II–1. STRUCTURE OF POTENTIAL EXPOSURES CONSIDERATION The regulations of Belarus relevant to the potential exposure comprise:
Safety requirements to ensure that the activities relating to the construction, operation and decommissioning of facilities are conducted to achieve the highest standards of safety that can be reasonably achieved;
Risk criteria which address the risk of mortality and of cancer from nuclear installations;
Dose and risk constraints for planned exposure situations and reference levels for emergency and existing exposure situations;
Emergency preparedness and response planning to mitigate the consequences of nuclear accidents.
For the NPP-2006 design which is being constructed in Belarus the acceptance criteria for design basis accidents are defined in accordance with the Russian Federation requirements [II–
1]:
Less than 1 mSv/event for the accidents with probability higher than 10-4 events/a;
Less than 5 mSv/event for the accidents with probability lower than 10-4 events/a.
The calculated probability of a beyond design basis accident for the Belarusian NPP is lower than 10-6 a-1.
TABLE II–1. NOMINAL RISK COEFFICIENTS TAKING INTO ACCOUNT CANCER RISK AND RISK OF HERITABLE EFFECTS [II–2]
Exposed population Cancer·10-2 Sv-1 Heritable effects·10-2 Sv-1 Total·10-2 Sv-1
Whole 5.5 0.2 5.7
Adults (Workers) 4.1 0.1 4.2
TABLE II–2. GENERIC CRITERIA FOR ACTIONS IN EMERGENCY EXPOSURE SITUATIONS TO REDUCE THE RISK OF STOCHASTIC EFFECTS [II–2]
Generic criteria Examples of protective actions and other response actions
Criteria for urgent protective actions
Hthyroid 50 mSv in the first 7 days Iodine thyroid blocking E 100 mSv in the first 7 days Sheltering; evacuation;
decontamination; restriction of consumption of food, milk and water;
contamination test; public information
Hfetus 100 mSv for the full period
of in utero development
Screening based on equivalent doses to specific radiosensitive organs (as a basis for medical follow-up), consulting
Hfetus 100 mSv for the full period
of in utero development
Counselling to allow informed decisions to be made in individual circumstances
Note: E – effective dose, H – equivalent dose
Ref [II–3] states that the risk of death for the public living in the vicinity of an NPP caused by a reactor accident should not exceed 0.1 % of the sum of all risks of death caused by other
accidents. Belarus regulation Ref [II–2] defines the nominal risk coefficients as sex-averaged and age-at-exposure-averaged lifetime risk estimates for a representative population. These coefficients are presented in Table II–1. Ref [II–2] further introduces:
Generic criteria for protective actions and other response actions in the emergency exposure situations to reduce the risk of stochastic health effects (Table II–2);
Generic criteria for acute doses for which protective actions and other response actions are to be undertaken under any circumstances, to avoid or to minimize severe deterministic health effects (Table II–3);
Guidance values for limiting exposure of emergency workers;
Limited dose rates for protective actions in radiation emergency exposure situation.
TABLE II–3. GENERIC CRITERIA FOR ACUTE DOSES FOR WHICH ACTIONS ARE TO BE TAKEN TO AVOID OR MINIMIZE SEVERE DETERMINISTIC EFFECTS [II–2]
External acute exposure (less than 10 hours)
Internal exposure in 30 days
AD Red marrow 1 Gy 0.2 Gy for radionuclides with atomic number Z ≥ 90;
2 Gy for radionuclides with atomic number Z ≤ 89
AD Fetus 0.1 Gy 0.1 Gy intervention levels for food, milk and drinking water.
II–2. DESCRIPTION OF MODELS OR METHODOLOGIES APPLIED
The International Radiological Assessment System (InterRAS) was used for dose assessment.
InterRAS is a computer-based tool that was developed for the IAEA to assist with the technical assessment of nuclear reactor accidents for the purpose of determining protective actions for the public and emergency workers. It is based on the RASCAL code Ref [II–5] and is consistent with the generic assessment procedures presented in Ref [II–6].
InterRAS is a set of three computer-based tools: Decay Calculator, Field Measurement to Dose and Source Term to Dose Ref [II–7]. The source term to dose model (ST-DOSE) is considered as the primary tool of InterRAS. It is designed to provide a rapid assessment of potential consequences from a set of information about the plant conditions or source term and meteorology at the accident site. The model generates estimates of integrated dose.
