solid breeder blanket design and tritium breeding

Fusion Engineering and Design 16 (1991) 73-84 73 North-Holland

Solid breeder blanket design and tritium breeding E. Proust a, L. Anzidei b, M. DaUe D o n n e c, U. Fischer c and T. K u r o d a d

a C.E.A. DRN/DMT, C.E.N. Saclay, France b ENEA/Fus., C.R.E. Frascati, Italy c KfK/ INR, Kernforschungszentrum Karlsruhe, Germany d jAERI, Naka FE.R., Tokyo, Japan

Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements--to breed one tritium per fusion nentron--are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li20, LiA102, Li2ZrO3, Li4SiO 4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the mount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket.

1. I n t r o d u c t i o n

The blanket of an electricity-generating deuterium- tritium fusion reactor is the nuclear component of the machine which surrounds the plasma chamber, and whose two basic functions are: - - to convert into heat the Kinetic energy o f the neutrons

created in the plasma chamber by the fusion reactions

O + T ~4He(3.5 MeV) + n(14.1 MeV)

and to transfer this heat (80% of the fusion energy) to a coolant under pressure and temperature conditions appropriate for driving a good efficiency ther- modynamic cycle.

- - to breed tritium to replace that burned in the plasma chamber.

Worldwide (in the EC, in Japan, in the USA, in USSR), the R&D activity on this component is organized around programs aiming at developing, for testing them in the next-step tokamak machine--ITER or NET-- , test articles representative of the most prom- ising blanket concepts [1].

One of the main objectives of these tests will be to make a substantial step towards the demonstration of

the feasibility of achieving the tritium self-sufficiency in a thermo-nuclear power plant.

This paper is an attempt to cover in a few pages the question of tritium breeding, and the influence of this issue on the design of, and the material selection for, blankets relying on the use of solid breeder materials.

In addition to this introduction and to a conclusion, the paper includes four sections dealing respectively with: - - tritium breeding requirements - - candidate blanket materials and selection criteria. - - design guidelines for achieving high tritium breed-

ing ratios, and their technological limitations. - - typical design concepts of solid breeder blankets for

power reactors.

2. T r i t i u m b r e e d i n g

2.1. Tritium production methods

An electricity-generating D - T fusion reactor will typically consume, per full power year and per GWth, about 50 kg of tritium (i.e. 150 kg T /Gwe year at a net efficiency of 33%).

0920-3796/91/$03.50 © 1991 - Elsevier Science Publishers B.V. All rights reserved

74 E. Proust et al. / Solid breeder blanket design and tritium breeding

This isotope of hydrogen being radioactive with a half-life of 12.35 years, does not occur naturally (but in infinitesimal amounts, created by interaction between cosmic rays and the upper atmosphere). Consequently, tritium, the fusion reactor fuel, has to be synthesized by nuclear reactions.

The most efficient tritium-producing nuclear reaction (the one having the largest cross-section *) is the (n, ~) reaction of lithium-6, the least abundant ( - 7.5%) of the two naturally occuring isotopes of lithium:

6Li + n ~ T +4He + 4.8 MeV.

The technology of large scale tritium production, developed in the 60's for military purposes, relies on this nuclear reaction, set up in dedicated heavy-water moderated fission reactors in which 6Li-containing aluminum targets are irradiated (the Savannah River reactors in the USA). Such reactors are featured by a production efficiency of typically one kg of tritium per year per GWth.

A simple comparison of these two figures (yearly consumption of a D - T fusion reactor: - 50 kg of T per produced GWth, yearly production of a dedicated fission reactor: - 1 kg of T per GWth) clearly shows that breeding in a fission reactor even a very small fraction of the tritium consumed by a fusion reactor would not be energy efficient.

Thus, there is no other economically viable choice for an electricity-generating D - T fusion reactor, but to breed all the tritium it burns using the 14 MeV neutrons that are produced in the plasma chamber by the fusion reactions, and making them react in the blanket with a lithiated compound.

2.2. Tritium breeding requirements

In fact, the tritium breeding performances required by a fusion reactor blanket exceed the simple replace- ment, one for one, of the tritium burnt in the plasma chamber. Indeed, the D - T fusion power plant, because it has to be tritium self-sufficient, will include on site, contrarily to a fission power, fuel (tritium) recovery, reprocessing and storage facilities, which, along with the blanket itself, will contain substantial amounts of tritium (on the order of ten kilos for the whole plant). To this inventory, are associated losses due to the natural radioactive decay of tritium (550 g of T per year for a

* 3He exhibits a high tritium-production cross section at low energy, but its use is impractical because of too limited ressources.

10 kg tritium inventory), losses which have to be compensated for. Furthermore, the plant, at start up, will require an initial tritium provision to feed the plasma chamber while the first amounts of tritium produced in the blanket will be absorbed in the building-up of the tritium inventories in the various tritium cycle units. Some of the tritium produced by the plant will therefore have to be levied in order to constitute the start-up tritium load of one or two plants of the following generation (about 1.5 kg per year to accumulate, over the 30-year lifetime of the plant, two 10 kg T loads).

All in all, the blanket will have to breed between 1.01 and 1.02 tritons per triton burnt in the plasma chamber, so as both to compensate for the decay losses. and to fabricate start-up tritium loads [2].

