Simplified environmental analysis (Sustainability Assessment)

This page is the "Appendix II" to Environmental Impact of Stressors

This appendix presents practical advice on how to perform simplified environmental analysis of radiological stressors based on IAEA and United States Department of Energy (DOE) methods. It further discusses briefly the analysis of the consequences of toxic chemical releases and presents the results of comprehensive environmental studies.

Introduction

The INPRO assessment approach using a comparable current facility which was presented for UR1 is not applicable when:

• New radionuclides, chemicals or other stressors different from current facilities exist in an innovatively

designed facility of the NES assessed;

• The level of a stressor in the NES facility assessed is higher than the value of a comparable current facility;
• Significant differences exist between the environmental characteristics of the site of the current facility and the one proposed for the NES facility;
• The regulatory requirements used for the current facility are significantly less stringent than those in the country of the NES assessed.

The INPRO assessor should use a simplified environmental analysis approach when an EIA is not yet available for the NES facility under assessment and a comparable current facility cannot be identified. The regulatory requirements relevant to the environmental impact of stressors usually relate to the potential impacts of pollutants or radionuclides than to the levels of stressors, i.e. the emission rates. In the case of radiological stressors, the regulatory standards are defined based on the impact of radiation in terms of exposures. The determination of the impact of stressors, e.g. doses, is required to compare it to current regulatory standards (e.g. dose limits).
Environmental standards are defined by Member States. In particular, for radioactive emissions from nuclear facilities, the regulatory body of a Member State is responsible for identifying the values of the dose constraints and sets discharge authorizations for nuclear facilities or sites, which are usually given in terms of annual limits of discharge in licensing and operation permits.
Summarizing the statements above, in such cases where no comparable current facility for an NES facility to be assessed can be found, the INPRO assessor should demonstrate that ALs for the stressors of the NES facility (current environmental standards including dose constraints in the country where the facility will be installed) are met, and the environmental impact of stressors should be estimated.
Simplified methods on how to calculate the environmental impacts (doses) of emitted radionuclides on humans and non-human biota (plants and animals) are presented in next two sections.

IAEA generic models for radiological impact on humans

A simplified calculation based on the IAEA generic models described in Ref. could be conducted to demonstrate compliance with dose limits (impacts of radiological stressors). This method, called a tiered approach, consists of a sequence of several tiers with increasing complexity and detail and is discussed in this section.
Similar simplified approaches have been proposed for calculation of impacts from radionuclides in different contexts. For example, in the DOE graded approach (DOE-GA) model (see Section 3.3), the RESRAD family of codes includes a tool for the calculation of doses to humans, which might be used for comparison with the calculations performed using the IAEA generic models. Reference examples of the application of these simplified approaches can be found in Refs for humans and in Refs, for biota.
With regard to doses to humans caused by the discharge of radionuclides into the environment, the IAEA has developed generic models for the purpose of screening of discharges. They are designed to determine the likely magnitude of the radiological impacts through a simplified but conservative method. The use of simple screening models for dose calculation is usually one of the first steps in licensing a nuclear facility.
Originally, these methods were expected to be applied mainly for small scale facilities in order to facilitate easy calculation at low cost, to determine whether further, more refined and expensive analyses could be avoided. According to Ref.:

“However, for many larger scale nuclear facilities the assessed doses from the screening models presented in this report are more likely to approach the dose limiting criteria set by the Regulatory Authority (e.g. dose constraint), and users are more likely to need to follow a screening calculation with a more realistic, site specific and detailed assessment.”

The goal of the method presented in Ref. is to preliminary calculate doses to the most exposed individuals in hypothetical critical groups. To achieve this goal, the last year of discharge (assumed the 30th year of operation in Ref.) is considered in order to maximize doses; the effective doses from external and internal radiation per unit of concentration of each radionuclide are then calculated.
The application of the method described in Ref. is relatively straightforward. The basic information required is:

• Quantities and types of radionuclides discharged;
• Mode of discharge and the discharge points, identifying separately discharges to the atmosphere, surface water or sewerage systems.

To ensure that the analysis is consistent, to the extent possible, with the approach for the biological effects on non-human biota based on the DOE-GA (see Section 3.3), it was decided to include both steps described in Ref. into the first level of the method, called tier 1.

Generic IAEA model tier 1: No dilution and generic environment

In tier 1, a calculation of effective individual doses (during the final year, which, in Ref., is assumed to be the 30th year of discharge) from external and internal exposure to the most exposed individuals in a hypothetical critical group is to be performed. There are two steps (or iterations).
Step 1 of tier 1: Application of the ‘No dilution (no dispersion) model’
The maximum predicted annual average radionuclide concentrations in Bq/m³ at the point of discharge in either air or water are to be used. Effective doses to a hypothetical critical group can be calculated by multiplying the concentrations at the point of discharge by dose factors (or dose coefficients) in Sv/a per Bq/m³, which are different for discharges to air and surface water. These are listed in tables I–I and I–II of Ref., and Excel sheets with these factors are available from the IAEA. These coefficients were derived using established methodologies, default parameters and dose coefficients.
The most exposed individual is hypothesized to be located at the discharge point. The maximum predicted annual average radionuclide concentrations in either the air or water at the point of discharge are considered. No dilution screening factors are used to convert exposures into doses.
The calculated maximum dose is compared with the acceptance limit, which is the dose constraint (see Section 2.3) for a specific nuclear facility. If no compliance with the acceptance limit is found, the analysis should be refined with step 2.
Step 2 of tier 1: Application of the ‘generic environmental model’
The concentrations at the location nearest to the facility at which a member of the public will be likely to have access, or from which a member of the public may obtain food or water, are needed in order to see whether they match the generic assumptions. The analyst can easily calculate doses for the generic environmental model by multiplying release rates with generic dose calculation factors (see tables I–III and I–V of Ref.), which are given in units of Sv/a per Bq/s for discharges into the atmosphere and surface water, and in Sv/a per Bq/a for discharges into sewers.
For the background calculations, the simplified methods for estimating radionuclide concentrations in air are outlined in Ref. and will not be reported here. The average radionuclide concentrations in air, food and water are combined with the annual rates of intake to obtain an estimate of the total radionuclide intake during that year. This total intake over the year is then multiplied by appropriate dose coefficients to obtain an estimate of the maximum effective dose, in one year, from inhalation or ingestion. In a similar manner, the concentrations of radionuclides in shoreline sediments and surface soils are used with appropriate dose coefficients to estimate the effective dose received from external irradiation during the final operational year of the facility. The effective dose, in one year, from immersion in a cloud containing radionuclides may be calculated by multiplying the average concentration in air by the appropriate external dose coefficients. To obtain the total maximum effective dose in the final operational year, the effective doses from all radionuclides and exposure pathways are summed.
Dilution and dispersion are somewhat taken into account; generic factors are derived using additional standardized assumptions about the discharge characteristics and location of a hypothetical critical group. Similar to the no dilution model, the calculated maximum dose is to be compared to an acceptance limit to confirm acceptable results. As suggested in section 8 of Ref., for the acceptance limit for the generic environment model, a more stringent value of dose should be used, i.e. one tenth of the dose constraint (see Section 2.3).
If no compliance with the acceptance limit (e.g. one tenth of the dose constraint) is met, Ref. suggests that the site specific discharge conditions and the actual critical group location be taken into account. Within an analysis, this step would correspond to tier 2.

