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Contact dose rates and relevant radioactive inventory in ITER TBM systems
Fusion Engineering and Design 86 (2011) 2690–2693
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Contact dose rates and relevant radioactive inventory in ITER TBM systems M. Zucchetti a,∗ , L. Guerrini b , Y. Poitevin b , I. Ricapito b , M. Zmitko b a b
EURATOM/ENEA Fusion Association Politecnico di Torino, Torino, Italy Fusion for Energy, ITER Department, Test Blanket Modules Group, Barcelona, Spain
a r t i c l e
i n f o
Article history: Available online 31 May 2011 Keywords: ITER Test Blanket Module Radioactive inventory Contact dose rates
a b s t r a c t The determination of the radioactive inventory and of the contact dose rates in the different ITER Test Blanket Modules systems is carried out, both for Helium-Cooled Lithium–Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated. © 2011 Elsevier B.V. All rights reserved.
1. Introduction The testing of Tritium Breeder Blanket concepts is one of the ITER missions and has been recognized as an essential milestone in the development of a future reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): • the Helium-Cooled Lithium–Lead (HCLL) concept which uses the eutectic Pb–15.7Li as both breeder and neutron multiplier, • the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles (Li4 SiO4 or Li2 TiO3 ) as breeder and beryllium pebbles as neutron multiplier. The determination of the radioactive inventory and of the contact dose rates in the different TBM systems is an essential step in view of the evaluation of the radiological safety of the blanket and of the ITER machine overall, both in case of routine operation and in case of accident. The following systems and locations have been addressed in this study: 1. For HCLL TBM: • PbLi system: piping in the port interspace and system/components in the Port Cell (PC), including the cold trap of the PbLi system.
∗ Corresponding author. Tel.: +39 011 5644464. E-mail addresses: [email protected] (M. Zucchetti), [email protected] (L. Guerrini). 0920-3796/$ – see front matter © 2011 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2011.01.014
• Tritium Removal System (TRS): system/components in the PC and in the Tritium building. • Helium Cooling System (HCS): piping in the port interspace, system/components in the PC and in the Chemical Volume and Control System (CVCS) area of the tokamak building. • Coolant Purification System (CPS): system/components in the CVCS area of the tokamak building. 2. For HCPB TBM: • HCS and CPS: same as for HCLL. • Tritium Extraction System (TES): piping in the port interspace, system/components in the PC and in the Tritium building. A scheme showing the layout of the ITER components after the TBM is shown in Fig. 1. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules [1–4]. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated: • For HCLL, EUROFER corrosion products contamination, and activation of PbLi. • For HCPB, contamination powder from the ceramic breeder material into the He gas, possible contamination of He purges gas by Be dust from Be pebble beds, and EUROFER corrosion products into the helium coolant or purge gas.
2. Main assumptions The main assumptions/input data are the following:
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Fig. 1. A scheme of the ITER components and the TBM. (1) TBM, (2) frame, (3) plug shield, (4) pipe system, (5) bioshield and (6) AEU.