2.3. Implication of tritium self-sufficiency requirements

Attaining a Tritium Breeding Ratio (TBR, ratio of the tritium produced to the tritium burnt) exceeding unity proves difficult.

Indeed, the production of one tritium atom by the 6Li (n, a) T reaction requires and consumes one neutron, while its fusion with a deuterium generates only one neutron, which exhibits a significant probability (typically about 30-35%) of not being available for tritium production [3], that is to say: - - of undergoing a parasitic absorption (10 15%),

especially in the structural material of the blankets. - - of streaming (10-20%) through the openings pro-

vided in the blanket for accommodating plasma exhausts pumping ducts and plasma heating and diagnostic devices. The geometrical coverage ratio of the plasma chamber by the blanket usually turns around 80-90% depending on reactor internals handling principles and blanket design

- - or of simply leaking (2-10%) the blanket at its back. Indeed, the blanket thickness being limited mainly for economical reasons (typically on the order of 500 and 900 mm for respectively the inboard and outboard sides) is only five to ten times the mean free path of a 14 MeV neutron

The design of a blanket and the materials to be used are thus dictated, among others, by the double necessity: - - of minimizing these neutron losses (search for com-

pactness and high plasma chamber coverage ratio while using a minimum amount of structures and coolant)

- - and of compensating for them by introducing in the blanket, in addition to a compound rich in 6Li, a material apt either at multiplying neutrons by (n, 2n) reactions (like beryllium or lead), or at producing

E. Proust et al. / Solid breeder blanket design and tritium breeding 75

tritium by (n, n'a) T reactions which do not consume neutrons (like 7Li).

3. Candidate b lanke t mater ia l s and se lec t ion cri teria

Five categories of materials enter into the constitu- tion of a solid breeder blanket: - - the breeder material, a solid lithiated compound

which generates tritium. - - the tritium transport fluid, required to transport on

line to the recovery and purification units the tritium produced within the blanket by the breeder material.

- - the coolant material, a fluid whose two functions are to maintain the blanket at a safe temperature level, and to transfer to a steam generator the thermal energy released within the blanket.

- - the structural material, which ensures the mechanical strength and the geometrical integrity of the blanket and its associated circuits.

- - the multiplier material, whose role is to multiply neutrons so as to permit to reach TBRs in excess of 1 despite neutron losses and parasitic absorptions.

This chapter ~s an attempt to briefly present, category by category, the materials usually proposed by solid blanket designers, and the rationale for their choice. When possible selection criteria related to tritium breeding have been separated from design-related ones.

3.1. Solid breeder materials

From the standpoint of tritium breeding only, the best breeder material is the one containing the largest amount of li thium per unit volume, and constituted, besides lithium, of elements having the lowest neutron absorption cross sections.

But other important criteria, related to the design, enter into the selection of the breeder material which, as far as possible, has also: - - to exhibit a low retention to tritium, so as to mini-

mize the amounts of trit ium immobilized in the blanket, and the associated safety problems and breeding requirements.

- - to be chemically compatible with the materials it is in contact with during normal operating conditions - - t h e structural material, the purge gas, the multiplier ma t e r i a l - - and during upset or accidental conditions - t h e coolant - .

- - to be mechanically and structurally stable at high fluence, high lithium burnup, and high temperature (to remain solid, not to swell excessively, no to fall into powder . . . . ).

- - to exhibit a long term nuclear activation as moderate as possible so as to minimize radioactive wastes storage problems.

- - to be industrially fabricable at a reasonable cost. Metallic lithiated compounds (LiA1, Li7Pb 2 . . . . )

having in general a relatively low melting point, the best candidate solid breeder materials are ceramics: lithium

Table 1 Typical properties of candidate ceramic breeding materials (at 500°C).

Li density (g/cm 3) Melting point (°C) Saturation temperature (°C) at

10-2 Pa total partial pressure 1020 Thermal conductivity (W/mK) 4 Lin. thermal expansion

Coef. (10-6/°C) 29 Young modulus (GPa) 60 Ultim. bending strength (MPa) 80 Temperature (°C) corresponding to

a tritium residence time of 1 day 325 *

$ 1 ~tm grain diameter, 20% porosity. • 10 I~m grain diameter.

• * 25 ~m grain diameter, 3% porosity.

Li 2 ° s LiA102 $

0.75 0.22 1430 1750

1270 2.8

12 75 60

430

Li2ZrO 3 s

0.30 1615

> 1350 1.4

11 90 60

310

L i 4 S i 0 4 * *

0.53 1255

< 1120 1.4

30 85 50

380


oxide Li20, or ternary ceramics Li X My O z among which the most actively studied are LiAIO 2, Li2ZrO 3 and LiaSiO 4. Table 1 compares these materials against the above mentioned selection criteria.

These lithium ceramics are generally used in the form either of sintered pellets or slabs, or of small pebbles (typically 0.5 to 2 mm in diameter) when small dimensions are required to limit thermal stresses because of poor thermomechanical properties of the material.