Generic IAEA model tier 2: Use of site specific characteristics

Application of tier 2 depends on the maturity level of the NES facility being analysed. If no specific environmental data of a proposed site exist, or if the NES has a too low level of maturity, tier 2 cannot be applied, and the need for RD&D should be identified by the analyst. If the maturity level is medium to high, the analysis could be refined by appropriately adjusting the key parameters to the specific situation (e.g. more specific regional dispersion parameters, local/regional food chains) or by using more refined pathway analysis tools. In the case where the site is known, the analyst could consider actual discharge conditions, take into account the location of the actual critical group and make use of specific parameters. If no compliance is achieved in this step, the analyst should identify and propose appropriate RD&D. Default values used within Ref. can be changed to adapt to the characteristics of the region or site and to suit the needs of the analysis.

Dose constraints for radiation protection of humans

In this section, the INPRO AL for impact of radiological stressors on humans is defined. This AL for radiation protection should be used in the case when no comparison of facilities is necessary, i.e. when an EIA is available or a simplified analysis has been applied (as discussed in Introduction).
Table 1 (taken from WS-G-2.3) shows values of annual dose constraints for a single source (facility) used in current practices by different Member States. Based on the consideration that the reported values are within a relatively narrow range, WS-G-2.3 recommends that the annual dose constraint should not exceed 0.3 mSv (see also Ref.).

Table 1. Dose constraints used in current practices by member states

Member state	Dose constraint (mSv/a)	Source
Argentina	0.3	Nuclear fuel cycle facilities
Belgium	0.25	Nuclear reactors
China	0.25	Nuclear power plants
Italy	0.1	Pressurized water reactors
Luxembourg	0.3	Nuclear fuel cycle facilities
Netherlands	0.3	Nuclear fuel cycle facilities
Spain	0.3	Nuclear fuel cycle facilities
Sweden	0.1	Nuclear power reactors
Ukraine	0.08	Nuclear power reactors
	0.02	Nuclear fuel cycle facilities
United Kingdom	0.3	Nuclear fuel cycle facilities
United States of America	0.25	Nuclear fuel cycle facilities

Noting that the INPRO methodology demands that the NES analysed should demonstrate an improved overall environmental performance compared to current systems, an assessor should strive for lower dose constraints than those used today.
Thus, for example, in the case where a level of a radiological stressor of the NES is higher than that in a comparable current NES or a new stressor exists in the NES analysed, the assessor should set the dose constraints for a single facility (component) of the NES assessed towards the minimum of the values in Table 1.
This leads to values recommended to be used as INPRO ALs, for the cases when dose constraints are not defined, as follows:

• A value of 0.08 mSv/a for nuclear power reactors;
• A value of 0.2 mSv/a for all other nuclear facilities of the NES assessed.

Method to set dose constraints in the United Kingdom

The safety assessment principles (SAPs) of the UK Office for Nuclear Regulation contain a set of principles that are structured into 12 separate sections, including, among others, fundamental principles, engineering principles, numerical targets and legal limits. The SAPs include eight fundamental principles that are “founded in UK health and safety law and international good practice”. They embody the requirements for the ALARP concept (see Appendix III) to be applied to radiation exposures resulting from normal operation and to the risks from accidents. The eight fundamental principles are:

(a) “The prime responsibility for safety must rest with the person or organisation responsible for the facilities and activities that give rise to radiation risks.”
(b) “Effective leadership and management for safety must be established and sustained in organisations concerned with, and facilities and activities that give rise to, radiation risks.”
(c) “Protection must be optimised to provide the highest level of safety that is reasonably practicable.”
(d) “Dutyholders must demonstrate effective understanding and control of the hazards posed by a site or facility through a comprehensive and systematic process of safety assessment.”
(e) “Measures for controlling radiation risks must ensure that no individual bears an unacceptable risk of harm.”
(f) “All reasonably practicable steps must be taken to prevent and mitigate nuclear or radiation accidents.”
(g) “Arrangements must be made for emergency preparedness and response in case of nuclear or radiation incidents.”
(h) “People, present and future, must be adequately protected against radiation risks.”

The UK Health and Safety Executive (HSE) risk management philosophy is based on the decision making process “Reducing Risks, Protecting People”. This defines individual and societal risks to workers and the public which are so high that they are intolerable, and risks which are so low that they may be considered broadly acceptable and no further regulatory pressure to reduce risks further would be applied. Between these levels, the risks must be reduced to ALARP,.
Licensees are required to assess the average effective dose equivalent (including any committed effective dose equivalent) to specified classes of persons.
The limits on radioactive discharges are set on the basis of the ‘justified needs’ of the licensees, i.e. the licensees must make a case that the allowed discharges are necessary to allow safe and continued operation of the plant.
In setting limits, the Environment Agency, the Scottish Environment Protection Agency and the Northern Ireland Environmental Agency use monitoring, discharge and plant performance data to ensure that the radiation exposure of the public as a consequence of the discharges would be less than the dose constraints and limits set by the UK Government. Currently, these are:

• A single source constraint of 0.3 mSv/a for an individual nuclear installation. Reference explains that in WS-G-2.3 “the experience of numerous countries in establishing dose constraints has been reviewed; the values are mainly between 0.1 and 0.3 mSv, which can be considered a relatively narrow range. Based on this analysis the Guide recommends that as a rule the annual dose constraint should not exceed 0.3 mSv” (see fig. 3 of WS-G-2.3). If the lower dose of the interval was used as the basic limit (BL), a factor of ten lower value for the basic objective (BO) would mean 10 μSv/a, which is the exemption level shown in fig. 3 of WS-G-2.3, and which is thus fully consistent with the current approach for dose control.
• A site constraint of 0.5 mSv/a for a site comprising more than one source.
• A dose limit of 1.0 mSv/a from all sources of human-made radioactivity including the effects of past discharges but excluding medical exposure.