• As a source of radioactivity, only the TBM itself is considered [1,2]. Radioactive decay patterns vary strongly depending on material composition (especially in LiPb) and time scenario [3,4]. • No neutron streaming or contributions from the ITER machine components is considered. • For HCPB TES powder from the ceramic breeder material is considered as a source of contamination. • Corrosion rate of EUROFER in PbLi to be considered in the estimations: 30 m/year for the HCLL TBM and 20 m/year for the rest of the system (lower temperature than in the TBM itself). 3. Activity and dose estimates for HCLL-TBM There will be two sources of radioactive material in the ancillary system, contributing to the dose rates: • Activated PbLi, • EUROFER corrosion products. All the other contributions may be neglected. For the corrosion products contamination, it is important to know the corrosion rate of EUROFER in PbLi to be considered in the estimations: we have assumed a 30 m/year for the HCLL TBM (consistent with that assumed for US DCLL of 20 m/year at 450 ◦ C) [5]. Over the 3-year operational lifetime, we may consider that LiPb will spend 50% of the time inside the blanket, so 1.5 years. The blanket internal wetted surface by LiPb is around 10.9 m2 , so according to our assumption we estimate the actual value to be 2.5 kg. These corrosion products will be partially captured by the cold trap, a dedicated component for removing corrosion products and impurities from the circuit. We may assume that nearly half of the activation products are held by the cold trap (50% efficiency): if we assume that, then around 1.25 kg of corrosion products will stay inside LiPb and about 1.25 kg will be captured by the cold trap. According to activation calculations, and accounting for radioactive decay, Mn-56 and V-52 will dominate the corrosion products (gamma emitters) at 0 s after shutdown. Concerning the Pb–Li system (piping in the port interspace and system/components in the Port Cell (PC)), the system comprises the following main components: the lithium–lead circuit, which includes a storage tank, a single circulating pump and piping to and from the TBM. The storage tank is designed to hold the almost entire inventory of lithium lead when the entire circuit (including the TBM) is drained. For the storage tank, the dominating source of radioactivity is the LiPb itself, while the activation/contamination of the structure can be considered negligible. Concerning the activation of PbLi, we will assume that activation/dose rate source in LiPb circuit can be taken as the values computed by [1], reduced by one half: this is justified by the fact
that the storage tank is designed to hold the entire inventory of lithium lead when the entire circuit (including the TBM) is drained, so we could think in principle that the LiPb spends half of its time in the blanket and half outside in the circuit. Moreover, one has to take into account the shielding effect of the tank structure, in order to compute the contact dose rate on the external surface of it. This effect has been considered by performing shielding calculations by means of the MICROSHIELD code [6]. The tank structure has been assumed to be a steel layer with 10 mm of thickness. Dose rate estimates have then been computed for LiPb: 1.31E+03 Sv/h at shutdown, 0.81E−01 Sv/h at 4 days and 2.60E−02 Sv/h at 12 days. For the pump and the piping, the main source of dose rate after draining is the LiPb film deposited on the inner surfaces. Based on US DCLL safety evaluations and on TRITEX experience, after draining the loop they found PbLi films on the pipe walls that were ∼45 mg/cm2 [5]. Independent evaluations estimate the thickness of the LiPb layer to be around 1 mm. This brings to a total mass of LiPb in the piping of about 3.6 kg. According to activation calculations, this film will be radioactive with the radioactive inventory after some days dominated, in terms of nuclides relevant for gamma dose rate, by Pb-203 and Ta-182: contact dose rate after 4 days of decay is 75% to Pb-203 and 20% to Ta-182 [1]. The dose rate on the outer surface of the pipes and pump, due to this inner LiPb film, may be estimated – included the shielding effect of the structures (pipe specification as follows: 1 in., schedule 40S, i.e. 3.38 mm wall thickness) – by means of an ad-hoc calculation with the MICROSHIELD code [6]. The zone has been modelled as a 1-mm source layer, shielded by a 3.38 mm shield layer. Pump and piping dose rate, due to LiPb activation, are the following: 3.78 Sv/h at shutdown, 4.57E−04 Sv/h at 4 days, and 9.34E−05 Sv/h at 12 days. Concerning the cold trap, for removing corrosion products and