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Study of the helium-cooled lithium lead test blanket module nuclear behaviour under irradiation in ITER
Fusion Engineering and Design 84 (2009) 2178–2186
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Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Study of the helium-cooled lithium lead test blanket module nuclear behaviour under irradiation in ITER P. Chiovaro, P.A. Di Maio ∗ , G. Vella Dipartimento di Ingegneria Nucleare, Università di Palermo, Viale delle Scienze, Edificio 6, 90128 Palermo, Italy
a r t i c l e
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Article history: Received 24 April 2008 Received in revised form 13 January 2009 Accepted 1 April 2009 Available online 2 May 2009 Keywords: HCLL test blanket module Neutronics Monte Carlo method
a b s t r a c t The present paper deals with the detailed investigation of the helium-cooled lithium lead test blanket module (HCLL-TBM) nuclear behaviour under irradiation in ITER, carried out at the Department of Nuclear Engineering of the University of Palermo adopting a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of the HCLL-TBM was set-up and inserted into an ITER 3D semiheterogeneous model that realistically simulates the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source was adopted for the calculations. The main features of the HCLL-TBM nuclear response were assessed, paying a particular attention to the neutronic and photonic deposited power, the tritium production rate and the spatial distribution of their volumetric densities. Structural material irradiation damage was also investigated through the evaluation of displacement per atom and helium and hydrogen production rates. © 2009 Elsevier B.V. All rights reserved.
1. Introduction The helium-cooled lithium lead (HCLL) blanket is one of the two EU breeding blanket lines selected for the DEMOnstration fusion reactor. It relies on the use of Pb–Li liquid eutectic alloy, both as tritium breeder and neutron multiplier, and on reduced-activation martensitic steel as structural material. Helium at a pressure of 8 MPa and at a temperature ranging from 300 to 500 ◦ C is envisaged as the coolant [1]. Within the framework of the research and development activities focussed on this blanket line, a program was launched aimed to the design and construction of several test blanket modules (TBMs), which were intended to be installed and irradiated within an ITER equatorial port, being suitably allocated in a supporting frame actively cooled by pressurized sub-cooled water [2,3]. The functions of the supporting frame are to minimize the TBM interactions with the surrounding shielding blanket and provide a common interface for all the TBMs. This program was regarded, since its first proposal, as one of the points of greatest strength of the whole ITER project. In fact, it would allow the investigation of the main nuclear, thermo-
Abbreviations: BC, bottom cap; BP, back plate; BU, breeder unit; BZ, breeder zone; CP, cooling plate; FW, first wall; SB, segment box; SPh, horizontal stiffening plate; SPv, vertical stiffening plate; SW, side wall; TC, top cap. ∗ Corresponding author. Tel.: +39 031 232227; fax: +39 091 232215. E-mail address: [email protected] (P.A. Di Maio). 0920-3796/$ – see front matter © 2009 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2009.04.001
mechanical and thermofluid-dynamic issues of the liquid metal breeding blanket under conditions suggestive of its actual operating environment inside the DEMO fusion reactor, in terms of coolant thermodynamic conditions and thermal and mechanical stress states. In particular, the HCLL-TBMs will be mainly devoted to the investigation of helium cooling efficiency, tritium transport, magneto-hydrodynamic and corrosion processes, effectiveness of manufacturing technologies and resistance to electro-magnetic and/or thermo-mechanical stresses [2,3]. At the Department of Nuclear Engineering (DIN) of the University of Palermo a research campaign was started with the specific aim of investigating the nuclear response of the integral HCLL-TBM (named HCLL-TBM in the following) when irradiated inside ITER, to provide useful data for the assessment of both its thermal-hydraulic and thermo-mechanical performances. A computational approach was followed based on the Monte Carlo method. The Monte Carlo N-Particle (MCNP) code version 4C was adopted using the FENDL-2 transport cross-section libraries [4,5]. In particular, a first detailed study was performed with reference to the TBM design described in ref. [6] and the results obtained as to power deposition by neutrons and photons, tritium breeding performances and structural material irradiation damage were reported in ref. [7]. A subsequent study was carried out to investigate the potential influence on the HCLL-TBM nuclear response of the supporting frame lay-out and composition. The results concerning the frame lay-out effects on power deposition, tritium breeding performances and irradiation damage, reported in refs. [8,9], show that a 20 cm thick frame allows a reasonable minimization of the
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TBM interaction with its surroundings in ITER. Moreover, the analysis of the frame composition effects on the overall nuclear response of the module shows that the typical assumption of cooling water uniformly distributed inside the frame results in unrealistic predictions which underestimate both deposited power and neutron irradiation damage in the module plasma-facing layers, if compared with those calculated assuming a more realistic non-uniform distribution [10,11]. The overall dimensions of the steel supporting frame were recently revised, causing a reduction of the space available for the TBM to 1.7 m (poloidal) × 0.524 m (toroidal) × 0.811 m (radial). Therefore, the HCLL-TBM design was deeply modified, [2,12], and a further study of its complete nuclear response under irradiation in ITER was carried out at DIN. The results obtained are herewith presented and critically discussed. 