ESC16/17 Council Directive 2013/59/EURATOM and Its Application

Transcription

1 ERMCO EUROPEAN READY MIXED CONCRETE ORGANIZATION ASSOCIATION EUROPEENNE DU BETON PRET A L EMPLOI E U R O P Ä I S C H E R T R A N S P O R T B E T O N V E R B A N D ESC16/17 Council Directive 2013/59/EURATOM and Its Application TC 351 WI Construction Products: Assessment of Release of Dangerous Substances - Radiation from Construction Products - Dose Assessment of Emitted Gamma Radiation

2

3 CEN/TC 351 Date: TC 351 WI CEN/TC Secretariat: NEN Construction products: Assessment of release of dangerous substances Radiation from construction products Dose assessment of emitted gamma radiation Einführendes Element Haupt-Element Ergänzendes Element 10 Élément introductif Élément central Élément complémentaire ICS: ; ; ; Descriptors: xxxxxxx Remark 1: STILL! to be converted into the new CEN.std template. 15 Remark 2: Still to be edited by a native speaker? 1

4 European foreword 20 This document (TC 351 WI ) has been prepared by Technical Committee CEN/TC 351 Construction products: Assessment of release of dangerous substances, the secretariat of which is held by NEN. This document is currently submitted to the Formal Vote. This document has been prepared under a mandate given to CEN by the European Commission and the European Free Trade Association

5 Contents European foreword 2 Contents 3 1 Scope 6 2 Terms and definitions 6 3 European Regulatory Framework Provisions on radiation protection Provisions on the marketing of construction products Radon exhalation from building materials 9 4 Provisions for dose assessment Principle of calculation Room model Basic assumptions Graded approach to dose assessment taking into account density and thickness Assessment of indoor gamma exposure due to building materials and construction products 15 5 Conclusions 16 Annex A Calculation of external gamma dose rate 17 A.1 Calculation of gamma dose rate 17 A.2 Parameters for a simple computer program 18 Annex B Examples of dose assessment 20 B.1 Example 1: Exposure to gamma radiation in a concrete room where the 226 Ra and 232 Th concentrations are slightly above average 20 B.2 Example 2: Exposure to gamma radiation in a room where the walls are made of material with elevated 226 Ra and 232 Th concentrations and the floor and ceiling of typical concrete 21 Annex C Estimate of indoor gamma dose based on mass per unit area as control parameter 24 Annex D Validation of the dose modelling and a density corrected index formula 27 D.1 General 27 D.2 Calculations 27 D.3 Conclusion 30 Annex E Considerations on and justification for choosing an appropriate room size 31 E.1 General 31 E.2 Chosen dimensions for the model room 32 E.3 Calculation of values given in Table E.1 32 E.4 Specific dose rates from different structures in the CEN/TS model room 34 Annex F Derivation of the unit conversion factor 37 Bibliography

6 Introduction The aim of this report is to propose a dose assessment methodology that accounts for factors such as density or thickness of the material as well as factors relating to the type of construction and the intended use of the material (bulk or superficial) as required by Annex VIII of the EU-BSS [5]. This approach is specially needed for building materials and construction products with an index exceeding 1 but that nonetheless may comply with the 1 msv per year reference level established in the EU-BSS. NOTE Although the methodology is centred around the reference level of 1 msv established in the EU-BSS, the methodology is also applicable if a reference value other than 1 msv per year is selected. In that case the selected dose value D and its corresponding index value I must be adjusted accordingly. In 1996, natural radiation sources were already included in the standards established by EURATOM as well as those established by the IAEA [1]. Since then, the European Commission has moved ahead publishing, on a regular basis, technical support guidance and recommendations on Naturally Occurring Radioactive Material (NORM) issues. In 1997, for instance, recommendations [2] were published to help deal with "significant increase in exposure due to natural radiations". In 1999, the European Commission published radiological protection principles concerning the natural radioactivity of building materials [6] and reference levels for workplaces processing materials with enhanced levels of naturally occurring radionuclides [3]. Lastly, in 2001 the European Commission published recommendations dealing with exemption and clearance levels for NORM residues [4]. These recommendations have provided Member States with criteria and a sound technical framework to help establish national regulations for NORM and building materials. Some Member States have already included all or parts of these recommendations in their regulatory framework anticipating the new EU directive. Subsequently, the European Commission decided to harmonize, promote and consolidate the main recommendations, introducing them into a new Council directive (2013/59/Euratom; EU- BSS [5]) laying down basic safety standards for the protection against the danger arising from exposure to ionising radiation. This BSS directive was officially issued in January Member States have four years to transpose and implement this directive and according to the Euratom treaty, these members shall before then, communicate to the Commission their existing and draft provisions. The Commission shall then make appropriate recommendations for harmonising the provisions amongst member States. Requirements of this directive (EU-BSS, [5]) dealing with building materials are hereby presented. They need to be taken into account along with the 2011 EU regulation laying down harmonised conditions for the marketing of construction products (EU no 305/2011) [6], so called CPR, containing many relevant articles which complement the aforesaid directive Both EU regulatory documents constitute the new basis for building material radiation protection regulation and should be soon followed by more detailed EU guidance and standards of which this document (TC 351 WI ) should be part. The European Commission (EC) has mandated the CEN to establish EU harmonized standards regarding dose assessment and classification of emitted gamma radiation from construction products. The EC has also informed CEN (CEN/TC 351, Berlin 11/02/2013) that the aim is to establish one test method per product, or product type, that the method should be demonstrably robust and should be adopted by all Member States as soon as the BSS comes into force. 4 4

