2 LIQUID METALS AS REACTOR COOLANTS

9

2

LIQUID METALS AS REACTOR COOLANTS

2.1 Introduction

The possibility of developing a nuclear reactor producing more fuel than it consumed (a “breeder reactor”) was first raised during II World War in the United States by scientists involved in the Manhattan project. This was connected to the fact that fissile isotopes are essential nuclear materials for reactors and nuclear weapons. In particular, at that time, it was unclear whether there was enough uranium to produce highly enriched uranium and plutonium for a significant number of nuclear weapons.

Fermi and his colleagues at the Metallurgical Laboratory of Chicago University looked for ways to produce maximum power (or plutonium for weapons) with minimal resources. They recognized that some reactor configurations might permit the generation of new fissile material at a greater rate than its consumption, hence the term “breeder reactor.”

In general, fissile isotopes undergo fission when they absorb neutrons and, on average, release more neutrons than they absorb. This makes a sustained chain reaction possible in a “supercritical mass.” This supercritical mass must contain a significant concentration of fissile isotopes and must be large enough so that only a small fraction of the neutrons escape without interacting. The most important fissile materials are U235 and Pu239.

U235is available in nature, constituting 0.7 percent of Unat. Pu239is created when U238 (99.3 % Unat) absorbs a neutron (Fig.2.1 [1]).

Fig.2.1. Plutonium breeding

Most of the power reactors around the world are fueled with low-enriched uranium and use water, which also serves as reactor coolant, as neutron “moderator”. This allows to slow the neutrons thus increasing the likelihood of their capture by U235 and then their fission.

In the LWRs, but also in heavy water reactors, neutrons lose most of their energy in collisions with hydrogen (heavy-hydrogen or deuterium in heavy-water reactors). In both types of reactors, some neutrons from U235 fissions are captured by U238, which is converted into chain-reacting Pu239. In spite of this Pu239is not enough to replace the fissioned U235.

On the light of this, all breeder reactor programs have focused on reactors that do not use water as a coolantbecause collisions with the hydrogen nuclei in water quickly remove most of the kinetic energy from the neutrons.

Nevertheless, the fissile material in a reactor core must be more concentrated in order to sustain a chain-reaction with fast neutrons. As a result of this, fast-neutron reactor cores are smaller than those of light-water reactors with the same power, which leads to use a coolant able to carry away efficiently the heat.

Summarizing, the choice of fast neutron reactors implies a cooling system able to remove heat safely without any softening of the spectrum. Consequently, on the one hand, the amount of coolant must be limited as much as possible and on the other hand the use of high mass number or low fluid density is requested.

On this basis, the choice is limited to gases or liquid metals, even though gas is not in the primary choice, because it is not able to provide an adequate cooling in case of loss of coolant accident (LOCA), since its heat transfer coefficient is low.

Thus, liquid metals are the only choice for this kind of reactors.

2.2 Motivations for choosing liquid metals

Mercury represented the first attempt of the liquid metals use as reactor coolant. It is a heavy metal that practically does not moderate and absorb neutrons and therefore, from the point of view of physics, it was at first an ideal choice. In addition, it does not react with air and water in a very pure state except at relatively high temperatures. It reacts vigorously with Na and K and in pure O2 above 350°C.

Clementine [2] was the code name for the world's first fast neutron nuclear reactor, which was built at Los Alamos National Laboratory in 1945-46. It was an experimental scale reactor with a maximum output of 25kWth that achieved criticality in 1946 and full power in 1949.

The core was fuelled byplutonium metal with natural uranium slugs at each end of the steel-clad rods and cooled by liquid mercury with a maximum output of 25kWth. The mercury was circulated through the core and through a mercury-water heat exchanger at a maximum flow rate of 2 kg/s by an induction type electromagnetic pump which had no moving parts.

Clementine operated successfully from 1946 until 1950 when the reactor was shut down to correct a problem with the control and shim rods. During this shut down it was noted that one of the natural uranium rods had ruptured. It was replaced and the reactor was restarted. It was again operated successfully until 1952 when then cladding on one of the fuel rods ruptured. This caused contamination of the primary cooling loop with plutonium and other fission products.

At this time it was decided that all the primary objectives of Clementine had been achieved and the reactor was permanently shut down and dismantled. The experience and data provided by operating the Clementine reactor was very useful for both military and civilian applications. One of the notable achievements of the Clementine project included measurements for the total neutron cross sections of 41 elements to a 10% accuracy. Additionally, Clementine provided invaluable experience in the control and design of fast neutron reactors. It also helped in understanding that mercury was not an ideal cooling medium for this type of reactor.

11 Coolant choice was set on mercury even for the first soviet fast reactor. In 1956, BR-2 (Breeding Reactor 2) began operation in Obninsk [1,3] but experience showed, also here, that mercury was not attractive as coolant for several reasons. Firstly, the metal plutonium fuel was not stable under irradiation even at low temperatures and mercury leaked from pipe joints produced a significant corrosive impact on the structures materials. Furthermore, mercury is dangerous for men's health because its vapor is very toxic, especially after long-term exposure. The physiological influences are still increased in a humid and CO2-containing atmosphere [4].

The following attempt of using liquid metals concerned the sodium-potassium alloy. In 1951 the experimental reactor EBR-I (acronym of Experimental Breeder Reactor I) was built in the Idaho National Laboratory, United States. It represents a milestone in developing nuclear power reactor because it was the first reactor that produced electricity, through a Rankine steam cycle, for lighting four light bulbs at first and then running the whole EBR-I building [5].

Also the Na-K cooled reactor DFR (Dounreay Fast Reactor), which was built in Dounreay, Great Britain in 1956, represented a very important step in the fast nuclear reactor because it was the first reactor to be connected to the national grid [1].

In addition to the demonstration’s feasibility of a safe and easily operable fast reactor, it provided essential information about measurements of reactor physics data, development of shielding calculation methods, effects on materials subject to irradiation in a high flux of fast neutrons and the condition of Na-K [6].