2. Outline of the HCLL-TBM The HCLL-TBMs aim at representing a typical module of the HCLL blanket. They should be cased with a poloidal lay-out in an ITER equatorial port and should be housed in a water-cooled steel frame directly supported by the vacuum vessel. In particular, the integral TBM should integrate all relevant DEMO technologies and features and should be tested during the ITER high duty D–T plasma phase. The HCLL-TBM has a poloidal height of 1.655 m, a toroidal width of 0.484 m and a radial depth of 0.575 m, and is characterized by a box-shaped structure composed of a segment box and a breeder zone [2,12]. The segment box (SB) is a directly cooled steel box having basically the function of Pb–Li container. It is mainly composed of a first wall (FW), two side walls (SWs), a top cap (TC), a bottom cap (BC) and a set of five back plates (BPs). According to ITER requirements, the FW has to be covered by a 2 mm thick beryllium layer, before the module can be irradiated [3]. The SB is reinforced by 11 mm thick toroidal–radial (horizontal) and poloidal–radial (vertical) stiffening plates (SPh, SPv) which subdivide the BZ into 16 radial cells. They are arranged with pitches of 200.7 and 196 mm in poloidal and toroidal directions, respectively. The reference structural material is reduced-activation 9% Cr martensitic steel called EUROFER. A set of 24 triple U-turn cooling channels (12.5 mm × 11 mm) is grooved within the FW and the SWs with a poloidal pitch of 22.3 mm. Moreover, three U-turn cooling channels (6 mm × 10 mm) are foreseen inside each SP. The breeder zone (BZ) is occupied by the Pb–Li liquid eutectic alloy enriched to 90% in Li6 . It has a modular structure articulated in 16 breeder units (BUs) housed in the aforementioned SB cells. Every BU has a poloidal height of 189.7 mm, a toroidal width of 206.5 mm and a radial depth of 360 mm. A set of three toroidal–radial cooling plates (CPs) subdivides each BU into four channels where Pb–Li alloy flows. Each CP is cooled by three doubleU channels (4 mm × 4 mm). Further details on the TBM lay-out can be found in ref. [12]. 3. Nuclear analyses A detailed study of the HCLL-TBM complete nuclear response under irradiation in an ITER equatorial port was performed focussing the attention on power deposited by neutrons and photons, tritium breeding performances and structural material damage induced by neutron irradiation. The study was carried out following a numerical approach based on the Monte Carlo method and adopting the Monte Carlo N-Particle code version 4C. MCPLIB04 photon transport cross-sections and recently assessed FENDL-2.1 neutron transport
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cross-section libraries [13] were implemented for the calculations. In order to speed up calculations, analyses were carried out on a cluster of eight workstations with a homogeneous operating system (Linux Fedora Core 8) by implementing the Parallel Virtual Machine software. 3.1. Model A 3D semi-heterogeneous model of ITER [14] was adopted which realistically simulates a blanket sector of the whole reactor (Fig. 1). In particular, it represents 1/18 (20◦ ) of the reactor’s toroidal extension, describing in detail the shielding blanket, the divertor cassette, the magnet system, the vacuum vessel with its three major ports, and the cryostat. Two symmetry conditions were imposed at the model toroidal boundaries to simulate reactor’s continuity in that direction. The neutron source developed by the Physics Unit of Naka ITER Joint Work Site [14] was taken into account to simulate the D–T plasma of ITER. This source is characterized by a Gaussianshaped energy spectrum and by a discrete distribution in the space domain. In particular, the plasma region is subdivided in 40 × 40 poloidal–radial elements, each one with its own neutron emission intensity. A detailed 3D heterogeneous model of the HCLL-TBM was setup and was inserted, in a vertical lay-out, into the steel supporting frame semi-heterogeneous model, developed in ref. [11] and properly located inside an equatorial port of the ITER outboard blanket. The adoption of the symmetry boundary condition at the vertical centre-plane of the frame (Fig. 1) implies that two identical TBMs would be located in the two halves of the steel supporting frame. Nevertheless, the influence of the neighbouring HCLL-TBM on the nuclear response of the considered TBM can be assumed to be negligible since, according to refs. [8,9], a 20 cm thick dividing plate was considered for the steel supporting frame. The HCLL-TBM was modelled in an almost fully heterogeneous way (Figs. 2 and 3) with some simplifications as far as the SPs and the CPs are concerned. These were modelled as proper homogeneous mixtures of EUROFER and helium, according to their mass fractions within the CPs and the SPs. Pb–15.7Li, with lithium enriched up to 90% in Li6 , was assumed as the tritium breeder. 3.2. Results The HCLL-TBM nuclear response was investigated paying particular attention to neutronic and photonic deposited power, tritium production rate and spatial distribution of their volumetric densities. Structural material radiation damage was also investigated through the evaluation of displacement per atom and helium and hydrogen production rates. The analyses were carried out by simulating a large number of histories (∼109 ) so that the results obtained are affected by relative errors lower than 1% even in the module regions more distant from plasma. Since negligible variations may be expected along the toroidal direction, only the radial and poloidal distributions of the aforementioned variables were assessed. To this purpose the model was subdivided into 18 radial volumes and each of them was further subdivided into three poloidal sub-volumes (Fig. 3). The upper and lower ones comprise two BUs, while the central one the remaining four. The calculated nuclear responses were normalised to the ITER nominal fusion power of 500 MW. A preliminary analysis was performed to assess the neutron wall loading along the HCLL-TBM plasma-facing surface, which was estimated to be 0.709 MW m−2 in the upper zone of the
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Fig. 1. The ITER model.
FW plasma-facing surface, 0.757 MW m−2 in the central zone and 0.748 MW m−2 in the lower zone, for an average value of 0.742 MW m−2 which agrees quite well with the value calculated in ref. [7]. This poloidal distribution indicates that, according to the particular shape of the neutron source adopted and to its relative position with respect to the module, the lower and central regions of the TBM undergo higher neutron fluence intensities than the upper one.
Fig. 2. Toroidal–radial section of the HCLL-TBM model.
3.2.1. Neutronic and photonic power deposition The power deposited in the module by neutrons and photons was evaluated in order to provide useful data for the investigation of the HCLL-TBM thermal-hydraulic performances. It was estimated that a total power of 429.58 kW is released within the module. A detailed description of its distribution is reported in Tables 1 and 2 where the component denominations are those reported in Figs. 2 and 3. The spatial distribution of the deposited nuclear power volumetric density, q (r), was evaluated to allow the study of the HCLL-TBM thermo-mechanical performances to be carried out. In particular, its radial profiles were assessed within each of the three module poloidal zones investigated as far as the SB and the BZ are concerned. They are shown in Figs. 4–6. It should be highlighted that the q (r) radial profiles of the two SWs are quite similar and, for the sake of clarity, only one of them is reported in Fig. 4.
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Fig. 3. Poloidal–radial section of the HCLL-TBM model.