7 This document can help Member State regulators to complete the EU-BSS and CPR regulatory framework covering a screening tool, dose modelling, and related technical information about radiation protection. Amongst others, the following recommendations were discussed by the CEN and the EC for the content of this document: The scope will exclude radon and thoron exhalation from building materials because this exhalation is dealt with in a different manner in the EU regulation. Regulatory explanations are given in Clause 3. Main assumptions, coefficients and conversion factors are taken into account. The methodology enables establishing which building materials may lead to a dose exceeding 1 msv per year for a member of the public or which building materials can be exempted from further restrictions Mass per unit area (kg/m²) of the material will be considered in the approach keeping a dose estimate model based on similar room models as the one used to establish the index mentioned in the EU-BSS. Additional sensitivity analysis regarding the room geometry is presented in Annex E to demonstrate that there is no more than 10 % of influence of such geometry upon the determination of doses. Lastly, it is important to underline that the EU regulatory philosophy is to ensure that gamma doses from building materials to a member of the public remain under 1 msv per year in addition to outdoor external exposure (EU-BSS Article 75) [5]. A simplified model, so called "index" in the EU-BSS is also proposed as a conservative screening tool ensuring that materials with an index less than 1 do not present any risk exceeding 1mSv per year of indoor gamma radiation, in any construction, to a member of the public. 5 5

8 1 Scope The aim of this Technical Report is to propose a methodology to determine indoor gamma dose from building materials and to help classify such a product as required in the Construction Products Regulation [6]. This first technical approach could be a precursor for the development of a harmonized European standard based on this methodology. NOTE 1 In this Technical Report, doses from radon and thoron exhalation are excluded. However, in 3.3, information is given on how radon exhalation is dealt with in (EU)2013/59/Euratom, the Basic Safety Standards Directive (EU-BSS) [5]. NOTE 2 Building materials considered in this Technical Report are the construction products used for buildings. Other construction products used for any other construction works (civil engineering, ) are not relevant and out of the purpose of the scope of this Technical Report. NOTE 3 Compliance with national exemption levels for NORM nuclides remains Terms and definitions For the use of this Technical Report, the terms and definitions in EN [7] apply along with the following terms and definitions. NOTE These terms and definitions are derived from the Basic Safety Standards Directive (EU 2013/59/Euratom) [5] authorisation registration or licensing of a practice 2.2 building material any construction product for incorporation in a permanent manner in a building or parts thereof and the performance of which has an effect on the performance of the building with regard to exposure of its occupants to ionising radiation NOTE 1 to entry Building materials considered in this Technical Report are the construction products used for building works. Other construction products used for any other construction works (civil engineering, ) are not relevant and out of the purpose of the scope of this Technical Report. 2.3 competent authority authority or system of authorities designated by Member States as having legal authority for the purposes of the EU-BSS [5] effective dose (E) sum of the weighted equivalent doses in all the tissues and organs of the body from internal and external exposure. It is defined by the expression E wt H T wt wr DT, R T T R 175 Where D T,R = the absorbed dose averaged over tissue or organ T, due to radiation R; w R = the radiation weighting factor, and w T = the tissue weighting factor for tissue or organ T. 6 6

9 The values for w T and w R are specified in Annex Ia of the BSS [5] The unit for effective dose is the sievert (Sv). 2.5 exemption level value established by a competent authority or in legislation and expressed in terms of activity concentration, total activity at or below which a radiation source is not subject to notification or authorisation 2.6 inspection investigation by or on behalf of any competent authority to verify compliance with national legal requirements practice human activity that can increase the exposure of individuals to radiation from a radiation source and is managed as a planned exposure situation 2.8 radon radionuclide 222 Rn and its progeny, as appropriate and exposure to radon means exposure to radon progeny 2.9 reference level level of effective dose or equivalent dose or activity concentration above which it is judged inappropriate to allow exposures to occur, even though it is not a limit that may not be exceeded 2.10 regulatory control any form of control or regulation applied to human activities for the enforcement of radiation protection requirements 2.11 Sievert (Sv) special name of the unit of equivalent or effective dose 210 NOTE 1 to entry One sievert is equivalent to one joule per kilogram: 1 Sv = 1 J/kg thoron radionuclide 220 Rn and its progeny, as appropriate 3 European Regulatory Framework Provisions on radiation protection Some new requirements have been established for building materials in the EU-BSS [5] but they derive from earlier EU principles and recommendations given in references [3], [4], [8], [9], [10], [11] and [12]. Such principles and recommendations were taken into consideration by some Member States but further harmonisation and consistency throughout Europe were to be established. Existing principles and recommendations were then reviewed and enhanced by EU Member States to be turned into proper harmonized EU regulations (EU- BSS, [5]) which was officially issued in January

10 Building materials of concern, which are to be identified by individual Member States, whether from natural origin or from those in which specific residues from identified NORM industries have been incorporated, need to comply with the reference level of 1 msv per year (compared to outdoor background dose). In addition also stress the implications for radon exposure if the external radiation from building materials exceeds 1mSv per year. NOTE 1 A list of NORM industries is included in EU-BSS Annex XIII [5] and has to be taken into account by Member States when establishing theirs. NOTE 2 In the case of composite material, like concrete, the activity concentration index may be calculated from the contribution of the constituents, and not every constituent should necessarily comply with the condition I < 1. NOTE 3 The architect or constructor person in charge of the building project is responsible for checking and getting acceptance from an authority whether the total dose in the building remains below 1 msv. The EU RP 112 principles [13] established the first non-prescriptive EU radiation protection framework concerning the natural radioactivity of building materials. This EU RP 112 was based on a publication [14] from the Finnish regulator (STUK) and provides EU Member States with a user friendly screening tool to evaluate building materials radiation gamma emissions and help check compliance with the maximum reference level mentioned above. To establish this screening tool, a conservative dose estimate model was first created. This model considered the activity concentrations of 226 Ra and 232 Th in secular equilibrium with the members of each decay chain (progenies) and 40 K. The calculations were based on a hypothetical room (with dimensions of 4 m 5 m 2,8 m) with walls, ceiling and floor of 20 cm thick and made of a material with a fixed density of kg/m 3 (similar to concrete). In this model, it is also assumed: an annual exposure time of 7000 hours a year; a dose conversion factor of 0,7 Sv/Gy and a background absorbed dose rate of 50 ngy/h. The doses were calculated according to the Berger approach with empirical build-up factors and selfattenuation. Considering all these assumptions, an activity concentration index (I) was then determined by the following simplified formula. Where C is the activity concentration in Bq/kg of 226 Ra, 232 Th or 40 K naturally contained in most building materials: C226 C232 C40 Ra Th K I (1) The dose estimate is close to 1 msv per year only when the index value is close to 1. An index < 1 with the conditions mentioned above means a dose estimate in compliance with the maximum reference level of 1 msv per year for a member of the public. This simplified model was deemed to be sufficiently conservative to be part of the EU-BSS [5] since most dwellings or buildings will not be designed to be as massive as the 'bunker' (hypothetical room) described above. In the EU-BSS [5], for building materials identified by the Member States as being of concern, it is required that the activity concentration of 226 Ra, 232 Th and 40 K be determined (EU-BSS article 75 and its annex VIII, [5]). The index can then be used as a screening tool to allow building material to be placed onto the EU market without any restrictions. National regulators and/or building codes may use this index to identify building materials which need no further analysis with respect to emitted gamma radiation. Although this screening tool should be sufficient for most building materials, it remains much too conservative for thin materials such as tiles, for light density products or for materials used Comment [Vnm1]: 19 Apr: Bernd to propose text, BUT any explanation will call for additional explanations! Comment [Vnm2]: Add examples to raise awareness. Besides, 1 msv is not a maximum nor a limit. 8 8