As a general policy, the limits in discharge authorizations are progressively reduced and are kept close to the levels of the actual discharges.

Concepts, models and tools for assessment of radiological impact on non-human biota

International Commission on Radiological Protection approach

Until 2003, the system for radiation protection of the environment was based on the assumption that the standards of environmental control needed to protect people to the degree currently thought desirable will ensure that other species are not put at risk. In Ref., the ICRP acknowledges that this system provided “indirect protection of the human habitat but that a more comprehensive approach to study the effects on, and thus the protection of, all living matter with respect to ionising radiation” should be developed. The 2007 recommendations of the ICRP effectively extend the system of protection to address protection of flora and fauna explicitly. These recommendations explore the objectives of environmental protection and explain the basis for the proposed RAP, which is a small set of hypothetical entities that are representative of animals and plants present in different environments (terrestrial, freshwater and marine) and which form the basis of a structured approach to the assessment of exposures to, and effects of, ionizing radiation.
The concept and use of RAPs are dealt with in more detail in Ref., which contains information on the assumed biology, dosimetry and the available effects database for these entities. A range of derived consideration reference levels (DCRLs) is also proposed for each of the RAPs as numerical guidance for evaluating the level of potential or actual radiological impacts and as an input to decision making. These values are defined in terms of bands of doses within which certain effects have been noted, with a focus on those that might have some impact on the population structures of the animals and plants under consideration. Reference was issued in 2009, and provided transfer parameters for the set of RAPs.
In 2014, the ICRP published the framework for protection of the environment and how it should be applied within the ICRP system of protection. The report discusses: the protection of animals and plants (biota) in their natural environment, and how it can be done with the use of RAPs; their DCRLs, which relate radiation effects to doses over and above their normal local background natural radiation levels; and different potential pathways of exposure. It provides information on the different types of exposure situation to which its recommendations apply and how reference values based on the use of DCRLs can be used to inform on the appropriate level of effort relevant to different exposure situations. Reference also describes existing types of environmental protection legislation currently in place in relation to large industrial sites and practices, and the various ways in which wildlife is protected from various threats arising from such sites.
Overall, these developments represent the latest recommendations and data that can be used for protection of biota species and are the basis for the Environmental Risk from Ionising Contaminants: Assessment and Management (ERICA) tool updates.

Environmental Risk from Ionising Contaminants: Assessment and Management tool

Organized between 2004 and 2007, the European Commission project ERICA developed the ERICA integrated approach to assessment and management of environmental risks from ionizing radiation, with emphasis on biota and ecosystems. The objectives of the ERICA project have been achieved through the development of a user friendly assessment tool with risk characterization methodologies coupled with communication strategies aimed at decision making.
The ERICA project was built partly on the achievements of the Framework for Assessment of Environmental Impact (FASSET) project, which provided a basic framework for the assessment of the environmental impact of radionuclides. The FASSET Radiation Effects Database (FRED) has been updated, and a part renamed FREDERICA is available on-line. ERICA moved towards a tiered assessment approach:

“Using the ERICA reference organisms, Tier 1 will make use of conservative assumptions to derive dose limiting activity concentrations in environmental media (e.g. soil, air or water) for comparing with measured or predicted environmental concentrations around the (proposed) site. The derivation of a screening activity concentration will make use of a number of evaluations of effects data but will be based on threshold values below which effects are unlikely to be observed.
“Tier 2 assessments will predict dose rates to the reference organisms in ERICA and these will be compared to the same dose rate as used to derive the screening level in Tier 1. However it is felt that assessors might find it useful to evaluate the types of effects that may occur in different wildlife groups for the predicted dose rates.…
“Tier 3 assessments will predict dose rates and the assessments will focus on the effects that may be observed within the different wildlife groups.”

Department of Energy graded approach and RESRAD-BIOTA tool

When a judgement based on a comparison with current comparable facilities/environmental conditions/ regulatory requirements is not possible for non-human biota, the INPRO assessor can use the DOE-GA, which is similar to the radiological impact on humans described in Section 2.
Table 2 shows an overview of the DOE model indicating the parts that should be applied in an environmental analysis. General screening may serve as tier 1, and site specific screening may serve as tier 2, as specified in Section 2 for radiological impact on humans.

Table 2. Overview of the department of energy graded approach for evaluating doses to aquatic and terrestrial biota (adapted from Ref.)

Phase*	Implementation
1. Datat assembly	Assemble environmental media data and define evaluation area
2. General screening	Compare media concentrations with the biota concentration guides
3(a). Analysis** — site specific screening	Employ site representative parameters and conditions
3(b). Analysis** — site specific analysis	Employ kinetic/allometric modelling tool
3(c). Analysis** — site specific biota dose assessment	Employ ecorisk framework

* - Only phases 1, 2 and 3(a) should be considered for the proposed first application of the method for a simplified environmental analysis in the framework of INPRO assessment.
** - ‘Analysis’ consists of three increasingly more detailed steps of analysis. The DOE-GA process has been designed in three phases.

Phase 1: Data assembly

This is a data assembly phase in which the evaluation area and its characteristics are defined, and radionuclide concentration data for water, sediments and soil are assembled for subsequent screening. This phase includes the consideration of species and media involved, generic site characteristics in case they are known or postulated, and, because measurements will not (yet) exist, calculated concentrations in air and water as well as in sediments (suspended, bottom and shore/beach sediments).

Phase 2: General screening (tier 1)

This is an easy to use general screening method that provides limiting radionuclide concentrations (biota concentration guides, BCGs) in soil, sediments and water such that the dose limits for protection of biota are not exceeded.