impurities from the circuit, the EUROFER corrosion products will – according to the assumptions – half (50%) be drained by this component, which however could be periodically purged and have the corrosion products removed: we may assume, that all the corrosion products (1.25 kg) are present in the cold trap. Consequently, taking into account the energy of the gamma emissions of the main activation products, the cold trap can be modelled by means of the MICROSHIELD code [6] as a sphere surrounded by a 10 mm layer of steel shield. The dose can be estimated as follows: 3.43 Sv/h at shutdown, 0.159 Sv/h at 4 days and 9.43E−02 Sv/h at 12 days. Concerning the shielding to reduce radiation fields outside the module containing all of the above equipment, the dose rate at the surface of this component is mainly due to the LiPb activity and not to the corrosion products: a simulation by means of the MICROSHIELD code has determined that a shield thickness of 5 cm of lead, or of 10 cm of steel, approximately performs a two-order of magnitude reduction of the dose rate. In particular, a 10-cm shield
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layer of steel reduces the outer dose rate, given the source nuclides characteristics, by 1/167th of the original. The most critical points should be the ones in the proximity of the storage tank. In that case, a dose rate at the surface of the shielding may be estimated as • 7.81E+00 Sv/h at 0 s; • 4.83E−04 Sv/h at 4 days; • 1.55E−04 Sv/h at 12 days. In the other points of the shielding outside the module, dose rate will be relevantly inferior. These data may possibly be reduced if necessary by a better shielding with a higher shield thickness. Concerning the Tritium Extraction System (TES) and TRS: system/components in the Port Cell and in the Tritium building, the proposed TES is split in two parts: the first one, which includes the Q2 O adsorption column, is located in the Port Cell. The second one, which includes the Q2-TSA with its regeneration circuit and the by-pass line with the metal getter bed, should be placed in a glove box in the Tritium Plant Building. Main source of contamination for this system is obviously tritium, which is not considered in this estimate, since it does not contribute directly to external dose rates. The input of TES is the He gas that has been in contact with the LiPb in order to extract the tritium. Some activation products of LiPb could then be present in the He stripping gas, and be transported to the TES. A filter (component no. 2 in the figure above) is placed in order to remove most of those particles, and should have a high efficiency, around or better than 99%. We can first of all deduce that all the other components but the filter has a negligible non-tritium contamination. Excluding again Tritium which actually contributes to the 98.3% of the activity of LiPb at 4 days, the other two main activation products in LiPb are Ta-182 and Pb-203. These are heavy metals, with such partial pressures that their presence in the He gas can be neglected. Other minor activation products, contributing to a very low quota of LiPb activation, are Na-24, Co-58, Na-22, In-113m, Sb-125 and Mn-54. For most of them (Co, In, Sb, Mn), the same assumptions as for Ta and Pb can be applied. According to previous experience in fission fast-breeder reactors, limited amounts of Na isotopes could pass into the He gas and then be transported up to the TRS filter, however with a negligible concentration. Concentration of Na isotopes in the filter due to transport from He should be determined with more accuracy once the layout of the filter is determined. Moreover, the shielding effect of the filter steel case (10 mm) must be taken into account. A MICROSHIELD simulation, taking into account the source nuclides (Na-22 and Na-24) has estimated this shielding effect to be of 81% on the dose rate, i.e., the dose rate at the surface must be multiplied for a factor 0.81. Concerning Helium Cooling System (HCS): piping in the port interspace, system/components in the PC and in the CVCS area of the tokamak building, and Coolant Purification System (CPS): system/components in the CVCS area of the tokamak building, activity in these components can be practically neglected. 4. Activity and dose estimates for HCPB-TBM system There will be three sources of radioactive material in the system: • Contamination powder from the ceramic breeder material into the He purge gas. • Possible contamination of He purge gas by Be dust from Be pebble beds.
• EUROFER corrosion products into the helium coolant or purge gas. This contribution may be neglected with respect to the first two ones.
