The Risks of Nuclear Energy Technology: Safety Concepts of Light Water Reactors

Part I

The Physical and Technical Safety Concept of Light Water Reactors

© Springer-Verlag Berlin Heidelberg 2014

Günter Kessler, Anke Veser, Franz-Hermann Schlüter, Wolfgang Raskob, Claudia Landman and Jürgen Päsler-SauerThe Risks of Nuclear Energy TechnologyScience Policy Reports10.1007/978-3-642-55116-1_1

1. Introduction

Günter Kessler¹  and Anke Veser²

(1)

Stutensee, Germany

(2)

Eggenstein, Germany

Abstract

This chapter lists the capacity of commercial nuclear power plants built and operated in different countries of the world in 2013. About 80 % of all operating nuclear power plants are Light Water Reactors (LWRs), predominantly Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). An additional 11 % are Heavy Water Reactors (HWRs) and 4 % are advanced gas cooled, graphite moderated nuclear power reactors (AGRs). Only about 3.4 % are Russian retrofitted RBMK1000 reactors still operating in Russia. One prototype Fast Breeder Reactor (FBR) was operating in Russia, one became operational in India and one experimental FBR was operated in Japan.

The resources of natural uranium were assessed in 2007 by IAEA and OECD/NEA to be 5.47 million tons (reasonably assured and inferred). An additional 7.77 million tons of speculative and about 4.2 million tons in the Chattanooga Shales in the USA are listed.

These uranium resources are then contrasted with the uranium consumption of each nuclear power reactor which is 171 tons per GW(e) and year for LWRs. If plutonium recycling in a closed fuel cycle is applied this uranium consumption is reduced by a factor of 1.55. FBRs would consume only 1.7 tons of U-238 per GW(e) and year which would extend the time period for nuclear energy application (uranium and thorium resources) to thousands of years.

For LWRs and other commercial nuclear reactors the natural uranium must be enriched. This is done predominantly by the gaseous diffusion and the gas centrifuge process. The laser enrichment process (SILEX) is still under deployment in the USA. Commercial spent fuel reprocessing facilities were built and are operated in France, Great Britain, Russia and Japan. This reprocessing capacity in the world can reprocess the spent fuel of about half of the presently operating LWR capacities. The majority of nuclear power plants built and operated in the world today is used for electricity generation. Such nuclear power reactors are built in unit sizes of about 1 and 1.6 GW(e) and operated for economical reasons mainly in the so-called base load regime.

In April 2013 there were 433 nuclear power reactors with a total power capacity of about 370 GW(e) operating in the world (Fig. 1.1). These nuclear power reactors produced about 16 % of the world’s electrical energy consumption. Those countries having the highest number of nuclear power reactors installed and operating by 2013 are listed in Table 1.1. However, there were also many countries in Central- and South-America, in Africa, Asia, Australia and Europe which had not decided yet to rely on electricity generation by nuclear power reactors. In Western Europe, e.g. such countries are Portugal, Denmark, Norway, Italy, Austria etc. for different reasons [1].

Fig. 1.1

Map of nuclear power reactors and commercial nuclear reprocessing facilities operating in the world by 2012 [2] adapted. Red circle—Nuclear Power Reactors, Black square—Commercial nuclear reprocessing facilities

Table 1.1

Nuclear power reactor capacity installed in the world by 2013

aAll reactors in Japan, except for two PWRs, were under safety review before restart in early 2013

About 80 % of all operating nuclear power reactors are Light Water Reactors (LWRs); predominantly Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). An additional 11 % are Heavy Water Reactors (HWRs) and 4 % of all nuclear power reactors are advanced gas cooled, graphite moderated nuclear power reactors (AGRs). Only 11 RBMK-1000 reactors (Chernobyl-type, graphite moderated light water cooled), i.e. 3.4 % are still operating near St. Petersburg, Smolensk, and Kursk (Russia). However, this type of nuclear power reactor will be taken out of operation in the near future [1].

One prototype power reactor of the future Fast Breeder Reactor (FBR) type was also operating in Russia and one experimental Fast Breeder operated in Japan.