Phase 3: Site specific screening analysis (tier 2)

This is an analysis phase containing three increasingly more detailed steps comprising site specific screening, site specific analysis and site specific biota dose analysis. This phase is limited to the first step only, and the other steps could be used only for a more mature NES and if some knowledge of the likely sites exists or if generic/typical site characteristics are postulated. The BCGs are very conservative concentration limits for environmental media (water, sediment or soil) defined by the DOE-GA model for a general screening process. The BCGs are calculated in such a way that total doses (i.e. from internal as well as external exposures) received by real biota exposed to these concentrations are not expected ever to exceed biota dose limits. As the DOE-GA model aims at protecting ‘all biota, everywhere’, the BCGs are restrictive, and, in many circumstances, are overly conservative with regard to specific environments. The potential dose rate to a receptor from both internal and external exposures is calculated for the unit concentration of a radionuclide in each medium, i.e. for 1 Bq/kg of sediment for soil and 1 Bq/m³ for water. This potential dose rate is termed dose conversion factor, and is measured in Gy/a per Bq/kg. From the dose conversion factor, the concentration of the contaminant in each medium that will generate a dose limit (DL) can be calculated using a relation of the type:

${\mathrm {BCG} }_{i}\left(Bq\cdot {\mathrm {kg} }^{-1}\right)={\frac {DL\left(Gy\cdot \mathrm {a} ^{\mathrm {-1} }\right)}{{\mathrm {DCF} }_{i}\left({\frac {Gy\cdot \mathrm {a} ^{\mathrm {-1} }}{Bq\cdot {\mathrm {kg} }^{-1}}}\right)}}$; (4)

The dose limits to be used in analysis as ALs are discussed in Section 3.4. The limits are 10 mGy/d for aquatic animals and terrestrial plants, and 1 mGy/d for terrestrial animals. The total dose conversion factor is made of a part related to external exposure and a part related to internal exposure. When multiple radionuclides are present in multiple environmental media, the screening is based on a summation of the fractions Ci/BCGi for all radionuclides, where Ci is the concentration of the specific radionuclide i, for each medium involved, i.e. water and sediment for aquatic biota, and soil and water for terrestrial biota. To demonstrate compliance with the dose limit, the sum must be less than 1:

$\sum _{r}{({\frac {C_{i}}{{\rm {BCG}}_{i}}})}_{water}+\sum _{r}{({\frac {C_{i}}{{\rm {BCG}}_{i}}})}_{sediment}<1.0\$ ; (5)
$\sum _{r}{({\frac {C_{i}}{{\rm {BCG}}_{i}}})}_{soil}+\sum _{r}{({\frac {C_{i}}{{\rm {BCG}}_{i}}})}_{water}<1.0\$ ; (6)

BCGs are calculated for several radionuclides i, and implemented in the RAD-BCG Calculator and RESRAD-BIOTA.
Reference notes that empirically based parameters that describe concentrations of contaminants in an organism relative to the surrounding medium are available for many radionuclides for plant/soil, for aquatic species/water and, in some cases, for animal/soil or sediment. The advantage of using one of these lumped factors is that it allows the prediction of tissue concentration based on simple measurements of contamination in environmental media such as water, sediment and soil.
When the initial (over conservative) screening exceeds the defined dose limits (see Section 3.4), non-human biota may still, in reality, receive doses below such limits. This is because actual concentration ratios for a single radionuclide may range over several orders of magnitude, depending upon various characteristics of the environment. The more site specific screening phases of the DOE-GA model would allow an analyst to examine the case with increasingly more accurate representative parameters for the site and receptors.
The environmental analysis should not go very far with details; rather, it should be kept simple, especially for low maturity NESs. Reference or hypothetical sites may serve for simplified calculations, and assumed contamination levels may be either estimated on the basis of analogy with similar current conditions (justification is up to the analyst) or with simplified tools.
The methods of the DOE-GA have been encoded in a simple way, enabling the use of calculation tools such as the RAD-BCG Calculator (electronic spreadsheets) and RESRAD, in particular RESRAD-BIOTA (software).
The RAD-BCG Calculator spreadsheets enable the user:

• To use site specific and receptor specific environmental transfer factor parameters (e.g. sediment and soil distribution coefficients, bioconcentration factors) in place of default values;
• To compare radionuclide specific data with radionuclide specific BCGs;
• To verify compliance with the total dose limit, i.e. determine whether the sum of fractions for all radionuclide concentrations/BCG is less than 1.0;
• To apply user specified or regulatory agency specified biota dose limits;
• To modify, when technically justified, the default parameters used in the general screening phase, and to calculate the site specific BCGs using site specific information representing the evaluation area and receptors;
• To modify the radiation weighting factor for alpha emitters;
• To opt into the inclusion of progeny in the calculation of internal dose factors;
• To apply correction factors for the fraction of time that contamination is present in an evaluation area, and for the fraction of time that an organism resides in the contaminated area.

Since 2001, the DOE has been working in partnership with offices of the United States Environmental Protection Agency and the NRC to develop RESRAD-BIOTA, which has wider modelling capabilities (even beyond the current DOE-GA methodology) and a more effective user interface than the RAD-BCG Calculator. The following additional analysis capabilities are incorporated into the code:

• The ability to apply dose conversion factors based on geometries for a variety of organisms;
• Uncertainty and sensitivity analysis for biota dose estimates and associated parameters;
• Improved tabular and graphical presentation of results;
• Capabilities to link and import environmental radionuclide concentration data generated by other radionuclide

transport codes for subsequent use in evaluating doses to biota within RESRAD-BIOTA. These codes are not further described in detail here because the method for the first stages of application proposed is quite simple. In addition, details can be easily found in the cited literature. In more advanced stages, a (generic) site description would be necessary, and specific expertise in site characterization and radionuclide transfer and transport in different media would also become necessary. However, addressing such issues is beyond the scope of this publication.
As described above, the DOE-GA provides flexibility and the ability to iterate through the evaluation process. Parameter values, radiation weighting factors and organism residence times can be modified.
Table 3 shows a summary of the DOE-GA for the evaluation of radiation doses to aquatic and terrestrial biota. Only phases 1, 2 and 3(a) should be considered for the proposed first application of the method for a simplified environmental analysis. Table 3 shows parameters that could, with technical justification, be modified, corresponding to each phase of the DOE-GA.

Table 3. Summary of the united states department of energy graded approach (adapted from Ref.)