For the contamination powder from the ceramic breeder material into the He gas, it is important to determine the quantity of powder effectively passed into the gas. Few relevant data about this are available in literature. Tests were performed at ENEA (Helica tests) [7]. In those tests, a fraction of 0.7% of the ceramic was dissolved into the gas, with a dust size of less than 130 m. Those results are not easily sizeable to “TBM relevant” conditions. We may consider that around 0.5% of the total ceramic breeder mass passes into the He gas and it is deposited on the TES filter of the purge gas circuit inside the Port Cell. Concerning the Tritium Extraction System (TES): piping in the port interspace, system/components in the Port Cell and in the Tritium building, as previously mentioned, the proposed TES is split in two parts: the first one, which includes the Q2 O adsorption column, is located in the Port Cell. The second one, which includes the Q2-TSA with its regeneration circuit and the by-pass line with the metal getter bed, should be placed in a glove box in the Tritium Plant Building. The main source of contamination for this system is obviously tritium, which is not considered in this estimate, since it does not contribute directly to external dose rates. The input of TES is the He gas that has been in contact with the ceramic breeder in order to extract the tritium. Some activation products of ceramic breeder could then be present in the He gas, and be transported to the TES. A filter is placed in order to remove most of those particles, and should have a high efficiency, around or better than 99%. We can first of all deduce that all the other components but the filter have a negligible non-tritium contamination. As we defined previously, we may consider that around 0.5% of the total ceramic breeder mass passes into the He gas and is deposited on the TES filter of the purge gas circuit inside the Port Cell. Moreover, it must be also taken into account that about 0.1% of the mass of Beryllium passes into the He gas and then it is deposited too on the TES filter. Concentration of activation isotopes in the filter due to transport from He should be determined with more accuracy once the layout of the filter is determined. However we can make the very conservative assumption that all of them are present in the filter that brings to a total quantity of 500 g of ceramic breeder, and 177 g of beryllium. As far as the dose rates are concerned, the filter has been modelled by means of the MICROSHIELD code, in order to estimate the contact dose rate at the surface of the filter container, taking also into account the presence of a filter container made of 10 mm of steel, and its consequent shielding. No information was available on the dimensions and the shape of the filter. Such informations are essential to compute the radioactivity concentration of the activation products in the filter itself, and therefore can relevantly influence the dose rates that are computed by the model. As a tentative solution, the filter has been modelled with a cylindrical shape, with a volume of approximately 10 l (10.000 cm3 ). Results are summarized as follows: the ceramic breeder contribution is 6.07 Sv/h at shutdown, 5.00E−05 Sv/h at 4 days, 1.64E−05 Sv/h at 12 days. Beryllium contribution is 1.42E−02 Sv/h at shutdown, 1.630E−04 Sv/h at 4 days, 1.564E−04 Sv/h at 12 days. Total dose rate is therefore 6.08 Sv/h at shutdown, 2.13E−04 Sv/h at 4 days and 1.73E−04 Sv/h at 12 days. The above values turn out to be relatively high since the quite conservative assumption that all the activation products are present in the filter.
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Also, lacking other specifications, the assumption that the activation products are uniformly distributed inside the filter has been made, and this also may contribute to overestimate the dose rate. When more informations are available on the actual amount of activation products in the filter (for example, frequency of filter purge), dose rate values can be reduced in consequence of these better estimates. Concerning the Helium Cooling System (HCS) and the Coolant Purification System (CPS), the activity in these components is practically negligible. 5. Conclusions The determination of the radioactive inventory and of the contact dose rates in the different TBM systems has been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, completely available for both the blanket modules. The possible sources of
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radioactive material in all the systems have been individuated and their contributes estimated. In general, for both HCLL and HCPB systems, radioactivity inventory and contact dose rates will be used for an ORE analysis. References [1] L. Petrizzi, R. Villari, L. Auditore, G. Cambi, D.G. Cepraga, “HCLL: TBM Design, Integration and Analysis. Activation and Decay Analysis of TBM in Iter Machine” Final Report: Task Tw6-Ttbc 001 Del. 2, ENEA, November 2007. [2] U. Fischer, P. Pereslavtsev, “Activation and decay heat analyses of the HCPB TBM with detailed output by radio-nuclides and sensitivity to the irradiation time in ITER (2 irradiation scenarios)”, Final report on the EFDA task TW6-TTBB-001, Deliverable 2, FZK Karlsruhe (D), March 2007. [3] L. Petrizzi, et al., Helium-cooled lithium lead: activation analysis of the test blanket module in ITER, Fusion Eng. Des. 83 (2008) 1244–1248. [4] P. Pereslavtsev, U. Fischer, Activation and afterheat analyses for the HCPB test blanket module in ITER, Fusion Eng. Des. 83 (2008) 1742–1746. [5] B. Merrill, L. Cadwallader, M. Dagher, A. Ying, “US DCLL TBM Safety Update”, TBM Safety Workshop, Aix-en-Provence, France, March 3rd, 2008. [6] See MICROSHIELD website: http://www.radiationsoftware.com/mshield.html. [7] I. Ricapito, Personal communication, email, October 26th, 2009.