In addition to these 433 nuclear power reactors presently in operation there are 103 nuclear power reactors with a power capacity of 103 GW(e) under construction in the USA (10), France (1), Belarus (2), Slovakia (2), Finland (1), Russia (11), Ukraine (3), Romania (2), India (7), China (42), Taiwan (2), Pakistan (2), South-Korea (5), Japan (3), Argentina (1), Brazil (1), United Arab. Emirates (4) and Turkey (4). Again 88 % of these nuclear power reactors are LWRs, predominantly of the PWR type. Modern LWRs have a yearly power availability factor of about 85–93 %. They are predominantly operated in the base load regime, but can also be operated in partial load. Especially in Russia they were and are also used for heat generation for district heating and desalination (BN-350) [1].

During the past decades nuclear power reactors were designed for an operation time of 35–40 years. Modern LWRs, however, are designed for an operation time of 60 years.

1.1 Uranium Resources

Natural uranium is found in uranium ores in concentrations from around fractions of a percent to several percent. Natural uranium can be bought on the world market from uranium resources and uranium mines in Australia, Canada, Kazakhstan, Niger, Namibia, Russia, Uzbekistan, USA and other countries. Natural uranium contains 0.7204 % of the isotope U-235, 99.2742 % of the isotope U-238 and 0.0054 % of the isotope U-234. For LWRs this natural uranium must be enriched in the isotope U-235 up to a concentration of 4–5 %.

The available uranium resources are assessed on a yearly basis by IAEA and OECD/NEA and listed in different categories. The uranium resources were assessed in 2007 by IAEA and OECD/NEA to be 5.47 million tons (reasonably assured and inferred). Reasonably assured means that these uranium resources can be mined, inferred means that additional investigations are required until the uranium ores can be mined. At the same time IAEA and OECD/NEA prognosticated additional speculative 7.77 million tons of uranium ores and further 4.2 million tons in the Chattanooga Shales in USA [3, 4].

1.2 Uranium Consumption

A present LWR with a power capacity of 1 GW(e) consumes about 171 tons of natural uranium (availability factor of 93 %) per year. This means, that e.g. about 370 GW(e) presently in operation (assumed all nuclear power reactors would be LWRs) will consume over 80 years about 5 million tons of natural uranium. Correspondingly a future 480 GW(e) nuclear power capacity (assumed all nuclear power reactors would be LWRs) would consume in 180 years about 15 million tons of natural uranium. Heavy Water Reactors or Light Water Reactors with plutonium recycling would have by a factor of 1.55 lesser natural uranium consumption and would extend the above time period correspondingly [5].

The fission neutrons originating from the fission process are moderated or slowed down by the collision with atoms of a moderator or coolant, e.g. light or heavy water, in the cores of LWRs and HWRs to so-called thermal energy of 0.025 eV. This corresponds to neutron velocities of 2,200 m/s. In liquid metal cooled Breeder Reactors the fission neutrons originating from the fission process are slowed down only to 0.2 MeV as the moderator or the coolant (sodium, lead or lead-bismuth) is of medium or high atomic mass. In this range of neutron energies of 0.2 MeV and higher the nuclear reactions for breeding of Pu-239 from U-238 are favorable. This newly generated Pu-239 can be utilized as artificial fissionable nuclear fuel in e.g. LWRs or FBRs.

Fast Breeder Reactors are started initially with uranium/plutonium fuel in their core and uranium fuel in their blankets. They consume per GW(e) and year only 1.7 tons of U-238 (either natural uranium or depleted uranium from uranium enrichment plants). The technical feasibility of sodium cooled Fast Breeder Reactors has been proven already in the USA, Russia, UK, France, India and Japan during the past decades. Fast Breeder Reactors require a closed fuel cycle with spent fuel reprocessing and fuel refabrication [5, 6].

This by a factor of about 100 lower fuel consumption (1.7 tons per year and GW(e) for Fast Breeder Reactors compared to 171 tons per year and GW(e) for Light Water Reactors) of fast Breeders would extend the above given time periods accordingly. As the nuclear breeding is also possible for the Th-232/U-233 nuclear fuel cycle, the available resources of natural uranium and thorium and the application of Fast Breeder Reactors would prolong the above discussed 180 years to many thousand years [5].

