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Preliminary safety concept for disposal of the very low level radioactive waste in Romania
ARTICLE IN PRESS Applied Radiation and Isotopes 67 (2009) 935–938
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Preliminary safety concept for disposal of the very low level radioactive waste in Romania O. Niculae a,, V. Andrei a, G. Ionita a, O.G. Duliu b a b
ANDRAD National Agency for Radioactive Waste, 1 Campului, 15400 Mioveni, Arges County, Romania University of Bucharest, Department of Atomic and Nuclear Physics, P.O. Box MG-11, 077125 Magurele (Ilfov), Romania
a r t i c l e in fo
Keywords: Very low level radioactive waste Radioactive waste disposal Safety assessment Safety indicators Safety concept
abstract In Romania, there are certain nuclear installations in operation or under decommissioning, all of them representing an important source of very low level waste (VLLW). This paper presents an overview on the approach of the VLLW management in Romania, focused on those resulted from the nuclear power plants decommissioning. At the same time, the basic elements of safety concept, together with some safety evaluations concerning VLLW repository are presented and discussed too. & 2009 Elsevier Ltd. All rights reserved.
1. Introduction
60
Co, 134Cs, 137Cs, 63Ni, 59Ni, 95Zr/ 95Nb, 94Nb, 90Sr, 99Tc, 129I, 3H, C will be considered and treated as VLLW. Therefore, in this paper we present the basic elements of safety concept, together with some preliminary results of evolution scenarios developed for the Saligny VLLW Repository. 14
The very low level waste (VLLW) represents a category of waste with activity varying between those of conventional waste, that can be disposed in a landfill disposal and those of low and intermediate level waste which need usually to be disposed in near surface disposal facilities provided with a multi-barrier disposal system. According to the National Commission for the Control of Nuclear Activities (CNCAN) VLLW are represented by the short lived radioactive waste, with the specific activity higher than the exclusion levels stipulated by the regulations, but with a radioactive content lower than the level established by the Romanian Regulatory Body for the low level radioactive waste (CNCAN, 2005). At the present time in Romania there are one operational Nuclear Power Plant (NPP) at Cernavoda, and one decommissioned research reactor at the Institute of Nuclear Physics and Nuclear Engineering in Bucharest. The VLLW resulted from the operation of Cernavoda NPP, i.e. (i) compactable waste, (ii) noncompactable waste, (iii) organic liquids, (iv)–(v) spent filter cartridges and the low activity spent resins will be deposited in the Saligny VLLW Repository, planned to be built in the vicinity of the Cernavoda NPP. The chosen site represents a 43 m thick plateau consisting of an alternation of silty loess, clayey loess, red clay and as well as pre-quaternary clay considered as unsaturated, all of them laying on a compact Baremian limestone fundament, host of the main aquifer, and considered as a saturated zone (Durdun et al., 2001) (Fig. 1). In accordance with the first results of the characterization process, the following radionuclides
Corresponding author. Tel.: +40 021 3140392; fax: +40 021 3140391.
E-mail address: [email protected] (O. Niculae). 0969-8043/$ - see front matter & 2009 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2009.01.061
2. Theoretical background 2.1. Prerequisites for development of the safety concept As a prerequisite for the development of the VLLW safety concept and disposal strategy, ANDRAD initiated in 2007, a preliminary research and development plan, with the following subjects: (i) documentation on the international solutions for disposal of VLLW, (ii) safety aspects, (iii) a preliminary inventory of VLLW resulted from Cernavoda NPP operation and decommissioning, based on the data obtained from Cernavoda NPP, and (iv) the development of a disposal concept. The initial study on the development of a VLLW repository has already been completed (SITON, 2007), and other researches will be concluded in 2008, and in the next years. Based on the international practice, and taking into account the recommendations of the Romanian Regulatory Body, it was proposed a trench type repository, located on Saligny site, in the Cernavoda exclusion area (SITON, 2007). The main objective in the design of the repository will be to minimise the impact of the ionising radiation emmited by gases, particulates and leachates. This objective is achieved by isolating waste from the accessible environment, containing the waste within the disposal facility, as well as attenuating the associated release of radionuclides from the facility as far as practicable.