Phase*	Description	Parameters*
1. Data assembly	Knowledge of sources, receptors and routes of exposure for the area to be evaluated is summarized. Measured radionuclide concentrations in water, sediment and soil are assembled for subsequent screening.	Size of evaluation area Radionuclide concentration in environmental media
2. General screening	Maximum measured radionuclide concentrations in an environmental medium (i.e. water, sediment, soil) are compared with a set of biota concentration guides (BCGs). Each radionuclide specific BCG represents the limiting radionuclide concentration in an environmental medium that would not result in recommended dose standards for biota to be exceeded.	Initial general screening using maximum radionuclide concentration: no parameter modifications are allowed
3(a). Analysis** — site specific screening	Site specific screening using more realistic site representative lumped parameters (e.g. bioaccumulation factors) in place of conservative default values used in the general screening phase. Use of mean radionuclide concentrations in place of maximum values, taking into account time dependence and spatial extent of contamination, may be considered.	Sediment values may be modified, with technical justification, for aquatic system evaluations where only water or only sediment concentration data are available for the screening process
3(b). Analysis** — site specific analysis	Site specific analysis employing a kinetic modelling tool (applicable to riparian and terrestrial animal organism types) provided as part of the graded approach methodology. Multiple parameters which represent contributions to the organism internal dose (e.g. body mass, consumption rate of food/soil, inhalation rate lifespan, biological elimination rates) can be modified to represent site and organism specific characteristics. The kinetic model employs allometric equations relating body mass to these internal dose parameters.	Exposure area or receptor residence time correction factors for all organism types may be considered For riparian and terrestrial animals the following parameters can be modified: food source; body mass; uptake fraction of radionuclide ingested/ absorbed; biological elimination rate constant of radionuclide exiting the organism; food/soil intake rates, inhalation and soil inhalation rates and supporting parameters; water consumption rate; maximum lifespan; allometric equation provided
3(c). Analysis** — site specific biota dose assessment	Actual site specific biota dose assessment involving the collection and analysis of biota samples. The dose assessment would involve a problem formulation, analysis and risk characterization protocol consistent with the widely used ecological risk assessment paradigm.	Design, collection and direct analysis of environmental media and biota

* - Parameters that can be modified. The RAD-BCG Calculator provides capabilities to modify the dose limits for aquatic and terrestrial organisms, to modify the relative biological effectiveness weighting factor for alpha emitters and to deselect inclusion of energies for progeny of chain decaying nuclides with regard to internal dose conversion factors. These default values are to be used in dose evaluations conducted for DOE sites.
** - ‘Analysis’ consists of three increasingly more detailed steps of analysis.

Limits for radiation protection of non-human biota and their application for environmental assessments

In the following, the corresponding limits for radiation protection of non-human biota (plants and animals) are presented to be used in the case when simplified analysis needs to be performed within the framework of INPRO assessment in the area of environmental stressors.
Increasing attention has been paid to radiological protection of non-human environmental components, in particular biota, independently from protection of humans. Avoiding measurable impairment of reproductive capability is deemed to be the critical biological end point of concern in establishing the dose limits for aquatic and terrestrial biota. Other aspects being considered are individual lethal doses, genetic modifications and biodiversity.
The DOE has defined a radiation dose limit for the protection of aquatic organisms, and has considered (absorbed) dose limits for terrestrial biota. The limits for absorbed dose, below which no deleterious effects on populations of aquatic and terrestrial organisms have been observed, as discussed by the National Council on Radiation Protection and Measurements (NCRP) and the IAEA, are:

• A limit of 10 mGy/d (1 rad/d) for aquatic animals, from exposure to radiation or radioactive material releases into the aquatic environment;
• A limit of 10 mGy/d (1 rad/d) for terrestrial plants, from exposure to radiation or radioactive material releases into the terrestrial environment;
• A limit of 1 mGy/d (0.1 rad/d) for terrestrial animals, from exposure to radiation or radioactive material releases into the terrestrial environment.

For implementation of these proposed limits for protection of biota, the DOE has developed methods, models and guidance within a graded approach for evaluating radiation doses to non-human biota. These methods are briefly described in Sections 3.1 – 3.3.

Study of radiological impact on non-human biota in the United Kingdom

As an example of application of simplified methodologies for the estimation of effects of radionuclides to non-human species, during the Environment Agency’s re-examination of 2000–2002 radioactive discharges and disposals from the Sellafield site, a specific study was performed. Internal dose rate conversion factors for non-human biota were determined using a relatively simple and conservative method. Utilizing modelled and measured radionuclide activity concentrations, these dose rate conversion factors were employed to estimate whole body doses. The highest doses to non-human biota due to measured and predicted concentrations were estimated to be at least an order of magnitude below the level of 1×10⁻³ Gy/d suggested by the IAEA as a threshold for chronic exposure to radiosensitive species in the terrestrial environment, below which measurable detrimental effects on populations are unlikely to be observe. It is concluded in Ref. that the result would suggest that measures taken to apply “best practicable means” may have been unnecessarily conservative. However, one order of magnitude may cover the added higher uncertainties of discharged amounts, dispersion patterns, biota involved and site characteristics.
In 2001, the Environment Agency conducted a review of available data on radiation effects on biota and concluded that it is unlikely there will be any significant effects at:

• Chronic dose rates of 400 μGy/h (= 10 mGy/d = 1 rad/d) for populations of freshwater and coastal organisms;
• Chronic dose rates of 400 μGy/h (= 10 mGy/d = 1 rad/d) for terrestrial plant populations;
• Chronic dose rates of 40 μGy/h (= 1 mGy/d = 0.1 rad/d) for terrestrial animal populations.

These findings are the same as those from the DOE and are largely consistent with the findings and biota dose recommendations of the NCRP, the IAEA and UNSCEAR.
Additionally, the Environment Agency concluded that it is unlikely that there will be any significant effects at:

• Chronic dose rates below 1000 μGy/h (25 mGy/d = 2.5 rad/d) for populations of organisms in the deep ocean.

International activities related to radiological impact on non-human biota

Sections 4.1 – 4.5 briefly describe international activities (and their background) regarding radiation dose limits for non-human biota. This introductory information can be used when further elaboration on the ALs for doses to the reference biota species is needed.

International Commission on Radiological Protection

Radiological protection has traditionally focused on the protection of humans. In the past, the ICRP has taken the position that “the standards of environmental control needed to protect man to the degree currently thought desirable will ensure that other species are not put at risk”. This position has now changed, and environmental effects, i.e. the impact of ionizing radiation specifically on the environment, and protection of the environment against its harmful effects is now part of the development activities of the ICRP.