1.3 Uranium Enrichment

For present Light Water reactors the initial fuel of the core must be enriched from the 0.72 % of U-235 of natural uranium to an enrichment of 4–5 % U-235 (depending on the fuel burnup) in U-235/U-238 uranium dioxide (UO2) fuel. This is achieved first by chemical conversion of the uranium ores U3O8 into uranium hexafluoride, UF6, which is gaseous above a temperature of 55 °C. This gaseous UF6 is enriched presently in essentially three different commercial enrichment processes:

gaseous diffusion process

gas centrifuge process

LASER enrichment process

The LASER enrichment process SILEX is deployed in a first commercial enrichment plant in the USA. The gas centrifuge process is already used in large scale plants in Russia, Europe, Japan, China and the USA. The earliest deployed large scale gaseous diffusion enrichment plants are still in operation in the USA, France and China. They will be replaced in the future by the more economical gas centrifuge enrichment plants and probably LASER enrichment plants.

The production capacity of enrichment plants is measured in kg or tons separation work units (SWU). An LWR of 1 GW(e) power requires a reload of 25 tons enriched uranium fuel with an enrichment of 4.4 % U-235. This requires 175 tons SWU [7, 8] (Table 1.2).

Table 1.2

Worldwide installed capacity of gaseous diffusion-, gas centrifuge- and LASER-enrichment plants [8]

After enrichment in U-235 the UF6 will be treated chemically to become UO2. In fuel fabrication facilities UO2 powder will be pressed and sintered to UO2 pellets of about 1 cm diameter and 1 cm height. These pellets are filled into about 4 m long Zircaloy (zirconium-aluminum alloy) tubes (fuel rods). The tubes are then filled with helium and welded gastight on both ends. On the upper end of these fuel rods an empty space of about 10 cm length remains where the fission gases can collect during reactor operation. Fission gas pressure can rise then up to several MPa.

A number of countries with nuclear power plants operate also UO2 fuel cycle plants. In total there were 37 uranium mines, 22 uranium conversion plants, 13 uranium enrichment plants, 40 uranium fuel fabrication plants and 5 spent fuel reprocessing plants commercially operating in the world in 2008 [5].

1.4 Spent Fuel Reprocessing

Commercial spent fuel reprocessing was deferred in the USA in 1982 for fear of proliferation of plutonium. Later this decision was revised by the US government but no commercial reprocessing industry developed in the USA up till now. Only intermediate storage and direct spent fuel disposal were pursued. Germany and Sweden did follow this example of the USA. Other countries like France, Great Britain, Russia or Japan did not follow this strategy but do reprocess spent fuel elements. This led to the situation that spent fuel reprocessing facilities were almost entirely built and operated in nuclear weapon states with Japan being the sole exception.

Table 1.3 shows the reprocessing capacities for spent fuel elements available in different countries of the world in 2012. The reprocessing capacities for LWR spent fuel add up to a total of 4,340 tons per year (see also Fig. 1.1). In addition reprocessing capacities of 1,200 tons per year for metallic spent fuel of AGRs, 100 tons per year capacities for CANDU spent fuel and 100 tons per year capacities for spent FBR fuel are in operation. As about 25 tons of spent uranium fuel are unloaded per GW(e) and year from a Light Water Reactor this world wide reprocessing capacity would be sufficient for spent fuel of e.g. 174 GW(e) LWR. This is about half of the presently available LWR capacities in the world.

Table 1.3

Worldwide spent fuel reprocessing capacity in tones per year [5]

LWR Light Water Reactor

AGR Advanced Gas cooled Reactor

CANDU Canadian Deuterium Uranium pressurized Heavy Water Reactor

FBR Fast Breeder Reactor

References

1.

American Nuclear Society (2013) World list of nuclear power plants. Nuclear News March 2013. American Nuclear Society, La Grange Park, IL

2.

International Nuclear Safety Center at ANL-Aug 2005 (2000) http://​www.​ne.​anl.​gov/​research/​ierc/​intnlcoop.​html

3.

OECD (2008) Nuclear Energy Outlook 2008. OECD-NEA No. 6348. OECD, Paris

4.

OECD (2008) Uranium 2007 – resources, production and demand. OECD-NEA-IAEA 6345. OECD, Paris

5.

Kessler G (2012) Sustainable and safe nuclear fission energy. Springer, Heidelberg

6.

Carré F et al (2009) Overview on the French nuclear fuel cycle strategy and transition scenario studies. In: Proceedings of Global 2009, Paris, Paper No. 9439

7.

Villani S (ed) (1979) Uranium enrichment: Topics in applied physics, Vol 35. Springer, Berlin

8.