In the former case, the fabrication process is con- ceived so as to obtain a porous product with fine grains (0.3/3 p,m in diameter) and substantial fully open porosity (18-23%), this in order to ease the tritium release by the ceramics. In the latter case, in order to improve, for tritium breeding purposes among others, the intrinsically low apparent density of pebble beds, lower porosity pebbles are usually searched for.

In many cases, in order to maximize tritium breeding and minimize the 6Li-burnup induced TBR decrease [4], these products are fabricated from lithium enriched at up to 95% in 6Li (and thus contain about 12 times more 6Li than non-enriched materials).

3.2. Tritium transport materials

Contrary to liquid breeder blankets, where the same material is used both as breeder material, and as fluid transporting bred tritium from the blanket to the tritium recovery unit, solid breeder blankets require a specific tritium transport fluid.

For this role, helium is by far the best candidate, first because of its chemical inertness, and second because tritium can be relatively easily separated from helium. Its total transparency to neutrons is also an advantage.

3.3. Coolant materials

A variety of materials are being used for heat transport in nuclear steam supply systems: pressurized water in PWRs (15.5 MPa, 290/325°C), gases like helium in HTRs (5 MPa, 200-260/680-825°C) or CO 2 in AGRs and Magnoxs, or liquid metals like sodium in LMFBRs.

Only water and gases (Helium essentially) are candidate coolant materials for solid breeders blankets, magnetohydrodynamic interactions with the strong magnetic field of the fusion machine making liquid metals not appropriate.

Helium offers very substantial advantages over pressurized water, among which some are related to tritium

breeding: total transparency to neutrons, moderate operating pressure resulting in a lower amount of neutron absorbing structures, and others which are related to safety and reliability: chemical inertness (total chemical compatibility with any candidate blanket material, no transport of activated corrosion products, no large pressure build-up and no hydrogen release in case of loss of coolant accident, possibility to tolerate coolant micro- leaks towards the purge gas circuit), relative ease of purification (with respect to tritiated impurities), and temperature level--up to 450-500°C or more--well adapted to the thermal conditioning of lithium ceramics, most of which requiring operating temperatures in excess of 300 to 400°C to exhibit no prohibitive tritium inventories (see Table 1).

In counterpart, helium implies structures working at higher temperatures (at least 450-550°C compared to 350°C with water), and because its density is much lower than water, it occupies a larger volume in the blanket (making it more leaky to neutrons, a drawback for tritium breeding) and requires a higher pumping power (which is detrimental to plant thermal efficiency).

From the standpoint of tritium breeding, pressurized water exhibits both advantages and drawbacks: in the front part of the blanket, its presence is disadvanta- geous, since elastic neutron scattering on hydrogen and inelastic scattering reactions (n, n ' ) with oxygen strongly slow down high energy neutrons, making them less apt to induce (n, 2n) reactions on the multiplier material. In the back part of the blanket on the contrary, the high slowing-down power of water is beneficial due to the presence of hydrogen which increases the probability of the neutrons to undergo a tritium breeding adsorption on 6Li.

One of the main reasons why pressurized water is often discarded in favour of helium is generally related to safety and reliability (even small water leaks toward the breeder material or to the plasma chamber will be difficultly tolerable) and not to tritium breeding considerations.

3.4. Structural materials

The most irradiated structures of the present nuclear power (fission) reactors--the cladding of the fuel ele- m e n t s - a r e made either of low neutron absorption zirconium base alloys (when the neutrons are thermal, and the temperature and fluence moderate as in PWRs: 400°C, 15 dpa max) or of austenitic stainless steel (Ti stabilized 316 LN when temperature and fluence are


higher as in LMFBRs -650°C, 140 dpa). For these latters, ferritic steels are also actively considered.

The temperature and irradiation conditions encountered in D - T power reactor blankets being closer to LMFBR ones than to PWR ones, and the presence of welds under irradiation preventing the use of low-swelling stabilized austenitic steels, the structural materials usually considered for blankets are ferritic or martensitic steels, although some advanced blanket concepts [5,6] rely on more heat-resistant, neutronically less ab- sorbant, and /o r lower activation materials [7] like molybdenum alloys, vanadium alloys or silicon carbide, which are still to be developed, or, at the very least, to be qualified under irradiation.

Here, the choice between existing structural materials is solely dictated by mechanical requirements. The substantial neutron absorption of steel has to be accepted, its negative effect on tritium breeding must be minimized as far as possible, by developing blanket designs requiring a low amount of structures, and compensated for using efficient neutron multipliers.

3.5. Neutron multiplier materials

As already mentioned in section 3.3, neutron multiplication is required in a D - T power reactor solid breeder blanket. A large number of non-fissionable nuclides exhibits neutron multiplication or (n, xn') reactions. However, in most cases, either their (n, 2n) cross section (XS) at energies below 14 MeV is not sufficient (XS below one tenth of a barn), or, more often, their neutron absorption XS is high. So that only three materials--Pb, Be, Z r - - c a n be considered as candidate neutron multipliers for breeding blanket ap- plications.

Lead is featured by the highest (n, 2n') XS at 14 MeV--2.15 ba rn - -bu t its reaction threshold--around 7 MeV--a lso is high so that secondary neutrons emitted in a (n, 2n') reaction have an energy below 7 MeV, and therefore cannot induce a further (n, 2n') reaction.