11 270 in marginal quantities. The EU-BSS [5] allows density, thickness and use of materials to be taken into account in an appropriate dose modelling approach if need be. 3.2 Provisions on the marketing of construction products The CPR [6] regulates the placing on the market of construction products. It establishes, in its Chapters II, III and IV, the requirements and obligations that have to be fulfilled where a construction product is covered by a harmonised technical specification (i.e. a harmonised standard or a European Technical Assessment). Generally for such products, a 'declaration of performance', which includes health and safety aspects (CPR Recitals 15 and 16) has to be drawn up, and this permits the affixing of the CE marking. The CE marking confirms that the product complies with its declared performance, and its harmonised technical specification, and permits free trans-boundary movement across the EAA. The manufacturer has to draw up such a declaration of performance with all related documentation and keep distributors informed. This declaration of performance should be accompanied by information on the content of all hazardous substances (Recital 25 of CPR [6] and Recitals of EC Regulation n 1907/2006 of the European Parliament and of the Council of 18/12/2006 REACH [24]). The content of the aforesaid declaration of performance (CPR article 6, [1]) should also include the construction product's uses along with its levels or classes (CPR articles 6.3d and 6.3g, [6]). It should be added that all the supply chain dealing with these construction products is responsible for the risks: manufacturers, importers and distributors. They all need to take account, appropriately, of the health and safety of people and the environment (CPR Article 28.2, [6]) within the declaration of performance. Moreover, such a responsibility will be monitored by "Market surveillance authorities" (CPR article 56, [6]) with technical support as appropriate (CPR articles 29-55, [6]). 3.3 Radon exhalation from building materials Regarding radon exhalation from building materials, Member States decided not to deal with this issue in the screening process, which, in consequence, addresses gamma radiation only. However, radon exhalation might be dealt with separately, including additional requirements, in national action plans. Additional strategies and methods for preventing radon ingress in new buildings, including identification of building materials with significant radon exhalation might be added by some Member States if need be (EU-BSS Annex XVIII.8). Specific EU-BSS parts deal with radon issues, fixing a maximum national reference level for all buildings, at 300 Bq.m -3 (EU-BSS Article 74.1, [5]), independent of the source of the radon (e.g. soil, building materials or water) It is important and an obligation on Member States to promote actions to identify dwellings with radon concentration (as an annual average) exceeding the reference level mentioned above and to encourage, where appropriate, by technical or financial means, radon concentration-reducing measures in these dwellings (BSS article 74.2, [5]). Action plans will have to be established by all EU Member States to seriously tackle this national health concern, including identification of building materials with significant radon exhalation. 4 Provisions for dose assessment In order to prospectively assess indoor external gamma doses resulting from the use of a given building material, mathematical models need to be used. For that purpose, the activity concentrations of three radionuclides in the building material have to be known, and a series 9 9

12 315 of assumptions need to be made regarding the room model (shape and size, thickness and densities of the walls, existence of door and windows, etc.). The main contributors to the gamma dose are 226 Ra, 232 Th and their progeny, and 40 K. The activity concentrations of these radionuclides can be determined in accordance with draft TS [15]. 320 The proposed method of dose assessment, together with the assumptions made on the room model, is described below. 4.1 Principle of calculation The method used in this report for calculating the external dose from building materials is based on the approach of STUK [14], which was also the basis for the document Radiation Protection 112 [13]. It consists of a point-kernel method that uses the Berger approximation for the build-up factor. Further details can be found in Annex A. As suggested by the EU-BSS directive, this document takes into account the real intended use of the building material (such as whether it is used as superficial or as bulk material) along with its thickness and density. It is important to bear in mind that the EU regulatory philosophy is to ensure that gamma doses from building materials to members of the public remain under 1 msv per year. 4.2 Room model 335 The typical room model used for gamma-radiation calculations is a rectangle parallelepiped in which all construction parts are made of 200 mm thick concrete with no windows or doors (Koblinger, 1978; Markkanen, 1995) [16], [14]. In particular, a room of 12 m 7 m 2,8 m was used in the document STUK-B-STO 32 [14]. The CEN/TS [17] reference room, on the other hand, has dimensions of 3 m 4 m 2,5 m, a door (1,6 m 2 ) in one of the long walls, and a window (2 m 2 ) in one of the short walls. For products used as a superficial layer with a thickness of 30 mm or less, it is assumed the layer behind is made of the reference concrete The absorbed dose rate in air is usually assessed in the middle of the room. For the STUK room model, the assumption of obtaining the dose rate at the centre of the room is valid as the dose rate is fairly constant in most of the room and only increases slightly (within 10 %) for areas within 50 cm of the walls [18]. Room size and dimensions have also very limited effect on the total dose, provided that the same material is used in all structures. Risica et al. (2001) [19] reported variations of less than 6 % when the width and length of the room were varied from 2 m to 10 m. Annex E provides more details. Exclusion of windows and doors in the room provides for a conservative approach regarding radiation protection (see Annex E). The CEN/TS reference room with no doors or windows is considered in this report (see figure 1). Whilst keeping consistency with other CEN standards, this room model is compatible with the activity concentration index formula, as expressed in EU-BSS 2013/59 Annex VIII [5]