Laughter M (2007) Profile of world uranium enrichment programs – 2007 ORNL/TM-2007/193. Oak Ridge National Laboratory, Oak Ridge, TN

© Springer-Verlag Berlin Heidelberg 2014

Günter Kessler, Anke Veser, Franz-Hermann Schlüter, Wolfgang Raskob, Claudia Landman and Jürgen Päsler-SauerThe Risks of Nuclear Energy TechnologyScience Policy Reports10.1007/978-3-642-55116-1_2

2. Some Facts About Neutron and Reactor Physics

Günter Kessler¹  and Anke Veser²

(1)

Stutensee, Germany

(2)

Eggenstein, Germany

Abstract

Chapter 2 describes some facts about neutron and reactor physics needed for the understanding of Chaps. 3–10. It starts with the radioactive decay and the definitions of the decay constant and the half-life. It continues with the explanation of the fission process for fissile nuclear isotopes, e.g. U-233, U-235, or Pu-239 and the fission energy release by creation of fission fragments (products), prompt fission neutrons and delayed neutrons and radiation (β-particles, γ-rays and antineutrinos). This is followed by the definition of reaction rates of neutrons with other atomic nuclei, the presentation of measured microscopic cross sections for absorption, capture and fission as well as the definition of the macroscopic cross section and the neutron flux.

In LWR cores the fuel is arranged heterogeneously in lattice cells together with a moderator (water) in order to slow down the fission neutrons with high kinetic energy to kinetic energies in the range of 0.025 eV (thermal energy). This is most effective if the enriched uranium fuel is put in cylindrical rods which are arranged in e.g. a square grid. The optimization of the geometrical distance between the fuel rods leads to important safety characteristics of LWR cores: the negative fuel Doppler coefficient and the negative coolant (moderator) coefficient.

The definition of the criticality factor or effective multiplication factor, keff, allows a characterization whether the reactor core is operated in steady state condition or whether it is subcritical or even supercritical. The criticality or effective multiplication factor, keff, can be changed by moving or by insertion or withdrawing of absorber material (boron, cadmium, gadolinium, indium, silver, hafnium, erbium) in the core. This allows control of the reactor. The reactor core is controlled always in a keff range where the delayed neutrons are dominating. The delayed neutrons are therefore of highest importance for the control of the reactor.

During reactor operation over months and years the initially loaded U-235 in the low enriched uranium fuel will be consumed, neutron absorbing fission products will build up or other heavy nuclei with masses above U-235 and Pu-239 will be created. This decreases the criticality of the effective multiplication factor keff. This burnup effect on the criticality factor keff is accounted for by the design of the reactor core. The enrichment of the initially loaded fuel is increased such that keff becomes slightly >1. This is balanced by absorber materials (moveable absorber rods, burnable neutron poisons, e.g. gadolinium or boric acid) which keep the reactor core always at keff ≥ 1.

After shutdown of the reactor the gradually decaying fission products and the radioactive decay of higher actinides creates afterheat in the reactor core. This afterheat (decay heat) must be transferred by the coolant water to outside coolant towers or to river or sea water.

Prior to the description of Light Water Reactor designs some basic characteristics of reactor physics and reactor safety will be presented. For a deeper understanding of these characteristics the literature given in the reference is recommended [1–8].

2.1 Radioactive Decay, Decay Constant and Half-Life

Radioactive decay changes the number of unstable (radioactive) isotopes, N(t), existing per cm³ as a function of time, t. This change can be described by the exponential law of

where λ is the decay constant and N0 the number of atomic nuclei per cm³ at the time t = 0. Instead of the decay constant, λ, one can also use the half-life, T1/2 = (ln2)/λ, which is the time by which half of the nuclei existing at t = 0 have decayed. The decay rate, λ ∙ N(t), is called the activity of a specimen of radioactive material. This activity is measured in units of Curie or Becquerel [1, 2].

One Becquerel, denoted Bq, is defined as one disintegration per second. One Curie, denoted Ci, is defined as 3.7 × 10¹⁰ disintegrations per second, which is approximately the activity of 1 g of radium. Low activities are also measured in mCi = 10−3 Ci or μCi = 10−6 Ci [1, 2].

