Jagersteig 4

Stutensee, Germany

Jagersteig 4

Stutensee, Germany

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Kessler G.,Jagersteig 4
Power Systems | Year: 2012

Plutonium isotopes, but also the isotopes of minor actinides: mainly neptunium, americium and curium can be fissioned by neutrons in the core of nuclear reactors. They also can be transformed as non-fissile isotopes by neutron capture into fissile nuclides (transmutation). Incineration of 99% of the plutonium, neptunium, americium and curium would decrease the long term radiotoxicity of the high active waste (HLW) such that the radiotoxicity level of natural uranium would be underrun already after about 3 × 104 years. This requires chemical separation of plutonium, neptunium, americium and curium and the fabrication of fuel elements with these actinides. These chemical separation methods and the fuel refabrication methods were already developed by research and development programs and demonstrated in pilot plants or at laboratory scale. The possible incineration rates for the different actinides in different reactor types (light water reactors, liquid metal cooled fast breeder reactors and accelerator driven systems) have been thoroughly investigated. Reactor strategies with light water reactors operating in symbiosis with liquid metal cooled fast breeders or accelerator driven systems are feasible. The different reactor and fuel cycle strategies have different radioactivity loads and different radiotoxicity levels within the different parts of their fuel cycle. Whereas the radiotoxicity can be drastically decreased in the back end of the fuel cycle, the masses of plutonium and minor actinides and their radioactivity and radiotoxicity can be higher during reprocessing and refabrication. Transmutation and destruction of long-lived fission products is only feasible with reasonable efficiency for Iodine-129 and Technetium-99. © Springer-Verlag Berlin Heidelberg 2012.


Kessler G.,Jagersteig 4
Power Systems | Year: 2012

The safety of light water reactors is based on long term international research programs. The objective is to protect the operational personnel, the environment and the population against radioactivity releases during normal operation and in case of accidents. The safety concept is based on multiple containment structures (multi-barriers) as well as engineered safeguards components and other measures combined in a staggered-in-depth concept of four safety levels. The light water reactor plant and its protection system must be designed and built according to the design basis concept. Those design basis accidents which are part of the licensing process must be accommodated by the protection system, the inherent safety features and by the emergency cooling systems of the nuclear plant. Probabilistic safety analysis are supplements to this deterministic approach. They show that European light water reactors have a frequency of occurrence of about 10-5 to 10-6 per reactor year for core meltdown. Reactor risk studies which had been performed during the 1970s (USA) and 1980s (Europe) showed that the risk arising from light water reactors as a result of core melt down is well below the risk of other power generating or traffic systems. However, the Chernobyl accident in 1986 resulted-in addition to a not well known number of fatalities-in large scale land contamination by cesium-137 with a half-life of about 29 years. Similarly, the Fukushima accident (2011) resulted in land contamination by radioactive cesium isotopes. New research programs on severe accident consequences were initiated around 1990s. Their results lead to a revision of the results of the early risk studies of the 1980s and a new safety concept for modern light water reactors, e.g. the European Pressurized Water Reactor (EPR) and the European Boiling Water Reactor (SWR-1000). This new reactor safety concept allows to limit the severe accident consequences to the plant site itself. Also the introduction of additional severe accident management measures for existing light water reactors resulted in a considerable improvement of the prevention and mitigation of severe accident consequences. The safety concept of fuel cycle plants, e.g. spent fuel storage facilities, reprocessing facilities and waste treatment facilities is based on similar containment and engineered safeguards measures. However, the risk of these fuel cycle facilities is much smaller as the fuel is at much lower temperatures in reprocessing and refabricataion plants. © Springer-Verlag Berlin Heidelberg 2012.


Kessler G.,Jagersteig 4
Power Systems | Year: 2012

According to IAEA and OECD/NEA the reasonably assured and inferred natural uranium resources amounted in 2007 to 5.47 million tonnes in the world for the cost category up to 130 US $/kg. The reasonably assured resources are based on high confidence estimates compatible with decision making standards for mining. Inferred resources still require additional measurements, before making decision for mining. Prognosticated resources amounted to another 2.8 million tonnes and speculative natural uranium resources were estimated to about 7.5 million tonnes. Together with so-called unconventional natural uranium resources the global amount of uranium is estimated to about 22 million tonnes. For a 400 GW(e) scenario the present LWRs operating in a once through fuel cycle mode would consume about 5.46 million tonnes of natural uranium over a time period of about 80 years. Plutonium recycling reactors, e.g. LWRs and FBRs, operating in a partially or fully closed fuel cycle would consume a reduced amount of natural uranium over a very long period of time. © Springer-Verlag Berlin Heidelberg 2012.