Table 1 Neutronic and photonic deposited power (kW).
SB
BZ
Be layer FW SW 1 SW 2 SPh SPv TC BC BP Rear slabs Stiffening rods SB total
6.923 95.656 24.565 23.060 3.215 6.379 5.472 5.361 4.718 5.721 1.482 182.551
Pb–Li CPs Pb–Li headers SS manifold BZ total
232.279 10.377 4.136 0.242 247.033
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As expected, the deposited power densities reach their highest values near the module plasma-facing region, both in the SB and the BZ, decreasing significantly along the radial direction. With regard to the SB (Fig. 4), according to the poloidal distribution of the neutron wall loading calculated, the q (r) radial distribution determined along the central poloidal zone may significantly differ from the corresponding distributions evaluated along both the upper and lower zones. In particular, as to the FW, the values predicted in the central zone are slightly higher (∼5%) than those of the upper and lower zones, while, concerning the SWs, a significantly larger difference (∼17%) may be observed. This difference seems to be due to a “boundary effect” induced by the interaction between SWs and neutrons arising from plasma and directed towards the gap between the SWs and the supporting frame. In fact, some of these neutrons reach directly the SW, where they release their energy, some other, interacting with the internal wall of the supporting frame, tend to be either reflected towards the SW or totally or “partially” absorbed by radiative capture (n, ) or inelastic scattering, respectively. In the former case they release their energy within the SW while, in the latter one, energy is released by the secondary photons produced. This “boundary effect” is more pronounced along the central region of the SWs where the source view factor is higher, therefore more nuclear power is deposited within this region. It has to be underlined that this boundary effect is not predictable along the FW where, due to the absence of toroidal segmentation within the model adopted, it vanishes as a consequence of toroidal averaging. Moreover, slight differences (∼3%) may be observed between the q (r) radial profiles of the two Caps (Fig. 5), according again to the particular poloidal distribution of the neutron wall loading. Concerning the BZ (Fig. 6), the q (r) radial distribution relevant to the central poloidal zone slightly differs from the corresponding distributions evaluated along both the upper and lower zones. In particular, the differences increase along the radial direction from about 3.5%, near the FW, up to 17% at the BZ rear region. From the analysis of Figs. 4–6 it can be deduced that the highest q (r) value calculated within the FW is 5.86 W cm−3 while, within the BZ, it is 6.17 W cm−3 . The radial distributions obtained as far as the SWs, the Pb–Li and the Caps are concerned were fitted by the least-square method imposing the following functional form: q (r) = ˛ exp(−ˇr) + exp(−ır)
(1)
where q (r) is expressed in W cm−3 and r represents the radial distance in centimetres from the plasma-facing surface of the beryllium layer covering the FW. A linear best fitting function, given by: q (r) = ωr +
(2)
Poloidal
1
2
was considered for the q (r) radial distributions within the FW. The minimum correlation factor obtained is higher than 0.998 and the best fitting functions determined, once properly integrated, preserve the overall deposited power. The functions relevant to the SWs, the Pb–Li and the Caps were reported in Figs. 4–6 in comparison with code predictions, while their coefficients are reported in Table 3. As to the q (r) radial distributions within the FW, their best fitting coefficients are reported in Table 4.
1 2 3 4 5 6 7 8
13.617 14.004 15.013 15.572 15.343 14.960 14.289 14.119
13.430 13.813 14.862 15.231 15.167 14.783 14.064 14.011
3.2.2. Tritium production Since one of the HCLL-TBM main goals is to test the tritium breeding, recovery and confinement capability of this blanket line, daily tritium production together with the radial distribution of the volumetric density of tritium production rate were investigated.
Total
429.584
Table 2 Nuclear deposited power within BU Pb–Li (kW). BU
Toroidal
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Fig. 4. Radial profiles of the q (r) function within the SB main components.
Fig. 5. Radial profiles of the q (r) function within the TC and BC.
Fig. 6. Radial profiles of the q (r) function within the BZ Pb–Li alloy.
P. Chiovaro et al. / Fusion Engineering and Design 84 (2009) 2178–2186 Table 3 Best fitting coefficients. ˛
Zones
ˇ
ı
SW 1
Upper Central Lower
q q q
2.542809 3.190300 2.469505
0.138064 0.120969 0.152151
3.158871 3.345689 3.517565
0.056961 0.053539 0.058601
SW 2
Upper Central Lower
q q q
2.622886 4.860578 3.229537
0.159705 0.101748 0.141843
3.244044 1.614184 2.750540
0.060659 0.042905 0.055823
Pb–Li
Upper
q T q T q T
110.029030 84.151954 109.091720 72.551199 111.580420 90.019943
0.985254 1.143682 0.982700 1.107535 0.987513 1.160833
4.213350 1.162590 4.414611 1.179487 4.356447 1.179777
0.085154 0.056485 0.080118 0.049838 0.085196 0.056105
q q
2.891763 3.371910
0.242052 0.185152
3.664381 3.100789
0.069370 0.062651
Central Lower Caps
Top Bottom
Table 4 Best fitting coefficients. Zones FW
Upper Central Lower
q q q
ω
−0.565294 −0.547137 −0.576974
5.906253 6.114500 6.046965
The daily tritium production obviously depends on the lithium enrichment in Li6 of Pb–Li eutectic alloy and on the ITER duty cycle. Assuming the nominal Li6 enrichment (90%) and a duty cycle of 0.22 (400 s burn length within a pulse of 1800 s), the daily tritium production was calculated to be 9.084 mg d−1 , including 0.239 mg d−1 produced within the Pb–Li headers. The contribution of each single BU to this overall value was also investigated and the results obtained are summarized in Table 5, where the BUs reported in the first toroidal column are those located near the supporting frame dividing plate, while the BUs belonging to the first poloidal row are those nearest to the TC (Figs. 2 and 3). As it can be expected, the BUs located in the module central region contribute to the overall daily tritium production much more (∼15% more) than the other ones, while the least contribution is given by those located in the TBM upper zone. The effect of the lithium enrichment in Li6 on the TBM tritium breeding performances was also investigated by means of Monte Carlo parametric analyses, specifically aimed to the assessment of the TBM daily tritium production dependence on the Li6 molar fraction, x. The numerical results obtained are reported in Fig. 7 together with their best fitting function given by: T (x) = 0.9588(x)0.5022
(3)
where T is expressed in mg d−1 and x in %.