International Atomic Energy Agency

Between 1986 and 1992, the IAEA examined the validity of the implicit assumption that protecting humans will also protect the environment, and published its conclusions for the case of radioactive releases to the terrestrial and freshwater environments, as well as solid waste disposal underground, and for the of case marine disposal of radioactive waste. The conclusion in Ref. was that there is no convincing evidence from the scientific literature that chronic radiation dose rates below 1 mGy/d will harm animal or plant populations. This should be about the upper value of the range for doses to biota in areas where critical groups live exposed to radiation, where the limit of 1 mSv/a applies. In the aquatic environment, it would appear that limiting chronic dose rates to 10 mGy/d or less to maximally exposed individuals in a population would provide adequate protection to the population.
However, reliance on human based radiological protection may not be adequate for all possible conditions, e.g. in situations where humans are not present or where they live far away from the sources of radionuclides.
Environmental Modelling for Radiation Safety
From 2003 to 2007, the IAEA conducted the Environmental Modelling for Radiation Safety (EMRAS) programme. This focused on the development, comparison and testing of environmental assessment models for estimating radiation exposure of humans and radiological impacts on flora and fauna due to actual and potential releases of radionuclides to terrestrial and aquatic environments. EMRAS was followed by EMRAS II, which continued some of the work of previous international exercises in the field of radioecological modelling and focused on areas where uncertainties remain in the predictive capability of environmental models.
Modelling and Data for Radiological Impact Assessments
The IAEA programme on Modelling and Data for Radiological Impact Assessments (MODARIA) was planned for 2012–2015. The general aim of the MODARIA programme was to improve capabilities in the field of environmental radiation dose assessment by means of: acquisition of improved data for model testing; model testing and comparison; reaching a consensus on modelling philosophies, approaches and parameter values; development of improved methods; and exchange of information. MODARIA has continued some of the work of previous international exercises in the field of radioecological modelling, and focused on areas where uncertainties remain in the predictive capability of environmental models. These previous international exercises include the Biospheric Model Validation Study (BIOMOVS) and BIOMOVS II, initiated by the Swedish Radiation Authority Biosphere Modelling and Assessment (1996–2001), EMRAS (2003–2007) and EMRAS II (2009–2011).
Coordination Group on the Radiation Protection of the Environment
The international Coordination Group on the Radiation Protection of the Environment (CGRPE) was established by the International Plan of Activities on the Radiation Protection of the Environment in September 2005. It serves as a mechanism to facilitate the coordination of activities among international organizations by reviewing their ongoing work related to protection of non-human species. The IAEA organizes the Secretariat of the CGRPE.
Database of Discharges of Radionuclides to the Atmosphere and the Aquatic Environment
The IAEA Member State Database on Discharges of Radionuclides to the Atmosphere and the Aquatic Environment (DIRATA) contains the annual records submitted on a voluntary basis by IAEA Member States. DIRATA also includes the historical discharge records, collected by UNSCEAR, the European Commission and other international and national organizations.

United Nations Scientific Committee on the Effects of Atomic Radiation

In 1996, UNSCEAR summarized and reviewed information on the responses to acute and chronic radiation of plants and animals, as individuals as well as populations. The conclusions were consistent with the findings and recommendations made earlier by the NCRP and IAEA concerning biota effects data and appropriate dose rate CR for protection of biota populations, i.e. 10 mGy/d for aquatic animals and terrestrial plants, and 1 mGy/d for terrestrial animals. Since 2005, UNSCEAR has been developing developing a new scientific annex that addresses the effects of radionuclides released to the environment as well as a methodology for dose assessment for biota.

International Union of Radioecology

The International Union of Radioecology (IUR) poses the problem of protection against ionizing radiation simultaneously for both biota (environment) and humans, i.e. it deals with biota protection by a combination of anthropocentric and ecocentric approaches.
The IUR’s main role is based on science, promoting its advancement, the dissemination of its knowledge and the communication of this to society. As such, the IUR does not set out to promote particular standards or regulations, but to contribute to the international effort aimed at developing a general framework that will allow such management tools to be derived from a sound scientific basis. A further strength of the IUR is the capacity of the multidisciplinary scientific community gathered within the IUR to provide expert, informed, up to date and independent scientific advice, incorporating the knowledge from wider environmental fields not focused uniquely on radioactivity. The ultimate goal of the IUR is to ensure that the radioprotection of both people and the environment are considered with a scientifically sound, balanced and appropriately precautionary approach that permits sustainable development and technical innovation. The basic document governing the IUR policy in the area of radiation protection is presented in Ref..

Advisory Committee on Radiation Protection

In 2002, the Advisory Committee on Radiation Protection (ACRP) provided recommendations to the Canadian Nuclear Safety Commission concerning dose rate limits for protection of biota. The ACRP recommended a generic dose rate criterion in the range 1–10 mGy/d. The ACRP specified that this dose rate CR is based on population level effects, and, given the current state of knowledge and consensus views of radiation effects on biota, represents the level at which ecosystems will suffer no appreciable deleterious effects. The criterion is specified in terms of daily dose to prevent cases for which this dose rate criterion would be received in a few days (as reported in Ref.).

Calculation of impacts of chemical stressors on humans and the environment

There are currently no internationally agreed computer models to calculate the impacts (caused by the release) of toxic chemicals on humans and the environment. Specific computer models can be presented on the web site of the national environmental protection agency that uses these tools (e.g. models used by the US Environmental Protection Agency are presented in Ref.).
In the European Union, a feasibility study was performed in 2007 that identified existing models to calculate the environmental concentrations of chemicals from emissions. The study identified a comprehensive list of existing models to calculate the environmental impacts of chemical stressors, most of which are available on-line. It verified the status of validation and general acceptance of these models.

RESRAD-Chem

RESRAD is a series of computer models designed by Argonne National Laboratories to estimate primarily radiation doses and risks from residual radioactive materials. One code of this family is called RESRAD-Chem, and it presents a model for evaluation of sites contaminated with hazardous chemicals.
RESRAD-Chem calculates cancer incidence risks and hazard indices to an on-site exposed individual and derives site specific soil cleanup CR for hazardous chemicals. It follows the United States Environmental Protection Agency Risk Assessment Guidance for Superfund (RAGS). It considers nine exposure pathways: inhalation of dust and volatiles, ingestion of plant foods, meat, milk, soil, aquatic food and water, and dermal absorption from soil and water contact.

Regulatory limits of chemical stressors on humans and the environment

As mentioned in Sustainability Assessment Volume, there are no internationally agreed regulatory limits for chemical stressors. Thus, an INPRO assessor performing an assessment of the environmental impact of release of chemicals has to apply the regulatory limits of the country where an NES facility is planned to be installed.
There are, however, some international guidelines available, such as the following World Health Organization (WHO) guidelines for limits of toxic chemicals stressors on humans:

• Air Quality Guidelines for Europe;
• Guidelines for Drinking-water Quality.

