0

Accepted Manuscripts

BASIC VIEW  |  EXPANDED VIEW
research-article  
Gueorgui Petkov
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039000
The experience of severe accidents shows that reliable determination of technological process parameters is necessary but not always sufficient to avoid catastrophic consequences. The accident measures should be considered in a broader context that includes the human factor, organization of the nuclear technology, external influences and safety culture. The ATWS (Anticipated Transient Without Scram) events were not considered in the original WWER (Russian PWR) design basis accidents. The design extension conditions scenarios progress in a context which is very uncertain and highly stressful for the operators. If a specific scenario requires some operators' actions as measures to mitigate, delay or distribute the accident consequences then the dynamics of accident context seem of primary importance for 'best estimate' evaluations and enhancing the plant's capability. The paper presents the capacities of the Performance Evaluation of Teamwork (PET) procedure for enhancing plant's capability for design extension conditions based on 'best estimate' context evaluation of human performance in ATWS events. The PET procedure is based on a thorough description of symptoms of various timelines and their context quantification. It is exemplified for different ATWS scenarios of the NPP with WWER-1000 based on thermal-hydraulic simulations with RELAP5/MOD3.2 code and models.
TOPICS: Design, Accidents, Dynamics (Mechanics), Safety, Simulation, Transients (Dynamics), Engineering simulation, Human factors, Delays, Nuclear power stations, Performance evaluation, Pressurized water reactors, Scram
Technical Brief  
Wenbin Xiong, Qin Xie, Huwei Li, Sengai Yang, Huan Mao and Jian Cao
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038999
The whole core model of CEFR is established according to the parameters of China Experimental Fast Reactor which are given by IAEA-TECDOC-1531, and the physical parameters of CEFR are simulated with the MCNP4a program. The calculation results are compared with the data contained in the safety analysis report of CEFR. The calculation results are consistent with the design values, which successfully demonstrates the acceptable fidelity of the MCNP model. The MCNP model will be further refined and applied for nuclear safety review of the CEFR in the future.
TOPICS: Simulation, China, Fast neutron reactors, Safety, Design
research-article  
Gueorgui Petkov and Monica Vela-Garcia
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038928
The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the Performance Evaluation of Teamwork (PET) procedure for dynamic context quantification and determination of alternatives, coordination and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions with the use of thermo-hydraulic model and severe accident codes (MELCOR and MAAP). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and an hypothetic unmitigated LT SBO at Peach Bottom #1 BWR Reactor NPPs. The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, "State-of-the-Art Reactor Consequence Analysis" and thermo-hydraulic calculations made by using MAAP code at the EC Joint Research Centre.
TOPICS: Accidents, Boiling water reactors, Nuclear power stations, Fukushima nuclear disaster, Japan, 2011, Safety, Performance evaluation, Uncertainty, Decision making
research-article  
Zachary Weems, Sedat Goluoglu and Mark DeHart
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038929
TREAT, a graphite moderated experimental reactor, is scheduled to restart in late 2017. There is now renewed interest in development of capabilities to model and simulate TREAT transients using three-dimensional coupled physics. To validate existing transient analysis tools as well as those under development, several temperature- limited transients have been modeled and analyzed. These transients are from the M8CAL experiment series, a set of experiments performed to calibrate the reactor detectors for the planned M8 series of fuel tests. Detailed reactor models were prepared that were than used to calculate the pre-transient and post-transient keff values as well as corresponding reactivity insertions. Alterations to modeled values of shutdown and initial transient rod insertion depths were made to better match the reported experimental values of reactivity insertions assuming just critical pre-transient states. It was found that two of the altered media inputs, fuel and Zircaloy-3 cladding, had significant effects on the keff. In addition, increasing shutdown rod insertion by 3 to 5 cm and decreasing initial transient rod insertion by 1 to 2 cm gave perfect pre-transient keff and total reactivity insertion values. However, the revised positions are as much as a factor of 3 to 20 different from reported uncertainty of 0.762cm. This suggests that boron concentration uncertainties may play a significant role in accurately modeling the TREAT transients and should be investigated thoroughly.