There are six ways of setting the source term. Three assume some sort of measurement of the radionuclide mix. The remaining three source term methods are specifically for use with reactor accidents. They generate a radionuclide mix based on the information provided about plant conditions, containment monitor reading, or irradiated nuclear fuel condition.
The 5 entries on the event time screen describe the sequence of the release events and also define when the cumulative dose should be calculated: Shutdown, Release to containment, Release to environment, End of release, End of calculation.
Meteorological conditions are required to be defined for the time of the radioactive release to the environment. ST-DOSE requires a minimum of one set of meteorological data to operate.
There is a default set (wind is 3 m/s from the west, D stability, 500 m mixing height, no
precipitation) which is used if the user makes no changes. Up to 4 sets of meteorological data may be entered. Effective release height as well as the release location must be defined by the modeller.
After defining the problem, the ST-DOSE model automatically generates a source term, models the transport and diffusion of the material in the environment, and estimates integrated doses.
The process of calculation is briefly described in the following paragraphs.
The first step in the calculation process is the generation of a source term if the user has not explicitly defined the isotopic mix. This mix of radionuclides starts with the core inventory and the severity of core damage. As needed, the mix is decayed, and reduction factors are applied to account for the removal processes (e.g. filters, sprays, and holdup). A final computed source term is created to be released at the specified leak rate over the specified release interval.
Depending on the distance from the point of release two transport and diffusion models are used. Close to the release point (inside 5 km) a straight-line Gaussian plume model is used. It computes doses at receptors arrayed on a polar grid, spaced every 10 degrees around at distances of 1, 2 and 5 km each. This grid provides better resolution close to the release point. Beyond 5 km a Lagrangian trajectory Gaussian puff model is assumed. This model computes doses at receptors arrayed in a square, 31x31 Cartesian grid with a 50 km radius. This gives a 3.33 km spacing between the receptor points.
Doses are calculated at the end of exposure which is to be set by the user. Inhalation doses are computed from the time-integrated air concentrations using dose factors and the breathing rates.
Ground shine dose is computed from the cumulative surface concentration assuming a surface correction factor of 0.7. Cloud-shine is computed using a finite puff approximation near the source; switching to a semi-infinite cloud model when the horizontal diffusion coefficient (sigma-y) exceeds 400 m. Cs-137 deposition can also be computed.
II–3. REPRESENTATIVE PERSON/CRITICAL GROUP FOR THE EXERCISE
To define the representative person for assessing doses in case of an emergency the data about population distribution around the nuclear installation is used together with the data about highest doses (TEDE and, in some cases, Thyroid doses).
For the purpose of dose assessment one age group is normally used (adults), but in some cases one more age group could be considered (1-year old children). For an assessment of the impact of accidental releases, the following doses are usually considered for atmospheric discharges:
Cloudshine dose from the plume;
Groundshine dose from deposited radionuclides;
Inhalation dose from the plume;
Ingestion dose;
Total effective dose;
Thyroid dose.Habit data for dose assessment must be chosen according to the survey data of the region of interest or, if not available, national statistical data can be used. In principle, preliminary dose assessment may be based on conservative assumptions (like consumption of local food only, no sheltering, etc.), whilst further assessment may include more realistic and site-specific information.
The actual location of the population is also taken into account. According to Belarusian legislation a Sanitary Protective Zone (SPZ) which is an area where dwellings, or any kind of
recreational or economic activities, are strictly prohibited is established around each nuclear installation. The size of such an SPZ depends on the type of installation and on estimated levels of public exposure during normal operation of the NPP. The representative person for assessing doses is normally assumed to be the person most affected by the discharges, located in the areas around the NPP. Thus, in Belarus, a representative person for assessing doses in case of potential emergency will be the person beyond the SPZ and receiving the maximum individual effective dose.
In the current exercise the maximum calculated dose was observed in the area 7, but as it is located at the distance 1 km from the NPP (and 1-km radius will be within the SPZ), the representative person for assessing doses is the person located in the area 29.
II–4. RESULTS OF ASSESSMENT
Dose assessment was made with the help of InterRAS tool [II–7]. As described previously in this report, in the InterRAS up to 4 sets of meteorological data can be entered, so all the meteorological data were averaged over a year and 4 meteorological scenarios were produced (Table II–4). The total effective dose calculated at different locations for different meteorological scenarios is presented in Table II–5. The total effective dose is assumed to be a combination of the effective dose from inhalation and the dose from cloud-shine and 7-day ground-shine exposure.