ARTICLE IN PRESS 936
O. Niculae et al. / Applied Radiation and Isotopes 67 (2009) 935–938
Fig. 1. A vertical section showing the geological profile of the place designed for repository location coresponding to the first and third scenarios (I) as well as to the second one (II).
To develop the safety concept for the VLLW repository, some basic questions were outlined and adequate responses are given in the present paper. The first question ‘‘what is necessary to demonstrate for each repository stage, in order to outline its safe evolution’’ will define the safety objective. Other questions like ‘‘what is the proof based on?’’, ‘‘which methodology is suitable for demonstration?’’ and ‘‘which are the parameters that illustrate the achivement of the safety objective?’’ define safety pillars, safety assessment methodology and safety indicators. Finally, the question ‘‘which are the possibilities to verify/maintain the repository safety?’’ will define the necessary management to sustain the repository safety. The answer to these questions are summary presented in the Section 3 of the present paper.
2.2. Safety evaluation The evolution of a repository is described by the evolution scenarios, developed both for normal and abnormal conditions for its entire life periods. These scenarios refer to all disposal system elements, and have as result the safety indicators. In our case we have used the ISAM methodology developed by the IAEA (2004) and applied previously for the planned Saligny Near Surface Repository (Niculae et al., 2008; Dogaru et al., 2008).
Consequently, by considering the water infiltration through the same reinforced concrete repository located in 65 m 45 m 14 m trench we developed three scenarios based on different construction variants but with some common elements, such as the compacted loess liner. As a result, when the infiltrated water reaches the waste, the radionuclides are washed out in both unsaturated geological layers and the saturated zone. Further, in the saturated zone we considered a well which represents the water source for a farm located at the limit of repository, following a similar scenario to the one proposed by Chen et al. (2007) and UKEA (2007). According to the geo-technical data presented by Dogaru et al. (2008) for the Saligny site, the water infiltration rate was considered to be about 0.02 m/y for the first two scenarios and 20 times higher then that for the third one (0.4 m/y). In the first scenario, the planned repository is located, as mentioned above, over a compacted loess liner of 3 m depth (Fig. 1I). In the second scenario, we considered the same trench but with a reinforced concrete liner of 1 m depth, a foundation ground of 3 m depth (compacted loess), as well as an upper cover with a thickness of about 1.5 m (Fig. 1II). The third scenario differed from the first one by the water infiltration rate. For the first and the third scenarios, we considered the infiltration starting at the moment when the repository was sealed. For the second scenario, the water infiltration rate was considered zero for the first 50 years after the repository closure (by taking into account the effectiveness of the final cover), and 0.02 m/y after that. All calculations were done by using the AMBER 5.0 (2006) computer code which considers a global compartment model and integrates all system components, migration processes as well as the exposure mechanisms. The simplified conceptual model of the repository has been developed using the interaction matrix method. The safety indicator was considered to be the dose received by a person (from the critical group) by water ingestion (Niculae et al., 2008). According to the hypotheses mentioned above, the following inventory was considered in the evaluation: 3H(100 GBq), 137Cs (1 MBq), 134Cs(1 MBq), 90Sr(1 MBq), 60Co(10 MBq), 59Ni(10 GBq), 63 Ni(10 GBq), 99Tc(1 GBq), 129I(10 MBq), 94Nb(100 MBq) and 14C (1 GBq). Tables 1 and 2 present the characteristic parameters of the Saligny site as well as the distribution coefficients for some radionuclides experimentally determined for the Saligny site (Dogaru et al., 2008) as well as presented in literature (Park et al., 2003). By taking into account the long lived radionuclides Table 1 Characteristic parameters of model compartments, Saligny site, as input data for safety indicators assessment. Horizon
Average values of dry density (kg m 3)
Average values of water filled porosity (%)
Depth of model compartments (m)
Waste form Compacted loess liner Concrete liner (only for scenario 2) Compacted loess foundation ground (only for scenario 2) Silty loess Clayey loess: Layer Iab1 Upper layer Ib Layer Iab2 Lower layer Ib Red clay Pre-quaternary clay
2500 1760 2500
15 32 15
14 3 1
1760
32
3
1540
12
1
1780 1570 1720 1690 1760 1760
26 14 25 25 32 31
6 4 2 6 8 16
Numerical values of all parameters are reproduced from Dogaru et al. (2008).
