Recent development and application of a new safety analysis code for fusion reactors (original) (raw)
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Trends in fusion reactor safety research
Fusion Engineering and Design, 1991
Fusion has the potential to be an attractive energy source. From the safety ar,6 environmental perspective, fusion must avoid concerns about catastrophic accidents and ,b unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control rouiine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex w-_th distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g. for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions.
Recent accomplishments of the fusion safety program at the Idaho National Laboratory
Fusion Engineering and Design, 2018
The Idaho National Laboratory (INL) Fusion Safety Program (FSP) is the Office of Fusion Energy Sciences' (FES) lead laboratory for Magnetic Fusion Energy (MFE) Safety. Our mission is to assist the US and international fusion communities in developing the inherent safety and environmental potential of fusion power by: 1) Developing fusion licensing data and analysis tools, 2) Participating in national and international collaborations and design studies, and 3) Assisting the US and international fusion community in licensing activities and guidance in operational safety. To achieve our mission, the FSP maintains core competencies in the several areas, two of which are: fusion safety code development and tritium retention and permeation in fusion plasma facing component (PFC) and blanket materials. This article details recent accomplishments of our program in these areas and future directions for fusion safety research and development at INL.
Recent accomplishments and future directions in the US Fusion Safety and Environmental Program
Nuclear Fusion, 2007
The US fusion program has long recognized that the safety and environmental (S&E) potential of fusion can be attained by prudent materials selection, judicious design choices, and integration of safety requirements into the design of the facility. To achieve this goal, S&E research is focused on understanding the behavior of the largest sources of radioactive and hazardous materials in a fusion facility, understanding how energy sources in a fusion facility could mobilize those materials, developing integrated state of the art S&E computer codes and risk tools for safety assessment, and evaluating S&E issues associated with current fusion designs. In this paper, recent accomplishments are reviewed and future directions outlined.
Fusion safety codes: international modeling with MELCOR and ATHENA–INTRA
Fusion Engineering and Design, 2002
For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA Á/INTRA codes and presents their modeling results for the following breaches of a water cooling line into the tokamak vacuum vessel, (1) water injection in vacuum with condensation of the flashed steam and (2) water injection in vacuum. The modeling predictions are compared with experimental results from the ingress-ofcoolant event (ICE) facility in Japan. It is observed that the codes' predictions exhibit good agreement with the experiment data, which suggests that the codes include the proper physical mechanisms associated with the chosen accident scenarios. It is thereby concluded that either of the codes can be used to model the defined transient.
Fusion Engineering and Design, 2016
Over the past five years, the fusion energy group at Lawrence Livermore National Laboratory (LLNL) has made significant progress in the area of safety and tritium research for Inertial Fusion Energy (IFE). Focus has been driven towards the minimization of inventories, accident safety, development of safety guidelines and licensing considerations. Recent technology developments in tritium processing and target fill have had a major impact on reduction of tritium inventories in the facility. A safety advantage of inertial fusion energy using indirect-drive targets is that the structural materials surrounding the fusion reactions can be protected from target emissions by a low-pressure chamber fill gas, therefore eliminating plasma-material erosion as a source of activated dust production. An important inherent safety advantage of IFE when compared to other magnetic fusion energy (MFE) concepts that have been proposed to-date (including ITER), is that loss of plasma control events with the potential to damage the first wall, such as disruptions, are non-conceivable, therefore eliminating a number of potential accident initiators and radioactive in-vessel source term generation. In this paper, we present an overview of the safety assessments performed to-date, comparing results to the US DOE Fusion Safety Standards guidelines and the recent lessons-learnt from ITER safety and licensing activities, and summarize our most recent thoughts on safety and tritium considerations for future nuclear fusion facilities.