Table 5 Daily tritium production within BUs (mg d−1 ). BU
Toroidal
Poloidal
1
2
1 2 3 4 5 6 7 8
0.519 0.537 0.574 0.592 0.587 0.573 0.541 0.532
0.511 0.529 0.567 0.582 0.579 0.565 0.532 0.525
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From the analysis of Fig. 7 it can be deduced that the daily tritium production increases more than linearly up to a lithium enrichment in Li6 of about 60%, while it behaves as a quasi-linear function when the Li6 molar fraction ranges from 60% up to 100%, when the production reaches the maximum value of 9.57 mg d−1 . In particular, it can be observed that, decreasing lithium enrichment in Li6 from 90% to 60% results in a 17.2% decrease in daily tritium production. In order to perform tritium transport analyses, the radial distributions of the volumetric density of the tritium production rate, T (r), were determined in the upper, central and lower zones of the BZ. The distributions obtained are shown in Fig. 8 together with their best fitting functions, which have the same analytical form given in Eq. (1) and the coefficients of which are shown in Table 3. The decrease of the volumetric density of the tritium production rate along the BZ radial depth (Fig. 8) is mainly due to the similar decrease of the flux of neutrons with energy lower than 10 keV, which contribute the most to the tritium production. Moreover, it has to be pointed out that the radial distribution of the T (r) function in the module central zone is higher than in the lower and upper zones, except for the first 2 cm of the BZ where they are almost the same.
3.2.3. Radiation damage During plant operation, the highly energetic fusion neutrons coming out from plasma continuously interact with the structural material atoms inducing two main damage mechanisms for the blanket materials. The first mechanism is due to the displacement of atoms from their lattice positions as a consequence of collisions, while the second is determined by the gas production as a result of various nuclear reactions mainly of (n, p), (n, n p), (n, ␣) and (n, n ␣) kind. While hydrogen isotopes diffuse out of the metallic lattice or form metal hydrides, ␣-particles remain trapped in the metal and generate helium gas bubbles. These processes lead to unfavourable changes of mechanical properties (such as embrittlement), limit the lifetime of the structural material and affect its reweldability [7]. Therefore, their prediction is pivotal to the assessment of the overall module’s lifetime. In order to investigate the level of the HCLL-TBM material damage due to the first mechanism, the displacement per atom (DPA) within the SB was evaluated focussing the attention on the Fe56 isotope and adopting the appropriate displacement crosssection taken from FENDL-2.1 [13]. The DPA distribution along the radial depth of the HCLL-TBM structural material (Fig. 9) was evaluated by assuming a reactor duty cycle of 0.22 and supposing a full power (500 MW) pulsed plant operation during a whole year. The results obtained indicate that, as expected, the maximum value of 0.8 DPA year−1 is reached within the FW, where the neutron flux energy distribution is particularly hard due to the presence of 14 MeV neutrons, arising directly from the D–T source, and fast neutrons back-scattered by the first layer of the Pb–Li alloy. In order to estimate the effect of the second damage mechanism, helium and hydrogen production rates were also evaluated along the radial direction within the SB structural material. The profiles obtained are reported in Figs. 10 and 11, where it is possible to observe that maxima are reached of about 12.94 and 53.27 appm year−1 for helium and hydrogen, respectively. The decrease of both helium and hydrogen production rates along the SB radial depth (Figs. 10 and 11) is mainly due to the similar decrease of the fast neutron flux, which mainly contributes to the aforementioned gas production.
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Fig. 7. Daily tritium production vs. lithium enrichment in Li6 .
Fig. 8. Radial profiles of the T (r) function within the Pb–Li alloy of the BZ.
Fig. 9. DPA radial distribution within the SB and the Caps.
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Fig. 10. He production rate radial distribution within the SB and the Caps.
Fig. 11. H production rate radial distribution within the SB and the Caps.
4. Conclusions A detailed investigation of the HCLL-TBM nuclear response in ITER was performed by means of a 3D-Monte Carlo neutronic and photonic analysis. A 3D heterogeneous model of the HCLLTBM was set-up assuming EUROFER as the structural material. It was inserted into a ITER semi-heterogeneous model with a proper D–T plasma neutron source. The main features of the HCLL-TBM nuclear response were assessed, focussing the attention on nuclear power deposition, tritium production and radiation-induced material damage. The nuclear power deposited in HCLL-TBM was estimated to be about 430 kW and its detailed spatial distribution was evaluated. A particular attention was paid to the radial distributions of the volumetric density of the nuclear deposited power in the module main components, obtaining their relevant best fitting functions which could be useful for the study of the module thermo-mechanical performances. The daily production of tritium and the radial distribution of the volumetric density of its production rate were evaluated and it was observed that the former reaches a value slightly higher than 9 mg d−1 . As far as irradiation damage is concerned, both the DPA and helium and hydrogen production rate distributions were calcu-
lated along the radial depth of the TBM structural material and it was found that their maxima are achieved in the FW proximity, where the neutron fluence intensity is higher. These maxima, in the hypothesis of one year of full power pulsed operation, were estimated to be 0.8 DPA year−1 and about 13 and 53 appm year−1 for helium and hydrogen, respectively.