ARTICLE IN PRESS O. Niculae et al. / Applied Radiation and Isotopes 67 (2009) 935–938
937
Table 2 Distribution coefficients (kd) for the model compartments. Radionuclide
137
Csa Coa 90 a Sr 14 a C 99 Tcb 129 b I 60
a b
kd in the model compartments (m3 kg
1
)
Waste form and concrete liner
Aquifer
Silty loess
Clayey loess
Red clay
Pre-quaternary clay
0.02 0.01 0.5 0.2 0.001 0.001
10 200 0.5 0.1 0.01 0.001
0.774 0.033 0.006 0.003 0.001 0.001
1.131 0.030 0.011 0.005 0.001 0.001
4.131 0.031 0.012 0.008 0.001 0.001
2.366 0.030 0.012 0.005 0.001 0.001
After Dogaru et al. (2008). After Park et al. (2003).
Fig. 2. The flow-chart of the safety concept proposed for VLLW repository.
from inventory (see Tables 1 and 2) a total period of 1 Ma was considered for the evaluation.
3. Results and discussion 3.1. Safety concept Fig. 2 presents a flow-chart of safety concept proposed for VLLW repository, with the purpose to integrate all factors, arguments and the evaluation methods used to demonstrate the safe evolution of the repository during 1 Ma. The safety objectives for each repository stage are the safety criteria, specified in the nuclear legislation as the practical way to fulfil the IAEA (2006) safety principles. According to (IAEA, 2003), the most important safety indicators are the radiological dose and radiological risk, their reference values being expressed by means of the dose limit or constraint as well as the risk limit or risk constraint. In Romania, according to (CNCAN, 2006), for non-professional irradiation, the reference value for the radiological dose received for public during one year, called the dose limit, is of 1 mSv/y, with a dose constraint of 0.3 mSv/y. This reference value was established for all the life periods of a near surface repository. 3.2. Safety evaluation In Fig. 3 we present the results of our calculations expressed by the annual water ingestion dose: Hwat received by a person from the critical group and corresponding to each of these scenarios. At the same time in Table 3 we have reproduced the Hwat maximum
values as well as the time these values are attained for each scenario too. In the first scenario, the early contribution comes from the mobile 3 H and 129I. Tritium reaches a peak value of 9.2 10 10 mSv/y at 14 years from the repository closure, while 129I presents two peaks: an early one of 3.8 10 13 mSv/y and a final one of 1.83 10 4 mSv/y at 8 y and, respectively, 5.4 ky after closure. Carbon and Technetium reach their maxima later: 29.5 ky for 14C (1.75 10 7 mSv/y) and 7.2 ky for 99Tc (3.13 10 4 mSv/y) (Fig. 3I) (Table 3). In the second scenario, tritium reaches its peak value of only 1.9 10 10 mSv/y at about 100 years after repository closure while 129 I and 99Tc remained unchanged, with 1.82 10 4 mSv/y at 5.4 ky, and 3.13 10 4 mSv/y at 7.2 ky, respectively. The 14C contributes with a dose value of 3.67 10 9 mSv/y at 35.5 ky, significantly lower than in the case of the first scenario (Fig. 3I) (Table 3). By comparing the results of Scenarios 1 and 2, one can see that the main effect of concrete liner is limited to 14C and 3H by reducing their annual water ingestion dose with about 98% and 79%, respectively, and by delaying the 3H peak value with 86 years. In the third scenario, the effect of the increased infiltration rate is visible as all peak values appear earlier and are significantly greater than in the previous scenarios. Therefore, 3H has an early peak value of 6.12 10 4 mSv/y at 39 years after closure, 129I presents a peak value of 1.2 10 3 mSv/y at 430 years, and 99Tc has the peak values of 6.4 10 4 mSv/y at 570 years. The late peak values are due to 14C, with 3.9 10 6 mSv/y at 7.9 ky. 59Ni and 94Nb also contributes, with 1.42 10 4 mSv/y at 129 ky and 8.74 10 7 mSv/y at 167 ky years, respectively (Fig. 3II) (Table 3). The corresponding annual dose values of the other radionuclides are totally insignificant (less than 10 15 mSv/y).