Kessler G.,Jagersteig 4
Power Systems | Year: 2012

The safety design concept of LMFBRs follows the same basic principles (multiple barrier concept and four level safety concept) as they were developed for light water reactors. This holds despite of the fact that LMFBRs have different design characteristics (fast neutron spectrum, liquid metal as coolant, plutonium-uranium fuel). It has been shown that LMFBRs possess a strong negative power coefficient and good control stability. The main design characteristics of control and shut-off systems do not differ much from those of light water reactors. For sodium cooled fast reactors the sodium temperature and sodium void coefficient can become positive above a power output of the core above about 350 MW(th). Therefore, special design provisions are taken for future LMFBR designs, e.g. flat and heterogeneous cores. The excellent cooling and natural convection properties as well as the low system pressure of about 1 bar of liquid metal cooled fast breeder allow the safe decay heat removal by a number of ways. The consequences of sodium fires or sodium water reactions can be prevented or limited by special design provisions. On the other hand, lead and lead-bismuth-eutecticum (LBE) as coolant do not chemically react neither with oxygen of the atmosphere nor with water in failing tubes of a steam generator. Historically the characteristics of homogeneous sodium cooled cores with a positive sodium void coefficient of the early prototype fast breeder reactors have lead to the analysis of core disruptive accidents with the objective to find a basis for the main design requirements of the containment. The discovery of the negative control rod drive line expansion coefficient in the early 1980s, changed this situation and lead to a new safety design which avoids anticipated transients without scram (ATWS) for future LMFBRs. The high boiling points of lead with 1,740°C LBE with 1,670°C offer an advantage with respect to safety concerns compared to sodium as coolant. © Springer-Verlag Berlin Heidelberg 2012.


Kessler G.,Jagersteig 4
Power Systems | Year: 2012

Nuclear power generation is currently mainly based on light water reactors, designed as pressurized water reactors and boiling water reactors. These are built by a number of manufacturers in various countries of the world. In this chapter, the standard German PWR of 1,300 MW(e) and the European Pressurized Water Reactor (EPR) will be described. In addition, the chapter deals with the German Standard BWR of 1,280 MW(e) and the newer design SWR-1,000 (KERENA). Gas cooled and graphite moderated commercial reactors with natural uranium were developed in the United Kingdom and in France and built in the 1950s and 1960s (MAGNOX reactors). Advanced gas cooled reactors (AGCRs) with graphite as moderator and carbon dioxide as coolant gas have been built in unit sizes up to 620 MW(e). High temperature gas cooled reactors with gas outlet temperature of 700-740°C use helium as a coolant gas. Their fuel elements have been developed as prismatic or spherical pebble fuel elements. High temperature gas cooled reactors with medium enriched uranium are now designed mainly as small modular reactors for safety reasons. Power reactors with heavy water as the moderator and heavy water or light water as coolant have been developed in Canada, Europe and Japan up to unit sizes of 630 MW(e). The advanced CANDU reactor (ACR) is developed currently to a unit size of up to 1,000 MW(e). Homogeneous core thermal breeders with molten salt and light water breeder reactors together with accelerator driven subcritical reactor cores are still in the design or development phase. © Springer-Verlag Berlin Heidelberg 2012.


During normal operation of nuclear power plants and facilities of the nuclear fuel cycle small amounts of radioactivity are released into the environment at a monitored and controlled rate. Men may be exposed to external radiation as well as radiation by inhalation and ingestion. Upper limits for the radiation exposure of individuals of the public as well as of employees during their occupational work time have been set by the International Commission on Radiation Protection as well as by state organizations. The nuclear fuel cycle begins with uranium mining and milling where the main effluents are radon and dust particles containing uranium and its decay products. Radioactive effluents are reported for both open pit and underground mining. This is followed by listing the radioactivity release and exposure rate of uranium conversion, enrichment and fuel fabrication facilities. For Pressurized Water Reactors the annual effective dose to the public is well below one micro-Sievert. For Boiling Water reactors the annual effective dose is somewhat higher. However, this is still more than a factor 100 lower than the permissible limit. Release data for radioactive nuclides from the European spent fuel reprocessing and waste treatment centers are collected by the European Commission. The radioactive exposures to the public from these facilities are well below the permissible effective radiation exposures as well. The same result is valid for the plutonium/uranium mixed oxide fuel refabrication plant MELOX in France. © Springer-Verlag Berlin Heidelberg 2012.


