As a result, the water level in core was reduced up to a level where the core is no longer covered by water. The uncovered core heats up to certain temperature where the oxidation of the cladding started to occur.  Afterwards the oxidation generated heat made fuel melting temperature reached and as consequences the release of radionuclide to the primary coolant.  The calculations show that in parallel with the increasing of fuel temperature, the radionuclide releases into the gap through diffusion started at time of 2000seconds after initial simulation but with a neglected concentration. Subsequently at the time of 29200seconds, the temperature reached more than 1000 K and the oxidation of the Zr-cladding material occurred which accelerated the fuel temperature increase and as well as radionuclide release. At34000seconds, maximum temperature of core reached 2800 K and radionuclide release into the primary cooling system started. At this time, accumulated dissolve fission product reached amount of 74.5 kg, while the non-condensable radionuclide reached 122 kg. However, these value need to be investigated further.</p>

Reactivity Accident Due to Control Rod Inadvertent Withdrawal of the 250MW High Temperature Gas-Cooled Reactor

2010 ◽  
Author(s):  
Yanhua Zheng ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Fission Products ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Control Rod ◽  
Detail Design ◽  
Fuel Particles ◽  
Reactivity Accident ◽  
High Temperature Gas

Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM) with single module power of 250MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper, e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not, and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature and Xenon concentration are studied and compared in detail between these different cases. The calculating results show that, the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.

Reactivity Accident in a High Temperature Gas-Cooled Reactor Due to Inadvertent Withdrawal of Control Rod

2010 ◽  
Vol 133 (5) ◽  
Author(s):  
Zheng Yanhua ◽  
Shi Lei
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Fission Products ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Control Rod ◽  
Detail Design ◽  
Fuel Particles ◽  
Reactivity Accident ◽  
High Temperature Gas

Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese pebble-bed modular high temperature gas-cooled reactor (HTR-PM) with single module power of 250 MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper (e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not) and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature, and xenon concentration are studied and compared in detail between these different cases. The calculating results show that the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way so that the temperature limits of the fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.

Integrated Multi-Modular Fast Reactor Concept Design

2012 ◽  
Author(s):  
Vishal Patel ◽  
Pavel Tsvetkov
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Negative Temperature ◽  
Thermal Hydraulics ◽  
Concept Design ◽  
Fuel Temperature ◽  
Feasibility Assessment ◽  
Balance Of Plant ◽  
Critical Reactor ◽  
Economics Analysis

The Integrated Multi-Modular Fast reactor is a pre-conceptual small modular fast reactor design consisting of 7 self-consist subcritical modules, each utilizing a BeO-MOX concept fuel with complete supercritical CO2 brayton cycle turbo-machinery. The subcritical modules, when brought into proximity of one another, form a complete critical reactor core. The feasibility of the reactor is assessed on a full core level, which includes a neutronics, thermal hydraulics, balance of plant, economics, and economics analysis. It has been shown that a critical configuration lasting for 14 years at 10 MWth can be achieved. A hot channel thermal hydraulics and safety analysis shows that the reactor can operate within safety limits with negative temperature coefficients of reactivity as well as stay within fuel temperature limits. A plant thermal efficiency of 36% was achieved and there is room for optimization to achieve higher efficiencies. An economical feasibility assessment shows that the reactor can be economical based on an economy of serial production argument. The analysis also leads to a licensing discussion.

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