TOPICS: Physics, Temperature, Sensors, Fuels, Transients (Dynamics), Cladding systems (Building), Modeling, Graphite, Transient analysis, Uncertainty, Boron
research-article  
Yuta Abe, Ikken Sato, Toshio Nakagiri, Akihiro Ishimi and Yuji Nagae
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038911
A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm x 107 mm x 222 mmh). An excellent perspective in terms of applicability of the non-transfer plasma heating to melting high melting-temperature materials such as ZrO2 has been obtained. In addition, molten pool was formed at the middle height of the test piece indicating its capability to simulate the initial phase of core degradation behavior consistent with the real UO2 fuel Phebus-FPT tests. Furthermore, application of EPMA, SEM/EDX and X-ray CT led to a conclusion that the pool formed consisted mainly of Zr with some concentration of oxygen which tended to be enhanced at the upper surface region of the pool. Based on these results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the Severe Accident (SA) experimental study was obtained. Keywords: Fukushima Daiichi decommissioning, core-material relocation (CMR), non-transfer type plasma heating, X-ray CT
TOPICS: Plasmas (Ionized gases), Accidents, Boiling water reactors, Heating, Computerized tomography, Melting, Temperature, Fuels, Oxygen, Zirconium, Nuclear decommissioning, Uncertainty, Fukushima nuclear disaster, Japan, 2011
research-article  
Mukesh Kumar Dhiman, Arun Nayak, Sumit Vishnu Prasad, Purnendra Verma, Raj Kumar Singh, Vikas Jain and Dinesh Kumar Chandraker
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038899
Detection of LOCA and generation of reactor trip signal for shutting down the reactor is very important for safety of the nuclear reactor. LBLOCA being the design basis accident in all reactors has attracted attention of the reactor designers. However, studies reveal that SBLOCA is sometimes much severe as it is difficult to detect SBLOCA with conventional methods and generate reactor trip signals for safety of the reactor. SBLOCA in channel type reactor is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. AHWR is a channel type BWR is prone to stagnation channel break conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprises of acoustic based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic based sensors system and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 seconds and trip can be initiated within 7 seconds followed by LOCA.
TOPICS: Safety, Signals, Acoustics, Sensors, Design, Pipes, Boiling water reactors, Failure, Nuclear reactor coolants, Nuclear reactors, Fuels, Accidents, Steam, Heat
research-article  
Satya Prakash Saraswat, Prabhat Munshi, Ashok Khanna and Chris Allison
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038823
This work attempts to investigate the thermal hydraulic safety of Lithium Lead Ceramic Breeder (LLCB) Test Blanket System (TBS) in ITER with the help of modified thermal hydraulic code RELAP/SCDAPSIM/MOD4.0. The design basis accidents, in-vessel and ex-vessel loss of coolant of First Wall (FW) of Test Blanket Module (TBM) are analyzed for this safety assessment. The sequence of accidents analyzed is start with Postulated Initiating events (PIEs). A detailed modeling of First Wall Helium Cooling System (FWHCS) loop and Lithium Lead Cooling System (LLCS) is presented. The analysis of steady state normal operation along with 10 s power excursion before the accident is also discussed in order to better understanding of initial condition of accidents. The analysis discusses a number of safety concerns and issues that may result from the TBM component failure, such as VV pressurization, TBM FW temperature profile, passive decay heat removal capability of TBM structure, pressurization of Port Cell and Tokomak Cooling Water System Vault Annex (TCWS-VA) and to check the capability of passive safety system (Vacuum Vessel Pressure Suppression System (VVPSS)). The analysis shows that in these accident scenarios the critical parameters have reasonable safety margins. Keywords TBM, ITER, LLCS, FWHCS, LOCA, Thermal hydraulics, Safety analyses, Decay heat removal, RELAP/SCDAPSIM/MOD4.0, VVPSS, Fusion, Plasma.
TOPICS: Safety, Accidents, Vessels, Lithium, Heat, Cooling systems, Ceramics, Vacuum, Cooling, Design, Modeling, Failure, Helium, Steady state, Temperature profiles, Thermal hydraulics, Water, Coolants, Plasmas (Ionized gases), Pressure
research-article  
Jian Li, Ding She and Lei Shi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038771
In depletion and transmutation calculation, it's important to solve detailed burnup chains with high computational accuracy and efficient. This requires the good performance of the burnup algorithms. NUIT (NUclide Inventory Tool) is a newly-developed nuclide inventory calculation code, which is capable of handling detailed depletion chains by implementing various advanced algorithms. Based on the NUIT code, this paper investigates the accuracy and efficiency of the Mini-Max Polynomial Approximation (MMPA) method, and compares it with other burnup solvers in NUIT code. It is concluded that the MMPA method is numerically accurate and efficient for dealing with detailed depletion chains with extremely short half-lived nuclides.