2.2 Fission Process

If a neutron of a certain velocity (kinetic energy) is absorbed by a fissile heavy nucleus, e.g. U-233, U-235 or Pu-239, the resulting compound nucleus can become unstable and split (fission) into two or even three fragments (Fig. 2.1). The fission fragments are created essentially according to a double humped yield distribution function with mass numbers between about 70 and 165. The mass yield distribution functions are similar for heavy nuclei fissioned by neutrons with kinetic energies of 0.0253 eV (thermal spectrum reactors) up to about 0.2 MeV for Fast Breeder Reactors¹ (Fig. 2.2). They depend slightly on the kinetic energy of the incident neutrons causing fission and on the type of heavy nuclei (U-233, U-235, Pu-239).

Fig. 2.1

Fission of U-235 nucleus by a thermal neutron

Fig. 2.2

Fission product yield (%) for fission reaction of different isotopes by thermal and fast (E > 0.2 MeV) neutrons [9]

In addition to the fission products (fragments), 2–3 prompt neutrons are emitted during the fission process. These prompt fission neutrons appear within some 10−14 s. They are created with different kinetic energies following a certain distribution curve around an average neutron energy of about 2 MeV. In some heavy nuclei with even mass numbers, e.g. Th-232 and U-238, nuclear fission can only be initiated by incident neutrons with a certain, relatively high, threshold kinetic energy (Table 2.1), whereas the uneven heavy nuclei, e.g. U-233, U-235, Pu-239 etc. can be fissioned by neutrons with all kinetic energies >0 eV. However, the even-uneven rule is not a rigorous one, e.g. Am-242m can also be fissioned by thermal neutrons.

Table 2.1

Threshold kinetic energy for incident neutrons causing substantial fission in different heavy nuclei [10]

The fission products can either be solid, volatile or gaseous. Many of the fission products decay further emitting so-called delayed neutrons, β-particles, γ-rays and antineutrinos. The delayed neutrons resulting from the decay of particular fission products—called precursors—represent less than 1 % of all released neutrons. The fraction β of delayed neutrons originating from fissioning by thermal neutrons (0.0253 eV) of U-235 is β = 0.67 %, and β = 0.22 % from fissioning of Pu-239. They appear following decay constants of 0.01–3 s−1 for U-235 and 0.01–2.6 s−1 for Pu-239. These delayed neutrons are of absolute necessity for the safe control and operation of nuclear fission reactors [6, 7, 11].

The total energy release per fission, Qtot, appears as kinetic energy of the fission products, Ef, of the prompt fission neutrons, En, as β−-radiation, Eβ, as γ-radiation, Eγ, or as neutrino radiation, Ev, (Table 2.2). The neutrino radiation does not produce heat in the reactor core due to the small interaction probability of neutrinos with matter. Table 2.2 also shows the total energy, Qtot, and the thermal energy, Qth, released during fission of a nucleus. Some of β−-radiation and γ-radiation of the fission products is not released instantaneously, but delayed according to the decay of the different fission products.

Table 2.2

Different components of energy release per fission of some heavy nuclei in MeV by incident neutrons of different kinetic energy (in the eV or MeV range) [10]

On the average, about 194 MeV or 3.11 × 10−11 J are released per fission of one U-235 atom. Most of the fission energy is released instantaneously.

Since 1 g of U-235 meal contains 2.56 × 10²¹ atoms, the complete fission of 1 g of U-235 results in:

For other fissile materials like U-233 or Pu-239 the energy release per fission is similar. Also fission by neutrons with thermal energies (0.025 eV) or by energies of 0.5 MeV leads to almost equal energy releases.

Therefore, it is usually assumed that the fission of the mass of 1 g of fissile material, e.g. U-235 or Pu-239 produces roughly 1 MWdth and the measure of burnup in MWdth per tonne of fuel also corresponds roughly to the number of grams of, e.g. U-235 fissioned in 1 ton of spent fuel [12].

2.3 Neutron Reactions

Neutrons produced in nuclear fission have a certain velocity or kinetic energy and direction of flight. In a fission reactor core, e.g. with U-235/U-238 fuel they may be scattered elastically or inelastically or absorbed by different atomic nuclei. In some cases the absorption of neutrons may induce nuclear fissions in heavy nuclei (U-235 etc.) so that successive generations of fission neutrons are produced and a fission chain reaction is established.