Even Soviet Union used sodium-potassium alloy, but just for equipping the secondary loop of the experimental reactor BR-5 launched in 1959 [7].

The choice of Na-K was suggested owing to several reasons. First of all, Na-K has a low melting temperature (-11Cº [4]) and consequently it is liquid at room temperature, as well as mercury. This feature allows at simplifying reactor design, which does not need to include a heating system aimed at maintaining coolant in a liquid state. Furthermore, it is not corrosive for steel as mercury is. But also Na-K that, at first, seemed to be the most suitable coolant, revealed its drawbacks.

Firstly, it demonstrated to be more inflammable and more aggressive than sodium. In fact, it reacts violently with water, liberating and igniting flammable hydrogen gas. It reacts also with air or oxygen, if exposed.

Then, inspection and repair work turned to be more complicated than expected just because of being liquid at room temperature.

For this reason, the Na-K alloy was abandoned in favor of the pure sodium [7], which played the major role as fast breeder reactor coolant in the ’60s and in the ‘70s.

During these two decades, the long-term energy supply became the most important problem that was thought to be solved by plutonium-fueled breeder reactors. They appeared to offer a way to avoid a potential shortage of the low-cost uranium, due to the estimationsof the uranium scarcity.

Thus, a big effort was undertaken in a sort of scientific-technological competition among several countries. Soviet Union, United Kingdom, France, Germany, Japan and India followed the United States in establishing national plutonium breeder reactor programs, while Belgium, Italy and the Netherlands joined the French and German took part in such programs as partners.

A lot of sodium cooled fast reactors were designed and built during those years, as can be seen inTable 2.1.

In USA, EBR-II reactor was built in 1963 after the promising experience with EBR-I, providing heat and power to the Idaho facility from 1964 to 1994 [5]. It was intended for demonstrating a complete sodium-cooled breeder reactor power plant with on-site reprocessing of metallic fuel. This was successfully done in 1964-69 and then it shifted to testing materials and fuels (metal and ceramic oxides, carbides and nitrides of U & Pu) for larger fast reactors.

The EBR-II was the basis of the US Integral Fast Reactor (IFR) [8,9], that is a liquid metal-cooled fast reactor using metal fuel and integrated with an on-site fuel recycling plant, considered by the National Academy of Sciences to be the nation's highest priority research program for future reactor types but opposed politically by US administrations and advisers seeking to minimize proliferation but not understanding the complex issues involved. Congress under the Clinton administration shut down EBR II in 1994 [9].

The first US commercial SFR (Sodium Fast Reactor), Fermi-1, was built in Michigan almost at the same time of EBR-II (1965), but it operated for only three years before a coolant problem caused overheating and it was shut down with some damage to the fuel. After repair it was restarted in 1970, but its license was not renewed in 1972.

The 400 MWth Fast Flux Test Facility (FFTF) was in full operation 1982-92 at Hanford as a major national research reactor [8].

In UK, the success of DFR ’s experience led to the following much larger Prototype Fast Reactor (PFR) which operated for 20 years until the government withdrew funding [1,6].

Germany also built sodium reactors. KNK-I (acronym of Kompakte Natriumgekülte Kernreaktoranlage I) was built in Karlsruhe and operated with a thermal core from 1971 to 1974 in order to gain experience on sodium coolant. From 1977 to 1991, KNK-I became KNK-II, a demonstrator fast breeder power plant running with a fast core. The plant was shut down in 1991. Meanwhile, in 1985, the construction of SNR-300 (Schneller Natriumgekülter Reaktor 300) was complete and the sodium coolant was already running through the coolant loop but, at the moment of receiving nuclear fuel, government blocked the actual opening of the plant that never operated for political reasons [10,11] .

Construction of France’s first experimental sodium-cooled reactor, RAPSODIE, started in 1962 and it went critical on 28 January 1967 with a nominal capacity of 20 MWth.

RAPSODIE was a loop-type reactor, with the heat exchanger between the primary and secondary sodium loops outside the reactor vessel. It was as close as possible to the basic design imagined for commercial applications (molten-sodium coolant, reactor material, power density, etc).The reactor operated until April 1983, when it was shut down permanently.

In February 1968, when RAPSODIE had been operating for one year, excavation work began at Marcoule for the construction of the 250 MWe (563 MWth) Phénix fast reactor. Phénix was connected to the grid on 13 December 1973 and generated power from that moment up to early 2009. It was a pool type reactor as well as Superphénix, the first commercial prototype SFR built in 1985, which had an output of 1240 MWe (3000 MWth).

Problems arose during its operation which was halted in May 1987 with the discovery of a major sodium leak in the fuel transfer tank or storage drum and in

13 1990 because of structural damages occurred following heavy snowfall. The plant was connected to EDF (Electricité De France) grid in December 1994 producing 4300 GWh of electricity, during the next 10 months of operation.

In September 1998, the plant was closed. Two incidents earlier in the year had culminated in a third, which triggered an automatic shutdown. This closure after a rather short operation period during 13 years but also due to political reasons set back developments. Research work on the 1450 MWe European Fast Breeder Reactor almost ceased [1].

The Soviet Union played a big role in developing these SFRs. In 1972, BR-10 was built based on the modifications on the reactor BR-5. This modifications allowed to get 10 MWth output with a new PuO2 core.

A higher power special fuel-testing reactor BOR-60, was designed and constructed in the Institute of Atomic Reactors (Dimitrovgrad) in five years and began operating in 1969. It is still operational and it represented the predecessor of BN-350 prototype SFR, which started to operate in 1973. It generated power in Kazakhstan for 27 years up to 1999 and about half of its output of 1000 MWth was used for water desalination. It used uranium enriched to 17-26%. Its design life was 20 years and after 1993 it operated on the basis of annual licence renewal. Both BOR-60 and BN-350 were loop type reactors.

BN-600 (600 MWe), a pool-type reactor built in Beloyarsk in 1980, is the only operational SFR all over the world that currently (2011) supplies electricity to the grid.