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[7] P. Chiovaro, P.A. Di Maio, G. Vella, On the nuclear response of the helium-cooled lithium lead test blanket module in ITER, Fusion Engineering and Design 75–79 (2005) 725–730. [8] P. Chiovaro, P.A. Di Maio, E. Oliveri, G. Vella, On the influence of the supporting frame on the nuclear response of the helium-cooled lithium lead test blanket module for ITER, Fusion Engineering and Design 81 (2006) 677–686. [9] P. Chiovaro, P.A. Di Maio, E. Oliveri, G. Vella, On the effects of the supporting frame on the radiation-induced damage of HCLL-TBM structural material, Journal of Nuclear Materials 367–370 (2007) 1344–1349. [10] P. Chiovaro, P.A. Di Maio, E. Oliveri, G. Vella, A study of the potential influence of frame coolant on HCLL-TBM nuclear response, Fusion Engineering and Design 82 (2007) 2359–2365.
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Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Study of the helium-cooled lithium lead test blanket module nuclear behaviour under irradiation in ITER P. Chiovaro, P.A. Di Maio ∗ , G. Vella Dipartimento di Ingegneria Nucleare, Università di Palermo, Viale delle Scienze, Edificio 6, 90128 Palermo, Italy
a r t i c l e
i n f o
Article history: Received 24 April 2008 Received in revised form 13 January 2009 Accepted 1 April 2009 Available online 2 May 2009 Keywords: HCLL test blanket module Neutronics Monte Carlo method
a b s t r a c t The present paper deals with the detailed investigation of the helium-cooled lithium lead test blanket module (HCLL-TBM) nuclear behaviour under irradiation in ITER, carried out at the Department of Nuclear Engineering of the University of Palermo adopting a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of the HCLL-TBM was set-up and inserted into an ITER 3D semiheterogeneous model that realistically simulates the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source was adopted for the calculations. The main features of the HCLL-TBM nuclear response were assessed, paying a particular attention to the neutronic and photonic deposited power, the tritium production rate and the spatial distribution of their volumetric densities. Structural material irradiation damage was also investigated through the evaluation of displacement per atom and helium and hydrogen production rates. © 2009 Elsevier B.V. All rights reserved.
1. Introduction The helium-cooled lithium lead (HCLL) blanket is one of the two EU breeding blanket lines selected for the DEMOnstration fusion reactor. It relies on the use of Pb–Li liquid eutectic alloy, both as tritium breeder and neutron multiplier, and on reduced-activation martensitic steel as structural material. Helium at a pressure of 8 MPa and at a temperature ranging from 300 to 500 ◦ C is envisaged as the coolant [1]. Within the framework of the research and development activities focussed on this blanket line, a program was launched aimed to the design and construction of several test blanket modules (TBMs), which were intended to be installed and irradiated within an ITER equatorial port, being suitably allocated in a supporting frame actively cooled by pressurized sub-cooled water [2,3]. The functions of the supporting frame are to minimize the TBM interactions with the surrounding shielding blanket and provide a common interface for all the TBMs. This program was regarded, since its first proposal, as one of the points of greatest strength of the whole ITER project. In fact, it would allow the investigation of the main nuclear, thermo-
Abbreviations: BC, bottom cap; BP, back plate; BU, breeder unit; BZ, breeder zone; CP, cooling plate; FW, first wall; SB, segment box; SPh, horizontal stiffening plate; SPv, vertical stiffening plate; SW, side wall; TC, top cap. ∗ Corresponding author. Tel.: +39 031 232227; fax: +39 091 232215. E-mail address: [email protected] (P.A. Di Maio). 0920-3796/$ – see front matter © 2009 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2009.04.001
mechanical and thermofluid-dynamic issues of the liquid metal breeding blanket under conditions suggestive of its actual operating environment inside the DEMO fusion reactor, in terms of coolant thermodynamic conditions and thermal and mechanical stress states. In particular, the HCLL-TBMs will be mainly devoted to the investigation of helium cooling efficiency, tritium transport, magneto-hydrodynamic and corrosion processes, effectiveness of manufacturing technologies and resistance to electro-magnetic and/or thermo-mechanical stresses [2,3]. At the Department of Nuclear Engineering (DIN) of the University of Palermo a research campaign was started with the specific aim of investigating the nuclear response of the integral HCLL-TBM (named HCLL-TBM in the following) when irradiated inside ITER, to provide useful data for the assessment of both its thermal-hydraulic and thermo-mechanical performances. A computational approach was followed based on the Monte Carlo method. The Monte Carlo N-Particle (MCNP) code version 4C was adopted using the FENDL-2 transport cross-section libraries [4,5]. In particular, a first detailed study was performed with reference to the TBM design described in ref. [6] and the results obtained as to power deposition by neutrons and photons, tritium breeding performances and structural material irradiation damage were reported in ref. [7]. A subsequent study was carried out to investigate the potential influence on the HCLL-TBM nuclear response of the supporting frame lay-out and composition. The results concerning the frame lay-out effects on power deposition, tritium breeding performances and irradiation damage, reported in refs. [8,9], show that a 20 cm thick frame allows a reasonable minimization of the
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TBM interaction with its surroundings in ITER. Moreover, the analysis of the frame composition effects on the overall nuclear response of the module shows that the typical assumption of cooling water uniformly distributed inside the frame results in unrealistic predictions which underestimate both deposited power and neutron irradiation damage in the module plasma-facing layers, if compared with those calculated assuming a more realistic non-uniform distribution [10,11]. The overall dimensions of the steel supporting frame were recently revised, causing a reduction of the space available for the TBM to 1.7 m (poloidal) × 0.524 m (toroidal) × 0.811 m (radial). Therefore, the HCLL-TBM design was deeply modified, [2,12], and a further study of its complete nuclear response under irradiation in ITER was carried out at DIN. The results obtained are herewith presented and critically discussed. 