Beryllium, although its (n, 2n') XS is four times lower than the one of lead, is a much better multiplier. Indeed, its (n, 2n') reaction threshold--l .85 M e V - - being much lower, secondary neutrons emitted in a (n, 2n') reaction are still energetic enough to induce a second (n, 2n'), enabling a neutron multiplication factor exceeding 2. Furthermore, 14 MeV neutrons having been slowed-down by e.g. inelastic scattering on the oxygen contained in the breeder material, can still par- ticipate in the multiplication.

The behaviour of Zirconium is similar to the one of lead, but with a twice lower cross section at 14 MeV,

which furthermore decreases much faster towards the reaction threshold. In addition, Zr shows a larger absorption cross section compared to lead.' So that its multiplying performances prove unsufficient.

Both beryllium and lead meeting the neutron multiplication minimum requirements of fusion reactor blankets, the fact that beryllium be the multiplier material selected in most solid blanket concepts is not due solely to its higher multiplying capability.

From the designer point of view, although it exhibits very substantial drawbacks such as its propensity to embrittlement and swelling under irradiation [8,9], or its fabrication cost, beryllium is indeed by far preferable to lead. The main reason is related not to the high thermal conductivity ( --- 100 W / m K ) , high modulus of elasticity ( - -270 GPa), or low density ( ~ 1.85 g / cm 3) of beryllium, but to the low melting point of lead (325°C) and to its volumetric expansion when it changes phase. The usual operating temperature range of candidate coolants (280/325°C for pressurized water, 250/500°C for helium) makes it extremely complicated to maintain lead entirely in one single s tate-- l iquid or sol id--during normal operation. The phase changes lead would unavoidably undergo within the blanket are considered likely to induce prohibitive stresses on the structures (lead compounds such as ZrsPb 3 or PbO initialy considered for their higher melting point prove to have a too low multiplying power). An additional drawback of lead, in the case of helium cooled blankets, is the need for some specific neutron moderating material at the back of the blanket for reducing the leakage of neutrons and increasing their absorption in 6Li [10], which com- plicates the design. Beryllium, which is also an efficient moderator, can serve both as multiplier in the front part of the blanket, and as moderating material in the back part.

It must be noted here that, because the other natural isotope of lithium 7Li exhibits an improperly called "neutron catalysed" nuclear reaction generating tritium without consuming neutron:

7Li + n ~ T + 4He + n' - 2.5 MeV

the use of neutron multipliers could, in theory, be avoided if a sufficiently large number of such (n, n 'a) T reactions could be induced on 7Li to compensate for neutron losses and parasitic absorptions.

In practice, because of the low cross section and of the high energy threshold (2.5 MeV), and also because of the insufficient density in lithium of usual solid lithiated compounds (except for Li20 ) 7Li only contrib- utes for a few percent to the tritium production.


3.6. Candidate materials summary 4.1. Optimal use of beryllium

The materials presently considered as best suited to a solid breeder blanket for a D - T power reactor are: - - as breeder material ceramics: Li20, LiA102, Li4SiO4,

Li2ZrO 3 fabricated from lithium enriched at up to 95% in 6Li.

- - as purge gas: helium. - - as neutron multiplier: beryllium. - - as structural material: ferritic/martensitic steel, or

advanced materials. - - as coolant: helium (5-10 MPa, 250/450-550°C),

pressurized water (15.5 MPa, 280/325°C). The choice between the breeder materials and between the coolants mentioned above being essentially a matter of design.

4 . D e s i g n g u i d e l i n e s f o r a c h i e v i n g h i g h T B R s

As already indicated in section 3, the blanket of an electricity-generating D - T fusion reactor is required to breed about 1.01 to 1.02 tritium atom per fusion neutron, while the production of one tritium atom consumes one neutron. Despite the existence of an efficient neutron multiplier--beryllium--, this requirement is quite demanding and consequently strongly influences the design of a solid breeder blanket.

Design guidelines given by the neutron physicists for achieving high tritium breeding ratios can be sum- marized as follows [11]: - - make an optimal use of beryllium by blending it

homogeneously with the breeding ceramics throughout the blanket in a Be to ceramics volume ratio of 80/20 to 90/10.

- - minimize the amount of structures. - - minimize the blanket volume fraction devoted to the

coolant. - - search for compactness. - - maximize the blanket coverage ratio.

Unfortunately, these design guidelines cannot be fol- lowed strictly because they often enter into conflict with technological requirements, The task of the blanket designer, which is to propose a design that not only meets tritium breeding requirements, but also exhibits reasonable development risks, is therefore challenging.

The purpose of this chapter is to tentatively explain the reasons for the above mentioned guidelines and to discuss the technological problems encountered by the designer who tries to follow them.

As already indicated in section 3.3, a neutron multiplier is required to enhance the TBR from say 0.65 (typical net TBR of a blanket containing only breeder material, steel structures and coolant) to the required 1.01 (i.e. a - 55%, increase).