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Contact dose rates and relevant radioactive inventory in ITER TBM systems M. Zucchetti a,∗ , L. Guerrini b , Y. Poitevin b , I. Ricapito b , M. Zmitko b a b
EURATOM/ENEA Fusion Association Politecnico di Torino, Torino, Italy Fusion for Energy, ITER Department, Test Blanket Modules Group, Barcelona, Spain
a r t i c l e
i n f o
Article history: Available online 31 May 2011 Keywords: ITER Test Blanket Module Radioactive inventory Contact dose rates
a b s t r a c t The determination of the radioactive inventory and of the contact dose rates in the different ITER Test Blanket Modules systems is carried out, both for Helium-Cooled Lithium–Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated. © 2011 Elsevier B.V. All rights reserved.
1. Introduction The testing of Tritium Breeder Blanket concepts is one of the ITER missions and has been recognized as an essential milestone in the development of a future reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): • the Helium-Cooled Lithium–Lead (HCLL) concept which uses the eutectic Pb–15.7Li as both breeder and neutron multiplier, • the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles (Li4 SiO4 or Li2 TiO3 ) as breeder and beryllium pebbles as neutron multiplier. The determination of the radioactive inventory and of the contact dose rates in the different TBM systems is an essential step in view of the evaluation of the radiological safety of the blanket and of the ITER machine overall, both in case of routine operation and in case of accident. The following systems and locations have been addressed in this study: 1. For HCLL TBM: • PbLi system: piping in the port interspace and system/components in the Port Cell (PC), including the cold trap of the PbLi system.
∗ Corresponding author. Tel.: +39 011 5644464. E-mail addresses: [email protected] (M. Zucchetti), [email protected] (L. Guerrini). 0920-3796/$ – see front matter © 2011 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2011.01.014
• Tritium Removal System (TRS): system/components in the PC and in the Tritium building. • Helium Cooling System (HCS): piping in the port interspace, system/components in the PC and in the Chemical Volume and Control System (CVCS) area of the tokamak building. • Coolant Purification System (CPS): system/components in the CVCS area of the tokamak building. 2. For HCPB TBM: • HCS and CPS: same as for HCLL. • Tritium Extraction System (TES): piping in the port interspace, system/components in the PC and in the Tritium building. A scheme showing the layout of the ITER components after the TBM is shown in Fig. 1. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules [1–4]. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated: • For HCLL, EUROFER corrosion products contamination, and activation of PbLi. • For HCPB, contamination powder from the ceramic breeder material into the He gas, possible contamination of He purges gas by Be dust from Be pebble beds, and EUROFER corrosion products into the helium coolant or purge gas.
2. Main assumptions The main assumptions/input data are the following:
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Fig. 1. A scheme of the ITER components and the TBM. (1) TBM, (2) frame, (3) plug shield, (4) pipe system, (5) bioshield and (6) AEU.