It is fuelled by uranium oxide fuel, enriched to 17, 21 and 26% (recently also MOX has been used) and it is equipped with heat exchangers for secondary coolant inside a pool of sodium around the reactor vessel and 3 steam generators outside the pool.

The sodium coolant delivers 525-550°C at little mor e than atmospheric pressure. Russia plans to reconfigure the BN-600 by replacing the fertile blanket around the core with steel reflector assemblies to burn the plutonium from its military stockpiles and to extend its life beyond the 30- year design span [1,3] .

A significant part of Japanese energy policy has been devoted to develop SFRs in order to improve uranium utilization dramatically.

From 1961 to 1994 there was a strong commitment to SFRs. Japan's JOYO experimental loop type reactor (140 MWth) has been operating since 1977 but without producing electricity. MONJU, a loop type demonstration reactor at Tsuruga, started in April 1994 but a sodium leakage in its secondary heat transfer system during performance tests in 1995 caused it to be shut down for almost 15 years. Its oversight passed to JNC, and a Supreme Court decision in May 2005 cleared the way for restarting it in 2008, but this was delayed and it restarted in May 2010. It has three coolant loops, uses MOX fuel, and produces 714 MWth, (246 MWe) [1].

India has been developing a Fast Breeder Reactors research program since 1950. At the Indira Gandhi Centre for Atomic Research a 40 MWth fast breeder test reactor (FBTR) has been operating since 1985. In addition, the tiny Kamini is devoted to explore the use of thorium as nuclear fuel, by breeding fissile U-233. In 2002 the regulatory authority issued approval to start construction of a 500 MWe prototype fast breeder reactor (PFBR) at Kalpakkam and this is now under construction by BHAVINI [1].

In China, R&D on fast neutron reactors started in 1964. Chinese Experimental Fast Reactor (CEFR), a 65 MWth (20 MWe) SFR was designed in

2003 and built near Beijing by Russia's OKBM Afrikantov in collaboration with OKB Gidropress, NIKIET and Kurchatov Institute. It achieved first criticality in July 2010 and will be connected to the grid by the end of 2011 [1].

Totally, 20 fast reactors operated from 1959 to 2009 gaining 390 reactor-years operating experience [9]. Sodium has represented the most widespread coolant for breeding reactors, despite its disadvantages that will be examined more in detail later in this chapter.

The updated overview of worldwide fast reactor experience is given in

Table 2.1.

15 Nevertheless, other coolants, such as lead and lead-bismuth eutectic (LBE), have been exploring since the late ’40s especially in IPPE (Institute for Physics and Power Engineering) in Obninsk, Soviet Union.

Lead and LBE, for reasons that will be explained more extensively later in this chapter, are (or appear to be) more attractive than sodium as fast reactor coolants. In fact, they do not practically moderate neutrons and do not react with air and water as sodium does.

At the same time, however, they are significantly corrosive to steel, with corrosion properties depending on the oxygen content.

About 80 reactor-years experience on these coolants has been gained building, in the ‘70s, nuclear submarines known as ‘Alfa class’, which represented a unique design among submarines.

Two types of Lead-bismuth Fast Reactor (LFR) were used in these submarines, the OK-550 and BM-40A both capable of producing 155MWth. The use of LBE allowed at reducing reactor’s size and therefore the overall submarine size with respect to conventional designs [12]. Based on the design of the two reactor models used in Alfa submarines, a joint venture between Rosatom and En+ Group called AKME Engineering decided to build a commercial 100 MWth bismuth reactor, SVBR-100 ('Svintsovo-Vismutovyi Bystryi Reaktor', lead-bismuth fast reactor) [13].

During the past decade, lead was considered as coolant in the design of the BREST type reactors (BREST 300 and BREST 1200) [14].

Unluckily, after the oil shock of the ‘70s and the consequent increase of the uranium spot price, the amount of uranium proved to be much more abundant than originally imagined and, after a fast start, nuclear power growth slowed dramatically in the late 1980s and nowadays global nuclear capacity is about one-tenth of the level that had been projected in the early 1970s.

The urgency of deploying fast neutron reactors for plutonium breeding has, therefore, considerably reduced at least in the western Organization for Economic Cooperation and Development (OECD) countries.

However, countries like India or China are developing very fast and, currently, they are building their electricity capacity up considering nuclear energy, too.

Generally speaking, all the developing countries are increasing their demand of energy (in particular of electricity) This implies, on the one hand, the availability of energy resources that might supply the demand and, on the other hand, to preserve the environment from climate changes, air pollution and, more generally, from greenhouse effect. For this reason, nuclear energy can play again a crucial role in future energy supplies.

The GIF, as mentioned in Chapter 1, has proposed the roadmap from current nuclear systems to Generation IV systems [15] in order to enhance the future contribution and benefits of nuclear energy utilization. The R&D of the new nuclear technologies based on the lines given by GIF is focused on increasing the use of fast neutron reactors because they allow for a more rational and efficient use of the fuel as a consequence of high conversion ratios. In addition, a real fuel optimization implies the reduction of the nuclear wastes. Thus, this objective resulted in studies concerning the “Partitioning and Transmutation”, that can be considered as the key technology to reduce the environmental burden of the HLW (High Level Waste) through the burning the Minor Actinides (MA) and reducing

repository needs.

In order to fulfil this goal, research efforts have been focused on the development of the ADS (Accelerator Driven System), as explained in Chapter 1. Theoretically, all the reactors types might burn MA but actually the transmutation of the MA in thermal reactors is not acceptable from an economic point of view because of the MA capability to act as a poison for thermal neutron. Therefore, fast neutron systems represent the most suitable choice.

Consequently, the choice of the coolant is focused on liquid metals again, because as demonstrated in the past, they allow to work at low pressure thus reducing the likelihood of LOCA. Currently, the selection criteria for the use of liquid metals as heat-transfer media in a nuclear environment may be summarized as in the following.