13 Figure 1 Room model; dimensions of 3,0 m 4,0 m 2,5 m 355 NOTE The room model for which the activity index formula was derived [13] and the model room used in CEN/TS [17] result in nominal dose rates differing by roughly 4 %, 5 % and 6 % for 226 Ra, 232 Th and 40 K, respectively. 4.3 Basic assumptions 360 The EU-BSS defines the reference value of 1 msv per year as a dose in addition to the natural background (e.g. 0,29 msv per year). Therefore, the total annual dose indoor as sum of building materials and background can be higher than 1 msv, e.g. 1,29 msv. There are two possible interpretations: a) Building materials shield all of the natural background. The total dose resulting from the building materials can be 1,29 msv. 365 b) There are no shielding effects. The dose resulting from a building material can be 1 msv per year. RP 112 follows the philosophy of option a). Therefore the basic approach in determining the excess exposure is as follows: 1) The total exposure caused by the building is calculated; 370 2) The exposure caused by terrestrial background gamma radiation is then subtracted from it; 3) The result is referred to as the excess exposure. 375 In the example calculations presented in the Annexes, a European surface area weighted average value of 60 ngy/h (corresponding to 0,29 msv per year for an occupancy time of h) is assumed for terrestrial background gamma radiation

14 NOTE 2 The UNSCEAR 2008 Report, Annex B [20] estimates an average worldwide value for outdoor dose rates of 58 ngy/h Shielding effect of materials for cosmic radiation 380 The possible shielding effect of materials for cosmic radiation is considered to be small, and therefore exposure originating from cosmic radiation is not considered in the assessments Conversion factor for absorbed dose in air A conversion factor of 0,7 Sv/Gy is used for converting the absorbed dose in air to the effective dose according to the UNSCEAR 2000 report [25] Occupancy 385 An occupancy time of h per year, as an average for the annual time spend indoors in Europe, is used in the calculations Activity concentrations for reference concrete in Europe Regarding activity concentrations for reference concrete in Europe, the values in Table 2, derived from the RP-112 [13], will be considered. 390 Table 1 Activity concentrations for reference concrete in Europe Radionuclide C 226 Ra 40 Bq/kg 232 Th 30 Bq/kg 40 K 400 Bq/kg 4.4 Graded approach to dose assessment taking into account density and thickness The gamma dose rate is calculated in the middle of the standard sized room presented in Figure 1. The specific dose rates contributed by the walls, floor and ceiling are given in Table 2. The total indoor dose rate is calculated by summing the separately calculated dose rates caused by walls, floor and ceiling. A variety of dose assessments resulting from the use of the specific dose rates given in Table 2 are described in examples in Annex B. The dose assessments provide examples relating to massive concrete structures (e.g. apartment blocks) and for a smaller simpler type of structure we may find in rural areas. Table 2 Specific dose rate in air from the different structures in the room of Figure 1 All calculations to be checked! 2 significant digits after the comma Comment [Vnm3]: Mika's Fortran Mass per unit area a of wall, ceiling or floor material Wall, ceiling or floor material (top layer) b Kg/m 2 pgy/h per Bq/kg pgy/h per Bq/kg 20 cm thick concrete behind the wall, ceiling or floor material Shielding factor c 226 Ra 232 Th 40 K 226 Ra 232 Th 40 K 226 Ra 232 Th 40 K Wall W 1 : Dimensions 4,0 m 2,5 m, distance to room centre 1,5 m 12 12

15 Wall W 2 : Dimensions 3,0 m 2,5 m, distance to room centre 2,0 m Floor or ceiling: Dimensions 4,0 m 3,0 m, distance to room centre 1,25 m a Mass per unit area of the wall, floor or ceiling is the product of the thickness and the density of the structure. For example, in the case of a 15 cm (= 0,15 m) thick wall made of building blocks whose density is kg/m 3, the mass per unit area of the wall is 0,15 m kg/m 3 = 300 kg/m 2. b This is the specific dose rate caused by the wall w 1 or w 2, or floor or ceiling having a certain mass per unit area. For example, if the wall w 1 has a mass per unit area of 300 kg/m 2 and its 226 Ra concentration is 100 Bq/kg, the dose rate caused by the 226 Ra in the wall w 1 is (140 pgy/h per Bq/kg) (100 Bq/kg) = pgy/h = 14,0 ngy/h = 0,014 μgv/h. The top layer in brackets refers to the case where the wall, ceiling or floor structure comprises two different material layers, for example a tile on the top and a concrete structure behind it. In such a case the specific dose rates given in this column should be used for the top layer material e.g. such as the tiles. c The shielding factor applies when the wall, floor or ceiling consist of two components. In that case the shielding factor describes the shielding from the inner surface to the outer surface. These shielding factors are applicable for example when computing the dose rate for a cavity wall. For such type of wall the dose rate from the outer and the inner surface are computed separately and subsequently accumulated. The dose rate for the outer surface is based on the dose rate for the outer wall multiplied with the shielding factor. The shielding factor is based on the surface thickness of the inner wall. The dose rate for the inner wall is obtained directly from this table. An example of a cavity wall is included in Annex B.3. NOTE Calculations are based on the procedure presented in Annex B. 405 The index I is used as a screening tool to release building materials from restrictions with regard to gamma radiation: 13 13