• As a source of radioactivity, only the TBM itself is considered [1,2]. Radioactive decay patterns vary strongly depending on material composition (especially in LiPb) and time scenario [3,4]. • No neutron streaming or contributions from the ITER machine components is considered. • For HCPB TES powder from the ceramic breeder material is considered as a source of contamination. • Corrosion rate of EUROFER in PbLi to be considered in the estimations: 30 m/year for the HCLL TBM and 20 m/year for the rest of the system (lower temperature than in the TBM itself). 3. Activity and dose estimates for HCLL-TBM There will be two sources of radioactive material in the ancillary system, contributing to the dose rates: • Activated PbLi, • EUROFER corrosion products. All the other contributions may be neglected. For the corrosion products contamination, it is important to know the corrosion rate of EUROFER in PbLi to be considered in the estimations: we have assumed a 30 m/year for the HCLL TBM (consistent with that assumed for US DCLL of 20 m/year at 450 ◦ C) [5]. Over the 3-year operational lifetime, we may consider that LiPb will spend 50% of the time inside the blanket, so 1.5 years. The blanket internal wetted surface by LiPb is around 10.9 m2 , so according to our assumption we estimate the actual value to be 2.5 kg. These corrosion products will be partially captured by the cold trap, a dedicated component for removing corrosion products and impurities from the circuit. We may assume that nearly half of the activation products are held by the cold trap (50% efficiency): if we assume that, then around 1.25 kg of corrosion products will stay inside LiPb and about 1.25 kg will be captured by the cold trap. According to activation calculations, and accounting for radioactive decay, Mn-56 and V-52 will dominate the corrosion products (gamma emitters) at 0 s after shutdown. Concerning the Pb–Li system (piping in the port interspace and system/components in the Port Cell (PC)), the system comprises the following main components: the lithium–lead circuit, which includes a storage tank, a single circulating pump and piping to and from the TBM. The storage tank is designed to hold the almost entire inventory of lithium lead when the entire circuit (including the TBM) is drained. For the storage tank, the dominating source of radioactivity is the LiPb itself, while the activation/contamination of the structure can be considered negligible. Concerning the activation of PbLi, we will assume that activation/dose rate source in LiPb circuit can be taken as the values computed by [1], reduced by one half: this is justified by the fact
that the storage tank is designed to hold the entire inventory of lithium lead when the entire circuit (including the TBM) is drained, so we could think in principle that the LiPb spends half of its time in the blanket and half outside in the circuit. Moreover, one has to take into account the shielding effect of the tank structure, in order to compute the contact dose rate on the external surface of it. This effect has been considered by performing shielding calculations by means of the MICROSHIELD code [6]. The tank structure has been assumed to be a steel layer with 10 mm of thickness. Dose rate estimates have then been computed for LiPb: 1.31E+03 Sv/h at shutdown, 0.81E−01 Sv/h at 4 days and 2.60E−02 Sv/h at 12 days. For the pump and the piping, the main source of dose rate after draining is the LiPb film deposited on the inner surfaces. Based on US DCLL safety evaluations and on TRITEX experience, after draining the loop they found PbLi films on the pipe walls that were ∼45 mg/cm2 [5]. Independent evaluations estimate the thickness of the LiPb layer to be around 1 mm. This brings to a total mass of LiPb in the piping of about 3.6 kg. According to activation calculations, this film will be radioactive with the radioactive inventory after some days dominated, in terms of nuclides relevant for gamma dose rate, by Pb-203 and Ta-182: contact dose rate after 4 days of decay is 75% to Pb-203 and 20% to Ta-182 [1]. The dose rate on the outer surface of the pipes and pump, due to this inner LiPb film, may be estimated – included the shielding effect of the structures (pipe specification as follows: 1 in., schedule 40S, i.e. 3.38 mm wall thickness) – by means of an ad-hoc calculation with the MICROSHIELD code [6]. The zone has been modelled as a 1-mm source layer, shielded by a 3.38 mm shield layer. Pump and piping dose rate, due to LiPb activation, are the following: 3.78 Sv/h at shutdown, 4.57E−04 Sv/h at 4 days, and 9.34E−05 Sv/h at 12 days. Concerning the cold trap, for removing corrosion products and