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4. RADIOLOGICAL CONSEQUENCES

4.3. On-site consequences outside the reactor building

OUTSIDE THE REACTOR BUILDING

In the event of radioactive releases to the reactor building interior, atmospheric releases to the exterior of the reactor building, apart from external direct radiation to the site, may occur. In particular, such releases may occur when the reactor building is not designed as a means of confinement. In this case, contamination in the vicinity of the building due to leakage through windows and doors may have to be considered in the consequence analysis.

Radioactive liquid spills may cause contamination in the vicinity of the building if no systems for draining liquid releases to liquid waste collection tanks are included in the design. It is important that consideration be given to the possibility of liquid radioactive material bypassing the designed barriers, or the possible relocation of radioactive material to areas with no retention or less retention than foreseen in the designed safety features. For screening purposes, the total critical group dose due to a particular accident scenario (or source term) is taken as the sum over the doses from all pathways and all radio-nuclides included in the assumed source term. One of these contributions is the dose due to external direct radiation fields.

It is clear that the relative importance of this contribution will depend on many factors; usually, it will be most important at points inside the reactor containment or outside but close to the reactor building, with decreasing importance as the evaluation point is moved farther from the facility.

It is advisable that, at this point, a detailed description of the source term be available. This will define the radiation source for the radiation transport calculations to be done.

On-site consequences may include release of radionuclides in gaseous, liquid or, in some cases, particulate form from the reactor building, and direct radiation from the containment. One of the on-site radiological hazards is intense direct gamma radiation from the reactor building/containment following a release to the building of even a relatively small fraction of core inventory. Dose determination will involve shielding calculations for a volume source and is both reactor and accident specific, depending on the nature of the release, the type of radionuclide released, the radionuclide distribution within the facility, and both the geometry and the materials of the building structure.

Liquid releases from the reactor building may have on-site consequences but usually do not pose a hazard comparable with that posed by gaseous releases. They need to be considered, however, if one of the following hazards cannot be excluded:

— Direct radiation from the released liquid;

— Evaporation of substantial amounts of radionuclides from the released liquid;

— Contamination of aquifers.

Usually, liquid releases occur to some type of reservoir where the liquid is collected before release. This reduces the risk of off-site contamination and leaves more time for enacting countermeasures. However, it should be considered that liquid releases of fractions of the core may produce significant radiation levels in areas not well protected or shielded. This may pose great problems concerning treatment of the radioactive liquid. Therefore, sufficient storage capacity with adequate shielding needs to be considered in the accident analysis.

4.3.1. On-site exposure resulting from containment release

Doses due to external irradiation from the containment on the site or in cases where the reactor building is only a short distance from the site boundary can be calculated using the shielding methodology discussed for consequences inside the building. In this case, the geometric modelling is not so important, and it is usually acceptable to use simplified geometric models.

For distances greater than 100 m, a simplified approach that treats the radioactivity inside the building as a point source is sufficiently accurate for the purpose of dose estimates. Furthermore, taking into account the accuracy of the evaluations of radioactive releases to the containment, the error caused by assuming a point source instead of a volume source may be acceptable at shorter distances as well. At points very close to the reactor building (i.e. for areas just outside the building, if this is a realistic exposure possibility to persons on the site), calculations ideally will take into account the actual volume source.

4.3.2. External exposure from airborne radioactive material

Because of the complexity of airflow near a building, there is no single model that is applicable to all situations. The models discussed in Refs [57, 59]

are applicable in the case of a source on or just above the building roof, or within the recirculation wake.

The effective dose rate from immersion in a radioactive cloud (for each radionuclide) of limited volume (Eim (Sv/h)) is given by:

Eim = CADFim (15)

where

CA is the specific activity of the radionuclide in air (Bq/m3);

DFim is the effective dose coefficient for immersion (Sv/h per Bq/m3).

The effective dose coefficients for immersion (for gamma and beta radiation) are summarized in Ref. [57]. To calculate the effective dose, the dose rate Eim must be integrated for the time period of exposure.

4.3.3. External exposure from deposited radioactive material

The effective dose rate from the ground deposition (for each radionuclide) (Egr (Sv/h)) is given by:

Egr = CgrDFgr (16)

where

Cgr is the surface specific activity of the radionuclide on the ground (Bq/m2);

DFgr is the effective dose coefficient for ground deposition (Sv/h per Bq/m2).

The effective dose coefficients for ground deposition are summarized in Ref.

[57]. To calculate the effective dose, the dose rate Egr must be integrated for the time period of exposure.

4.3.4. Internal exposure from airborne radioactive material

Internal doses arising from the inhalation of radionuclides are a function of the time integrated activity concentration of the air at the location of interest:

Hin = CBDin (17)

where

Hin is the dose in the specified organ (Sv);

C is the integrated activity concentration (Bq·s·m–3);

B is the breathing rate (m3/s);

Din is the inhalation dose conversion factor (Sv/Bq).

The breathing rate for an adult can be assumed to be 3.30 × 10–4 m3/s [19]. The inhalation dose conversion factors are summarized in Refs [57, 59, 60].

Exposure to dose conversion factors are available for a wide variety of tissues and organs, such as bone marrow, bone surface, breast, gonad, lung, skin, spleen, stomach, thyroid and uterus.

4.3.5. External exposure due to liquid releases

External doses due to liquid releases from the reactor building usually do not pose a hazard comparable with that posed by gaseous releases. They need to be considered, however, if any one of the following hazards cannot be excluded:

— Direct radiation from the released fluid;

— Evaporation of substantial amounts of radionuclides from the released fluid;

— Contamination of aquifers (on the site or off the site).

Usually, liquid releases occur to some type of reservoir where the liquid is collected before release. This reduces the risk of off-site contamination and increases the length of time for enacting countermeasures. However, it must be considered that liquid releases of fractions of the core may produce significant radiation levels in areas that are not well protected or shielded. This may pose great problems during treatment of the radioactive liquid. Therefore, it is important that sufficient storage capacity with adequate shielding be considered in the accident analysis.

Liquid releases into the reactor building are usually collected at special dedicated and shielded reservoirs designed to contain a specified activity. For a given accident scenario, the shielding provided must be verified. If such reservoirs are not available at the facility, doses due to liquid releases into the reactor building must also be assessed. The activity concentration of the liquid release must be determined; a simplified analytical point kernel buildup factor method with the estimated equivalent surface source can be used to evaluate the dose rate.

When evaporation releases substantial amounts of radionuclides to the atmosphere, the radiation source can be determined by using the appropriate atmospheric dispersion models [59].