16 C 226 C 232 C 40 Ra Th K I (1) Where C x = the activity concentration of the radionuclide x, in Bq/kg. 410 However, the above screening tool results may be too conservative in two types of situations: Building materials used in bulk amount with substantially lower densities than standard concrete, such as lightweight concrete, where the index may overestimate the resulting dose up to a factor 3 or 4 [18]. 415 Building materials used as thin covering materials, such as tiles, whose small thickness results in several-times-lower doses than predicted by the index. A dose model formula which could account for both situations and provide a more accurate assessment tool for light-density and thin building materials is presented in this report. 420 To combine density and thickness, the adapted formula uses the mass per unit area as control parameter. Each radionuclide is normalized to an activity concentration of 1 Bq/kg (see Annex C, Figure C.1). The isolines of equal dose plotted on a density vs thickness diagram are approximately hyperbolic suggesting proportionality between dose and the product of density and thickness with minor correction in higher orders. The resulting dose per year can be estimated using equation (2): 2 d ] C 226 Ra 2 6 d ] C [28116,3d 0,0161 D [319 18,5d 0, [22,3 1,28d 0,00114( d) ] C Th 40 K (2) 425 Where ρ = density, in kg/m 3 ; d = thickness, in m; C = activity concentration, in Bg/kg NOTE The dimensions of the reference room required parameters in the calculation of dose are 3 m 4 m 2,5 m without door and window and the chosen occupancy time is h per year. EXAMPLE 1 The dose resulting from a material with activity concentrations of 80 Bq/kg for radium-226, 80 Bq/kg for thorium-232 and 800 Bq/kg for potassium-40, with a density ρ = kg/m 3 and a thickness d = 0,2 m (ρ d = 470 kg/m 2 ) (see Example 1 of Annex B) can be calculated to be 1,1 msv per year. After subtraction of a background of 0,29 msv, the remaining dose is around 0,8 msv per year. The index value is 0,93. EXAMPLE 2 A material with activity concentrations of 110 Bq/kg for 226 Ra, 90 Bq/kg for 232 Th and Bq/kg for 40 K, with a density ρ = kg/m 3 and a thickness d = 0,32 m (ρ d = 320 kg/m 2 ) would have an index value of 1,2 but the resulting dose (after subtraction of a background) would be less than 1 msv per year, namely 0,85 msv per year. 440 This formula is valid up to a mass per unit area of 500 kg/m². Due to the self-attenuation the dose resulting from materials with higher mass per unit areas stay sufficiently constant

17 Therefore for those materials the dose can be estimated using a maximum value of 500 kg/m² Assessment of indoor gamma exposure due to building materials and construction products The assessment of indoor gamma exposure due to building materials of concern is a multistep process. For building materials (construction products) used in their intended use as a final product in a permanent manner in a building or parts thereof a flow-chart following the EU-BSS [5] Annex VIII and describing this process is given in Figure 2a. 450 The first step is to determine the activity concentrations of the relevant natural radionuclides in the building material, as described in prcen/ts [15]. The resulting activity concentrations will then be used to determine the activity concentration index, I, calculated according to the EU-BSS formula (1) In the case of composite material, like concrete, the activity concentration index can be calculated from the contribution of the constituents (components), as described in Annex C of prcen/ts :2016 [15]. It should be noted that every constituent may not necessarily comply with the condition I < 1. A flow-chart describing their assessment process is given in Figure 2b. NOTE 1 In case of not complying with the condition I < 1 restrictions for placing such constituents on market should not be set as indoor gamma exposure is assessed for the final product (as concrete). NOTE 2 Constituents (e.g. aggregates, fly ash, slag) in many cases are construction products that are CE marked under CPR. Their intended use must be defined in conjunction of determination of radionuclides 226 Ra, 232 Th and 40 K and calculation of index I for assessment of indoor gamma exposure purposes. 465 To be inserted: Figure 2a Assessment of indoor gamma exposure due to building materials (construction products) used in their intended use as a final product in a permanent manner in a building or parts thereof and 470 Figure 2b Assessment process for building materials (construction products) used as constituents in a final product Remark: Figures shall be language neutral If the index value is greater than 1, a more accurate gamma dose (D) estimate could be made using equation (2) considering thickness and density. For indoor thin materials (d 30 mm) and for dose modelling purposes a model room composed of a 20 cm thick bunker made of a reference concrete (40 Bq/kg for 226 Ra, 30 Bq/kg for 232 Th and 400 Bq/kg for 40 K, with a density ρ = kg/m 3, see Table 1) and coated inside with this thin material will be considered. To make it simple and conservative enough, the calculated gamma dose from the concrete (0,19 msv per year) will be added to the calculated dose produced by the thin material without any consideration of shielding aspects. The tiles on the other side are shielded by the wall. Then, if the dose estimate is less than the reference level of 1 msv per year, the building material may be fully exempted from regulatory control. Finally, unlike indoor thin materials mentioned above, the modelling approach for roof tiles would be different. To be realistic, no dose from a bunker made of concrete will be added but equation (2) will be used again with the right material thickness and density to determine the 15 15

18 Determination of radionuclides 226 Ra, 232 Th (or its decay product 228 Ra) and 40 K I C Ra C Th C 40 K 3000 YES Index I 1? NO TS CPR (art.18) + BSS (art Annex VIII and XIII) YES Indoor tiles or indoor thin materials? To consider a 20 cm thick reference room made of concrete and coated inside with such a material for dose modelling purposes. (no shielding considerations) NO Is the dose estimate 1 msv/year? YES NO No restrictions regarding radiation protection YES Restrictions in the declaration of performance to be established regarding the use of the building material NO To consider a reference room with a thickness d with no door and no window and made of this material with a density ρ for dose modelling purposes. Dose estimate (msv/year): Is the dose estimate 1 msv/year? Figure 2a Assessment of indoor gamma exposure due to building materials (construction products) used in their intended use as a final product in 16 a permanent manner in a building or parts thereof

19 Constituent/component 1 (building material, construction product) to be assessed Constituent/component 2 (building material, construction product) to be assessed Determination of radionuclides Ra 226, Th 232 (or its decay product Ra 228) and K 40 Determination the activity concentration index I, if required I 1 or I > 1 Determination of radionuclides Ra 226, Th 232 (or its decay product Ra 228) and K 40 Determination the activity concentration index I, if required I 1 or I > 1 Final product (building material, construction product) to be assessed Determination of mass activity concentration (radionuclides Ra 226, Th 232 and K 40) according to TS Annex C Determination of activity concentration I I C Ra C Th C 40 K 3000 Index I 1? no restrictions Index I > 1? further assessments according to Figure 2 a) Figure 2b Assessment process for building materials (construction products) used as constituents in a final product 17