Its multiplying power decreasing with the neutron energy to become zero below - 1.7 MeV, Be would be best used, a priori, when put directly behind the first wall (the region where neutrons are the most energetic) in the form of a slab in front of a lithium ceramics zone. However, in this arrangement, the high slowing-down power of Be (it is a light nucleus and the metal exhibits a high atomic density) results in a strong back scattering of moderated neutrons easily absorbed in the steel first wall, so that Be in this location, although it multiplies neutrons by 2, increases the TBR by 40% only, which is insufficient in most cases (with the possible exception of gi2 ° breeder blankets).

Two solutions for an optimal use of beryllium, which can be applied to all ceramic breeders, are proposed by the neutron physicists [12]: - - to put a thin ( - 1 cm) layer of breeder material in

between the first wall and the beryllium slab so as to absorb the back scattered neutrons in 6Li without significantly interfering with (slowing down) the 14 MeV neutrons. This solution is generally not used for power blankets because the strong neutron absorption in this front breeder material layer results in very high power densities (requiring complex thermal-hydraulics designs) and lithium burnups (and thus in short ceramics lifetimes).

- - to switch from the above "sandwich" arrangement to one where beryllium and the ceramics are "blended homogeneously" throughout the blanket with a beryllium to ceramics volume ratio of 70/30 to 90/10. Designs relying on the use of a bed of mixed beryllium and ceramics pebbles (see section 5.1) are a typical example of the implementation of this arrangement which, although ideal from the TBR point of view, raises two severe technological problems: - - the oxygen rich ceramics are poorly chemically

compatible with Be at the high temperatures--up to 800°C--encountered in "mixed" pebble beds (which are poor thermal conductors).

- - Be swelling, which strongly increases with temperature [9], is expected to be very high at the temperature level typical of "mixed" pebble beds, while the swelling accommodation capability of pebble beds is not yet demonstrated.


These problems can be avoided without sacrifying tritium breeding by adopting a design like the one presented in section 5.2, where Be and breeding ceramics are arranged in an heterogeneous way (ceramics pebbles layers in between internally-cooled radially-oriented Be plates), but where the heterogenity features are small compared to the mean free path of neutrons.

Such "low heterogeneity" designs, requiring in most cases a very segmented cooling circuit (large number of cooling tubes), are demanding in terms of reliability, thus much more heterogeneous arrangements are also considered (see section 5.4), whose lower--but still acceptable--tri t ium breeding performances are tentatively compensated for by a better potential reliability and a better Be-swelling accommodation capability.

4.2. Minimization of the amount of structures

As already indicated in section 3.4, besides advanced materials never yet used in nuclear (fission) power reactors, the only candidate structural materials of solid breeder blankets are steels, which, unfortunately absorb neutrons.

The amount of structures the blanket contains must therefore be minimized.

The first wall and the blanket structures proper must here be distinguished, since they have a different influence on the TBR.

The first wall is the first structure encountered by the 14 MeV neutrons, its negative effect is thus not only to absorb back scattered neutrons (see section 4.1), but also to slow down 14 MeV neutrons before they can interact with Be, making them less apt to induce (n, 2n) reactions. The first wall thickness impact on the TBR is therefore strong: typically 3 to 10% per cm of steel (depending on the coolant material, on the kind of steel, and on the design concept). The minimization of its thickness passes by a careful sizing on the basis, among others, of detailed calculations of the disruption effects.

The effect of the blanket inner structures (and segment box side walls) is mainly to absorb the neutrons reemitted by Be. The importance of this effect increases with the volume fraction of structures, but depends also on the blanket arrangement. Configurations where neutrons reemitted by Be are obliged to cross thick steel layers for reaching the breeder material and produce tritium are penalized. Such a penalization, if reasonable, is a price designers sometimes accept to pay for improv- ing the technological feasibility of their blanket. The design concept presented in section 5.4 illustrates this situation: for reasons related to the thermal conditioning of Be (swelling minimization) and of the ~ithium

ceramics (tritium inventory minimization) and to the accommodation of Be swelling, Be and the lithium ceramics are separated by several layers of steel, resulting in a 4 point penality on the TBR.

4.3. Search for a low coolant volume fraction

4.3.1. Case of a water coolant In the front part of the blanket, where the neutron

energy is highest and consequently where essentially neutron multiplication occurs, water is detrimental to tritium breeding mainly because elastic scattering on hydrogen and inelastic scattering on oxygen break the energy of 14 MeV neutrons, reducing thereby the probability of inducing (n, 2n) reactions on Be. Unfor- tunately, it is also the blanket region where coolant is most necessary (because of the neutron attenuation, the power density peaks just behind the first wall and steeply decreases to fall at the back to about 1/10 of its peak value). Minimization of the coolant volume fraction can be obtained through the adoption of high flow velocities (limited by corrosion and vibration phenomena), of low inlet and high outlet temperatures (limited respectively by thermal efficiency requirements and by water saturation pressure), and of configurations ex- hibiting a good thermal conductivity and allowing high peak temperatures.

In the back part of the blanket, water like beryllium, acts as an efficient moderating and reflecting material and its presence is thus useful as it increases neutron absorption on Li 6. However, due to the low heat deposi- tion in this region, the obtainment of a high coolant fraction is usually incompatible with the requirement of a low amount of water in the front part of the blanket.