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Table 3 The maximum attained values of annual ingestion dose as well as time of attainment for the most significant radionuclides and for each scenario. Radionuclide
Scenario
Hwat,
3
1 2 3
9.2 10 10 1.9 10 10 6.12 10 4
14 y 100 y 390 y
1 2 3
1.75 10 7 3.67 10 9 3.9 10 6
29.5 ky 35.5 ky 7.9 ky
1 2 3
510 9 510 9 1.42 10
– – 129 ky
1 2 3
510 9 510 9 8.74 10
1 2 3
3.13 10 4 mSv/y 3.13 10 4 mSv/y 6.4 10 4 mSv/y
7.2 ky 7.2 ky 570 y
1 2 3
1.83 10 4 mSv/y 1.82 10 4 mSv/y 1.2 10 3 mSv/y
5.4 ky 5.4 ky 430 y
H
14
C
59
Ni
94
Nb
99
Tc
129
I
max
(mSv/y)
4
7
Time
– – 167 ky
References Fig. 3. Graphical representation of the time evolution of the annual water ingestion dose Hwat received for a person from the critical group and corresponding to first (continuous line) and second (dashed line) scenario (I) as well as to the third scenario (II). It can be remark that in all three scenarios, the 0.3 mSv/y level is never attained.
By analyzing these data one can see that in each scenario the annual dose Hwat received for a person from critical group is well under the dose constraint of 0.3 mSv/y imposed by the national nuclear regulations (CNCAN, 2006) for the entire period of 1 Ma considered for the repository. For all radionuclides considered in this evaluation, the proposed disposal system seems to be very adequate. However, it is necessary to mention that this is a general purpose simulation, and that some input data specific for engineered barriers and waste form were taken from literature. The site specific distribution coefficients were determined only for 14 C, 60Co, cesium and strontium radioisotopes during the Saligny site characterization process (Dogaru et al., 2008).
4. Conclusions By considering the annual water ingestion dose as an important safety indicator, it was possible to demonstrate the fulfillment of the safety criteria established for a future Romanian VLLW repository for the 11 most important radionuclides generated by the Cernavoda Nuclear Power Plant. The proposed disposal system appears to be sufficient of coping with the given radionuclides as the peak values generated from the assessment of the radionuclides migration from the repository are all the time lower than the dose constraint of 0.3 mSv/y for the entire expected period of 1 Ma.
Amber 5, 2006. AMBER Users and References, Enviros Consulting Limited Building D5, Culham Science Centre, Culham, Oxfordshire, OX14 3DB, UK /www.enviros. com/amber/S. Chen, Q.Q., Kowe, R., Mobbs, S.F. Jones, K.A., 2007. Radiological Assessment of Disposal of Large Quantities of Very Low Level Waste in Landfill Sites, HPARPD-20. The study was funded by DEFRA and the Environmental Agency and was published by Health Protection Agency, Centre for Radiation, Chemical and Environmental Hazards, Radiation Protection Division, Chilton, Didcot, Oxfordshire OX110RQ, ISBN 978-0-85951-588-7. CNCAN, 2005. NDR-03, Regulations regarding radioactive waste classification, approved by CNCAN President Order No. 156/14.06.2005 (in Romanian). CNCAN, 2006. Regulations regarding the near-surface disposal of the radioactive wastes (NDR-05) approved by the CNCAN President’s Decree No. 400/ 13.12.2005 (in Romanian). Dogaru, D., Niculae, O., Jinescu, G., Duliu, O.G., 2008. The influence of the chemical properties of the geological formations on the migration parameters of the radionuclides from Saligny repository. Scientific Study & Research—Chemistry & Chemical Engineering, Biotechnology, Food Industry IX (2), 153–168. Durdun, I., Marunteanu, C., Andrei, V., 2001. The adaptation of natural (geological) barriers for radioactive LILW near surface disposal, WM’01 Conference, Tucson, AZ, USA, February 25–March 1, 2001 (in Romania). IAEA, 2003. IAEA-TECDOC-1372, Safety indicators for the safety assessment of radioactive waste disposal, IAEA, Vienna. IAEA, 2004. Improvement of safety assessment methodologies for near surface disposal facilities. Review and Enhancement of Safety Assessment Approaches and Tools, vol. I, IAEA-ISAM-1, Vienna. IAEA, 