• Neutronics related to the fast spectrum necessary for breeding, fuel conversion and MA transmutation in the next generation fast reactors and ADS concepts. In this respect, the coolant features should be:

o small capture cross-section for reducing small parasitic loss of neutrons;

o high scattering cross-section for avoiding leakage of neutrons from the core;

o small energy loss per collision for avoiding spectrum softening and moderating effect;

o high boiling temperature for preventing reactivity effects from boiling and the related coolant voiding.

• Materials:

o acceptable corrosion and mechanical degradation of structural and containment materials and lifetime of equipment;

o high stability of the liquid metal (e.g., limited chemical reactions with secondary coolants and air or formation of spallation products, etc.).

• Thermal-hydraulics:

o moderate power requirements for circulating the liquid metal;

o high heat transfer coefficient and small size of heat exchanger;

o large margin between coolant outlet temperature and boiling temperature, thus virtually eliminating safety concerns related to coolant boiling.

• Safety:

o controllable chemical and radioactive hazards;

o simple and reliable safety measures and systems.

• Economics.

2.3 Characteristics of liquid metals

As mentioned before, LMFRs (liquid metal cooled fast reactor) have been under development for more than 50 years. In some cases the overall experience with this kind of reactor has been extremely fruitful, especially for gaining experience and knowledge on material’s behaviour under irradiation, components and compatibility between structure material and coolant, e.g. depending on its oxygen concentration, showing good performance also with respect to design expectations.

Liquid metal coolants have also been shown to have very attractive safety characteristics, resulting to a large extent from being low pressure system with large thermal inertia and negative power and temperature reactivity coefficients.

17 suitable compromise from the view point of neutronics, costs, availability and cooling charateristics.

The main features of sodium are summarized in Table 2.2 - Table 2.4 and in [16,18].

Table 2.2. Sodium isotope characteristics [16]

Table 2.4. Thermophysical properties of sodium at different temperatures [16]

This coolant has been chosen for LMFRs because of several reasons. Firstly, it is among the most abundant elements, with its content on the earth’s crust of about 2.5 wt.%. The most important sodium compound, produced in the amount of millions of tons, is NaCl and its content in seawater is about 3%. Moreover, sodium can be produced by high-temperature reactions from almost any its compound (NaCl, NaOH, Na2CO3 and Na2S). In addition, it has good thermal and physical properties. In fact, sodium offers the possibility to achieve high specific power densities in the reactor core (about 500 kWth/l for MOX fuel) resulting in a short fuel doubling time. This characteristic parameter of a reactor design was considered an essential issue in the ‘70s.

Finally, the need of a weakly moderating material with good heat transfer properties may be satisfied. This matched the world-wide strategic line of fast reactor development in the 1960s and it has continued in some countries nowadays, too. For this very reason, when various liquid metal coolants for FBR were considered in the 50s, sodium was given preference.

As can be seen in Table 2.2 sodium is highly reactive with air and water, therefore, this might represent a big problem for reactors. Its chemical peculiarities are shown in the following.

Sodium is the most electropositive metal and displaces H2 out of water with production of hydroxide (NaOH). The formation of sodium hydride (NaH) is due to sodium interaction with dry hydrogen, which is soluble in sodium.

2Na(s) + H2(g) → 2NaH(s) + 57.4 kJ/mole

At the temperature of 420 oC, NaH dissociates with release of hydrogen. When sodium interacts with the small amount of oxygen, Na2O oxide is produced, whereas its burning in the air results in Na2O2 peroxide:

4Na(s) + O2(g) → 2Na2O(s) + 416 kJ/mole 4Na(l) + O2(g) → 2Na2O(s) + 436 kJ/mole 2Na(s) + O2(g) → Na2O2(s) + 499 kJ/mol

1

In the molten sodium, only the Na2O oxide is stable, while Na2O2 dissociates as follows

1

The indexes (s), (l), (g) represent solid, liquid and gas states, respectively. In some equations these indexes are missing because they are not available in the reference text [16]. In addition, some minor mistakes in the equations included in [16] have been corrected.

19 Na2O2 + 2Na → 2Na2O

When sodium interacts with water, some reactions occur and, depending on reaction temperature, impurities of different composition would be present in sodium. Sodium-water chemical interaction proceeds in two stages. During the first stage the reaction proceeds at a high rate with release of gaseous hydrogen and heat.

Hydroxide is produced by the following reaction at 201–300 oC 2Na (s) + 2H2O (l) → 2NaOH (aq) +H2 (g)+142 kJ/mole

This reaction is practically irreversible because of a very high hydrogen equilibrium pressure.

During the second stage a chemical interaction takes place between the first stage’s products of the reaction and the sodium excess, as it may happen for instance in case of water leak in the steam generator.

Oxide as well as hydroxide are produced at 350–400oC 2Na +NaOH → Na2O + NaH

2Na + H2 → 2NaH

For temperatures greater than 420 oC, the reaction with steam is 2Na(s) + H2O (g) → Na2O (s) + H2 (g)

The burning reaction is characterized by a zone of small flames at the sodium-air interface, by the formation of Na2O on the sodium surface and a vigorous emission of high density white oxide fumes. Therefore, the use of sodium as a coolant poses fire danger in case of leakage and interaction with air or water.

However, this burning causes relatively low heat release, namely 420 kJ/mole by Na2O or 500 kJ/mole by Na2O2. This is equivalent to about 10 kJ (~2 kcal) per gram of sodium burnt. Considering the same fuel volume, sodium burning results in the energy release equal to 50% of that for sulphur, 30% of that for gas/oil, 25% of that for magnesium and slightly over 10% of that for aluminum .

Another sodium feature is its radioactivity due to the neutron capture reaction (see also Table 2.1). The natural isotope of sodium is 23Na (abundance 100%). Neutron capture processes in sodium [(n,γ) reaction resulting in 1.4 MeV γ -quanta emission] lead to formation of 24Na isotope with half-life time of 15 hours, while sodium is flowing through the core. Besides, there is (n, 2n) threshold reaction producing 22Na with 2.6 years half-life time. 22Na emits 1.3 MeV γ-quanta; its activity is proportional to thermal power of the reactor plant. Neutrons are practically not generated by sodium radionuclides with half-life time exceeding 2.6 years value, even after 50 years of exposition to intensive fast neutron flux.