2.3.1 Reaction Rates

If n ( $$ \overrightarrow{r} $$ , v, $$ \overrightarrow{\Omega} $$ ) is the number of neutrons at point $$ \overrightarrow{r} $$ , with velocity v and the direction of flight $$ \overrightarrow{\Omega} $$ , then these neutrons can react within a volume element dV with N ∙ dV atomic nuclei (N being the number of atomic nuclei per cm³ of reactor volume). The number of reactions per second e.g. scattering or absorption, is then proportional to

The proportionality factor σ(v) is a measure for the probability of the nuclear reactions and is called microscopic cross section of the nucleus for a specific type of reaction. The microscopic cross section σ(v) is measured in 10−24 cm² ≙ 1 barn. It is a function of the velocity, v, or kinetic energy, E, of the neutron and of the type of reaction and differs for every type of atomic nucleus. As for an absorption reaction the neutron can either remain captured or lead to fission of a heavy nucleus the relation

Is valid with

σa(v) microscopic absorption cross section

σc(v) microscopic capture cross section

σf(v) microscopic fission cross section

The reaction rate can be written

The quantity $$ \Sigma \left(\overrightarrow{r},\mathrm{v}\right)=\mathrm{N}\cdot \upsigma \left(\mathrm{v}\right) $$ is called macroscopic cross section.

The quantity $$ \upphi \left(\overrightarrow{r},\mathrm{v}\right)=\mathrm{v}\cdot \mathrm{n}\left(\overrightarrow{r},\mathrm{v}\right) $$ is called the neutron flux.

Figure 2.3 shows the microscopic fission cross section as a function of the neutron kinetic energy for the heavy nuclei U-235, U-238 and Pu-239. The fission cross sections for U-235 and Pu-239 increase with decreasing kinetic energies. In the energy region of about 0.1–10³ eV this behavior is superposed by resonance cross sections [12–14].

Fig. 2.3

Microscopic fission cross sections (measured in barn) for U-235, U-238 and Pu-239 [13]

The capture cross section for U-238 is shown in Fig. 2.4.

Fig. 2.4

Microscopic capture cross section for U-238 [13]

Neutron capture in U-238 leads to U-239 and after β−-decay to Np-239 which again decays to Pu-239.

Such microscopic cross sections (measured in barn) are steadily compiled, evaluated, supplemented and revised in nuclear data libraries [15–17].

The capture cross section of U-238 (Fig. 2.4) shows distinct narrow resonance peaks above about 5 eV. At medium neutron energies (keV range) the resonance peaks become smaller and above about 10 keV—in the so-called unresolved resonance energy range—they cannot be resolved any more by experiments because of resonance overlapping. These resolved and unresolved resonance peaks broaden if the temperature of the U-238 fuel increases. This phenomenon is very important for the fuel-Doppler-temperature coefficient, which determines—among other temperature coefficients—the safety characteristics of nuclear power reactors [12, 18].

The microscopic fission cross sections of U-235 or Pu-239 (Fig. 2.3) become higher—with the exception of the energy range where the resonances occur—at low neutron energies (about <0.1 eV).

Therefore, the U-235/U-238 fuel in Light Water Reactor cores is mixed with materials of low mass number (moderator) in order to slow down the fission neutrons having high kinetic energy by a number of elastic and inelastic collisions to kinetic energies in the range <1 eV (Fig. 2.3). This is most effective if the enriched uranium fuel is put in cylindrical rods which are arranged in e.g. a square grid. This lattice of fuel rods arranged with certain distances must be cooled by a flowing coolant which can be identical with the moderator as in case of water in Light Water Reactors. The fission neutrons after a few collisions with the fuel atoms then fly with high velocity into the surrounding water/moderator. They are slowed down by collisions within a short distance to so-called thermal energies of 0.025 eV (Fig. 2.5). The neutrons are then in thermal equilibrium with the relatively low kinetic energies of the water molecules. Another advantage of the deceleration of the neutrons in the surrounding water is given by the fact that the probability is lower that the neutrons can be captured in the resonance region of U-238 inside the fuel rods (Fig. 2.4) [1–5].

Fig. 2.5

Square lattice cell with fuel rod, cladding and moderator of a LWR-fuel element

After the neutrons are thermalized within the moderator region they migrate back by diffusion processes into the fuel rods. As they have lower kinetic energies now also the microscopic fission cross sections are much higher (in the 0.025 eV energy range) than those for fission neutrons. The consequences are more fission reactions. Also the ratio between fission and capture reactions becomes more favorable in the fuel.

