Operational conditions, radionuclides

Spent fuels vary in microstructure, and phase and elemental distribution depending on the in-core reactor operating conditions and reactor history. The chemical stability of spent U oxide fuel is described by local pH and Eh conditions, redox being the most important parameter. However, the redox system will also evolve with time as various radionuclides decay and the proportion of oxidants and reductants generated at the fuel/water interface changes with the altering a-, (J-, y-radiation field and with the generation of other corrosion products that can act as... 

Methodological artefacts may arise for a number of reasons, most notably as a result of specific interactions of species with the filter membrane. Therefore the choice of the ultrafiltration system, the properties and influence of the membrane and the operating conditions must be carefully considered before the ultrafiltration technique is applied for the separation of different radionuclide species in environmental samples. 

Nuclear reactors and reprocessing plants are constructed and operated in such a way that the radioactive inventory is confined to shielded places. Only limited amounts of radionuclides are allowed to enter the environment. The amounts of T and produced in nuclear reactors vary with the reactor type, between about 10 and 10 Bq of T and about lO Bq of per GWg per year. Tritium is released as HTO and about one-third of the is in the form of " C02. Under normal operating conditions, very small amounts of fission products and radioelements are set free from nuclear reactors and reprocessing plants. In this context, the actinides and long-lived fission products, such as °Sr, Tc, I, and Cs, are of greatest importance. 

Provide additional design features or systems to ensure containment of radionuclides in the event that normal operating conditions cannot be maintained and/or plant protection is not assured. 

A fuel performance analysis was conducted to predict the core temperature distributions, fuel particle failure, and gaseous and metallic fission product release under normal operating conditions at full power. The calculated fission product releases were then compared with the radionuclide design criteria, summarized In Section 4.2.3 and presented In detail In Section 11.1 to determine the adequacy of the fuel and core designs with regard to the radionuclide control requirements. 

The risk that, through a deterioration in their condition, the SNF containment barriers fail during recovery operations, releasing radionuclides either at the scene or during subsequent recovery and transportation to the land disposal site. 

A nuclear reactor is a device in which the fission process is controlled, either to produce power, radionuclides, or both. All nuclear reactors depend upon an initial load of fuel that contains fissile materials. Absorption of a neutron by a fissile nucleus produces another fission event with high probability, accompanied by the emission of more neutrons. If one of the neutrons emitted in each fission induces another fission, the number of neutrons in each succeeding generation will remain constant and the neutron economy is balanced. This is referred to as a self-sustaining chain reaction, and is the normal operating condition of a nuclear reactor. (See O Chaps. 57 and O 58 in this Volume.)... 

During reactor operation, small amounts of C are also formed in nuclear fuels. With regard to the radioactivity balance of the fuel during plant operation, this radionuclide is of no relevance however, considering its comparatively long halflife of 5736 years and its postulated release behavior from the fuel matrix under storage or accident conditions, it is of definite interest with respect to reprocessing or final disposal of the spent fuel. For these reasons, it is of importance to know the inventories as well as the chemical state of this radionuclide in the fuel. 

Under normal operating conditions, the water-steam circuit (or secondary circuit) of pressurized water nuclear power reactors is completely free of radionuclides. Activated corrosion products and tritium, which have been reported in very low activity concentrations from the water-steam circuits of some high-temperature reactors and which are caused by the neutron Held reaching into the steam generator or by diffusion through intact steam generator heating tubes, do not appear at the PWR secondary side. 

The carry-over of these radionuclides from the reactor water to the steam depends on both the steam moisture content and the solubility of the individual elements in the steam. With elements forming strong electrolytes in the reactor water (such as Cs" ), solubility in the steam under BWR operating conditions is very low this means that cesium carry-over is exclusively controlled by the moisture content of the steam. On the other hand, elements which are present in the water phase in the form of weakly dissociated compounds will show a noticeable solubility in steam even at BWR conditions (i. e. operating pressure in the range 7 MPa), resulting in a carry-over which may be markedly higher than the moisture... 

In a PWR primary circuit, the in-core and the out-of-core surfaces are wetted by the same coolant further, the temperature differences over the entire primary circuit amount only to about 35 K it can, therefore, be assumed that the metal release rates of identical materials are identical over the whole circuit. In full confirmation of this assumption, the comparative investigations of Lister and Davidson (1989) of the corrosion behavior of the materials inside a neutron field (NRX research reactor) and outside of it, when PWR operating conditions were applied, yielded no indication of radiation-induced differences in the metal release rates. As a consequence of these findings, the metal release rates obtained in out-of-pile experiments can be expected to represent the in-pile behavior of these materials with sufficient reliability. As was emphasized above, this conformity is restricted to the corrosion release rates of the elements from the materials and does not give any indication of the rates of supply of radionuclides to the coolant from different locations of the materials. 

The investigations cited above (like others performed in this field) yielded results on the amounts of corrosion products formed under PWR operating conditions, but not on the amounts of the associated radionuclides. As was emphasized above, Co source must not be confused with the term of essential interest Co source . Moreover, it has to be pointed out that the conclusions drawn from such evaluations of corrosion rates are only fully valid if the deposition of corrosion products on the fuel rod surfaces and subsequent neutron activation there, i. e. mechanism 1 in Fig. 4.26., were the most important contributor to the production of Co carried in the coolant. If this assumption does not apply, as will be discussed below, then these arguments would be of less significance. In order to correlate the two figures Co supply to the coolant and Co supply to the coolant , the activation period and the neutron flux density to which the different potential sources are exposed also have to be taken into account. Such calculations are comparatively... 

The carry-over of corrosion product radionuclides with the main steam in the direction of the turbine is effected, on the one hand, by droplet entrainment with the residual moisture content of the steam and, on the other, by steam volatility. Usually, droplet carry-over is the most significant transport mechanism however, the oxides of the primary system metals show a measurable solubility in steam even at BWR operating conditions. At different plants, concentrations of dissolved cobalt on the order of 60 ng/kg were measured in condensed samples of main steam, i. e. significantly higher than could be explained by droplet entrainment (e. g. Hepp et al., 1986). These observations are consistent with the fundamental results on steam volatility of weakly dissociated compounds under BWR operating conditions which were reported by Styrikovich and Martynova (1963). Since only non-dissociated substances are volatile with steam, it has to be assumed that a fraction of the cobalt present as dissolved ions in the reactor water at ambient temperature is converted to non-dissociated oxide, hydroxide or ferrite at the plant operating temperature. 

The approach to BMN-170 plant safety is based on the concept of retaining radionuclides in the fuel during normal operation and accidents, so that radiation impact on personnel and population in the NPP area lies within allowable limits. Barriers for fission products release are mainly claddings of fuel elements and reliable operation is provided under the specified operating conditions. 

Indeed the overall effect of such chemicals on the surface chemistry of soils and sediments remains poorly understood. Despite the difficulties this type of technique is still widely used. As with many operationally defined methods, it is difficult to make direct comparisons between independent studies however, for individual studies sequential extraction can play a useful role in an overall programme of study regarding the chemical forms of radionuclides bound to the solid phase under different environmental conditions so long as it is realised that in this type of study the species are themselves really defined by the nature of the extraction system employed (see Chapter 10). 

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