24

Na is the main isotope giving rise to requirement of protection against γ -radiation. Approximately 10 days after reactor shutdown the primary circuit activity is mainly determined by 22Na. This feature, together with the fact that sodium interacts chemically with water and air results in three coolant loops in NPP design, including:

• a primary loop containing radioactive sodium heated up in the core;

• an intermediate non-radioactive sodium loop coupled with the primary loop through the intermediate heat exchanger;

Within the primary circuit radioactive coolant is protected against air by steel barriers and cover argon gas. Radioactive sodium of the primary circuit is separated from nonradioactive sodium of the secondary circuit by the steel tubes of the intermediate heat exchangers. Operational experience gained on LMFRs, such as BN-350, Phénix, PFR, BN-600, etc., has proven that sodium is practically non-corrosive to stainless steel, and content of impurities, mainly oxygen and carbon, are maintained at acceptably low level by the cold traps installed in the bypass of the main coolant circuit.

The amount of the long lived radioactivity generated in sodium by neutrons is negligible as also Usanov claimed in [17] “It is found that the radiation properties

of the coolants differ substantially, which could be decisive for handling them as radioactive wastes. It is concluded that sodium is closest to a radiationally waste-free technology”.

Activation of sodium reaches equilibrium state in about ten years of the first cycle of its use and will never exceed this level. Radioactive hazardous isotopes (cesium, tritium, strontium and iodine) are retained by sodium.

The primary sodium activity in the reactor under operation, mainly determined by 24Na (from BN-350 reactor, Kazakhstan) is about 10 Ci per kilogram of sodium [19] while after the reactor has been shut down for decommissioning residual activity of 22Na is ~1⋅10-4 Ci per kilogram of sodium [19].

The long-lived radionuclides produced by fission products, sodium impurities and corrosion activation products are chemical elements that are chemically not alike but rather dissimilar to sodium that makes possible its external decontamination at the reactor plant decommissioning stage.

If any component of the primary sodium circuit is to be removed from the reactor for the purpose of repair/maintenance, sodium stuck to their surface must be removed because of the chemical reaction of sodium with oxygen and moisture in air and of the radioactivity of sodium.

The cover gas is another source of LMFR radioactivity. Primary gas activity is to a considerable extent determined by the impurities in sodium and activation of 40

Ar and 41Ar. As a result of (n, p) reaction, radioactive 23Ne with short half-life of 38 s is produced from 23Na [19,20].

Xe and Kr isotopes are other gaseous fission products that might be present in the cover gas, although only traces of 133Xe isotope produced can be considered because it has a relatively short halflife of 5.27 days. Other fission product noble gases generated are either stable or leave relatively short halflives.

In any case, it is necessary to purify argon eliminating impermissible release to the environment of these isotopes.

The problem of Xe and Kr isotopes separation from the atmosphere within a reactor containment dome, which is of interest also for thepurification of the off-gases from plants for reprocessing spent nuclear fuels, has been solved using a two step process that, firstly, let the gaseous effluents flow through a condenser and then let the remaining aerosols remain trapped in high efficiency filters. 85Kr (halflife = 10.8 years) can be stored under pressure in special-purpose cylinders cooled by air.

Nevertheless, some drawbacks of sodium technology have emerged during the operation of several plants. In particular, the problem of the sodium reactivity with water and air and the related problem of sodium fire resulted from leakages in the reactor components (pipes, valves, vessel, heat exchangers and pumps) mainly during the reactor startup, day-to-day operation, and refueling and

21 maintenance [21]. In order to minimize the consequences of such accidents, a more complicated layout of the NPPs has been studied, thus contributing to the delay of fast reactors commercial introduction. This has lead to investigate on the use of others liquid metal as coolant.

At first, the choice has been focused on LBE, mainly because of the experience gained by the Russian Federation organizations.

The thermo-hydraulic features of LBE and lead coolants are high boiling temperatures and the relative inert chemical behaviour compared with sodium (see

Table 2.5 and Table 2.8 [22]). The boiling points of both are well above cladding

failure temperatures. High boiling temperature eliminates pressurization and boiling concerns and enhances inherent safety of reactor cores. Higher allowable operating temperatures improve efficiency and feasibility of other energy products. For LBE, the relatively low melting point reduces the risk of uncontrolled freezing. High density and wider range of possible operating temperature offer increased design space for passive safety.

Lead and LBE specific heats per unit volume are similar to sodium but the conductivities are about four times smaller. The γ-radioactivity induced in both coolants is low enough to allow the accessibility to the coolant circuit after a 24 hours shutdown period. Other potentially favourable features of HLM are represented by

• lower reactivity associated with hypothetical voiding of the coolant;

• better shielding against gamma rays and energetic neutrons;

• high solubility of the actinides in the coolant, which could help to minimize the potential for re-criticality events upon core melting, and no energetic reaction with air and water, thereby eliminating the possibility of fires.

With respect to spallation neutron sources in ADS systems, there is a general consensus that above 1 MW of beam power, solid targets are hardly feasible from a heat removal point of view. Therefore, liquid metals targets are the best choice (see e.g. [23]); among the liquid metals lead-alloy-based liquid metal targets are to be preferred if high operating temperatures are required.

In spite of all these features, LBE presents some drawbacks principally due to the amount of bismuth in the alloy. Bismuth has the important shortcoming of high costs and limited availability. In fact, the bismuth content in the earth’s crust is about 8.5·10-3 mg/kg and 2·10-5 mg/l in earth’s oceans [24]. The deposits are few and far each other and the recovery from the bismuth bearing ores is a very complicated multistage process. Therefore, on this basis, suspicions could be justified that the use of bismuth might be restricted to a limited number of reactors.

Furthermore, irradiated bismuth produces α-radioactive 210Po, which has a high volatility. This poses problems because of its migration from the coolant to the cover gas and formation of aerosols; any leakage from the cover gas poses in turn hazards to the plant operators and the environment. Therefore polonium activity is one of the important operational problems in case of lead-bismuth used as a coolant.