TOPICS: Algorithms, Chain, Nuclides, Polynomial approximation
research-article  
Yi-Ning Zhang, Hao-Chun Zhang and Ke-Xin Wang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038772
Point reactor neutron kinetics equations describe the time dependent neutron density variation in a nuclear reactor core. These equations are widely applied to nuclear system numerical simulation and nuclear power plant operational control. This paper analyses the characteristics of 10 different basic or normal methods to solve the point reactor neutron kinetics equations. The accuracy after introducing different kinds of reactivity, stiffness of methods and computational efficiency are analyzed. The calculation results show that: considering both the accuracy and stiffness, implicit Runge-Kutta method and Hermite method are more suitable for solution on these given conditions. The explicit Euler method is the fastest, while the power series method spends the most computational time.
TOPICS: Neutrons, Stiffness, Density, Computer simulation, Nuclear reactors, Nuclear power stations, Runge-Kutta methods
research-article  
Xie Yang, Ding She and Lei Shi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038774
Due to the advantages of small volume, light weight and long-time running, nuclear reactor can provide an idea energy source for submarines, ships and even space crafts. In this paper, two small compact prismatic nuclear reactors with different core block material are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with high enriched uranium nitride fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa respectively. High temperature helium flowing out of the reactor can be used as the working medium for Closed Brayton Cycle (CBC) power conversion to generate at least 1 MW electricity, due to the high efficiency of CBC. Neutronics analyses of reactors for the preliminary design in this paper are performed using Reactor Monte-Carlo (RMC) code developed by Tsinghua University. Both the two reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, which have at least $1 reactivity shutdown margin, and remains at least $1 subcritical in the submersion accident as well as one control drum failed accident. Finally, the optimization design is determined after comparing the 235U mass and the total reactor mass of the above two prismatic reactors.
TOPICS: Design, Nuclear reactors, Accidents, Helium, Ships, Underwater vehicles, Uranium, High temperature, Graphite, Energy resources, Optimization, Brayton cycle, Weight (Mass), Pressure, Temperature, Alloys, Fuels, Thermal energy, Energy conversion
research-article  
Pan Qingquan, Lu Haoliang, Li Dongsheng and Wang Kan
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038653
Solving the SP3 equation is the key point of the development in the advanced reactor calculation method and has been widely concerned. The semi-analytical nodal method (SANM) based on transverse-integrated diffusion equation has the advantages of high accuracy and convenience for multi-group calculation. Due to its advantages, the method is expected to be used in solving the SP3 equation. However, the traditional SANM is not rigorous since the expansion process does not take the special modality of SP3 equation and its analytical solution into consideration. There are two modalities of SP3 equation, so there are two traditional SANM forms on solving the SP3 equation, and the differences between the two forms will be very important in farther research on the SANM. A code is developed to solve the SP3 equation under the two different forms. After the calculation of the same benchmark, the difference between the two forms on solving the SP3 equation is found. According to the results, and in view of the special modality of the SP3 equation, points on a more rigorous SANM for solving SP3 equation are discussed.
TOPICS: Diffusion (Physics), Computational methods
research-article  
Kevin Fernández-Cosials, Gonzalo Jimenez, César Serrano, Luisa Ibanez and Ángel Peinado
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038595
During a severe accident in a nuclear power plant, there are several challenges that need to be faced. To coup with a containment overpressure, the venting action will lower the pressure but it will release radioactivity to the environment. In order to reduce the radioactivity released, a filtered containment venting system can be used to retain iodine and aerosols radioactive releases coming from the containment atmosphere. However, during a severe accident, large quantities of hydrogen can also be generated. Hydrogen reacts violently with oxygen and its combustion could impair systems, components or structures. For this reason, to protect the integrity of the filtered containment venting system against hydrogen explosions, an inertization system is found necessary. This system should create an inert atmosphere previous to any containment venting that impedes the contact of hydrogen and oxygen. In this paper, the inertization system for Cofrentes nuclear power plant is presented. It consists of a nitrogen injection located in three different points. A computational model of the filtered containment venting system as well as the inertization system has been created. The results show that if the nitrogen sweeps and the containment venting are properly synchronized, the hydrogen risk could be reduced to a minimum and therefore, the integrity of the filtered containment venting system would be preserved.
TOPICS: Nuclear power stations, Containment, Hydrogen, Nitrogen, Accidents, Radioactivity, Oxygen, Risk, Combustion, Explosions, Aerosols, Pressure
research-article  
Massimo Di Pietro, Sergio Pistelli, Eugenio Garneri, Rosa Lo Frano and Riccardo Ciolini
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038556
In this paper a new technique to handle solid radioactive materials inside a liquid matrix is presented. The conceptual design of the device profits of the experience and know-how gained in decontamination procedures. The proposed system makes use of an ejector for the suction of a water-highly radioactive swarf mixture from the purifiers pool of the Italian E. Fermi nuclear power plant and moving it in a suitable container for the subsequent conditioning. A dedicated circuit with an ejector to demonstrate the feasibility of the method was realized. A minimum inlet flow rate was found to have swarf suction. The feasibility of the method was demonstrated, even if it is required to homogenize the inlet mixture to avoid swarf packing conditions inside the ejector.