Precipitation None None Light rain None
TABLE II–5. TOTAL EFFECTIVE DOSE, Sv Scenario/
The maximum doses (see Table II–6) were observed in case of meteorological scenario 1.
9At the same time, there could be exceptional cases when people from the category “public” could be located
TABLE II–6. MAXIMUM DOSES FOR ADULTS (FIRST 7 DAYS), Sv
Distance, km 0.3 1 2.5 3.5 5 6.5 7 8 10
Total Effective dose 1.1E+01 2.6E+00 1.2E+00 8.7E-01 6.5E-01 5.2E-01 4.9E-01 4.1E-01 3.3E-01 Thyroid CDE 7.4E+02 2.4E+01 1.1E+01 8.0E+00 6.2E+00 5.1E+00 5.0E+00 4.4E+00 3.5E+00 Inhalation Eff 3.0E+01 9.9E-01 4.4E-01 3.4E-01 2.6E-01 2.1E-01 2.1E-01 1.8E-01 1.5E-01 Cloud-shine 5.3E+00 6.9E-01 3.1E-01 2.3E-01 1.6E-01 1.2E-01 9.4E-02 7.1E-02 5.9E-02 7-day Ground-shine 2.8E+01 9.1E-01 4.0E-01 3.1E-01 2.3E-01 1.9E-01 1.9E-01 1.6E-01 1.3E-01 ET per Annum EP 4.8E+01 1.6E+00 6.8E-01 5.2E-01 4.0E-01 3.3E-01 3.3E-01 2.8E-01 2.3E-01
TABLE II–6. MAXIMUM DOSES FOR ADULTS (FIRST 7 DAYS), Sv (cont.)
Distance, km 15 20 25 30 35 40 45 50
Total Effective dose 1.8E-01 1.3E-01 8.9E-02 6.4E-02 4.7E-02 4.0E-02 3.0E-02 2.3E-02 Thyroid CDE 1.7E+00 1.3E+00 9.6E-01 7.3E-01 5.5E-01 4.8E-01 3.6E-01 3.0E-01 Inhalation Eff 7.2E-02 5.4E-02 4.0E-02 3.0E-02 2.2E-02 2.0E-02 1.5E-02 1.2E-02 Cloud-shine 5.6E-02 3.8E-02 2.6E-02 1.7E-02 1.2E-02 1.0E-02 6.7E-03 5.5E-03 7-day Ground-shine 5.0E-02 3.5E-02 2.4E-02 1.7E-02 1.2E-02 1.0E-02 7.1E-03 5.6E-03 ET per Annum EP 8.7E-02 6.1E-02 4.3E-03 3.0E-02 2.2E-02 1.8E-02 1.3E-02 1.0E-02
REFERENCES TO ANNEX II
[II–1] ROSATOM, AES-2006. Terms of reference for the concept design development (in Russian), Rosatom, Moscow (2006)
[II–2] MINISTRY OF HEALTH OF BELARUS, Criteria for Radiation Impact Assessment, Hygienic standard (in Russian), Ministry of Health of Belarus, Minsk (2012).
http://radbez.bsmu.by/library/GN_2012.pdf
[II–3] MINISTRY OF HEALTH OF BELARUS, Requirements to the radiation safety, Sanitary standard and regulation (in Russian), Ministry of Health of Belarus, Minsk (2012).
http://minzdrav.gov.by/ru/static/acts/normativnye/postanovlenia_ministerstva
[II–4] THE COUNCIL OF MINISTERS OF THE REPUBLIC OF BELARUS, On the approval of the protective action plan for a radiation accident at a Belarusian nuclear power plant (external emergency plan), Resolution № 211 (in Russian), The Council of Ministers of the Republic of Belarus, Minsk (2018).
[II–5] U.S. NUCLEAR REGULATORY COMMISSION, RASCAL 3.0.5: Description of Models and Methods. NUREG-1887. DC 20555-0001. U.S. NRC, Washington (2007).
[II–6] INTERNATIONAL ATOMIC ENERGY AGENCY, Generic Assessment Procedures for Determining Protective Actions during a Reactor Accident. IAEA-TECDOC-955. IAEA, Vienna (1997).
[II–7] INTERNATIONAL ATOMIC ENERGY AGENCY, INTERRAS 1.2. IAEA, Vienna (2000).