2006. Fundamental Safety Principles. Safety Fundamentals No. SF-1, IAEA, Vienna. Niculae, O., Vladescu, L.G., Naum, M., 2008. Application of the IAEA safety assessment methodology during the decision making and licensing process for the Romanian radioactive waste final repository. Romanian Reports in Physics 60 (2), 299–313. Park, J.B., Park, J-W., Lee, E-Y., Kim, C-L., 2003. Experiences from the source-term analysis of a low and intermediate level radwaste disposal facility, WM’03 Conference, Tucson, AZ, USA, February 23–27. SITON, 2007. Conceptual design study for the disposal of very low level radioactive waste resulted from nuclear facilities decommissioning, Internal Report no. 1595/2007, SITON, Romania (in Romanian). UKEA, 2007. Environmental Agency, Scottish Executive and DEFRA, Nov. 2007, Dose implications of very low level radioactive waste disposal, Project UKRSR09, Final Report /www.fwr.org/snifrprt.htmS.
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Preliminary safety concept for disposal of the very low level radioactive waste in Romania O. Niculae a,, V. Andrei a, G. Ionita a, O.G. Duliu b a b
ANDRAD National Agency for Radioactive Waste, 1 Campului, 15400 Mioveni, Arges County, Romania University of Bucharest, Department of Atomic and Nuclear Physics, P.O. Box MG-11, 077125 Magurele (Ilfov), Romania
a r t i c l e in fo
Keywords: Very low level radioactive waste Radioactive waste disposal Safety assessment Safety indicators Safety concept
abstract In Romania, there are certain nuclear installations in operation or under decommissioning, all of them representing an important source of very low level waste (VLLW). This paper presents an overview on the approach of the VLLW management in Romania, focused on those resulted from the nuclear power plants decommissioning. At the same time, the basic elements of safety concept, together with some safety evaluations concerning VLLW repository are presented and discussed too. & 2009 Elsevier Ltd. All rights reserved.
1. Introduction
60
Co, 134Cs, 137Cs, 63Ni, 59Ni, 95Zr/ 95Nb, 94Nb, 90Sr, 99Tc, 129I, 3H, C will be considered and treated as VLLW. Therefore, in this paper we present the basic elements of safety concept, together with some preliminary results of evolution scenarios developed for the Saligny VLLW Repository. 14
The very low level waste (VLLW) represents a category of waste with activity varying between those of conventional waste, that can be disposed in a landfill disposal and those of low and intermediate level waste which need usually to be disposed in near surface disposal facilities provided with a multi-barrier disposal system. According to the National Commission for the Control of Nuclear Activities (CNCAN) VLLW are represented by the short lived radioactive waste, with the specific activity higher than the exclusion levels stipulated by the regulations, but with a radioactive content lower than the level established by the Romanian Regulatory Body for the low level radioactive waste (CNCAN, 2005). At the present time in Romania there are one operational Nuclear Power Plant (NPP) at Cernavoda, and one decommissioned research reactor at the Institute of Nuclear Physics and Nuclear Engineering in Bucharest. The VLLW resulted from the operation of Cernavoda NPP, i.e. (i) compactable waste, (ii) noncompactable waste, (iii) organic liquids, (iv)–(v) spent filter cartridges and the low activity spent resins will be deposited in the Saligny VLLW Repository, planned to be built in the vicinity of the Cernavoda NPP. The chosen site represents a 43 m thick plateau consisting of an alternation of silty loess, clayey loess, red clay and as well as pre-quaternary clay considered as unsaturated, all of them laying on a compact Baremian limestone fundament, host of the main aquifer, and considered as a saturated zone (Durdun et al., 2001) (Fig. 1). In accordance with the first results of the characterization process, the following radionuclides
Corresponding author. Tel.: +40 021 3140392; fax: +40 021 3140391.
