General Aspects of High-Temperature Reactors Abstract In this chapter, the principles of high-temperature reac-tors, especially the application of coated particles as fuel, are explained. The concept of reactors with spherical or prismatic fuel elements is fully developed. Special safety aspects like the self-acting decay heat removal and limitation of the fuel temperature even in case of extreme loss of cooling accidents characterize these concepts of modular HTR. It is possible to retain the radioactive substance practically totally inside the plant in these cases. The reactors can be applied to produce electricity with high ef.ciency and to supply cogeneration plants using modular units. The possibility to apply dry air cooling makes the system independent from the avail-ability of cooling water and preferable for arid regions. Different fuel cycles have been developed for this type of reactors: low enriched Uranium; use of Thorium; and even breeding is principally possible in the future. High burnup value and good neutron economy are character-istic for this reactor. So far as waste management is considered, no compact storage of spent fuel elements in a water pool (as in case of LWR) is necessary. The spent HTR fuel elements can be stored directly in air-cooled intermediate storage vessels for many decades. Direct .nal storage of spent fuel elements in geological systems has been developed, although reprocessing will be possible in future too. The HTR-PM for the generation of hot steam for a steam turbine process will be a next step to introduce this system into the energy market. Different large programs have been carried out and are under work today, to establish this technology. A special aspect is the use of modular HTR for future application as heat source for many chemical processes, especially conversion of hydrocarbons, oil processing, and hydrogen production. Keywords HTR principles .Coated particles .Spherical fuel elements .Modular HTR .Helium cooling Safety aspects .Fuel cycles .Applications Waste management .Concept of HTR-PM 1.1 Overview The use of nuclear energy worldwide in the last decades has grown up in the .rst time rapidly and later with a somewhat smaller rate. Today worldwide a nuclear capacity of around 400 GWe with an installation of 440 nuclear power plants is in operation. The electricity production worldwide was around 20 ℅1012 kWh/year in the Year 2010. A share of 12% was produced by nuclear power plants. On the other hand, the catastrophic accidents in Chernobyl (USSR, 1986) and Fukushima (Japan, 2011) have demonstrated the risks of the today introduced commercial reactors. Together with the accident in Three Mile Island (USA, 1979) by these events, the introduction of nuclear energy in the world has been slowed down. In some countries like Germany, a decision for .nishing the operation of the existing nuclear power plants was made. Many reactor concepts have been developed in the past, and just some of the possible variants until now became important. Table 1.1 gives a short overview on Table 1.1 Some aspects of reactor types Aspect Reactor Unit PWR BWR RBMK CANDU AGR HTR LMFR Characteristic of concepts Moderator H2O H2O H2O/C D2O C C Neutron spectrum Thermal Thermal Thermal Thermal Thermal Thermal Fast Fuel UO2 PuO2 UO2 PuO2 UO2 PuO2 UO2 PuO2 UO2 UO2 PuO2 UO2 PuO2 Type of fuel Rods Rods Rods Rods Rods Pebbles blocks Rods Coolant H2O H2O H2O D2O CO2 He Na Status of coolant Liquid Liquid/steam Liquid/steam Liquid Gas Gas Liquid Special aspects Zircaloy Zircaloy Pressure Pressure Steel Coated particles Steel of fuel canning canning tubes tubes canning canning Parameters of design Enrichment % 3每4 3每4 2 <1.5 2 8 10 Fuel burnup (average) MWd/kg 45 40 30 10 20 80 100 Core power density MW/m3 100 50 4 15 2 3 400 Coolant temperature ∼C 290每325 200每285 200每285 200每305 250每650 250每750/950 380每540 Coolant pressure MPa 16.0 *7.0 7.0 9.5 4.0 6.0 1.0 Steam pressure MPa 6.5 7.0 7.0 4.3 18.0 18.0 17.0 Steam temperature ∼C 280 285 285 250 530 530/600 500 Ef.ciency % 33 33 32 30 40 40/45 40 Thermal power MW 3800 3800 3000 1500 1500 200每600 750 Special aspects Natural uranium Gas turbine process heat Breeding PWR pressurized water reactor, BWR boiling water reactor, RBMK Russian boiling water reactor with graphite structures, CANDU Canadian heavy water reactor, AGR advanced gas-cooled reactor, HTR high-temperature reactor, LMFR liquid metal cooled fast reactor characteristics of reactor types, which are already introduced in the energy economy and some data of future HTR for comparison. 1.2 Future Sustainable Energy Technologies There are some general aspects important for the future use of all energy carriers and of nuclear energy worldwide. Nuclear energy has to ful.ll the requirements of sustain-ability in the future as any other energy technology too. Figure 1.1 summarizes some important aspects regarding these features. As far as economic conditions are considered today the production of electricity in nuclear power plants is compet-itive to coal-.red and gas-.red power plants in many countries. Especially the weak dependence of generation cost from escalation of Uranium ore prices during the whole operation time of decades makes nuclear power plants very attractive. Compared to renewable, especially the solar options photovoltaic and solar thermal power plants, nuclear energy is much more attractive. If the storage problem