The Impact of Reactor Operating Time on Activation Levels for Safety Analyses (original) (raw)
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Long-term radioactivity in fusion reactors
Fusion Engineering and Design, 1988
The specific activity limits for shallow land ("Class C") waste disposal of all long-lived radionuclides with atomic number less than 88 have been calculated using the 10 CFR 61 methodology. These specific activity limits were used to determine the concentration limits of nearly all naturally-occurring elements in fusion reactor blanket materials. Of the elements that could be constituents of or impurities in blanket materials, aluminum, silicon, nickel, zirconium, tantalum, and tungsten were found to be limited to concentrations of 0.1 to 10%, and niobium, molybdenum, silver, gadolinium, terbium, and holmium were found to be restricted to 0.1 to 10 parts per million.
Fusion Engineering and Design, 2018
Fusion waste will not contain transuranics and fission products like fission. Other than volumes and total radioactive inventories, a comparison among radioactive material production in different energy sources should take into account the different nature of radioactive nuclides, in terms of their hazard potential. A convenient comparison can be made with the total radiotoxicity indices. We have performed such a comparison considering fusion power plant models, GEN II PWR and GEN IV fission reactors, and ashes from a coal-fired power plant, than bear naturally occurring radioactive materials. The results are normalized, and total indices have been calculated considering the complete materials and fuel cycle for the reactors. Fusion compares favorably with fission, as expected. If low activation materials are used, fusion radiotoxicity indices, after a cooling time of the same order of magnitude as the interim storage envisaged, are also even lower than that of coal ashes.
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.
Long-term radioactive waste from fusion reactors: Part II
Fusion Engineering and Design, 1990
In Part I we calculated 10 CFR 61 "Class-C" specific activity limits for all long-lived radionuclides with atomic number less than 88 (Ra). These calculations were based on the whole-body dose. We also estimated the production of these radionuclides from all naturally occurring elements with atomic numbers less than 84 (Po) in the first wall of a typical fusion reactor, and thereby derived concentration limits for these elements in first-wall materials, if the first wall is to be suitable for Class-C disposal. In Part II we use the "effective dose equivalent" (EDE), which is a much better indication of the risk from radiation exposure than the whole-body dose, to calculate specific activity limits for all long-lived radionuclides up to Cm-248. In addition, we have estimated the production of long-lived actinides and fission products from possible thorium and uranium impurities in first-wall structures. This completes our study of long-lived radionuclides that are produced from all elements that occur in the earth's crust at average concentrations greater than one part per trillion.
Comparative Radiation Dose Study of a Hypothetical Accident in a Research Reactor
Kemija u industriji
This study is a contribution for radiation dose calculations of a hypothetical accident of a 1 MW research reactor Triga Mark II using HotSpot code. A postulated accidental release of noble gases and halogens were considered. The total effective dose (TED) was estimated for 1 day and 50 years after release. The total damage of fuel element cladding with a maximum radioactivity was considered. The obtained results show minimal TED values at the beginning of the release and at a shorter distance from the source. The maximum calculation results are acceptable and below the recommended public dose limit.
Radioactivity Buildup During Fuel Irradiation in Light Water Reactors (LWR)
CHIMIA, 2005
Power production in LWRs is economic and very efficient. A by-product of the fission process, which generates the thermal power, is radioactivity associated with fission and activation nuclides. This article describes the ways in which radionuclides are produced and then dissipated in the systems of a LWR. Radionuclides in and on the surface of components create radiation fields which can cause harm to workers. The article describes methods which help to minimize the buildup of these radiation fields. The difficulties in determining the radionuclide inventories for structural materials in LWRs are discussed and appropriate code packages are discussed. The basic principles adopted in the context of treatment of radioactive waste are summarized, including an overview of typical conditioning strategies. Radiation protection issues are briefly described and methods used to protect workers are discussed.
Nuclear Engineering and Design, 2021
In order to perform this comparison, some simple common criteria related to accidental behavior of the reactors have been developed. The first kind of criteria aims at assessing the main physical thresholds to exceed in order to have a core degradation: phase changes of coolant and of core materials (including the effect of chemical reactions) for the various reactor concepts considered. The second set of criteria deals with kinetics aspects of the accident. On the basis of core power (after scram and without scram), on the coolant inventory and on the reactor capability to be passively cooled, the heating rate of the coolant and of the core materials are assessed thanks to simplified energy balances presented in the paper. As a result, for each reactor concept, the time to reach the physical thresholds defined above is evaluated. A third set of criteria deals with core features and aims at assessing the possible reactivity insertion that withstands each concept up to core melting (or boiling for the MSR) and the associated expected power peaks in case of coolant voiding/depressurization and in case of fissile material compaction. Finally, a last criterion set deals with the assessment of the possibility to jeopardize physical barriers confining fission products. These criteria deal with the risk of barrier loadings due to coolant and core material vaporization depending on the features of the coolant/fuel and on the operating point of each reactor concept. In the last part of the paper, a synthesis is made in order to underline the weak and strong points of each of the reactor concepts investigated in terms of severe accident prevention and mitigation.
Estimation of Routine Discharge of Radionuclides on Power Reactor Experimental Rde
Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir, 2017
Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source term released into the environment based on postulation of normal or routine operations of RDE. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitation during the fabrication. Mechanism of fission products release from the fuel to the environment was cr...
Critical review of the reactor-safety study radiological health effects model. Final report
1983
This review of the radiological health effects models originally presented in the Reactor Safety Study (RSS) and currently used by the U.S. Nuclear Regulatory Commission (NRC) was undertaken to assist the NRC in determining whether or not to revise the models and to aid in the revision, if undertaken. The models as presented in the RSS and as implemented in the CRAC (Calculations of Reactor Accident Consequences) Code are described and critiqued. The major elements analyzed are those concerning dosimetry, early effects, and late effects. The published comments on the models are summarized, as are the important findings since the publication of the RSS. We conclude that the following areas deserve priority for consideration in any revision: dosirnet?: the likely particle size distributions and chemical forms of radioactive aerosols, the behavior of radioactive materials on and in soil, the uptake of radioactive material by various organs, the dosimetry of radioactive iodine in the thyroid; surly effects: the existence of especially sensitive or insensitive pDpulation subgroups, the availability of various levels of medical treatment for accidents of different magnitudes, dose-rate effects, situations in which lung damage may contribute substantially to the number of fatalities, broadening the list of early morbidities evaluated, interactions among various types of injuries; Late e f l e c t s : re-evaluation of the risks of various cancer types in the light of the BEIR I11 (NAS, 1980) report, the mortality rate for thyroid cancer victims, the doubling dose for genetic effects, "non-specific aging," and interactions between radiation and other carcinogens, genetic resistance or predisposition to cancer.
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