E-mail address: [email protected] (O. Niculae). 0969-8043/$ - see front matter & 2009 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2009.01.061
2. Theoretical background 2.1. Prerequisites for development of the safety concept As a prerequisite for the development of the VLLW safety concept and disposal strategy, ANDRAD initiated in 2007, a preliminary research and development plan, with the following subjects: (i) documentation on the international solutions for disposal of VLLW, (ii) safety aspects, (iii) a preliminary inventory of VLLW resulted from Cernavoda NPP operation and decommissioning, based on the data obtained from Cernavoda NPP, and (iv) the development of a disposal concept. The initial study on the development of a VLLW repository has already been completed (SITON, 2007), and other researches will be concluded in 2008, and in the next years. Based on the international practice, and taking into account the recommendations of the Romanian Regulatory Body, it was proposed a trench type repository, located on Saligny site, in the Cernavoda exclusion area (SITON, 2007). The main objective in the design of the repository will be to minimise the impact of the ionising radiation emmited by gases, particulates and leachates. This objective is achieved by isolating waste from the accessible environment, containing the waste within the disposal facility, as well as attenuating the associated release of radionuclides from the facility as far as practicable.
ARTICLE IN PRESS 936
O. Niculae et al. / Applied Radiation and Isotopes 67 (2009) 935–938
Fig. 1. A vertical section showing the geological profile of the place designed for repository location coresponding to the first and third scenarios (I) as well as to the second one (II).
To develop the safety concept for the VLLW repository, some basic questions were outlined and adequate responses are given in the present paper. The first question ‘‘what is necessary to demonstrate for each repository stage, in order to outline its safe evolution’’ will define the safety objective. Other questions like ‘‘what is the proof based on?’’, ‘‘which methodology is suitable for demonstration?’’ and ‘‘which are the parameters that illustrate the achivement of the safety objective?’’ define safety pillars, safety assessment methodology and safety indicators. Finally, the question ‘‘which are the possibilities to verify/maintain the repository safety?’’ will define the necessary management to sustain the repository safety. The answer to these questions are summary presented in the Section 3 of the present paper.
2.2. Safety evaluation The evolution of a repository is described by the evolution scenarios, developed both for normal and abnormal conditions for its entire life periods. These scenarios refer to all disposal system elements, and have as result the safety indicators. In our case we have used the ISAM methodology developed by the IAEA (2004) and applied previously for the planned Saligny Near Surface Repository (Niculae et al., 2008; Dogaru et al., 2008).
Consequently, by considering the water infiltration through the same reinforced concrete repository located in 65 m 45 m 14 m trench we developed three scenarios based on different construction variants but with some common elements, such as the compacted loess liner. As a result, when the infiltrated water reaches the waste, the radionuclides are washed out in both unsaturated geological layers and the saturated zone. Further, in the saturated zone we considered a well which represents the water source for a farm located at the limit of repository, following a similar scenario to the one proposed by Chen et al. (2007) and UKEA (2007). According to the geo-technical data presented by Dogaru et al. (2008) for the Saligny site, the water infiltration rate was considered to be about 0.02 m/y for the first two scenarios and 20 times higher then that for the third one (0.4 m/y). In the first scenario, the planned repository is located, as mentioned above, over a compacted loess liner of 3 m depth (Fig. 1I). In the second scenario, we considered the same trench but with a reinforced concrete liner of 1 m depth, a foundation ground of 3 m depth (compacted loess), as well as an