Validation of European computer codes used for fusion safety analysis
The development of safety files for fusion plants rely on several calculations and safety codes used for computing various situations and events. Within the future safety files, several safety related simulation codes are used for demonstrating the safe behaviour of the concerned machine and the limited impact on the population and environment in various situations from the normal operation up to the largest credible accident. These codes are ranging from neutronics calculation and dose assessment to the modelling of accident sequences and thermo hydraulics analysis in normal, incidental and accidental situations. An important aspect for the safety analysis process is the validation of the models and the reliability of the calculation results. This necessitates simulating in actual facilities physico-chemical situations related to the specific validation needs of the concerned codes. Several of these computer codes were already applied for long time in the fission industry and have ...
Laser and Particle Beams, 2005
In the field of computational modelling for S&E analysis our main contribution refers to the computational system ACAB [1] that is able to compute the inventory evolution as well as a number of related inventory response functions useful for safety and waste management assessments. The ACAB system has been used by Lawrence Livermore National Laboratory (LLNL) for the activation calculation of the National Ignition Facility (NIF) design [2] as well as for most of the activation calculations, S&E studies of the HYLIFE-II and Sombrero IFE power plants . Pulsed activation regimes can be modeled (key in inertial confinement fusion devices test/experimental facilities and power plants), and uncertainties are computed on activation calculations due to cross section uncertainties. In establishing an updated methodology for IFE safety analysis, we have also introduced time heat transfer and thermalhydraulics calculations to obtain better estimates of radionuclide release fractions. Off-site doses and health effects are dealt with by using MACSS2 and developing an appropriate methodology to generate dose conversion factor (DCF) for a number of significant radionuclides unable to be dealt with the current MACSS2 system. We performed LOCA and LOFA analyses for the HYLIFE-II design. It was demonstrated the inherent radiological safety of HYLIFE-II design relative to the use of Flibe. Assuming typical weather conditions, total off-site doses would result below the 10-mSv limit. The dominant dose comes from the tritium in HTO form. In the Sombrero design, a severe accident consisting of a total LOFA with simultaneous LOVA was analyzed. Key safety issues are the tritium retention in the C/C composite, and the oxidation of graphite with air that should be prevented. The activation products from the Xe gas in the chamber are the most contributing source to the final dose leading to 47 mSv. We also analyzed the radiological consequences and the chemical toxicity effects of accidental releases associated to the use of Hg, Pb, and
Recent Upgrades at the Safety and Tritium Applied Research Facility
Fusion Science and Technology, 2017
This paper gives a brief overview of the Safety and Tritium Applied Research (STAR) facility operated by the Fusion Safety Program (FSP) at the Idaho National Laboratory (INL). FSP researchers use the STAR facility to carry out experiments in tritium permeation and retention in various fusion materials, including wall armor tile materials. FSP researchers also perform other experimentation as well to support safety assessment in fusion development. This lab, in its present two-building configuration, has been in operation for over ten years. The main experiments at STAR are briefly described. This paper discusses recent work to enhance personnel safety at the facility. The STAR facility is a Department of Energy less than hazard category 3 facility; the personnel safety approach calls for ventilation and tritium monitoring for radiation protection. The tritium areas of STAR have about 4 to 12 air changes per hour, with air flow being once through and then routed to the facility vent stack. Additional radiation monitoring has been installed to read the laboratory room air where experiments with tritium are conducted. These ion chambers and bubblers are used to verify that no significant tritium concentrations are present in the experiment rooms. Standby electrical power has been added to the facility exhaust blower so that proper ventilation will now operate during commercial power outages as well as the real-time tritium air monitors.
The ACAB system [1] to compute the inventory evolution as well as a number of related inventory response functions useful for safety and waste management has been used by Lawrence Livermore National Laboratory (LLNL) for the activation calculation of the National Ignition Facility (NIF) design as well as for most of the activation calculations, S&E studies of the HYLIFE-II and Sombrero IFE power plants with a severe experimental testing at RTNS-II of University Berkeley. Pulsed activation regimes can be modelled (key in inertial confinement fusion devices test/experimental facilities and power plants), and uncertainties are computed on activation calculations due to cross section uncertainties. In establishing an updated methodology for IFE safety analysis, we have also introduced time heat transfer and thermal-hydraulics calculations to obtain better estimates of radionuclide release fractions.
Fusion Engineering and Design, 2011
Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al 2 O 3 , SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.