Kessler G.,Jagersteig 4
Power Systems | Year: 2012

After discharge from the reactor core, the fuel elements are stored in a fuel element storage pool onsite for several years to allow radioactivity decay and after heat decrease. Spent fuel elements are shipped then in special fuel transport casks to either intermediate storage facilities or to the storage pool of a reprocessing plant. After a total cooling period of about 7 years LWR spent fuel elements can be chemically reprocessed. The spent fuel elements are moved from the storage pool into the disassembly cell, where they are cut up by large bundle shears into small pieces. These pieces fall into a dissolver basket filled with boiling nitric acid. The PUREX process is used to chemically separate the dissolved spent fuel into uranium, plutonium and higher actinides with fission products. The final products are uranylnitrate, plutonium nitrate and high level waste. The total capacity of commercial reprocessing facilities is currently about 4,500 tHM/year in France, UK, Russia, Japan and India. The uranium and plutonium products can be converted into oxides and fabricated into Uranium/Plutonium mixed oxide fuel elements. The latter can be loaded into light water reactor or fast breeder reactor cores. Thorium/uranium fuel can be reprocessed using the THOREX process. The thorium/uranium-233 fuel also can be fabricated into mixed oxide fuel elements and loaded into light water reactors or fast breeder reactors. The remaining wastes are classified into high level waste, medium level waste and low level waste. The high level waste after concentration is vitrified by giving it first into a calcinator and then mixing it with borosilicate glass frits and melting this mixture to a glass. The result is a vitrified high level glass in a steel container. The fuel rod hulls and end pieces of fuel elements as well as insoluble residues are compacted by a 250 MPa press into a cylindrical container. Low level organic waste is sent to a medium temperature pyrolysis system and then to a calcination system. The end product is mixed with pastes, grouts or concrete and filled into low level waste containers. The medium and low level waste packages are sent to medium/low level waste repositories which are already in operation in France, Japan, Spain, Sweden, Finland and the USA since the early 1990s. High level waste packages are foreseen to be disposed into deep geological repositories. For the direct disposal concept of spent fuel elements either the fuel elements or only the fuel rods are loaded in high level waste containers and foreseen to be disposed in a deep geological repository. Up to now no deep geological repository is in operation, but test sites are explored and under investigation. © Springer-Verlag Berlin Heidelberg 2012.


Kessler G.,Jagersteig 4
Power Systems | Year: 2012

In 2011 there were about 436 commercial nuclear power reactors operating in the world with a total capacity of 370 GW(e). Nuclear energy supplied about 16% of the world electricity. The countries with the largest nuclear energy generating capacities were the USA, France, Japan, Russia, South Korea, UK, Canada, Ukraine, China, Spain. About 81% of the operating nuclear reactors were light water cooled and moderated reactors. About 11% were pressurized heavy water moderated reactors and about 3.4% graphite moderated and gas cooled reactors. Another about 4% light water cooled and graphite moderated reactors of the Chernobyl type were still operating in Russia. The remaining 0.6% were new prototype power reactors. Besides the use of nuclear power for electricity generation, wider application directly using the nuclear heat as process heat with temperatures of about 800°C (gas cooled reactors) is possible in future. In the past BN 350 situated on the shore of the Caspian Sea was already used as a dual purpose plant, supplying in addition to electricity (150 MW(e)) also fresh water (120,000 m3/d) by desalination. The economic advantages of nuclear power is based on the relatively low fuel cycle costs. However, nuclear power plants have capital costs higher than those of e.g. fossil fired power plants, but a much more pronounced cost degression for larger units. Nuclear power avoids to a large extent the emission of CO2, SO2, NOx and also particulate emissions. © Springer-Verlag Berlin Heidelberg 2012.


Kessler G.,Jagersteig 4
Power Systems | Year: 2012

Breeder reactors with a fast neutron spectrum have a sufficiently high breeding ratio to attain a fuel utilization of more than 60% which is almost by a factor of 100 higher than that of present light water reactors. They can operate on the U-238/Pu-239 or on the Th-232/U-233 breeding process and utilize depleted or natural uranium or thorium. In this way they can open up an energy potential with the existing uranium and thorium reserves which can last for many thousand years. Construction of breeder reactors began in the USA, the UK and the former Soviet Union already before 1960. Their development started with small test reactors and continued with the construction and operation of prototype power reactors of unit sizes of 300 MW(e) up to 1,250 MW(e) in the USA, Europe, Russia, India and Japan. This proved their technical feasibility. Fast breeder reactors with a fast neutron spectrum use sodium or in more recent designs lead or a lead-bismuth-eutecticum (LBE) as a coolant. Plutonium-uranium mixed oxide fuel, but also metallic alloys and nitride fuel were developed for the fuel of fast breeder reactors. At present the small test reactors JOYO (Japan) and BOR 60 (Russia) and the fast breeder reactor BN 600 in Russia are operating since several decades whereas MONJU (Japan) is close to full power operation and BN 800 as well as SVBR/75/100 in Russia and PFBR (India) are under construction. © Springer-Verlag Berlin Heidelberg 2012.


Kessler G.,Jagersteig 4
Power Systems | Year: 2012

The most important reactor physics characteristics needed for the understanding of the design and operation of nuclear reactors and of their fuel cycle are presented. This comprises the criticality factor, the neutron and temperature distributions in the reactor core and reactivity effects to be controlled by the safety systems. The evolution of the isotopic composition during burnup, i.e., the buildup of fission products and actinides in the reactor fuel, and the importance of conversion and breeding ratios are discussed together with the fuel utilization. Inherent safety characteristics like the negative fuel Doppler coefficient and the negative coolant temperature coefficient are essential for the safe operation and control of nuclear reactors. © Springer-Verlag Berlin Heidelberg 2012.

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