These publications establish guideline values for chemicals in air and drinking water which are of significance to health. The report of the INPRO assessment of the planned NES of Belarus presents an application of national standards (limits) on the release of toxic chemicals from a nuclear power plant.

Representative examples of environmental studies

In the following, in addition to the simplified methods of analyses discussed in Sections 2 and 3, selected examples of environmental studies are presented. These studies could be used to identify comparable current facilities for the NES facilities to be assessed and to extract an appropriate list of stressors (primarily radiological), a measure of the levels (and impacts) of the stressors and, in some cases, the applicable regulatory standards.
An incomplete overview of radiological emissions and doses involving estimations performed for milling facilities, MOX fuel fabrication and fuel reprocessing is reported in Ref. and is reproduced here for illustration in Tables 4–7.

Table 4. Selected normalized radiological emissions and collective doses from milling

Radionuclide	Normalized release* (GBq/(GW(e)∙a))			Normalized collective effective dose* (man Sv/(GW(e)∙a))
	Mill	Mill tailings		Mill	Mill tailings
		In operation	Abandoned		In operation	Abandoned
Pb-210	0.02			0.00002
Po-210	0.02			0.00002
Rn-222	3000	20000	1000**	0.045	0.3***	150****
Ra-226	0.02			0.00001
Th-230	0.02			0.0006
U-234	0.02			0.003
U-238	0.02			0.003
Total				0.05

* - Normalized emissions in liquid effluents (0.01 for Pb-210 and Th-230; 0.02 for Ra-226; 0.3 for U-234 and U-238) contribute negligibly to the collective dose.
** - Annual activity released; the rate of activity is assumed to remain constant over more than 10 000 years.
*** - Dose commitment corresponding to a five year release.
**** - Dose commitment corresponding to a 10 000 year release.

Table 5. Selected normalized radiological emissions from mixed oxide fuel fabrication

Type of emission	Value
Annual atmospheric emissions (Bq/a)	107 - 108
Normalized annual emissions (Bq/(GW(e)∙a))	107 - 108

Table 6. Doses from mixed oxide fuel fabrication

Effective dose	Existing plants (experience)	Modern plants (design)
Annual effective dose per monitored worker (Sv/a) (average)	0.007	0.003
Collective effective dose per produced mixed oxide tonne (man Sv/t)	0.07	0.005-0.020
Normalized collective effective dose (man Sv/(GW(e)∙a))	2.0	0.15 - 0.60

Table 7. Selected normalized radiological emissions and doses from fuel reprocessing (oxide fuel)

Time period	Monitored workers (thousands)	Annual collective effective dose (man Sv)	Annual effective dose per monitored worker (mSv)
1975 - 1979	0.1	0.36	4.0
1980 - 1984	1.0	2.4	2.3
1985 - 1989	4.0	5.7	1.4

Values of actual annual emissions from currently installed nuclear facilities can be found in several publications (e.g. Ref.). Annex C of Ref. includes mining and milling (emissions per tonne of produced uranium, using a model mine and mill), uranium enrichment, fuel fabrication, nuclear reactor operation, fuel reprocessing and solid waste disposal. For power plants, power production and emission data in recent years are included for noble gases, tritium, ¹³¹I and particulates to air, and tritium and other radionuclides to water. Reference also reports the specific normalized collective effective doses in terms of the units man Sv/GW∙a.
The normalized collective effective dose of 7.5 man Sv/GW∙a is calculated in Ref. for mill tailings of radon releases integrated over 10 000 years. Incidentally, the releases to groundwater from low and intermediate disposal in shallow depositories of waste from reactor operation produce 0.5 man Sv/GW∙a, mainly from ¹⁴C.
A review of five year annual emissions from individual nuclear power plants in the European Union is reported in Ref.. Data were taken from plant annual reports and comprise, for airborne emissions: the noble gases, tritium, ¹³¹I and other beta and gamma emitters; for emissions to water: tritium, other beta and gamma emitters (including spectra of individual radionuclides when available) and alpha emitters when available. Reference also reports discharges from reprocessing sites in Europe, La Hague and Marcoule (France), and Sellafield and Dounreay (United Kingdom), which include aerial discharges of tritium, total beta and gamma emitters (without tritium), ¹⁴C, ⁸⁵Kr, iodine isotopes and total alpha emitters (individual radionuclides are also given for Dounreay and Sellafield). Data on liquid discharges from reprocessing sites include tritium, total beta and gamma emitters (without tritium), ¹⁴C and total alpha emitters. Liquid discharge spectra are also included for the four sites. Portions of discharge limits are indicated for individual species or groups of radionuclides.
Emission data as well as doses to critical groups can be found in annual environmental reports of several nuclear facilities, e.g. for reprocessing at La Hague, the annual report includes radioactive, as well as non-radioactive emissions, and the site authorization levels for all discharge classes. Airborne radioactive emissions given in Ref. are tritium, iodine, noble gases including ¹⁴C, ⁸⁵Kr, other beta and gamma emitters, as well as alpha emitters.
Reference provides data on radioactive emissions to the sea: tritium, iodine, ¹³⁴Cs, ¹³⁷Cs, ¹⁴C, ⁶⁰Co, ¹⁰⁶Ru and ⁹⁰Sr, other beta and gamma emitters, as well as alpha emitters. Non-radiological emissions to air comprise: CO, NO_X, SO₂ and particulates. Reference reports volatile organic compounds, ammonia, calcium, hydrogen chloride, hydrogen fluoride, silicon and sodium, and the heavy metals iron, nickel, titanium and vanadium. Non-radiological emissions to water comprise: aluminium, ammonia, barium, cadmium, chemical oxygen demand, chromium, cobalt, fluorine, hydrazine, iron, lead, manganese, mercury, nickel, nitrate, nitrite, phosphorous, sulphur, tributylphosphate, zinc and zirconium. Annual amounts of industrial and hazardous wastes are presented. In addition, the demands of chemicals (sodium hydroxide, nitric acid, formaldehyde), water, fossil fuels and electricity are provided, which can be useful for life cycle accounting of secondary material flows.
The assessment of radiological impacts, in terms of biota dose rates and their related potential biological effects on marine biota arising from radioactive discharges to the sea (as liquid effluents) of the La Hague facility can provide a part of the necessary input data for the INPRO assessment of a planned NES facility. In Ref., a representative set of local marine biota is selected for assessment. The guidance values are based on Refs, and on a review of the database developed in the FASSET project. The derived generic guidance values are similar to those published in the MARINA II study on radioactive discharges into north European marine waters, concentrations of radionuclides in the environment and assessment of their impact. The study provides dose rate results for selected marine biota categories for the coastal areas of La Hague and Sellafield. MARINA II also includes generic guidance values for the protection of marine biota. The estimated marine biota dose rates are low (by at least two to three orders of magnitude) compared with the lowest generic guidance values for the protection of populations of marine biota.
A summary of an environmental assessment of Bruce, a Canadian nuclear power plant, can be found in Ref., which also includes non-radioactive stressors that are, however, producing negligible impacts.
A generic comparative study has been performed by the OECD Nuclear Energy Agency (NEA) on the radiological impact of spent nuclear fuel options. The options considered were a pressurized water reactor (PWR) with an open (once through) uranium fuel cycle and the same PWR with monorecycling of uranium fuel. Generic release rates of radionuclides of all fuel cycle facilities including the power plant can be found in Ref..