Some studies [25,26] performed in the Russian Federation have pointed out that, in the reactors cooled by lead-bismuth, the polonium activity is determined by reaction on 209Bi isotope st

Pb

d

Po

d

Bi

n

Bi

8

.

138

0

.

5

)

,

(

γ

β

α

and equilibrium activity is ~10 Ci/kg. A small amount of Po209results from the reaction Po210(n,2n)Po209.

It was found [26] that the specific α -activity of the LBE is defined by the Bi210 (half-life equal to 3.6⋅106 years), which is generated by the reaction Bi209(n,γ) Bi210m. The long-lived β-activity of Bi208 (half-life equal to 3.65⋅105 y) is generated in the reaction Bi209 (n, 2n) Bi208. It may reach a value of 5·10-4 Ci/kg by the end of the reactor service life.

Furthermore, it has been noted that such lead-bismuth coolant activity gives rise to problems even under normal operating conditions. In fact, in case of a cover gas leak rate of 0.01% of its volume per day, the release of 210Po to the central hall (if gas circuit is not cleaned from polonium) may cause a 200 fold excess of the maximum permissible concentration (mpc) [6]. To ensure that the mpc value is not exceeded for people in the central hall, it is necessary to comply with very strong requirements for the reactor over gas circuit leak tightness.

Experience gained on LOCA studies [26, 27] has shown that application of protective coatings allows polonium’s confinement preventing its release to the environment. This coating effect is based on sorption and dissolution of impurities in a dispersed medium, like a charcoal bed or even a cryogenic trap [1,22], and fixing in the coating. In case of steam generator tube rupture (SGTR) and depressurization of the secondary circuit, LBE can enter the secondary circuit and water can be contaminated by polonium. In this case, a basic amount of polonium is rfretained in the alloy, a condensate saturation activity of 10-104 Bq/kg is reached and SG inner surfaces become contaminated owing to polonium sorption from the water. Water evaporation determines unacceptable radioactivity levels in the turbine hall.

Contamination of the inner surface of the secondary circuit constitutes a danger in case of equipment repair, since the alloy kept in the secondary circuit is a permanent source of water contamination. Water replacement and SG inner surface decontamination without any alloy removal would not give the desired result. The secondary circuit decontamination turns out to be possible only after complete removal of the alloy from the secondary circuit. Obviously, this cannot be assured; if SG tubes failures are assumed possible, the alloy penetration to the secondary circuit should not be neglected.

Tupper’s experiments, focused on Po evaporation from LBE [28], have shown that LBE evaporation rate is within the limits given by the Raoult law. In addition, Feuerstein [29] has shown that within the temperature range between 300 and 800 oC an inert gas atmosphere (Ar, Ne) may reduce evaporation rate of Po more than 103 times.

Po may be evaporated from LBE in the form of PbPo or BiPo. The rate of evaporation into vacuum follows the Lenghmure law

T

M

kP

G

=

/

where P = pressure, k = 0.0044, M = molecular mass, T = temperature [K]. If Po is in solution, the evaporation rate is proportional to mole part “m”: G = mGo.

The saturation pressure for polonide vapour, that is chemical compounds of the radioactive element polonium with any other element except halogens, follows the relationship

)

/

7270

(

94

.

6

log

P

=

−

T

23 Calculations from these relationships give a polonium yield rate equal to 6.7⋅105 Bq/kg for LBE with activity of 1.85 1013 Bq/l at 300oC.

Aiming at reducing the problems related to polonium lead has been proposed instead of lead-bismuth because of its similarity with LBE, as it can be observed comparing Table 2.5 and Table 2.8 where the main correlations of thermophysical properties of lead and LBE are summarised.

In similarity with LBE, pure lead cooling has been firstly studied in the already quoted IPPE and then in Europe and USA as well.

One of the arguments in favor of using lead instead of sodium has been related to severe accidents. Assuming that damaged fuel elements still keep their integrity, and no radioactivity leakage occurs from the fuel, studies carried out at Research and Development Institute of Power Engineering (RDIPE) in Russia has demonstrated that radioactivity release in SFR would be 3 to 4 orders higher than that in lead cooled reactor because of sodium burning. However, if the fuel element failures take place (and this seems to be the most probable event) coolant radioactivity becomes less important, and fission products would contribute essentially.

Another important aspect to be taken into account are costs. Lead is much cheaper than LBE because of its higher availability. In fact, lead content in the earth’s crust is about 1.4·10-1mg/kg and 3·10-5 mg/l in earth’s oceans [30]. The most important lead minerals are: PbS (86.6% Pb), 3PbS·Sb2S3 (58.8% Pb), PbCO3 (77.55% Pb), 2PbS·Cu2S·Sb2S3 (42.4% Pb) and PbCl2·PbCO3 (76.0% Pb). Only sulphide ores (PbSO4, PbS) are reprocessed in the lead-zinc industry using or the hydrometallurgical or the pyrometallurgical way, which are the preferred ones [27, 31]. Nevertheless, also lead presents some disadvantages (see also Table

2.9).

First of all, control and maintenance of coolant quality, its compatibility with structural materials and their corrosion strength should be studied to a greater extent at higher temperatures. A careful control of the purity of the coolant is required to avoid the formation of deposits. This problem has been particularly investigated since the early stages of development of the design of the HLM reactors.

In addition, it is necessary to develop corrosion resistant steels, pre-treatments of the component surfaces and special inhibitors in heavy liquid metals coolant. More extensive studies are required for lead coolant to demonstrate the corrosion-resistance of structural material.

Another drawback associated with the use of lead as a coolant (as well as LBE) is the potential complexity of in-service inspection and repair because its high density.

Furthermore, although Po activity in lead is about 1000 times lower than that in LBE, polonium yield from lead is approximately the same due to higher temperature.

Tupper’s experiments [28] have highlighted that lead in air may produce the same danger for health as 210Po does. In fact, as can be seen below, the problem of polonium contamination exists also for lead owing to Bi209 formation from neutron capture in Pb208, which represents almost the 52.3% of naturally-occurring lead [16] st

Pb

d

Po

d

Bi

n

Bi

d

Pb

n

Pb

8

.