4.3.2. Case of a gaseous coolant (hefium) Gaseous coolants like helium being fully transparent

to neutrons, their presence is detrimental to tritium breeding only because it makes the blanket more leaky and thus increases neutrons losses. Minimization of gaseous coolant volume fraction in the blanket can be obtained in the same way as in the case of a water coolant. However the phenomena limiting this minimization, as well as the sensitivity of the TBR to the presence of the coolant are basically different: - - because of higher power densities, mainly the front

part of the blanket will contain a relatively large amount of coolant. Both multiplication and tritium production will be lower in this region, but consequently more neutrons will be available in the much less leaky back region for reacting with Be and 6Li


and thus for partly compensating for the lower reaction rate of the front region

- - the increase of the coolant velocity is limited here by pumping power requirements which are much higher for a gas than for a liquid like water, and therefore have a much stronger impact on the thermal efficiency of the plant.

For these reasons, both the coolant velocities and operating pressure adopted for fusion reactor blankets generally are higher than those adopted in gas-cooled fission reactors.

4.4. Search for a high compactness

The space available for the blanket and the radiation shield of a tokamak reactor is restricted for machine cost considerations. The minimization of neutron leakage through narrow blankets, made necessary mainly by tritium breeding requirements, but also by shielding requirements, passes by the adoption of as compact as possible blanket configurations.

The examples of solid breeder blanket design concepts presented in section 5 however clearly show that some technical considerations oppose to the attainment of a high compactness.

Thus, the design concept presented in section 5.2 is featured by a large space devoted to collectors at the back of the blanket, which is due to the adoption of helium as coolant (mainly for safety and reliability considerations, see section 3.3) in a radial (in the breeder modules) routing scheme (search for a good thermal conditioning of the breeding ceramics so as to minimize the tritium inventory, see section 3.1). The penalty on the TBR is however very low here, owing to the high compactness achieved in the (front) breeding zone proper.

Similarly, the use of breeding ceramics in the form of pebble beds instead of 35% denser massive blocks can be required for some materials so as to avoid they fall into pieces and powder under the effect of thermal stresses. This penalty, in terms of TBR, is generally affordable because the breeder materials accounts for only 10% (often less) of the total blanket volume.

Finally, most solid breeder blankets exhibit a neutron leakage rate ranging typically from 0.05 to 0.10 per 14 MeV neutron for inboard/outboard blanket thick- nesses of 50/90 cm.

4.5. Search for a high blanket coverage ratio

The blanket cannot entirely surround the plasma chamber, because, among others, of the need to accom- modate plasma exhausts pumping ducts and plasma

heating, fuel feeding, diagnostics and impurity control devices.

The areas not covered by the blanket give rise to neutron losses by neutron streaming and thus degrade the TBR. Because of the elongated shape of the plasma - - the neutron source--, openings are more penalizing when located in the equatorial plane. Fortunately, the largest openings (those associated with the impurity control and exhausts pumping system when pumped divertors are used) are located at the top and bottom of the plasma chamber. They nevertheless can lead to a loss of as much as 0.15 neutron per 14 MeV neutron.

The maximization of the blanket coverage ratio is a matter not only of blanket design (poloidally segmented (see section 5.2) or poloidal-box shaped (section 5.1) designs are preferable in this respect) but also of reactor design. Thus, of special importance is to design the reactor and its handling strategy so as to allow the location of breeding zones behind the divertors, which, despite the screening effect of absorbing divertor plates. can boost the TBR by up to 5% (10% for liquid breeder blankets).

5. Typical design concepts of solid breeder blankets for power reactors

The previous section was an attempt, on the basis of a discussion of the design guidelines given by the neutron physicists for achieving high tritium breeding ratios, to show that to design a blanket essentially consists in finding the right compromise between tritium breeding and technological requirements.

The purpose of this section is to show a few typical design concepts of solid breeder blankets for power reactors in order to illustrate the compromises proposed by various design teams. The selection is arbitrary, and no claim is made that all the most important concepts are included.

As a foreword, it must be pointed out that the large differences between these design concepts are due of course to the difference in sensibility, background and engineering judgement of the designers, but also, and likely mainly, to the present lack of knowledge on the behaviour of the considered materials under the irradiation and temperature conditions encountered in fusion reactor blankets.

5.1. Water-cooled poloidal pressure tubes blanket concept (Japan)

In this blanket concept [5], the segment box, cooled by toroidally-running cooling channels connected to a


poloidal manifolding, is used as a poloidal container filled-up either with a mixture of Be and Li20 pebbles or with only Li20 pebbles. In the latter case, illustrated in Fig. 1, neutron multiplication is ensured by a beryllium plate located between the first wall and the breeder region. The pebble bed is cooled in parallel with the segment box by embedded poloidal tubes within which pressurized water (15.5 MPa, 280/320°C inlet /out le t temperature) is flowing. Coolant supply and return being both realized in the upper part of the blanket. Each of these poloidal tubes is provided, on its outer wall, of a thermal resistant layer so as to maintain the ceramics at the high temperature level required to minimize the tritium inventory in this material.