impurities from the circuit, the EUROFER corrosion products will – according to the assumptions – half (50%) be drained by this component, which however could be periodically purged and have the corrosion products removed: we may assume, that all the corrosion products (1.25 kg) are present in the cold trap. Consequently, taking into account the energy of the gamma emissions of the main activation products, the cold trap can be modelled by means of the MICROSHIELD code [6] as a sphere surrounded by a 10 mm layer of steel shield. The dose can be estimated as follows: 3.43 Sv/h at shutdown, 0.159 Sv/h at 4 days and 9.43E−02 Sv/h at 12 days. Concerning the shielding to reduce radiation fields outside the module containing all of the above equipment, the dose rate at the surface of this component is mainly due to the LiPb activity and not to the corrosion products: a simulation by means of the MICROSHIELD code has determined that a shield thickness of 5 cm of lead, or of 10 cm of steel, approximately performs a two-order of magnitude reduction of the dose rate. In particular, a 10-cm shield
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layer of steel reduces the outer dose rate, given the source nuclides characteristics, by 1/167th of the original. The most critical points should be the ones in the proximity of the storage tank. In that case, a dose rate at the surface of the shielding may be estimated as • 7.81E+00 Sv/h at 0 s; • 4.83E−04 Sv/h at 4 days; • 1.55E−04 Sv/h at 12 days. In the other points of the shielding outside the module, dose rate will be relevantly inferior. These data may possibly be reduced if necessary by a better shielding with a higher shield thickness. Concerning the Tritium Extraction System (TES) and TRS: system/components in the Port Cell and in the Tritium building, the proposed TES is split in two parts: the first one, which includes the Q2 O adsorption column, is located in the Port Cell. The second one, which includes the Q2-TSA with its regeneration circuit and the by-pass line with the metal getter bed, should be placed in a glove box in the Tritium Plant Building. Main source of contamination for this system is obviously tritium, which is not considered in this estimate, since it does not contribute directly to external dose rates. The input of TES is the He gas that has been in contact with the LiPb in order to extract the tritium. Some activation products of LiPb could then be present in the He stripping gas, and be transported to the TES. A filter (component no. 2 in the figure above) is placed in order to remove most of those particles, and should have a high efficiency, around or better than 99%. We can first of all deduce that all the other components but the filter has a negligible non-tritium contamination. Excluding again Tritium which actually contributes to the 98.3% of the activity of LiPb at 4 days, the other two main activation products in LiPb are Ta-182 and Pb-203. These are heavy metals, with such partial pressures that their presence in the He gas can be neglected. Other minor activation products, contributing to a very low quota of LiPb activation, are Na-24, Co-58, Na-22, In-113m, Sb-125 and Mn-54. For most of them (Co, In, Sb, Mn), the same assumptions as for Ta and Pb can be applied. According to previous experience in fission fast-breeder reactors, limited amounts of Na isotopes could pass into the He gas and then be transported up to the TRS filter, however with a negligible concentration. Concentration of Na isotopes in the filter due to transport from He should be determined with more accuracy once the layout of the filter is determined. Moreover, the shielding effect of the filter steel case (10 mm) must be taken into account. A MICROSHIELD simulation, taking into account the source nuclides (Na-22 and Na-24) has estimated this shielding effect to be of 81% on the dose rate, i.e., the dose rate at the surface must be multiplied for a factor 0.81. Concerning Helium Cooling System (HCS): piping in the port interspace, system/components in the PC and in the CVCS area of the tokamak building, and Coolant Purification System (CPS): system/components in the CVCS area of the tokamak building, activity in these components can be practically neglected. 4. Activity and dose estimates for HCPB-TBM system There will be three sources of radioactive material in the system: • Contamination powder from the ceramic breeder material into the He purge gas. • Possible contamination of He purge gas by Be dust from Be pebble beds.
• EUROFER corrosion products into the helium coolant or purge gas. This contribution may be neglected with respect to the first two ones.
