Volatile radionuclide

According to researchers, limitations of DC arc systems include the corrosive nature of the vitrified material, limitations on salt and water content, and uncertain performance in the destruction of organic wastes. The addition of flux materials may be required to allow the vitrified material to be poured and to allow the final waste form to meet performance goals. Volatile radionuclides and metals may accumulate in the off-gas treatment system. 

Low-temperature treatment of low-level mixed wastes has also been accomplished by solidification/stabilization with chemically bonded phosphate ceramics (CBPC). These are made by hydrothermal chemical reaction rather than by sintering. Chemical bonding develops when acid phosphates react with oxides to form crystalline orthophosphate (Singh et al. 1997). The ceramic matrix stabilizes the wastes by microencapsulation. The low temperature of the reaction allows volatile radionuclides to be treated (Singh et al. 1997). 

A gradient of Cs/Mn ratios with distance from the crater beyond 1500 feet is present owing to the fractionation of those radionuclides during the detonation, and subsequent deposition of ejecta and fallout. Figure 14 shows Cs/Mn ratios for 6-inch depth ejecta samples collected at Sedan crater from the lip to 3000 feet from ground zero. Any increase in Cs/Mn ratios observed at sites distant from the crater lip must be evaluated in view of the fractionation that is evident in the ejecta with respect to these two radionuclides. The Cs/Mn ratio at crater lip sites and those out to 1500 feet are all close to 0.01-0.02. A sample of fused silicate material, which represents a class of ejecta extremely depleted in volatile radionuclides, had a low Cs-Mn ratio of 0.006. 

Vented Underground Burst. An underground detonation which produces no visible fireball, but which results in the release of volatile radionuclides through fissures or other vents, produces a single particle class— atmospheric aerosol particles with condensable radionuclides deposited on their surface. Radionuclide abundance is independent of particle size. 

Assessment of Waste Management of Volatile Radionuclides. PB80-147754. The Mitre Corporation, McLean, VA, 1979. 

The behavior of release intensity and the activity accumulation in the atmosphere are demonstrated in Fig.7. The release intensity reaches its maximum of 8-10" Bq/s dming the first day after the pulse. The major contribution is due to long-lived Kr (Ti/2= 10.7 year) generated over previous operation of the reactor. The release of volatile radionuclide (mainly long-lived Cs) is considerably less. 

Most models assessing the long-term behaviour of Cs and Sr from fallout include processes relevant to LMM ionic species only. However, radionuclides may be associated with particles due to (a) release of fuel matrix or clusters, (b) condensation of volatiles on available particle surfaces after the release of (c) interactions with aerosol particulates during atmospheric transport. For volatile radionuclides (e.g., Cs, °Sr), all three mechanisms may be equally relevant, while the deposition of nuclides of refractory elements (e.g., " Ce, Nb, Zr) indicates the release of fuel particles. For fuel particles, depletion of volatiles relative to refractory elements would be expected to depend on the temperature reached during the releases, whereas the activity ratios for refractory elements should reflect the reactor fuel bum-up. 

In a high containment plant, the release of fuel solution would probably be below current detection limits. However, if minimal off-gas treatment is applied (i.e. caustic scrubbing and filtration is omitted) then there would be potential for the release of fine droplets of fuel solution. The release fractions of volatile radionuclides will depend on the type of off-gas treatment process used. An indication of the activity release for a 50 kg dissolution of 1000 MWd/t fuel, involving the use of different off-gas treatment system (consider, scrubber), is presented in Table 12.7 (after McMahon et al., 1993). 

Table 8,11 Volatile radionuclides in discharge fuel from neutron activati< it...

Another volatile radionuclide formed in HTGR fuel is 3.1 X 10 year Cl, formed by neutron activation of chlorine contaminant in the fuel, according to the reaction... 

In preparation for dissolution, step 1, cladding is opened to permit subsequent dissolution of the oxide fuel. For steel or zircaloy this is done by mechanical shearing or sawing. Off-gases from decladding contain up to 10 percent of the radiokrypton and xenon in the fuel and some of the 002," tritium, and other volatile fission products. If voloxidation (Sec. 4.3) is used after decladding to remove tritium, more of the other volatile radionuclides will then be evolved also. 

Disadvantages of the process are loss of volatile radionuclides at the elevated temperature and the large amount of the fusion reagent added to the sample. Handling the crucibles at the specified high temperatures is a safety concern. Platinum crucibles are expensive, while other crucibles (e.g., nickel and iron) are attacked by some reagents and contribute a contaminant to the sample. 