20 gamma dose and related classification. Such an approach should help being realistic and consistent with an apartment made within the roof structure for instance. 490 If the material does not comply with formula (2), the possibility remains to demonstrate compliance with the reference level of 1 msv per year by performing a specific assessment with a dedicated modelling tool. In all these cases, where the annual gamma dose is less than or equal to 1 msv per year, the building material should be exempted from regulatory control. Otherwise, the material should require restrictions for use as appropriate taking into account the restrictions which are inherent to the material itself. Worked examples are included in Annex B Conclusions to be completed 16 18

21 Annex A 500 Calculation of external gamma dose rate A.1 Calculation of gamma dose rate The geometry used in the calculation is shown in Figure A.1. The absorbed rate in air D 1 (Gy/h) originating from the top layer at point P (x p,y p,z p ) can be calculated by using formula (A.1) 505 D 1 5, C11 4 en i Ei e B ( 1) i i ( l ) s1 where the build-up factor B i (1) is written according to the Berger model: l 2 dv (A.1) D( Ei ) i ( 1) s1 1e B ( 1) 1 C( E ) ( 1) s (A.2) i i i with z s1 l (A.3) z z p 510 and where x x 2 y y 2 z z 2 l (A.4) p p p C 1 activity concentration of the top layer, in Bq/kg; ρ 1 bulk density of the top layer, in kg/m 3 ; 515 γ i gamma intensity of gamma line i; E i gamma energy of gamma line i, in MeV; (μ en /ρ) i energy absorption coefficient in air for gamma energy E i, in cm 2 /g; μ i (1) attenuation coefficient in the top layer for gamma energy E i, in cm -1 ; 520 C(E i ) D(E i ) l s l coefficient in the Berger model, numerical values are given in Table A.1; coefficient in the Berger model, numerical values are given in Table A.1; distance between point P (x p,y p,y p ) and the point of integration Q (x,y,y), in cm; fraction of l within the top layer, in cm; a, b height and width of the object, in cm. NOTE The constant 5, is derived from conversions of units as explained in Annex F

22 525 The integration limits for the x-, y- and z-directions are a/2 a/2, -b/2 b/2 and h 1 0, respectively. The absorbed dose rate in air D 2 at point P (x p,y p,z p ) due to the bottom layer can be calculated by using formula (A.5) D 2 5, C 22 4 en i Ei e B ( 2) i ( i ( 1) s1 i ( 2) s2 ) l 2 dv (A.5) 530 With h1 s1 l (A.6) z z p and z s l (A.7) 2 s 1 zp z Where 535 h 1 s l thickness of the top layer, in cm; is the fraction of l within the bottom layer, in cm; μ i (2) is the attenuation coefficient in the bottom layer for gamma energy E i, in cm -1. The integration limits for the x-, y- and z-directions are a/2 a/2, -b/2 b/2 and -(h 1 +h 2 ) -h 1, respectively. 540 In this case, the estimate of the build-up factor is not as obvious as the attenuation term. The first approximation would be a product of the two build-up factors calculated separately for both layers. This would, however, overestimate the dose rate because the energy distribution of the flux has changed when the flux enters the upper layer. The following approximation is therefore used: 545 B (2) i i(2) s B i(1) 1 i(1) s1 i(2) s 2 D( Ei ) i (2) s2 i(2) s2e 2 (A.8) This is a product of the two build-up factors but in the bottom layer only the fraction proportional to the mean free paths in the different layers is considered. A.2 Parameters for a simple computer program In the cases of 238 U and 232 Th the absorbed dose rate should be calculated separately for every gamma line and then summed. It was demonstrated, however, by various comparisons that sufficient accuracy for practical assessment was achieved by using only one computational averaged gamma line for the U-238-series and two gamma lines for the 232 Thseries. This is possible because the energy absorption and the attenuation coefficients are rather smooth functions of energy in the interval kev. Only the intensive kev gamma line of the thorium series is treated separately because this single line causes over 40 % of the thorium series dose rate

23 The average gamma energy was calculated by using the emission probability as a weighting factor. The emission probability for the computational gamma line is the sum of the emission probabilities. The values for energy absorption, attenuation and the energy-dependent coefficients in the build-up factor were chosen on the basis of the calculated average gamma energy. The values used are given in Table A.1. The geometry used is illustrated in Figure A.1. Table A.1 Averaged gamma energies, attenuation coefficients in concrete, energy absorption coefficients in air, emission probabilities and coefficients C and D in the build-up factors Radionuclide Gamma energy a kev Gamma intensity γ Averaged values used in calculations Attenuation coefficient in concrete µ cm -1 Energy absorption coefficient in air µ e /ρ cm 2 /g Coefficient C Coefficient D 238 U 810 2,12 0,166 0,0285 1,161 0, Th 587 2,05 0,193 0,0295 1,279 0, Th ,356 0,0927 0,0217 0,724 0, K ,107 0,124 0,0257 0,946 0,0755 a The gamma energies used for averaging are taken from STUK 32 [14]. Figure A.1 The geometry used in calculation of external gamma dose rate. Origin is the middle of the surface of the top layer 19 21

24 570 Annex B Examples of dose assessment B.1 Example 1: Exposure to gamma radiation in a concrete room where the 226 Ra and 232 Th concentrations are slightly above average 575 The walls, floor and ceiling of a room are constructed of concrete. The concrete is assumed to have slightly elevated levels of radionuclides of natural origin. The room is assumed to have the specifications as shown in Table B.1. Table B.1 Specifications of the room as shown in Figure 1 Activity concentration Floors, ceiling, walls (concrete) Radionuclide 226 Ra 80 Bq/kg 232 Th 80 Bq/kg 40 K 800 Bq/kg Other parameters Density of concrete kg/m 3 Thickness of concrete 20 cm 580 For the walls, the floor and the ceiling, the specific dose rates for a mass per unit area of 500 kg/m 2 from Table 2 are used. The dose rate in the room is calculated as shown in Table B.2. Table B.2 The dose rate in the room as shown in Figure 1 Source Calculation Dose rate w 1 (concrete) 2 ( ) ,0806 µgy/h w 2 (concrete) 2 ( ,4 800) ,0448 µgy/h Floor and ceiling (concrete) 2 ( ) ,0909 µgy/h Total dose rate in the room (cosmic radiation excluded) Terrestrial gamma radiation outdoors: the concrete structures of the building shield against this source Excess dose rate caused by building materials Excess effective dose 0,2195 µgy/h -0,06 µgy/h 0,1595 µgy/h 0,7 Sv/Gy 0,1595 µgy/h 0,112 µsv/h 585 The annual excess effective dose to an occupant depends on the annual occupancy time: hours per year 0,113 μsv/h = 791 μsv per year = 0,8 msv per year 20 22