TOPICS: Nuclear power stations, Nuclear decommissioning, Scrap metals, Ejectors, Suction, Radioactive substances, Packing (Shipments), Flow (Dynamics), Decontamination, Containers, Circuits, Packings (Cushioning), Water, Conceptual design
research-article  
Frederic Salaun and David Novog
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038557
The Canadian Super Critical Water Reactor (SCWR) design is part of Canada's Generation IV reactor development program. The reactor uses batch fueling, light water coolant above the thermodynamic critical point and a heavy water moderator. The design has evolved considerably and is currently at the conceptual design level. As a result of batch fueling a certain amount of excess reactivity is loaded at the beginning of each fueling cycle. This excess reactivity must be controlled using a combination of burnable neutron poisons in the fuel, moderator poisons and control blades interspersed in the heavy water moderator. Recent studies have shown that the combination of power density, high coolant temperatures and reactivity management can lead to high maximum cladding surface temperatures (MCST) and maximum fuel centerline temperatures (MFCLT) in this design. This study focuses on improving both the MCST and the MFCLT through modifications of the conceptual design including changes from a 3 to 4 batch fueling cycle, a slightly shortened fuel cycle (although exit burnup remains the same), axial graded fuel enrichment, fuel-integrated burnable neutron absorbers, lower reactivity control blades, and lower reactor thermal powers as compared to the original conceptual design. The optimal blade positions throughout the fuel cycle were determined such as to minimize the MCST and MFCLT using a genetic algorithm and the reactor physics code PARCS. The final design was analyzed using a fully coupled PARCS-RELAP5/SCDAPSIM model to accurately predict the MCST as a function of time during a fueling cycle.
TOPICS: Physics, Optimization, Supercritical water reactors, Design, Water, Fuels, Temperature, Blades, Cycles, Conceptual design, Nuclear fuel cycle, Neutrons, Coolants, Cladding systems (Building), Genetic algorithms, Generation IV reactors, Power density
research-article  
Lori Walters, Michael Wright and D. Guzonas
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038367
The Canadian Super Critical Water-Cooled Reactor (SCWR) concept requires materials to operate at higher temperatures than current Generation III water-cooled reactors. Materials performance after radiation damage is an important design consideration. Materials that are both corrosion resistant and radiation damage tolerant are required. This paper summarizes the operating conditions including temperature, neutron flux and residence time of in-core Canadian SCWR components. The focus is on the effects of irradiation on in-core components, including those exposed to a high neutron flux in the fuel assembly, the high pressure boundary between coolant and moderator, as well as the low-temperature, low-flux calandria vessel that contains the moderator. Although the extreme conditions and the broad range of SCWR in core operating conditions present significant materials selection challenges, candidate alloys that can meet the performance requirements under most in-core conditions have been identified. However, for all candidate materials, insufficient data are available to unequivocally ensure acceptable performance and experimental irradiations of candidate core materials will be required. Research programs are to include out-of-pile tests on un-irradiated and irradiated alloys. Ideally, in-flux studies at appropriate temperatures, neutron spectrum, dose rate, duration, and coolant chemistry will be required. Characterization of the microstructure and the mechanical behavior including strength, ductility, swelling, fracture toughness, cracking and creep on each of the in-core candidate materials will ensure their viability in the Canadian SCWR.
TOPICS: Irradiation (Radiation exposure), Supercritical water reactors, Temperature, Alloys, Neutron flux, Water, Coolants, Radiation damage, High pressure (Physics), Ductility, Cracking (Materials), Corrosion, Design, Fracture (Process), Low temperature, Mechanical behavior, Chemistry, Fracture toughness, Creep, Vessels, Fuels, Manufacturing, Neutrons
Technical Brief  
Chu Rainer Kwang-Hua
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038336
We adopted the verified absolute-reaction theory which originates from the quantum chemistry approach to explain the anomalous plastic flow or plastic deformation for Si nanowires irradiated with 100 keV (at room temperature regime) Ar$^+$ ions as well as the observed amorphization along the Si nanowire [Nano Lett. {\bf 2015}, 15, 3800-3807]. We also demonstrate some formulations which can help us calculate the temperature-dependent viscosity of flowing Si in nanodomains.