The choice of a lithium ceramics in the form of small diameter ( < 1 mm) pebbles aims at avoiding thermal cracking of the breeder material and at obtaining a good packing fraction in the narrow gaps between cooling tubes. The adoption of a mixed bed of Li20 and Be pebbles reflects the concern to optimally use Be by blending it homogeneously with the ceramics throughout the blanket. The first wall is made structurally integral with the blanket proper also with a view to enhancing tritium breeding by minimizing the amounts of structures.

It should be noted that this design concept is also proposed with helium cooling [5]. The main differences being the suppression of the insulating layer around the

tubes, and the switch from austenitic steel to a molybdenum alloy as structural material so as to allow helium inlet /outlet temperatures of 400/700°C.

5.2. Helium-cooled radial pressure-tubes blanket concept (EEC)

This blanket concept [13] relies on the use of breeder modules made-up of nearly rectangular self-supporting canisters mounted on the back plate of the segment box, the ones over the others.

Each canister (Fig. 2), whose walls are cooled by brazed pressure-tubes, contains the beryllium multiplier in the form of vertically arranged plates separated by gaps. These gaps are filled with a bed of 0.35 to 0.6 mm in diameter lithium orthosilicate (Li4SiO4) pebbles through which a low-pressure purge gas flows. Cooling of the Be/ceramics assembly is ensured by helium (8 MPa 250/450°C inlet /outlet temperature), which after a preheating through the first wall, flows through cooling coi ls--embedded in each Be plate--which are connected to toroidally-running subheaders routing it to poloidal manifoldings. The canisters are furthermore provided with radial stiffening plates so as to withstand the pressure build-up consecutive to the hypothetical failure of a cooling coil.

This general arrangement allows to achieve a high compactness in the breeding zone proper, which com-

(PCA RESISTANCE

SECTION A

Fig. 1. Schematic view of the water-cooled poloidal pressure tubes blanket concept (Japan).

82 E. Proust et a L / Solid breeder blanket design and tritium breeding

850

He Cooling System 1 EB Welded He Coolinq System 2 Purge Gas

Outlet Helium Coolant

InLet

Breeder Coolant Tubes

Pebble Bed Breeder Canister

Coolant

'~ P'urge L~as bupply 18,5 19,75

'Segment Box with First Wall 34

~..24 ~__

Copper Saddle

and Shield ~

Fig. 2. Helium-cooled radial pressure tubes blanket concept (EEC). Outboard segment isometric view and mid plane cross section.

pensates for the large space occupied by the manifoldings required, at the back part of the blanket, to accom- modate the radial coolant routing scheme. Radial coolant routing makes it possible to obtain rather ho- mogeneous temperatures in the canisters despite the steep power-density radial grandients.

5.3. Helium-cooled radial pressurized module blanket concept (USA)

The general architecture of this blanket concept [14] with radial breeder canisters is similar to the one of the previous concept, with nevertheless a major difference: these canisters are not contained within a segment box as in the EC design, their nose directly receives the flux of particles.

The design of the canisters is however basically different as shown in Fig. 3: beryllium and the breeding ceramics (LiA102) are used in the form of bare toroidal rods and cladded radial plates respectively. They are externally cooled with helium (275/510°C) flowing radially, so that the canister enveloppe is internally pressurized at the coolant pressure (5 MPa). A separate purge gas stream of low pressure helium passes through the breeder plates for tritium extraction.

The selection of a plate geometry for the solid breeder

reflects the concern to maximize the breeder volume fraction in the blanket and to minimize the structure volume fraction. The suppression of the segment box

STRONG BACK END PLATE3 /

INSULATED DIVIDER P L A T E ~

FLOW

Fig. 3. Schematic view of the breeder module generic of the helium-cooled radial pressurized canister blanket concept

(USA).


FIRSt WALL OUTLET COQLANT HEADER

FIRST WALL SEGHENT INLET BOX PRESSURE COOLANT BACKPLATE VESSEL OAFFLE HEAOER

D.IOUDI

CCHrRAt 00~Y Roo pu~G~ GA~ ~CHAINEL B~EEO~ CER~NIC A~NULAR PELUET StPlRATO~ ( BAFFL[ I

pRESSURt

Fig. 4. Helium-cooled poloidal breeder rod blanket concept (EC). Vertical and mid plane blanket cross sections and details of the breeder modules.

aims, among others, at reducing the amount of structures in the front part of the blanket, while the radial coolant routing scheme provides the advantages already mentioned.

5.4. Helium-cooled poloidal breeder rod blanket concept (EEC)

This concept [15] adopts a poloidal modular architecture (Fig. 4). Poloidally running breeder modules of two different designs are associated inside the segment box. At the back of the blanket, they consist of a steel pressure tube closed at its lower end (thus acting as a pressure vessel), and on which beryllium blocks are either brazed or mechanically attached. This pressure tube contains a bundle of breeder rods surrounded by a baffle. Each one of the rods is made-up of a tube containing a stack of annular pellets of lithium gamma aluminate LiAIO 2 through which the purge gas flows. The modules are connected at their upper end to a coolant supply and return (coaxial) duct. The coolant,

helium at 6 MPa, 300/530°C inlet/outlet temperature, first flows downwards in the annular space between pressure tube and baffle, and then upwards through the bundle.

This coolant routing scheme allows to maintain both beryllium and the pressure tube at moderate temperature while taking advantage of the heat deposited in these materials for thermally conditioning the breeding ceramics, minimizing thus its tritium inventory.