25 B.2 Example 2: Exposure to gamma radiation in a room where the walls are made of material with elevated 226 Ra and 232 Th concentrations and the floor and ceiling of typical concrete 590 The floor and ceiling of the room in Figure 1 are made of concrete containing weighted average concentrations of radium, thorium and potassium of 33 Bq/kg, 45 Bq/kg and 420 Bq/kg for 226 Ra, 232 Th and 40 K, respectively. The walls are made of bricks with elevated levels of radionuclides of natural origin. The material specifications are as shown in Table B.3. Table B.3 Specifications of the room as shown in Figure 1 Activity concentration Floors, ceiling, walls (concrete) Walls (brick with elevated natural activity) Radionuclide 226 Ra 33 Bq/kg 200 Bq/kg 595 Other parameters 232 Th 45 Bq/kg 300 Bq/kg 40 K 420 Bq/kg Bq/kg Density kg/m kg/m 3 Thickness 20 cm 15 cm For the walls the specific dose rates for a mass per unit area of 300 kg/m 2 from Table 2 and for the floor and ceiling the specific dose rates for a mass per unit area of 500 kg/m 2 from Table 2 are used. The dose rate in the room is calculated as shown in Table B.4. Table B.4 The dose rate in the room as shown in Figure Source Calculation Dose rate w 1 (brick) 2 ( ) ,1534 µgy/h w 2 (brick) 2 ( ,5 1500) ,0887 µgy/h Floor and ceiling (concrete) 2 ( ) ,0456 µgy/h Total dose rate in the room (cosmic radiation excluded) Terrestrial gamma radiation outdoors: the concrete structures of the building shield against this source Excess dose rate caused by building materials Excess effective dose 0,2877 µgy/h -0,06 µgy/h 0,2277 µgy/h 0,7 Sv/Gy 0,2277 µgy/h 0,1594 µsv/h The annual excess effective dose to an occupant staying hours per year in this room would then be hours per year 0,1594 μsv/h = μsv per year = 1,2 msv per year 605 B.3 Example 3: Exposure to gamma radiation in a room with concrete floor and ceiling, and cavity walls with brick and limestone 21 23

26 610 The floor and ceiling of the room in Figure 1 are made of regular concrete. The cavity walls are made of an outer brick wall and an inner lime stone wall. The material specifications are as shown in Table B.5. Activity concentrations Other parameters Table B.5 Specifications of the room as shown in Figure 1 Floor, ceiling Cavity wall with brick and limestone Radionuclide Concrete Brick Limestone (outer wall) (inner wall) 226 Ra 33 Bq/kg 74 Bq/kg 10 Bq/kg 232 Th 45 Bq/kg 86 Bq/kg 9 Bq/kg 40 K 420 Bq/kg 720 Bq/kg 230 Bq/kg Density kg/m kg/m kg/m3 Thickness 20 cm 10 cm 24 cm The mass per unit area of the outer brick walls is kg/m 3 0,10 m = 137 kg/m 2 and the inner limestone wall is kg/m 3 0,24 m = 437 kg/m 2. Based on those values the specific dose rates in Table 2 for brick are based on a mass per unit area of 150 kg/m 2, and for limestone they are based on a surface mass of 400 kg/m 2. For the floor and ceiling the mass per unit area is kg/m 3 0,2 m = 470 kg/m 2 and the specific dose rates are based on a surface mass of 500 kg/m 2. The dose rate in the room is calculated as shown in Table B.6. For the outer brick wall the shielding effect from the inner limestone wall is considered. This effect is represented by a shielding factor s (-) for each of the three nuclides n and is obtained using the shielding factors in Table 2. The factor is based on the surface thickness of the inner wall, which is ρ d,shield = 400 kg/m 2 (lime stone). Shielding factors are obtained through interpolation

27 Table B.6 The dose rate in the room as shown in Figure 1 Source Calculation Dose rate w 1 (Brick) s 226 Ra 0.08 s 232 Th 0.09 s 40 K x (89x74x x86x x720x0.10) x μgy/h w 1 (Limestone) 2 x (168x x9 + 14x230) x μgy/h w 2 (Brick) s 226 Ra 0.09 s 232 Th 0.10 s 40 K x (50x74x x86x x720x0.10) x μgy/h w 2 (Limestone) 2 x (100x x x230) x μgy/h Floor and ceiling (concrete) 2 x (188x x x420) x μgy/h Total dose rate in a room (cosmic radiation excluded) μgy/h Terrestrial gamma radiation outdoors: the concrete structures of the μgy/h building shield against this source Excess dose rate caused by building materials μgy/h Excess effective dose 0.7 Sv/Gy x μgy/h μsv/h 630 The annual excess effective dose to an occupant staying hours per year in this room would then be hours per year 0,0088 μsv/h = 61 μsv per year = 0,06 msv per year 23 25

28 635 Annex C Estimate of indoor gamma dose based on mass per unit area as control parameter With the approach of Annex A, the gamma dose of a building material can be calculated taking into account its thickness and density. Normalized to an activity concentration of 1 Bq/kg for 226 Ra and assuming an occupancy time of h per year, the resulting doses in µsv per year (1.000 µsv = 1 msv) are plotted in Figure C.1. These contour plots can also be used for a simplified dose estimation. The hyperbolic-like forms of the contour lines suggest that the dose depends on the product of density and thickness, also known as mass per unit area. For any given mass per unit area, the dose is acceptably independent from the actual thickness, at least in the range of values for typical building materials. Due to this independency, the data from any actual thickness (e.g. 0,2 m) can be used to quadratic fit (see Figure C.2) and to approximate the dose in respect to the mass per unit area for all three nuclides. Ra : Th : K : ,3 d 0,0161d ,5 d 0,0178d ,3 1,28d 0,00114( d) 10 (C.1) NOTE The intercept values from each quadratic fit (i.e. 281, 319 and 22,3) have no physical meanings. Therefore, for sufficiently large 'mass per unit area' values, they can be ignored