LS counters are suitable for measuring radionuclides that emit only very low-energy beta particles or electrons, notably tritium (Emax = 18.6 keV). When tritium is measured as tritiated water and its activity is reported relative to water weight or volume, no yield measurement is needed. Liquid samples, e.g., water from the environment, process streams, urine, or dissolved solids, can be counted directly or purified by distillation. Results for purified samples are more reliable to the extent that the radioelement can be identified, quenching is stabilized, and luminescent contaminants are removed. Reagents may have to be added to the distillation flask to hold back other potentially volatile radionuclides, such as radioactive iodine, carbon, ruthenium, or technetium. 

Water reaches the disposal drifts via small fractures, and saturates bentonite in a few decades. Minimum container lifetime due to anaerobic general corrosion is 20,000 years, although in the evaluation it is assumed that a few (up to 10) canisters fail much earlier due to a fabrication defect. After canister failure, and since no credit is given to the cladding as a barrier, there is an instantaneous release of some volatile radionuclides, such as C1 and Cs. When the water reaches the waste, the gradual release of the radionuclides in the UO2 matrix starts. 

Certain waste treatments reduce multiple hazards in one step. For example, incineration can destroy oxidizable organic chemicals and infectious agents, waste feed rates can be controlled to meet emission limits for volatile radionuclides, and radioactive ash can be disposed of as a dry radioactive waste. Likewise, some chemical treatment methods (e.g., those using bleach) both oxidize toxic chemicals and disinfect biological hazards. Such treatment could convert a chemical-radioactive-biological waste to a radioactive waste. 

There is no universal concentration process for measuring the radioactivity of water. The evaporation method has proven suitable for measuring radioactivity. Detecting the radioactivity of volatile radionuclides is... 

As only small absolute amounts of radionuclides are to be determined, and thus only small amounts of elements are to be separated, separation cannot usually be properly performed by methods whose success depends on the amount of the component to be isolated (e.g., as in precipitation). Therefore, methods that are independent of amount (such as liquid-liquid extraction and ion-exchange methods) are more advantageous. Extraction procedures very often take advantage of additions of chelating components. For separating volatile radionuclides (such as iodine or ruthenium) from the sample matrix, distillation methods can be used advantageously. Electrolytic deposition has been shown to be applicable in the separation of polonium. 

For refractory radionuclides such as Zr and Ce, 50% are assumed to be deposited locally and a further 25% is deposited regionally, while for volatile radionuclides such as °Sr, Cs, and 50% of the fission yield is assumed to be deposited locally and regionally. The remainder of the debris and all of the debris from airbursts is widely dispersed in the atmosphere. 

This report examines the severe accident sequences and radionuclide source terms at the Sizewell pressurised water reactor with a piestressed concrete containment, the Konvoi pressurized water reactor with a steel primary contaimnent, the European Pressurised water Reactor (EPR) and a boiling water reactor with a Mark 2 containment. The report concludes that the key accident sequences for European plant designs are transient events and small loss-of-coolant accidents, loss of cooling during shutdown, and containment bypass sequences. The most important chemical and transport phenomena are found to be revaporisation of volatile radionuclides from the reactor coolant system, iodine chemistry, and release paths through the plant. Additional research is recommended on release of fission products from the fuel, release of fission products from the reactor coolant system, ehemistry of iodine, and transport of radionuclide through plants. 

These various processes that lead to radionuclide release into the contaimnent atmosphere are discussed individually in the subsections that follow. It should be remembered that while ex-vessel releases from core debris are taking place, there can be continuing releases from residual core debris in the reactor vessel and revaporisation releases of volatile radionuclides deposited in the reactor coolant system. 

Tritium (H-3) and carbon-14 (C-14) are also volatile radionuclides but H-3 has a short half-life (12.3 yr) and its low beta decay energy (18.5 keV) makes it a relatively mild... 

McKay, H.A.C. 1980. Background considerations in the immobilization of volatile radionuclides, management of gaseous wastes from nuclear facilities. MEA report, LAEA-SM-245/8. Vienna, Austria International Atomic Energy Agency, pp. 59-80. 

Law, R. J., Indig, M. E., Lin, C. C., Cowan, R. L. Suppression of radiolytic oxygen produced in a BWR by feedwater hydrogen addition. Proc. 3. BNES Conf Water Chemistry in Nuclear Reactor Systems, Bournemouth, UK, 1983, Vol. 2, p. 23-30 Lin, C. C. Chemical behaviour and distribution of volatile radionuclides in a BWR system with forward-pumped heater drains. Proc. 3. BNES Conf. Water Chemistry in Nuclear Reactor Systems, Bournemouth, UK, 1983, Vol. 1, p. 103—110 Lin, C. C. Chemical behaviour and steam transport of nitrogen-13 in BWR primary systems. 

Lin, C. C. Chemical behaviour and distribution of volatile radionuclides in a BWR system with forward-pumped heater drains. Proc. 3. BNES Conf. Water Chemistry of Nuclear Reactor Systems, Bournemouth 1983, Vol. 1, p. 103-110... 

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