An optimum volume ratio between moderator and fuel for the grid of fuel rods is found around 2 for light water (H2O). This optimal volume ratio can be achieved by adaption of the distance between the fuel rods. For heavy water (D2O) as a moderator this optimal volume ratio is about 20 and for graphite as a moderator it is found to be around 54 [1–5].

Light water (H2O) has a higher microscopic capture cross section than heavy water or pure graphite. Consequently it is possible to build and operate nuclear power reactors with natural uranium (0.72 % U-235 enrichment) if heavy water or graphite are used as moderator. In fact the first reactor used natural uranium as fuel and graphite as moderator. With light water as moderator in Light Water Reactors the uranium fuel must be enriched to 3–5 % in U-235 (depending on the burnup of the uranium fuel). Structural materials which must be used for the design of the reactor core should also have low microscopic capture cross sections. Light Water Reactors, therefore, use an alloy of zirconium and aluminum (Zircaloy) for the cladding of the fuel rods and grid spacers of the fuel elements [1–4].

Uranium dioxide (UO2) with its high melting point (2,865 °C) and its good irradiation properties in the neutron field of the reactor core is used as fuel in LWR cores.

2.4 Criticality or Effective Multiplication Factor keff

The ratio between the number of newly generated neutrons by fission and the number of neutrons absorbed in the reactor core or escaping from the reactor is called the criticality factor or effective multiplication factor, keff.

When keff = 1, the reactor core is critical and can be operated in steady state. At keff < 1 the reactor core is subcritical, e.g. with control or absorber rods fully inserted in the core.

Boron, Cadmium or Gadolinium etc. can be used as absorber materials, either as metallic alloys in control rods or as burnable poisons in ceramic form in fuel rods and special poison rods or as a fluid, e.g. boric acid in the coolant of an LWR [1–4].

For a keff > 1 the reactor core is supercritical. More neutrons are produced than are absorbed in the reactor core or do escape from the core. The neutron chain reaction is ascending (reaction rates and the number of neutrons and, thus, reactor power increase as a function of time).

The criticality or effective multiplication factor keff of a reactor core is determined by the proper choice of its geometrical dimensions (diameter and spacing of the fuel rods, diameter and height of the reactor core), by the choice of the moderator and coolant as well as by the choice of the fuel and structural materials. The choice of the U-235 enrichment of the fuel is of decisive importance for LWRs.

2.5 Neutron Density and Power Distribution

Figure 2.6 displays the spatial distribution of the neutron density in the range of thermal energies for a Pressurized Water Reactor (PWR). The distribution of fission reaction rates or of the power generated by fissions is essentially proportional to the neutron density distribution. The absorber- or control-rods are partially inserted in axial direction in the PWR core. The control rods absorb neutrons and are responsible for the spatial distortions of the thermal neutron flux. They influence the criticality level and the spatial power distribution.

Fig. 2.6

Spatial distribution of the thermal neutron density in a Pressurized Water Reactor core with partially inserted control rods [11]

The spatial distribution of the neutrons with a certain velocity or kinetic energy and flight direction can be described by the Boltzmann neutron transport equation or by Monte Carlo methods [1–4]. For both cases, numerical methods in one-, two- or three-dimensional geometry were developed. Computer program packages (deterministic codes for the solution of the Boltzmann transport equation and Monte Carlo codes using stochastic solution methods) are available for various applications [1, 3–5, 18, 38].

For many practical applications it is sufficient to solve the neutron diffusion equation which is an approximation to the Boltzmann neutron transport equation.

The microscopic cross sections shown in Figs. 2.3, 2.4 and 2.7 are collected in a special format in cross section libraries, e.g. JEFF [15], ENDF/B [16], JENDL [17]. Their continuous energy range can be approximated and divided into a number of energy groups with specifically defined microscopic group cross sections applying codes, e.g. NJOY [19] or MC²-3 [20, 21]. The heterogeneous cell geometry of the reactor core (Fig. 2.5) can be accounted for by codes, e.g. WIMS [22] or MC²-3 [20, 21]. These computational methods are summarized in [23, 24].

Fig. 2.7

Microscopic capture cross sections of