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3. DERIVATION OF THE SOURCE TERM

3.1. Factors influencing the source term

The release of radioactive substances from a research reactor to the environment (the source term) depends on the following factors:

— The inventory of fission products and other radionuclides in the core (or the inventory in experimental devices or other locations such as the spent fuel pool or isotope production facilities);

— The progression of core damage (or failure of experimental devices or isotope production facilities);

— The fraction of radionuclides released from the fuel (or from experi-mental devices or other locations), and the physical and chemical forms of released radioactive materials;

— The retention of radionuclides in the primary cooling system;

— The performance of means of confinement (e.g. emergency ventilation rate, filter efficiency, leak rate, liquid effluent release rate, radioactive decay due to time delay of release, deposition on surfaces and resuspension).

In addition, the doses associated with the source term depend on the release mode (single puff, intermittent, continuous) and the release point (stack, ground level, confinement bypass). The above factors are taken into

consideration by the methods presented in detail in this section; first, however, it is pertinent to discuss the following general considerations.

3.1.1. General considerations for source term calculations

Over the past few decades, considerable work has been undertaken to realistically evaluate source terms for core meltdown accident sequences for power reactors and, in some cases, for research reactors. Several important conclusions have been drawn from these research efforts [15]:

— Fuel inventory source terms can be calculated reasonably well using analytical methods.

— Generally, source terms have tended to be grossly overestimated.

— Source terms are facility specific; generalization of source term values from one reactor to another is not possible; each research reactor must be considered individually according to its design and mitigatory features.

— The magnitude of releases depends on the research reactor performance characteristics, in particular the performance of the means of confinement.

— Uncertainties may be significant in predicting building release source terms, and the significance of many of these uncertainties varies with the timing and mode of confinement failure.

— Realistic assumptions will usually result in lower source terms and less severe consequences.

The level of detail to which the safety analysis, including the derivation of the source term, has to be carried out will depend on the potential hazard presented by the research reactor, which to some extent depends on the operating power, the fission product inventories, the engineered and inherent (physical) safety features, and the nature of the accident sequences that are considered credible. The sophistication of the calculation methods employed to determine source terms and doses will be influenced by the magnitude of the potential consequences, the complexity of the techniques, the availability of data and codes, and the regulatory requirements. Where realistic assumptions or data are not readily available, conservative assumptions are to be used.

Conservative assumptions will greatly simplify the calculation effort, but often can lead to predicted consequences that are more severe than can realisti-cally be expected. Realistic assumptions, on the other hand, will usually result in source terms and consequences that are less severe. However, in some instances the use of realistic data and assumptions will be complex and may involve substantial effort. This is particularly true for determining fission product

releases within the reactor, the primary cooling system and the reactor building.

In addition, the results of such calculations are highly specific to the individual reactor/facility design and to particular accident sequences. To avoid sophisti-cated calculations (unless they are considered necessary to determine a realistic source term for a particular reactor, mainly owing to potentially severe conse-quences), the use of conservative release fractions may be preferable. For the same reason, the use of various conservative assumptions often is a regulatory requirement. In other areas, however, realistic calculation methods are normally used, and validated computer codes are available. Examples include the determi-nation of realistic fission product inventories, atmospheric dispersion modelling and dose calculations.

It is also usual to assess reactor specific building/containment designs and associated ESFs on a case-by-case basis. Performance of confinement and air cleaning features can be crucial to mitigating releases, even where the building is not designed for pressure loads as a tight containment system. This point is apparent from the limited releases associated with the severe reactivity excursion accidents at the NRX reactor at Chalk River [16] and the SL-1 reactor at Idaho Falls [17], and others discussed in Ref. [18]. Confinement performance is specific to the facility design and the nature of the accident sequence. Event tree analysis can be used to assess the structural and system response and performance to ultimately characterize release mechanisms.

As research reactors usually have much lower fission product inventories than power reactors, as well as less stringent requirements for post-shutdown heat removal, many of the more complex analytical techniques have not generally been applied to research reactor analyses, and the methods currently applied often are relatively simple. The following factors should be considered in determining the need for an extensive analysis:

— Reactor type, associated irradiation and experimental facilities, and specific hazards posed by them;

— Reactor design features, including fuel type, coolant/moderator systems and other features influencing releases to the building;

— Reactor operating history and modifications made, if any;

— Type of accident sequence and factors that would influence releases (fire, pressure pulse, loss of pool water, structural integrity of the confinement building, operator actions, etc.);

— Availability of validated modelling methods and empirical data for realistic release assumptions.

Fuel melting as a result of LOCAs or LOFAs is not a concern for low power research reactors. However, the extent of the analysis required for

higher power research reactors will depend critically on the reactor design and the limiting DBA. Design features such as routing of coolant piping and beam tube locations will influence the fraction of the core exposed to air during a LOCA, and thus the fraction of the core that might degrade. This complexity arises because it is more usual to assume overly large, conservative release fractions and to show that the resulting consequences are acceptable. However, the use of realistic assumptions, supported by experimental or accident data, may substantially reduce the source term.

Some estimates of fission product releases may be obtained from experi-mental or accident data. However, this type of information may be highly specific to the reactor type and to the experimental or accident sequence conditions. Thus, extrapolations are to be made with caution. Some examples of radioactive releases from materials testing reactors (MTRs) are given in Appendix II.

Factors important in determining the source term — namely, fission product inventory, fuel damage progression, radionuclides released from the fuel, retention of fission products in the primary coolant and confinement performance — are discussed in detail below.

3.1.2. Atmospheric dispersion modelling

The dispersion and deposition of material released to the atmosphere are typically modelled as a plume. One such model is based on a Gaussian plume [19] with Pasquill–Gifford dispersion parameters. Simple plume models can simulate phenomena such as buoyant plume rise, wake effects on plume dispersion caused by obstructions such as buildings, and wet and dry deposition. Time dependent radioactive buildup and decay in the plume can also be calculated. Some newly developed codes on accidental dispersion, transportation and deposition of radioactive material are described in Refs [20, 21].

3.2. FISSION PRODUCT INVENTORY