138

0

.

5

)

,

(

14

.

0

)

,

(

γ

β

γ

β

α

Calculations performed with equations 2.1 and 2.2 have given a polonium yield rate equal to 1.15⋅106 Bq/kg for Pb with activity of 1.85 1010 Bq/l at 500 oC.

Concerning its specific long-lived residual radioactivity the most important contributor of lead is Pb205 (half-life equal to 1.51⋅107 y) which is generated in the reaction Pb204(n, γ)Pb 205. The specific β-activity of a pure lead coolant is significantly less than for LBE. It must be noted that activation of lead coolant (and LBE also) increases in every cycle of use, if re-use is possible in principle [32].In addition, after decommissioning of lead cooled nuclear power plant the coolant most probably has to be treated as high level waste.

Another of the most important features of lead to be taken into account is its toxicity, which can cause poisoning. Stable and radioactive lead’s isotopes, such as Pb210, Pb212 and Pb214, may be released to the atmosphere because of industrial use. Therefore, lead may be accumulated in the human body and expelled slowly together with the living activity products. It attacks the nervous system, marrow, blood and the vascular system, disturbing albumen synthesis and genetic structure of cells when it is introduced in the human body. Because of this, its maximum permissible concentration values are 0.003 mg/m3 and 0.03 mg/l respectively in air and water.

In normal reactor operation lead is kept within the gas-tight circuit. Enhancement of lead concentration in working rooms is possible as a result of accidents causing leaks through the reactor upper plate or its disassembly.

Table 2.5. Recommended correlations for main thermophysical properties of molten LBE (p~0.1 atm) [22]

25

Table 2.6. Stable isotope of lead [16]

Table 2.7. Unstable isotope of lead [16]

Table 2.8. Recommended correlations for main thermophysical properties of molten LBE (p~0.1 atm) [22]

A comparison of the basic characteristics of liquid metal reactor coolants has been made on the basis of the previous overview of the main features of sodium, LBE and lead and it has been summarized in Table 2.9.

Coolant Advantages Disadvantages

Sodium

• best thermohydraulic characteristics

• low reactor vessel pressure (close to atmospheric pressure)

• good breeding characteristics

• non-corrosive to stainless steel and any fuel compositions

• decay heat removal by passive means

• low density, preventing fuel SA from floating up, and allowing passive safety rod operation (by gravity force); all the ‘dirt’ resulted from the assembly and operation is precipitated in the stagnant zones

• retention of cesium, strontium, iodine and tritium hazardous isotopes by sodium

• no generation of radionuclides with a half life more than 2.6 years (Na22₎

from neutrons

• chemical activity with respect to oxygen and water with explosion potential

• high γ-activity of Na24 needs of intermediate circuit

• needs of heaters aiming at maintaining liquid phase of coolant (Na freezing temperature about 100 ºC, temperature for refueling/repairing between 150 – 200 ºC)

• problems with sodium removal and disposal during reactor plant operation and decommissioning

Lead, LBE

• lack of high γ-activity

• chemical interaction with water and air without explosion

• low moderation and absorption of neutrons

• good reflecting properties

• good neutronic performance

• reactor decay heat removal by passive means

• high boiling temperature, coolant void reactivity is negative

• production of α-radioactive, volatile polonium, any leakage from the cover gas poses a hazard to operators and environment (in particular LBE)

• high corrosive to steels and some fuel compositions

• high density, probability of core pollution by suspensions (assembling and operating ‘dirt’) in heavy coolants

• high freezing temperature(327 ºC for lead)

• high residual activity of coolant (T1/2 –1.5⋅10

7

y of nuclide Pb205 and 3.6⋅106 y for Bi210)

• formation of products from water - heavy metal interaction (SG leaks) which might block flow channels of the core two circuit design

• risk of leak-before-break because of corrosion damage

• problems with radioactive waste management and coolant disposal during and after decommissioning

Table 2.9. Advantages and disadvantages of using several liquid metals

27

2.4 Ongoing R&D activities

Several R&D activities are being carried out all over the world aiming at developing reliable liquid metal systems.

Referring to Europe, studies concerning SFR and LFR are ongoing with comparable timing, as can be illustrated in Fig. 2.2.

Fig. 2.2. European industrial initiative of the plan dedicated to the demonstration of Gen IV liquid metal technologies [32]

With reference to sodium cooled fast reactors, ESFR is the ongoing project of the EU under the aegis of the 7th Framework Program [33].

The ESFR uses liquid sodium as the reactor coolant, allowing high power density with low coolant volume. It builds on more than 300 reactor-years experienced with sodium-cooled fast neutron reactors over five decades and in eight countries and it represents the main technology of interest in GIF.

Most plants so far have had a core-plus-blanket configuration, but new designs are studying also the possibility of having a mixture of fissile-fertile fuel in the core. Additional R&D efforts are focused on safety in loss of coolant scenarios and improved fuel handling.

Three variants of reactor concepts are proposed: 1) a 50-150 MWe type with actinides incorporated into a U-Pu metal fuel, requiring electrometallurgical processing (pyroprocessing) integrated on site; 2) a 300-1500 MWe pool-type version of this reactor, and 3) a 600-1500 MWe type with conventional MOX fuel and advanced aqueous reprocessing in central facilities elsewhere.

At the same time, the French Atomic Energy Commission (CEA) has stressed on the high priority of developing ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), which should be built at Marcoule.

ASTRID should have an output ranging from 250 to 600 MWe as prototype of a commercial series which is likely to be deployed from about 2050.

It will have high fuel burnup, including minor actinides in the fuel elements in order to work as actinide-burner, too. It will be equipped with an intermediate sodium loop, while the choice of tertiary coolant (water/steam or gas) is still under debate. Four independent heat exchanger loops are likely and it will be designed to reduce

the probability and consequences of severe accidents to an extent that is not now done with previous SFRs.