2. Outline of the HCLL-TBM The HCLL-TBMs aim at representing a typical module of the HCLL blanket. They should be cased with a poloidal lay-out in an ITER equatorial port and should be housed in a water-cooled steel frame directly supported by the vacuum vessel. In particular, the integral TBM should integrate all relevant DEMO technologies and features and should be tested during the ITER high duty D–T plasma phase. The HCLL-TBM has a poloidal height of 1.655 m, a toroidal width of 0.484 m and a radial depth of 0.575 m, and is characterized by a box-shaped structure composed of a segment box and a breeder zone [2,12]. The segment box (SB) is a directly cooled steel box having basically the function of Pb–Li container. It is mainly composed of a first wall (FW), two side walls (SWs), a top cap (TC), a bottom cap (BC) and a set of five back plates (BPs). According to ITER requirements, the FW has to be covered by a 2 mm thick beryllium layer, before the module can be irradiated [3]. The SB is reinforced by 11 mm thick toroidal–radial (horizontal) and poloidal–radial (vertical) stiffening plates (SPh, SPv) which subdivide the BZ into 16 radial cells. They are arranged with pitches of 200.7 and 196 mm in poloidal and toroidal directions, respectively. The reference structural material is reduced-activation 9% Cr martensitic steel called EUROFER. A set of 24 triple U-turn cooling channels (12.5 mm × 11 mm) is grooved within the FW and the SWs with a poloidal pitch of 22.3 mm. Moreover, three U-turn cooling channels (6 mm × 10 mm) are foreseen inside each SP. The breeder zone (BZ) is occupied by the Pb–Li liquid eutectic alloy enriched to 90% in Li6 . It has a modular structure articulated in 16 breeder units (BUs) housed in the aforementioned SB cells. Every BU has a poloidal height of 189.7 mm, a toroidal width of 206.5 mm and a radial depth of 360 mm. A set of three toroidal–radial cooling plates (CPs) subdivides each BU into four channels where Pb–Li alloy flows. Each CP is cooled by three doubleU channels (4 mm × 4 mm). Further details on the TBM lay-out can be found in ref. [12]. 3. Nuclear analyses A detailed study of the HCLL-TBM complete nuclear response under irradiation in an ITER equatorial port was performed focussing the attention on power deposited by neutrons and photons, tritium breeding performances and structural material damage induced by neutron irradiation. The study was carried out following a numerical approach based on the Monte Carlo method and adopting the Monte Carlo N-Particle code version 4C. MCPLIB04 photon transport cross-sections and recently assessed FENDL-2.1 neutron transport
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cross-section libraries [13] were implemented for the calculations. In order to speed up calculations, analyses were carried out on a cluster of eight workstations with a homogeneous operating system (Linux Fedora Core 8) by implementing the Parallel Virtual Machine software. 3.1. Model A 3D semi-heterogeneous model of ITER [14] was adopted which realistically simulates a blanket sector of the whole reactor (Fig. 1). In particular, it represents 1/18 (20◦ ) of the reactor’s toroidal extension, describing in detail the shielding blanket, the divertor cassette, the magnet system, the vacuum vessel with its three major ports, and the cryostat. Two symmetry conditions were imposed at the model toroidal boundaries to simulate reactor’s continuity in that direction. The neutron source developed by the Physics Unit of Naka ITER Joint Work Site [14] was taken into account to simulate the D–T plasma of ITER. This source is characterized by a Gaussianshaped energy spectrum and by a discrete distribution in the space domain. In particular, the plasma region is subdivided in 40 × 40 poloidal–radial elements, each one with its own neutron emission intensity. A detailed 3D heterogeneous model of the HCLL-TBM was setup and was inserted, in a vertical lay-out, into the steel supporting frame semi-heterogeneous model, developed in ref. [11] and properly located inside an equatorial port of the ITER outboard blanket. The adoption of the symmetry boundary condition at the vertical centre-plane of the frame (Fig. 1) implies that two identical TBMs would be located in the two halves of the steel supporting frame. Nevertheless, the influence of the neighbouring HCLL-TBM on the nuclear response of the considered TBM can be assumed to be negligible since, according to refs. [8,9], a 20 cm thick dividing plate was considered for the steel supporting frame. The HCLL-TBM was modelled in an almost fully heterogeneous way (Figs. 2 and 3) with some simplifications as far as the SPs and the CPs are concerned. These were modelled as proper homogeneous mixtures of EUROFER and helium, according to their mass fractions within the CPs and the SPs. Pb–15.7Li, with lithium enriched up to 90% in Li6 , was assumed as the tritium breeder. 3.2. Results The HCLL-TBM nuclear response was investigated paying particular attention to neutronic and photonic deposited power, tritium production rate and spatial distribution of their volumetric densities. Structural material radiation damage was also investigated through the evaluation of displacement per atom and helium and hydrogen production rates. The analyses were carried out by simulating a large number of histories (∼109 ) so that the results obtained are affected by relative errors lower than 1% even in the module regions more distant from plasma. Since negligible variations may be expected along the toroidal direction, only the radial and poloidal distributions of the aforementioned variables were assessed. To this purpose the model was subdivided into 18 radial volumes and each of them was further subdivided into three poloidal sub-volumes (Fig. 3). The upper and lower ones comprise two BUs, while the central one the remaining four. The calculated nuclear responses were normalised to the ITER nominal fusion power of 500 MW. A preliminary analysis was performed to assess the neutron wall loading along the HCLL-TBM plasma-facing surface, which was estimated to be 0.709 MW m−2 in the upper zone of the
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Fig. 1. The ITER model.
FW plasma-facing surface, 0.757 MW m−2 in the central zone and 0.748 MW m−2 in the lower zone, for an average value of 0.742 MW m−2 which agrees quite well with the value calculated in ref. [7]. This poloidal distribution indicates that, according to the particular shape of the neutron source adopted and to its relative position with respect to the module, the lower and central regions of the TBM undergo higher neutron fluence intensities than the upper one.
Fig. 2. Toroidal–radial section of the HCLL-TBM model.
3.2.1. Neutronic and photonic power deposition The power deposited in the module by neutrons and photons was evaluated in order to provide useful data for the investigation of the HCLL-TBM thermal-hydraulic performances. It was estimated that a total power of 429.58 kW is released within the module. A detailed description of its distribution is reported in Tables 1 and 2 where the component denominations are those reported in Figs. 2 and 3. The spatial distribution of the deposited nuclear power volumetric density, q (r), was evaluated to allow the study of the HCLL-TBM thermo-mechanical performances to be carried out. In particular, its radial profiles were assessed within each of the three module poloidal zones investigated as far as the SB and the BZ are concerned. They are shown in Figs. 4–6. It should be highlighted that the q (r) radial profiles of the two SWs are quite similar and, for the sake of clarity, only one of them is reported in Fig. 4.