upper cover with a thickness of about 1.5 m (Fig. 1II). The third scenario differed from the first one by the water infiltration rate. For the first and the third scenarios, we considered the infiltration starting at the moment when the repository was sealed. For the second scenario, the water infiltration rate was considered zero for the first 50 years after the repository closure (by taking into account the effectiveness of the final cover), and 0.02 m/y after that. All calculations were done by using the AMBER 5.0 (2006) computer code which considers a global compartment model and integrates all system components, migration processes as well as the exposure mechanisms. The simplified conceptual model of the repository has been developed using the interaction matrix method. The safety indicator was considered to be the dose received by a person (from the critical group) by water ingestion (Niculae et al., 2008). According to the hypotheses mentioned above, the following inventory was considered in the evaluation: 3H(100 GBq), 137Cs (1 MBq), 134Cs(1 MBq), 90Sr(1 MBq), 60Co(10 MBq), 59Ni(10 GBq), 63 Ni(10 GBq), 99Tc(1 GBq), 129I(10 MBq), 94Nb(100 MBq) and 14C (1 GBq). Tables 1 and 2 present the characteristic parameters of the Saligny site as well as the distribution coefficients for some radionuclides experimentally determined for the Saligny site (Dogaru et al., 2008) as well as presented in literature (Park et al., 2003). By taking into account the long lived radionuclides Table 1 Characteristic parameters of model compartments, Saligny site, as input data for safety indicators assessment. Horizon
Average values of dry density (kg m 3)
Average values of water filled porosity (%)
Depth of model compartments (m)
Waste form Compacted loess liner Concrete liner (only for scenario 2) Compacted loess foundation ground (only for scenario 2) Silty loess Clayey loess: Layer Iab1 Upper layer Ib Layer Iab2 Lower layer Ib Red clay Pre-quaternary clay
2500 1760 2500
15 32 15
14 3 1
1760
32
3
1540
12
1
1780 1570 1720 1690 1760 1760
26 14 25 25 32 31
6 4 2 6 8 16
Numerical values of all parameters are reproduced from Dogaru et al. (2008).
ARTICLE IN PRESS O. Niculae et al. / Applied Radiation and Isotopes 67 (2009) 935–938
937
Table 2 Distribution coefficients (kd) for the model compartments. Radionuclide
137
Csa Coa 90 a Sr 14 a C 99 Tcb 129 b I 60
a b
kd in the model compartments (m3 kg
1
)
Waste form and concrete liner
Aquifer
Silty loess
Clayey loess
Red clay
Pre-quaternary clay
0.02 0.01 0.5 0.2 0.001 0.001
10 200 0.5 0.1 0.01 0.001
0.774 0.033 0.006 0.003 0.001 0.001
1.131 0.030 0.011 0.005 0.001 0.001
4.131 0.031 0.012 0.008 0.001 0.001
2.366 0.030 0.012 0.005 0.001 0.001
After Dogaru et al. (2008). After Park et al. (2003).
Fig. 2. The flow-chart of the safety concept proposed for VLLW repository.
from inventory (see Tables 1 and 2) a total period of 1 Ma was considered for the evaluation.
3. Results and discussion 3.1. Safety concept Fig. 2 presents a flow-chart of safety concept proposed for VLLW repository, with the purpose to integrate all factors, arguments and the evaluation methods used to demonstrate the safe evolution of the repository during 1 Ma. The safety objectives for each repository stage are the safety criteria, specified in the nuclear legislation as the practical way to fulfil the IAEA (2006) safety principles. According to (IAEA, 2003), the most important safety indicators are the radiological dose and radiological risk, their reference values being expressed by means of the dose limit or constraint as well as the risk limit or risk constraint. In Romania, according to (CNCAN, 2006), for non-professional irradiation, the reference value for the radiological dose received for public during one year, called the dose limit, is of 1 mSv/y, with a dose constraint of 0.3 mSv/y. This reference value was established for all the life periods of a near surface repository. 3.2. Safety evaluation In Fig. 3 we present the results of our calculations expressed by the annual water ingestion dose: Hwat received by a person from the critical group and corresponding to each of these scenarios. At the same time in Table 3 we have reproduced the Hwat maximum