For the contamination powder from the ceramic breeder material into the He gas, it is important to determine the quantity of powder effectively passed into the gas. Few relevant data about this are available in literature. Tests were performed at ENEA (Helica tests) [7]. In those tests, a fraction of 0.7% of the ceramic was dissolved into the gas, with a dust size of less than 130 m. Those results are not easily sizeable to “TBM relevant” conditions. We may consider that around 0.5% of the total ceramic breeder mass passes into the He gas and it is deposited on the TES filter of the purge gas circuit inside the Port Cell. Concerning the Tritium Extraction System (TES): piping in the port interspace, system/components in the Port Cell and in the Tritium building, as previously mentioned, the proposed TES is split in two parts: the first one, which includes the Q2 O adsorption column, is located in the Port Cell. The second one, which includes the Q2-TSA with its regeneration circuit and the by-pass line with the metal getter bed, should be placed in a glove box in the Tritium Plant Building. The main source of contamination for this system is obviously tritium, which is not considered in this estimate, since it does not contribute directly to external dose rates. The input of TES is the He gas that has been in contact with the ceramic breeder in order to extract the tritium. Some activation products of ceramic breeder could then be present in the He gas, and be transported to the TES. A filter is placed in order to remove most of those particles, and should have a high efficiency, around or better than 99%. We can first of all deduce that all the other components but the filter have a negligible non-tritium contamination. As we defined previously, we may consider that around 0.5% of the total ceramic breeder mass passes into the He gas and is deposited on the TES filter of the purge gas circuit inside the Port Cell. Moreover, it must be also taken into account that about 0.1% of the mass of Beryllium passes into the He gas and then it is deposited too on the TES filter. Concentration of activation isotopes in the filter due to transport from He should be determined with more accuracy once the layout of the filter is determined. However we can make the very conservative assumption that all of them are present in the filter that brings to a total quantity of 500 g of ceramic breeder, and 177 g of beryllium. As far as the dose rates are concerned, the filter has been modelled by means of the MICROSHIELD code, in order to estimate the contact dose rate at the surface of the filter container, taking also into account the presence of a filter container made of 10 mm of steel, and its consequent shielding. No information was available on the dimensions and the shape of the filter. Such informations are essential to compute the radioactivity concentration of the activation products in the filter itself, and therefore can relevantly influence the dose rates that are computed by the model. As a tentative solution, the filter has been modelled with a cylindrical shape, with a volume of approximately 10 l (10.000 cm3 ). Results are summarized as follows: the ceramic breeder contribution is 6.07 Sv/h at shutdown, 5.00E−05 Sv/h at 4 days, 1.64E−05 Sv/h at 12 days. Beryllium contribution is 1.42E−02 Sv/h at shutdown, 1.630E−04 Sv/h at 4 days, 1.564E−04 Sv/h at 12 days. Total dose rate is therefore 6.08 Sv/h at shutdown, 2.13E−04 Sv/h at 4 days and 1.73E−04 Sv/h at 12 days. The above values turn out to be relatively high since the quite conservative assumption that all the activation products are present in the filter.
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Also, lacking other specifications, the assumption that the activation products are uniformly distributed inside the filter has been made, and this also may contribute to overestimate the dose rate. When more informations are available on the actual amount of activation products in the filter (for example, frequency of filter purge), dose rate values can be reduced in consequence of these better estimates. Concerning the Helium Cooling System (HCS) and the Coolant Purification System (CPS), the activity in these components is practically negligible. 5. Conclusions The determination of the radioactive inventory and of the contact dose rates in the different TBM systems has been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, completely available for both the blanket modules. The possible sources of
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radioactive material in all the systems have been individuated and their contributes estimated. In general, for both HCLL and HCPB systems, radioactivity inventory and contact dose rates will be used for an ORE analysis. References [1] L. Petrizzi, R. Villari, L. Auditore, G. Cambi, D.G. Cepraga, “HCLL: TBM Design, Integration and Analysis. Activation and Decay Analysis of TBM in Iter Machine” Final Report: Task Tw6-Ttbc 001 Del. 2, ENEA, November 2007. [2] U. Fischer, P. Pereslavtsev, “Activation and decay heat analyses of the HCPB TBM with detailed output by radio-nuclides and sensitivity to the irradiation time in ITER (2 irradiation scenarios)”, Final report on the EFDA task TW6-TTBB-001, Deliverable 2, FZK Karlsruhe (D), March 2007. [3] L. Petrizzi, et al., Helium-cooled lithium lead: activation analysis of the test blanket module in ITER, Fusion Eng. Des. 83 (2008) 1244–1248. [4] P. Pereslavtsev, U. Fischer, Activation and afterheat analyses for the HCPB test blanket module in ITER, Fusion Eng. Des. 83 (2008) 1742–1746. [5] B. Merrill, L. Cadwallader, M. Dagher, A. Ying, “US DCLL TBM Safety Update”, TBM Safety Workshop, Aix-en-Provence, France, March 3rd, 2008. [6] See MICROSHIELD website: http://www.radiationsoftware.com/mshield.html. [7] I. Ricapito, Personal communication, email, October 26th, 2009.