ASTRID is called a "self-generating" fast reactor rather than a “breeder” in order to demonstrate low net plutonium production. This is aimed at meeting the stringent criteria of the Generation IV International Forum in terms of safety, economy and proliferation resistance [8].

In Japan, on the one hand, the 280 MWe (714 MWth) MONJU demonstration SFR at Tsuruga restarted in May 2010, as mentioned before. On the other hand, Mitsubishi Heavy Industries is involved in a consortium in order to build the JSFR (Japan Standard Fast Reactor) concept, with breeding ratio less than 1. Its size should range from 500 to 1500 MWe and it should burn actinides with uranium and plutonium in oxide fuel [8].

In any case, new demonstration breeder reactors are being built in India and Russia and China.

CIAE (China Institute of Atomic Energy) program has established as a priority the construction of 1000 Chinese Demonstrator Fast Reactor (CDFR) starting from 2017, while in October 2009 it signed an agreement with Russia's Atomstroy-export to begin design works for a commercial nuclear power plant with two BN-800 reactors.

The construction should start in 2013 and commissioning should be around 2018-19. This reactor would be similar to the OKBM Afrikantov design being built at Beloyarsk 4 and due to start up in 2012 [1,8].

Construction is well advanced on Beloyarsk-4 which is the first BN-800 from OKBM Afrikantov. This reactor represents a new, more powerful (880 MWe) FBR, based on the BN-600 design.

Several features are improved including flexibility in using different fuel (U+Pu nitride, MOX, or metal), safety and reducing costs that are expected to be only 15% more than VVER.

The Pu burning capability is up to 2 ton per year from dismantled weapons. Furthermore, it will test the recycling of minor actinides in the fuel. Russia expected to have 40 tonnes of separated plutonium stockpiled by 2010, and after some furnishes the initial core load, the rest was expected to be burned in the BN-800 by 2025.

Regarding lead-cooled fast reactors, flexible fast neutron reactors are envisaged which may use depleted uranium or thorium fuel matrices and may also burn actinides from LWR fuel. Liquid metal (Pb or Pb-Bi eutectic) cooling is at low pressure by natural convection (at least for decay heat removal). The international R&D programs envisage a wide range of unit sizes ranging from a factory-built "battery" with 15-20 year life for small grids or developing countries, to modular 300-400 MWe units and large single plants of 1400 MWe.

These large size plants should be similar to Russia's BREST fast reactor design [1,8,14], which is lead-cooled and builds on a 80 reactor-years experience of lead or lead-bismuth cooling, mostly in submarine reactors. Its fuel is made of U+Pu nitride. No weapons-grade plutonium may be produced (since there is no uranium blanket), and spent fuel can be recycled indefinitely through on-site facilities. A pilot unit of 1200 MWe is planned at Beloyarsk.

Also US are involved in developing LFRs. The main project is represented by SSTAR (Small Secure Transportable Autonomous Reactor), which is being developed by a joint venture between USA and Japan. It should have an output of 20 MWe and should be equipped with an integral steam generator inside the sealed

29 unit, which would be installed below ground level. The expected thermal efficiency is 44% and one future target is the coupling with a Brayton cycle turbine using supercritical carbon dioxide with natural circulation to four heat exchangers [8,35].

In Europe, European Lead-cooled SYstem (ELSY) of 600 MWe is the project [36] led by Ansaldo Nucleare from Italy and is being financed by Euratom in the frame of 6th FP. It runs on MOX fuel and the molten lead (inlet temperature 480 °C, outlet temperature 400 °C) is pumped to eight s team generators. A pilot plant is planned in operation by 2020, followed by a prototype of a large unit and deployment of small transportable units.

In the frame of 7th EU FP, LEADER (Lead cooled European Advanced DEmonstration Reactor) can be seen as the continuation of the ELSY project. In fact, starting from the analysis of the hard points of the ELSY design, the LEADER project is aimed at defining ,at first, a conceptual design of a demonstrator facility (size of about 100 MWth) and then at developing a LFR reference plant.

At the same time ESNII (European Sustainable Nuclear Industrial Initiative) is also going to include an LFR technology demonstrator known as ALFRED, also about 100 MWth, which is seen as a prelude to an industrial demonstration unit of

about 600 MWe. Construction on ALFRED might begin in 2017 and the unit might

start operating in 2025.

Concerning LBE technology [8], in Europe, Belgium's SCK.CEN is planning to build the MYRRHA (Multipurpose Hybrid Research Reactor for High-tech Applications) research reactor at Mol. Initially it will be a 57 MWth ADS, consisting of a proton accelerator delivering a 600 MeV, 2.5 mA (or 350 MeV, 5 mA) proton beam to a LBE spallation target that in turn couples to a Pb-Bi cooled, subcritical fast nuclear core. This plant is aimed at being a multipurpose plant. In fact, as an ADS it will be used to prove that technology and to study transmutation of long-lived radionuclides and then it is intended to be run as a critical fast neutron facility, decoupling the accelerator and removing the spallation loop from the reactor core. Then MYRRHA will be used for fuel research, for materials research for Generation IV reactors, and for the production of radioisotopes and doped silicon (an essential component of high-grade electronic circuits).

In Russia, a smaller and newer design is represented by SVBR (Lead-Bismuth Fast Reactor) of 75-100 MWe. This reactor has an integral design, with the steam generators placed in the LBE pool at 400-495°C where also the reactor core, which might use a wide variety of fuels, is placed. A power station with 16 such modules is expected to supply electricity at lower cost than any other new Russian technology as well as achieving inherent safety and high proliferation resistance [8].

Japan's LSPR (LBE-Cooled Long-Life Safe Simple Small Portable Proliferation-Resistant Reactor) is a lead-bismuth cooled reactor design of 150 MWth (53 MWe. Fuelled units would be supplied from a factory and operate for 30 years, then be returned. This concept is intended for developing countries.

In addition, a joint venture among Toshiba, CRIEPI and STAR (USA) is designing a version of the already quoted SSTAR cooled by LBE, which is named L-4S [8].

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