values as well as the time these values are attained for each scenario too. In the first scenario, the early contribution comes from the mobile 3 H and 129I. Tritium reaches a peak value of 9.2 10 10 mSv/y at 14 years from the repository closure, while 129I presents two peaks: an early one of 3.8 10 13 mSv/y and a final one of 1.83 10 4 mSv/y at 8 y and, respectively, 5.4 ky after closure. Carbon and Technetium reach their maxima later: 29.5 ky for 14C (1.75 10 7 mSv/y) and 7.2 ky for 99Tc (3.13 10 4 mSv/y) (Fig. 3I) (Table 3). In the second scenario, tritium reaches its peak value of only 1.9 10 10 mSv/y at about 100 years after repository closure while 129 I and 99Tc remained unchanged, with 1.82 10 4 mSv/y at 5.4 ky, and 3.13 10 4 mSv/y at 7.2 ky, respectively. The 14C contributes with a dose value of 3.67 10 9 mSv/y at 35.5 ky, significantly lower than in the case of the first scenario (Fig. 3I) (Table 3). By comparing the results of Scenarios 1 and 2, one can see that the main effect of concrete liner is limited to 14C and 3H by reducing their annual water ingestion dose with about 98% and 79%, respectively, and by delaying the 3H peak value with 86 years. In the third scenario, the effect of the increased infiltration rate is visible as all peak values appear earlier and are significantly greater than in the previous scenarios. Therefore, 3H has an early peak value of 6.12 10 4 mSv/y at 39 years after closure, 129I presents a peak value of 1.2 10 3 mSv/y at 430 years, and 99Tc has the peak values of 6.4 10 4 mSv/y at 570 years. The late peak values are due to 14C, with 3.9 10 6 mSv/y at 7.9 ky. 59Ni and 94Nb also contributes, with 1.42 10 4 mSv/y at 129 ky and 8.74 10 7 mSv/y at 167 ky years, respectively (Fig. 3II) (Table 3). The corresponding annual dose values of the other radionuclides are totally insignificant (less than 10 15 mSv/y).
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Table 3 The maximum attained values of annual ingestion dose as well as time of attainment for the most significant radionuclides and for each scenario. Radionuclide
Scenario
Hwat,
3
1 2 3
9.2 10 10 1.9 10 10 6.12 10 4
14 y 100 y 390 y
1 2 3
1.75 10 7 3.67 10 9 3.9 10 6
29.5 ky 35.5 ky 7.9 ky
1 2 3
510 9 510 9 1.42 10
– – 129 ky
1 2 3
510 9 510 9 8.74 10
1 2 3
3.13 10 4 mSv/y 3.13 10 4 mSv/y 6.4 10 4 mSv/y
7.2 ky 7.2 ky 570 y
1 2 3
1.83 10 4 mSv/y 1.82 10 4 mSv/y 1.2 10 3 mSv/y
5.4 ky 5.4 ky 430 y
H
14
C
59
Ni
94
Nb
99
Tc
129
I
max
(mSv/y)
4
7
Time
– – 167 ky
References Fig. 3. Graphical representation of the time evolution of the annual water ingestion dose Hwat received for a person from the critical group and corresponding to first (continuous line) and second (dashed line) scenario (I) as well as to the third scenario (II). It can be remark that in all three scenarios, the 0.3 mSv/y level is never attained.
By analyzing these data one can see that in each scenario the annual dose Hwat received for a person from critical group is well under the dose constraint of 0.3 mSv/y imposed by the national nuclear regulations (CNCAN, 2006) for the entire period of 1 Ma considered for the repository. For all radionuclides considered in this evaluation, the proposed disposal system seems to be very adequate. However, it is necessary to mention that this is a general purpose simulation, and that some input data specific for engineered barriers and waste form were taken from literature. The site specific distribution coefficients were determined only for 14 C, 60Co, cesium and strontium radioisotopes during the Saligny site characterization process (Dogaru et al., 2008).
4. Conclusions By considering the annual water ingestion dose as an important safety indicator, it was possible to demonstrate the fulfillment of the safety criteria established for a future Romanian VLLW repository for the 11 most important radionuclides generated by the Cernavoda Nuclear Power Plant. The proposed disposal system appears to be sufficient of coping with the given radionuclides as the peak values generated from the assessment of the radionuclides migration from the repository are all the time lower than the dose constraint of 0.3 mSv/y for the entire expected period of 1 Ma.
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