29 Figure C.1a Annual Dose, µsv per Bq/kg, Ra Figure C.1b Annual Dose, µsv per Bq/kg, Th 25 27

30 Figure C.1c Annual Dose, µsv per Bq/kg, K Figure C.1 Contour plots of normalized doses Ra, Th and K 655 Figure C.2 Quadratic fit of the normalized dose as a function of mass per unit area 26 28

31 Annex D 660 Validation of the dose modelling and a density corrected index formula D.1 General The index formula (1) described in Clause 3 is meant as a simple screening tool. With the help of this tool, it is possible to estimate whether or not a building material used in the construction of a dwelling could potentially lead to an annual dose for the occupants in excess of 1 msv per year in addition to the natural radiation level. Certain limitations are incorporated in the approach to achieve a conservative approximation, as it is a complex set of variables, like material properties, density and thickness that determine the real dose rate. The formula on which these calculations are based stems from a time in which calculation power was very limited by today s standards. Modern computer systems offer enough power to perform a so called Monte Carlo calculation employing more accurate physics by simulating the individual fate of e.g to 10 9 randomly created gamma rays. To test the validity of the results of the formula based on the Berger model, calculations have been performed using the Monte Carlo Code PENELOPE. It is widely used in a variety of radiation scenarios, is validated for medical applications [21]. With the code, a 3D model of the RP 112 model room was created and the natural radiation from the walls was simulated for different photon energies. D.2 Calculations 680 In a first step, it is necessary to adjust the attenuation coefficient for different densities and energies. The energy dependency of the attenuation coefficient can be seen in Figure D.1. For the purpose of natural radioactivity the range between 0,1 and 3 MeV is of interest

32 Figure D.1 Values of the mass attenuation coefficient, μ/ρ, and the mass energyabsorption coefficient, μ en /ρ, for concrete as a function of photon energy [22] 685 The values necessary for the linear adjustment of the Berger model for a selection of different densities based on the data described above is summarized in the table D.1. Table D.1 Values of the mass attenuation coefficient for different densities and relevant photon energies U Th 40 K 810 kev 587 kev 2615 kev 1461 kev 0, ,042 0,0202 0,027 0, ,067 0,0322 0,043 0, ,092 0,0433 0,059 0, ,117 0,0564 0,075 0, ,143 0,0685 0,092 0, ,168 0,0806 0,108 0,166 0,193 0,0927 0, Figure D.2 shows an uncorrected extrapolation of the Berger model towards lower densities (solid line) as well as the results that are based on the modified attenuation coefficients (dashed line). It can be seen that the dose rates for rooms built from walls with a lower density are much lower than what would be assumed by the index formula in the RP 112 (density of kg/m³), up to a factor of 4. It can also be seen that a simple linear extrapolation to lower densities underestimates the dose

33 TC 351 WI :2016 (E) 0,28 Air kerma [mgy/7000h] 0,23 0,18 Th 232 0, U K 232 Th corrected 0, U corrected 40 K corrected 0, Concrete density [kg/m3] 700 Figure D.2 Comparison between uncorrected and corrected extrapolation of the Berger model to lower densities, calculated for an activity concentration of 40 Bq/kg U-238, 30 Bq/kg 232Th and 400 Bq/kg 40K. The wall thickness is 20 cm The final comparison between the density corrected Berger model and the Monte Carlo simulation is shown in Figure D.3. 0,28 Air kerma [mgy/7000h] 0,23 0, ,13 Th PENELOPE 238U 40K PENELOPE PENELOPE 232Th 0,08 238U 40K corrected corrected corrected 0, Concrete density [kg/m3] Figure D.3 Comparison between corrected Berger model and Monte Carlo simulation (40 Bq/kg U-238, 30 Bq/kg Th and 400 Bq/kg K, wall thickness 20 cm) 29 31

34 710 For the U-238 series and 40 K the PENELOPE results confirm the results of the Berger model. For the 232 Th series, which is a combination of two different photon energies, the results from PENELOPE show lower doses than those calculated with the Berger model. Therefore the results from the Berger model for this 232 Th series are more conservative. The Index-Formula assumes a wall thickness of 20 cm, but anything between 10 cm and 40 cm is to be expected in real life construction. The results from PENELOPE calculations as well as those from the Berger model are compared in Figure D.4. Annual dose (7000h) [mgy] 0,27 0,25 0,23 0,21 0,19 0, U PENELOPE 232 Th PENELOPE 40 K PENELOPE 238 U 232 Th 40 K 715 0, Wall thickness [cm] Figure D.4 Dependency of the resulting dose for different wall thickness. (40 Bq/kg U-238, 30 Bq/kg 232 Th and 400 Bq/kg 40 K, wall density kg/m³) The results from PENELOPE and the Berger model are consistent. There is a saturation setting in at a wall thickness of about 20 to 25 cm. In principle, the dose rate function which is used for the density correction could be calculated for a wall thickness of 40 cm to be most conservative, but for a 20 cm-thick bunker the index formula underestimates the dose by only 10 %. This underestimate could be considered to compensate, to a degree, for the overestimate obtained when ignoring the dose reduction introduced by model rooms which include 'holes' made by doors and windows. For a wall thickness below 20 cm, a linear correction factor could be considered in the index formula. D.3 Conclusion 730 The comparison between the calculations of the RP 112 [13], based on the Berger model and the simulations with a validated Monte Carlo Code shows a reasonable consensus. Within the scope of validating the 1 msv per year-criterion for building products in the context of the CPR [6] and the EU-BSS [5], the use of the more practicable Berger model, corrected for different attenuation coefficients for different densities and a linear scaling factor for a thickness of less than 20 cm, leads to results with acceptable uncertainties