The breeder modules located in the front part of the blanket closer to the plasma are exposed to a much higher neutron flux. Their slightly different design (smaller rods, larger coolant volume fraction, beryllium located inside the pressure tube) provides a better accommodation of higher power densities and fluence.

6. Conclusion

Thermonuclear D - T power plants will have to be tritium self-sufficient.

84 E, Proust et al. / Solid breeder blanket design and tritium breeding

In addition to recovering the energy carried by the fusion neutrons (more than 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burned in the fusion process.

Trit ium breeding requirements (TBR - 1 , 0 1 ) are very severe and therefore dictate, to a large extent, the choice of some materials (mainly the neutron multiplier and the breeder materials) and of the design concept. However the blanket operating conditions and func- tional specifications are also quite demanding in terms of temperature, neutron fluence, and design complexity, so that the task of the designer, which has to conceive a solid blanket meeting not only tritium breeding requirements, but also technological and economic requirements, proves to be quite difficult.

This is all the more true considering that, if the tritium breeding performances of a blanket design concept can be evaluated with sufficient accuracy, this is not the case of its feasibility (the behaviour of considered blanket materials under the conditions of a fusion power reactor are indeed still largely unknown). To progress in the demonstrat ion of the feasibility of achieving trit ium self-sufficiency in the fusion power plant will require important technological efforts.

Over the past 40 years, fusion research has been focused on plasma physics. Progress has been such that breakeven and ignition are now within hand reach. In the future, the emphasis will therefore have to be put more and more on addressing the technological problems associated with obtaining electric power from con- trolled thermonuclear fusion.

References

[1] ITER Test Program, ITER Documentation series No, 24 (International Atomic Energy Agency, Vienne, 1990).

[2] F. Carrr, E. Proust and A. Rocaboy, Analysis of the tritium requirements for a power reactor, Fusion Technol. 4, (1983) 805-810.

[3] F. Barr6, F. Gervaise and L. Giancarli, Fusion reactor blanket neutronic studies in France, Fusion Technol. 4 (1983) 799-804.

[4] L. Giancarli, Time-dependent neutron-induced effects in a

DEMO fusion reactor blanket: Burnup calculations, Fus. Technol. 10 (1986) 100-107.

[5] T. Kuroda, Technical considerations on breeding blanket systems for fusion power reactors, Fusion Engrg. Des. 8 (1989) 219-226.

[6] C.P.C. Wong, E.T. Cheng, B. McQuillan, E.E. Reis, K.R. Schultz, S.P. Grotz, M.Z. Hasan, R. Martin, F. Najmabadi, S. Sharafat, T. Kunugi, J.S. Herring, D.K. Sze and the ARIES Team, ARIES-I SiC composite low activation blanket design, Proc. 9th Topical Meeting Technology of Fusion Energy, Oak Brook, Illinois, October 7-11, 1990, in Fusion Technology 19 (1991) 938-943.

[7] G.J. Butterworth, Low Activation Structural Material, Fur sion Technol. 1 (1988) 231-244.

[8] J.B. Rich, G.P. Walter and R.S. Barnes, The mechanical properties of some highly irradiated beryllium, J. Nucl. Mater. 4 (1961) 287-294.

[9[ J.M. Beeton and L.G. Miller, Comparison of compression properties and swelling of beryllium irradiated at various temperatures, J. Nucl. Mater. 122-123 (1984) 802-812.

[10] L. Anzidei, M. Gallina, L. Petrizzi and R. Gatto., A helium-cooled ceramics breeder blanket concept with lead/graphite as multiplier/moderator, Proc. 16th Sym- posium on Fusion Technology, London, UK, September 3-7, 1990 (Elsevier, Amsterdam, 1991).

[11] S. Taczanowski, Guidelines for fusion reactor blanket nuclear design, Fusion Technol. 1 (1982) 687-692.

[12] U. Fischer, Optimal use of beryllium for fusion reactor blankets, Fusion Technol. 13 (1988) 143.

[13] M. Dalle Donne, E. Bojarski, U. Fischer, M. Kiichle, P. Norajitra, G. Reimann, H. Reiser, G. Sordon, H.D. Baschek and E. Bogusch, The Karlsruhe helium cooled ceramic breeder blanket design for a demonstration reactor, Proc. 16th Symposium Fusion Technology, London, UK, September 3-7, 1990 (Elsevier, Amsterdam, 1991).

[14] D.L. Smith, G.D. Morgan et al., Blanket Comparison and Selection Study. Final Report, ANL report FPP-84-1 (Sept 1984).

[15] E. Proust, L. Giancarli, X. Raepsaet, J. Szczepanski, k. Baraer, B. Bielak, F. Gervaise, J. Mercier, F. Valette, k. Anzidei, P. Cecchi, S. Cevolani, M. Gallina, k. Petrizzi, V. Rado, V. Violante, V. Vettraino and V. Zampaglione, Status of the design and feasibility assessment of the European ceramics B.I.T. test blanket, Proc. 9th Topical Meeting Technol. of Fusion Energy, Oak Brook, Illinois, October 7-11, 1990, in Fusion Technology 19 (1991) 944-950.

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