Comprehensive environmental studies

The nuclear fuel cycle has been the object of a few life cycle assessment (LCA) studies focused on greenhouse gases or energy expenditures, but relatively few comprehensive studies have been published on current nuclear fuel cycles, and even fewer on future NESs. Known published examples of comprehensive studies on current operating NESs are given below.
Within the Externalities of Energy (ExternE) project of the European Commission, the French nuclear fuel cycle was evaluated. All stages were included in the assessment, from mining uranium ores through to waste disposal. The general methodology adopted in the study was the ‘impact pathway analysis’, modelling the path from discharges into the environment through dispersion and deposition to potential doses to humans. The damages (collective doses to workers and the public) were converted to costs, using a single framework devised for all the fuel cycles considered in the ExternE project.
The Swiss LCA study on current (reference year 2000) European energy systems included the nuclear fuel cycles for PWRs, boiling water reactors (BWRs) and country specific LWR mixes. Both open and closed fuel cycles (one time reprocessing of spent fuel, at equilibrium) were analysed. All stages of the front and back ends were included. Approximately a thousand elementary environmental flows (requirements as well as emissions) per unit of net electricity produced at the power plant are available in the ecoinvent LCA database, described in Section 6.2. Known examples of comprehensive environmental assessments of possible future nuclear fuel cycles are given below.
An extrapolation of LCA data of currently installed technology (Swiss LWR) to new designs (AP600 and advanced BWRs) was performed in Ref.. Also taken into account after a priority analysis were: improvements in mill tailing management; centrifuge/atomic vapour laser isotope separation enrichment plants substituting the units based on the gaseous diffusion process once they were phased out; and reduction of waste volumes at reprocessing. A semistatic approach was defined to calculate cumulative burdens at discrete time points. An assessment of five different scenarios (once through, single plutonium recycling, plutonium multirecycling, 45% thermal 55% breeder, and breeder burning neptunium, americium and copper homogeneously mixed with plutonium) within the twenty-first century in the French context was published in 2004. The EPR and European fast reactor types were modelled. Conventional emissions and the health effects of ionizing radiation (local impacts) were analysed. Application of LCA, covering the global environmental analysis part of the study, actually focused on a few emission species to air, was complemented by the assessment of local and regional radiological impacts on humans in terms of annual doses, and by the inventorying of nuclear materials using the material flow accounting tool COSI.
A project of the European Commission addressed future technologies up to the year 2050, integrating LCA with external cost estimation and an optimization energy economy code (Times-MARKAL). The nuclear industry is represented with Generation III and Generation IV power plants, in addition to fossil fuel and renewable technology power plants. A multicriteria decision assessment (MCDA) methodology will be used, as well as external costings for comparison and ranking of the technology options.

Life cycle assessment ecoinvent database

The commercial, centralized, web based LCA database ecoinvent was developed and implemented by the Swiss Centre for Life Cycle Inventories and supported by Swiss Federal Offices. It has been available on-line since September 2003.
The aim of developing the centralized LCA ecoinvent database was to achieve consistency throughout different life cycle inventory (LCI) databases maintained by the participating organizations. The sectors included are: energy systems (Paul Scherrer Institute); materials and metals, waste treatment and disposal (Swiss Federal Laboratories for Materials Testing and Research); transport systems and chemicals (Swiss Federal Institute of Technology in Zurich); and agricultural products (Swiss Federal Research Station for Agroecology and Agriculture). About 2750 individual processes (datasets) have been modelled and about 1000 elementary environmental flows inventoried, including emissions to air, water, soil, solid wastes, land use, and biotic and abiotic resources. The environmental cumulative burdens associated with the processes are integrated based on an algorithm, thus reflecting the interactions of industrial activities within an economy system. In addition to LCI, ecoinvent also includes the results of LCA, using current methods developed by different organizations.
The assessed energy systems, constituting about half of the processes available in the database, include electricity and heating systems. Electricity transmission and distribution, as well as country specific production and supply mixes, have also been modelled. Fossil fuel, nuclear and renewable energy systems associated with Swiss and European power plants, boilers and cogeneration plants have been assessed, reflecting conditions around the reference year 2000.
The nuclear fuel cycles modelled in ecoinvent are those associated with power generation at LWRs currently installed in the Union for the Co-ordination of Transmission of Electricity (UCTE). The focus was on two 1000 MW units operating in Switzerland: Gösgen PWR and Leibstadt BWR. The conditions of the fuel cycles for the Swiss reactors were assessed, modelled and extrapolated to France, Germany and the UCTE (which can be assumed to be representative of LWRs in western Europe). In addition to describing the nuclear cycle as such, the assessment served the estimation of cumulative environmental burdens of European electricity mixes, along with other electricity systems in ecoinvent. The stages separately modelled were: mining, milling, conversion, enrichment (diffusion and centrifuge), fuel fabrication, PWRs, BWRs, national mixes of PWRs and BWRs, reprocessing, spent fuel conditioning, interim storage, low level radioactive waste depository and geological final repositories.

See also

• Appendix I
• Appendix III
• Appendix IV

	[ + ] Assessment Methodology
Areas of INPRO Sustainability Assessment Overview • Economics • Safety (Nuclear Reactors) • Safety (NFCF) • Waste management • Environmental Impact on Stressors • Environmental Impact from Depletion of Resources • Infrastructure Requirements Basic Principle • User requirements • Criteria