TOPICS: Deformation, Nanowires, Silicon, Temperature, Ions, Viscosity, Quantum chemistry
research-article  
Laurent Cantrel, Thierry Albiol, Loic Bosland, Juliette Colombani, Frédéric Cousin, Anne-Cécile Grégoire, Olivia Leroy, Sandrine Morin, Christian Mun, Ohnet Marie Noëlle, Sidi Souvi, Céline Monsanglant-Louvet, Florent Louis, Bruno Azambre and Christophe Volkringer
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038223
This paper deals with near past, ongoing and planned R&D works on fission products (FPs) behavior in Reactor Cooling System (RCS), containment building and in Filtered Containment Venting Systems (FCVS) for severe accident (SA) conditions. The FCVS topic, in link with the post-Fukushima accident management, is again on the agenda [1]. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in simulation softwares. After being initiated in the International Source Term Program (ISTP), researches on iodine transport through the RCS are still ongoing and for containment, the last advances are linked to the Source Term Evaluation and Mitigation (STEM) OECD/NEA project. The objective is to improve the evaluation of Source Term for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major FPs: iodine and ruthenium. A follow-up, called STEM2 has just started to reduce some remaining issues and be closer to reactor conditions. For FCVS works, the efficiencies for trapping iodine covering scrubbers and dry filters are examined to get a clear view of their abilities in SA conditions. Another part is focused on specific porous materials able to trap volatile iodine. Influence of zeolites materials parameters (nature of the counter-ions, structure, Si/Al ratio...) will be tested. New kind of porous materials constituted by Metal organic Frameworks will also be looked at because they can constitute a promising way of trapping.
TOPICS: Accidents, Containment, Porous materials, Simulation, Nuclear fission, Ions, Metals, Chemistry, Filters, Nuclear reactor cooling, Nuclear power stations, Ruthenium, Containment buildings, Uncertainty, Accident management, Fukushima nuclear disaster, Japan, 2011
research-article  
Robin McDougall, Scott B. Nokleby and Ed Waller
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038185
This paper presents a novel methodology for generating radiation intensity maps using a mobile robotic platform and an integrated radiation model. The radiation intensity mapping approach consists of two stages. First, radiation intensity samples are collected using a radiation sensor mounted on a mobile robotic platform, reducing the risk of exposure to humans from an unknown radiation field. Next, these samples, which need only to be taken from a subsection of the entire area being mapped, are then used to calibrate a radiation model of the area. This model is then used to predict the radiation intensity field throughout the rest of the area that could not be directly measured. The performance of the approach is evaluated through experiments. The results show that the developed system is effective at achieving the goal of generating radiation maps using sparse data.
TOPICS: Radiation (Physics), Robotics, Risk, Sensors
Expert View  
Samuel Miranda
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038160
Nuclear safety criteria are based upon the concept that plant situations that are expected to have a high frequency of occurrence must not pose a danger to the public, and that plant situations that pose the greatest danger to the public must be limited to situations that have a very low expected frequency of occurrence. This concept is implemented by grouping postulated plant situations (or events) into categories that are defined according to their expected frequencies of occurrence. Events in each category must be shown to yield consequences that remain within the limits that are defined for that category. Events must not be allowed to develop into the more serious events that belong in other, higher-consequence categories. In other words, nuclear plant designs must not allow high-frequency, low-consequence events to degrade into high-frequency, high-consequence events. The development of this system of frequency-based categorization is discussed, followed by an evaluation of various methods that have been proposed and applied to show how minor events could be prevented from becoming major events.
TOPICS: Safety, Accidents, Nuclear power stations
research-article  
Rafael Bocanegra, Valentino Di Marcello, Victor H. Sánchez-Espinoza and Gonzalo Jiménez
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038062
A VTT Fukushima Daiichi Unit 3 MELCOR model [1] was modified to simulate the Fukushima Daiichi Unit 2 accident. Five simulations were performed using different modeling approaches. The model 1F2 v1 includes only the basic modifications to reproduce the 1F2 accident. The model 1F2 v2 includes the same modifications used in 1F2 v1 plus the WW improvement. In the 1F2 v3 model, the RCIC logic was modified to avoid the use of tabular functions for the mass flow inlet and outlet. As a result of this analysis it is concluded that there is a strong dependency on parameters which still have many uncertainties, such as the RCIC two-phase flow operation, the alternative water injection, the suppression pool behavior, the rupture disk behavior and the containment failure modes which affect the final state of the reactor core.
TOPICS: Simulation, Accidents, Fukushima nuclear disaster, Japan, 2011, Flow (Dynamics), Failure mechanisms, Modeling, Two-phase flow, Underground injection, Disks, Rupture, Containment, Uncertainty

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In