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Fig. 3. Poloidal–radial section of the HCLL-TBM model.
Table 1 Neutronic and photonic deposited power (kW).
SB
BZ
Be layer FW SW 1 SW 2 SPh SPv TC BC BP Rear slabs Stiffening rods SB total
6.923 95.656 24.565 23.060 3.215 6.379 5.472 5.361 4.718 5.721 1.482 182.551
Pb–Li CPs Pb–Li headers SS manifold BZ total
232.279 10.377 4.136 0.242 247.033
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As expected, the deposited power densities reach their highest values near the module plasma-facing region, both in the SB and the BZ, decreasing significantly along the radial direction. With regard to the SB (Fig. 4), according to the poloidal distribution of the neutron wall loading calculated, the q (r) radial distribution determined along the central poloidal zone may significantly differ from the corresponding distributions evaluated along both the upper and lower zones. In particular, as to the FW, the values predicted in the central zone are slightly higher (∼5%) than those of the upper and lower zones, while, concerning the SWs, a significantly larger difference (∼17%) may be observed. This difference seems to be due to a “boundary effect” induced by the interaction between SWs and neutrons arising from plasma and directed towards the gap between the SWs and the supporting frame. In fact, some of these neutrons reach directly the SW, where they release their energy, some other, interacting with the internal wall of the supporting frame, tend to be either reflected towards the SW or totally or “partially” absorbed by radiative capture (n, ) or inelastic scattering, respectively. In the former case they release their energy within the SW while, in the latter one, energy is released by the secondary photons produced. This “boundary effect” is more pronounced along the central region of the SWs where the source view factor is higher, therefore more nuclear power is deposited within this region. It has to be underlined that this boundary effect is not predictable along the FW where, due to the absence of toroidal segmentation within the model adopted, it vanishes as a consequence of toroidal averaging. Moreover, slight differences (∼3%) may be observed between the q (r) radial profiles of the two Caps (Fig. 5), according again to the particular poloidal distribution of the neutron wall loading. Concerning the BZ (Fig. 6), the q (r) radial distribution relevant to the central poloidal zone slightly differs from the corresponding distributions evaluated along both the upper and lower zones. In particular, the differences increase along the radial direction from about 3.5%, near the FW, up to 17% at the BZ rear region. From the analysis of Figs. 4–6 it can be deduced that the highest q (r) value calculated within the FW is 5.86 W cm−3 while, within the BZ, it is 6.17 W cm−3 . The radial distributions obtained as far as the SWs, the Pb–Li and the Caps are concerned were fitted by the least-square method imposing the following functional form: q (r) = ˛ exp(−ˇr) + exp(−ır)
(1)
where q (r) is expressed in W cm−3 and r represents the radial distance in centimetres from the plasma-facing surface of the beryllium layer covering the FW. A linear best fitting function, given by: q (r) = ωr +
(2)
Poloidal
1
2
was considered for the q (r) radial distributions within the FW. The minimum correlation factor obtained is higher than 0.998 and the best fitting functions determined, once properly integrated, preserve the overall deposited power. The functions relevant to the SWs, the Pb–Li and the Caps were reported in Figs. 4–6 in comparison with code predictions, while their coefficients are reported in Table 3. As to the q (r) radial distributions within the FW, their best fitting coefficients are reported in Table 4.
1 2 3 4 5 6 7 8
13.617 14.004 15.013 15.572 15.343 14.960 14.289 14.119
13.430 13.813 14.862 15.231 15.167 14.783 14.064 14.011
3.2.2. Tritium production Since one of the HCLL-TBM main goals is to test the tritium breeding, recovery and confinement capability of this blanket line, daily tritium production together with the radial distribution of the volumetric density of tritium production rate were investigated.
Total
429.584
Table 2 Nuclear deposited power within BU Pb–Li (kW). BU
Toroidal
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Fig. 4. Radial profiles of the q (r) function within the SB main components.
Fig. 5. Radial profiles of the q (r) function within the TC and BC.
Fig. 6. Radial profiles of the q (r) function within the BZ Pb–Li alloy.
P. Chiovaro et al. / Fusion Engineering and Design 84 (2009) 2178–2186 Table 3 Best fitting coefficients. ˛
Zones
ˇ
ı
SW 1
Upper Central Lower
q q q
2.542809 3.190300 2.469505
0.138064 0.120969 0.152151
3.158871 3.345689 3.517565
0.056961 0.053539 0.058601
SW 2
Upper Central Lower
q q q
2.622886 4.860578 3.229537
0.159705 0.101748 0.141843
3.244044 1.614184 2.750540
0.060659 0.042905 0.055823
Pb–Li
Upper
q T q T q T
110.029030 84.151954 109.091720 72.551199 111.580420 90.019943
0.985254 1.143682 0.982700 1.107535 0.987513 1.160833
4.213350 1.162590 4.414611 1.179487 4.356447 1.179777
0.085154 0.056485 0.080118 0.049838 0.085196 0.056105
q q
2.891763 3.371910
0.242052 0.185152
3.664381 3.100789
0.069370 0.062651
Central Lower Caps
Top Bottom
Table 4 Best fitting coefficients. Zones FW
Upper Central Lower
q q q
ω
−0.565294 −0.547137 −0.576974
5.906253 6.114500 6.046965
The daily tritium production obviously depends on the lithium enrichment in Li6 of Pb–Li eutectic alloy and on the ITER duty cycle. Assuming the nominal Li6 enrichment (90%) and a duty cycle of 0.22 (400 s burn length within a pulse of 1800 s), the daily tritium production was calculated to be 9.084 mg d−1 , including 0.239 mg d−1 produced within the Pb–Li headers. The contribution of each single BU to this overall value was also investigated and the results obtained are summarized in Table 5, where the BUs reported in the first toroidal column are those located near the supporting frame dividing plate, while the BUs belonging to the first poloidal row are those nearest to the TC (Figs. 2 and 3). As it can be expected, the BUs located in the module central region contribute to the overall daily tritium production much more (∼15% more) than the other ones, while the least contribution is given by those located in the TBM upper zone. The effect of the lithium enrichment in Li6 on the TBM tritium breeding performances was also investigated by means of Monte Carlo parametric analyses, specifically aimed to the assessment of the TBM daily tritium production dependence on the Li6 molar fraction, x. The numerical results obtained are reported in Fig. 7 together with their best fitting function given by: T (x) = 0.9588(x)0.5022
(3)
where T is expressed in mg d−1 and x in %.
Table 5 Daily tritium production within BUs (mg d−1 ). BU
Toroidal
Poloidal
1
2
1 2 3 4 5 6 7 8
0.519 0.537 0.574 0.592 0.587 0.573 0.541 0.532
0.511 0.529 0.567 0.582 0.579 0.565 0.532 0.525
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From the analysis of Fig. 7 it can be deduced that the daily tritium production increases more than linearly up to a lithium enrichment in Li6 of about 60%, while it behaves as a quasi-linear function when the Li6 molar fraction ranges from 60% up to 100%, when the production reaches the maximum value of 9.57 mg d−1 . In particular, it can be observed that, decreasing lithium enrichment in Li6 from 90% to 60% results in a 17.2% decrease in daily tritium production. In order to perform tritium transport analyses, the radial distributions of the volumetric density of the tritium production rate, T (r), were determined in the upper, central and lower zones of the BZ. The distributions obtained are shown in Fig. 8 together with their best fitting functions, which have the same analytical form given in Eq. (1) and the coefficients of which are shown in Table 3. The decrease of the volumetric density of the tritium production rate along the BZ radial depth (Fig. 8) is mainly due to the similar decrease of the flux of neutrons with energy lower than 10 keV, which contribute the most to the tritium production. Moreover, it has to be pointed out that the radial distribution of the T (r) function in the module central zone is higher than in the lower and upper zones, except for the first 2 cm of the BZ where they are almost the same.
3.2.3. Radiation damage During plant operation, the highly energetic fusion neutrons coming out from plasma continuously interact with the structural material atoms inducing two main damage mechanisms for the blanket materials. The first mechanism is due to the displacement of atoms from their lattice positions as a consequence of collisions, while the second is determined by the gas production as a result of various nuclear reactions mainly of (n, p), (n, n p), (n, ␣) and (n, n ␣) kind. While hydrogen isotopes diffuse out of the metallic lattice or form metal hydrides, ␣-particles remain trapped in the metal and generate helium gas bubbles. These processes lead to unfavourable changes of mechanical properties (such as embrittlement), limit the lifetime of the structural material and affect its reweldability [7]. Therefore, their prediction is pivotal to the assessment of the overall module’s lifetime. In order to investigate the level of the HCLL-TBM material damage due to the first mechanism, the displacement per atom (DPA) within the SB was evaluated focussing the attention on the Fe56 isotope and adopting the appropriate displacement crosssection taken from FENDL-2.1 [13]. The DPA distribution along the radial depth of the HCLL-TBM structural material (Fig. 9) was evaluated by assuming a reactor duty cycle of 0.22 and supposing a full power (500 MW) pulsed plant operation during a whole year. The results obtained indicate that, as expected, the maximum value of 0.8 DPA year−1 is reached within the FW, where the neutron flux energy distribution is particularly hard due to the presence of 14 MeV neutrons, arising directly from the D–T source, and fast neutrons back-scattered by the first layer of the Pb–Li alloy. In order to estimate the effect of the second damage mechanism, helium and hydrogen production rates were also evaluated along the radial direction within the SB structural material. The profiles obtained are reported in Figs. 10 and 11, where it is possible to observe that maxima are reached of about 12.94 and 53.27 appm year−1 for helium and hydrogen, respectively. The decrease of both helium and hydrogen production rates along the SB radial depth (Figs. 10 and 11) is mainly due to the similar decrease of the fast neutron flux, which mainly contributes to the aforementioned gas production.
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Fig. 7. Daily tritium production vs. lithium enrichment in Li6 .
Fig. 8. Radial profiles of the T (r) function within the Pb–Li alloy of the BZ.
Fig. 9. DPA radial distribution within the SB and the Caps.
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Fig. 10. He production rate radial distribution within the SB and the Caps.
Fig. 11. H production rate radial distribution within the SB and the Caps.
4. Conclusions A detailed investigation of the HCLL-TBM nuclear response in ITER was performed by means of a 3D-Monte Carlo neutronic and photonic analysis. A 3D heterogeneous model of the HCLLTBM was set-up assuming EUROFER as the structural material. It was inserted into a ITER semi-heterogeneous model with a proper D–T plasma neutron source. The main features of the HCLL-TBM nuclear response were assessed, focussing the attention on nuclear power deposition, tritium production and radiation-induced material damage. The nuclear power deposited in HCLL-TBM was estimated to be about 430 kW and its detailed spatial distribution was evaluated. A particular attention was paid to the radial distributions of the volumetric density of the nuclear deposited power in the module main components, obtaining their relevant best fitting functions which could be useful for the study of the module thermo-mechanical performances. The daily production of tritium and the radial distribution of the volumetric density of its production rate were evaluated and it was observed that the former reaches a value slightly higher than 9 mg d−1 . As far as irradiation damage is concerned, both the DPA and helium and hydrogen production rate distributions were calcu-
lated along the radial depth of the TBM structural material and it was found that their maxima are achieved in the FW proximity, where the neutron fluence intensity is higher. These maxima, in the hypothesis of one year of full power pulsed operation, were estimated to be 0.8 DPA year−1 and about 13 and 53 appm year−1 for helium and hydrogen, respectively.
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