for electricity produced by these systems has to be taken into account, there is more than one order of difference for the generation costs in comparison with nuclear energy. Hydropower is very attractive, and production of electricity by wind energy shows large potential. However, the storage problem has to be solved too for a large share of wind energy converters, which are integrated into the grid. The radioactive emissions of nuclear power plants have been reduced by technical measures very much during the past decades. Today in Germany as example nuclear power plants 1.2 Future Sustainable Energy Technologies 3 contribute less than 0.5% to the burden of the population by radioactivity. The amounts coming from natural sources or caused by medicine are much larger. The doses acting upon the operation personal have dropped also very much in the past. Therefore, it can be stated that the natural conditions of life are not changed by the operation of nuclear power plants and other installations of the fuel cycle, if additionally the normal local variations of radiation doses are taken into account. Certainly further improvements of technology can help to reduce the doses even stronger in the future. This is a question of optimization of efforts for .ltering and saving costs by the better protection of environment. The assurance of a long-term supply of fuel to the nuclear system is another important requirement. The today known very cheap resources of Uranium of around 5 million tons are suf.cient to feed the today installed nuclear capacity for nearly 70 years. It is well known that there is much more Uranium available, but this will be more expensive. A factor of 10 in the price of Uranium ore means just 30% higher electricity generating costs related to today cost structures. The estimation is that under this condition Uranium ore is available more by a factor of 10每20. As is well known, breeding opens the possibility to use the Uranium much bet-ter, a factor of 30每40 will be realistic. Considering these large resources, it can be understood that the fuel supply of nuclear systems could be guaranteed for 1000 years and longer. Finally, there is the option to generate energy by the spallation process, and this option does not cause questions of available material but just economical ones compared to other options. Some important aspects are to ful.ll the requirements of non-proliferation of .ssile materials and the aspects regard-ing safety of nuclear plants and of .nal disposal of radioactive waste. These aspects are discussed in more detail in later chapters. Since some time work is going on worldwide to develop reactors of a Generation IV. Table 1.2 gives a principal overview on the reactors systems belonging to the different generations (Table 1.2). In the actual phase of nuclear development improvements of PWR, BWR, CANDU, and fast reactors are of central interest. It is a phase of enlarging the power, and one tries to reduce the speci.c investment cost. Partly aspects of improvement of safety like the introduction of core catchers, double-walled containments, and improvements of decay heat removal systems are characteristic examples. As already indicated in Table 1.2 for the future nuclear technology work is going on in the Generation IV systems, in which many countries are working together. The following types of reactors are analyzed in the road maps of the Generation IV systems. . GFR: gas-cooled fast reactor . LFR: lead-cooled fast reactor . MFR: molten salt reactor . SFR: sodium-cooled fast reactor . SCWR: supercritical water-cooled reactor . V/HTR: very high-and high-temperature reactor Table 1.2 Reactor systems belonging to the different generations of reactor development Characteristics Examples Generation I Experimental reactors prototype,.rst phase of commercial reactors Prototypes and .rst commercial plants Generation II Commercial reactors Large commercial plants (mainly PWR, BWR, CANDU, AGR, RBMK) Generation III Improved commercial systems Improvement of safety (EPR, ABWR, AP1000, ESBWR, CANDU+ , system 80+) Generation IV Future reactors Innovative concepts Mostly they are based on well-known concepts, which already have been tested in the past. A way or parameter is the maximal outlet temperature of the coolant. In the most case, this parameter is limited by corrosion effects or by the characteristics of the fuel element design. Just the VHTR promises helium temperature till 1000 ∼C, and therefore this reactor has been highest potential to realize high ef.ciencies, as example. The new published GIF technology road map update rede.ned the VHTR to V/VHTR and the outlet temperature to 700每1000∼C. 1.3 Principle Characteristics of HTR A reactor concept, which can produce electrical energy with high ef.ciency and is able to deliver nuclear process heat with high temperatures, is the high-temperature reactor [1每8] (Fig. 1.2). Originally, the goal was to use the high coolant temper-ature for the production of hot steam above 500 ∼C and to realize high ef.ciencies. Furthermore, the aspect of a good neutron economy because of the small absorption of neu-trons in graphite favored this system. In the course of further development of nuclear technology and energy economy, aspects like the use of nuclear process heat and the chance to realize systems with very advantageous safety behavior became more important. Here, the pebble-bed reactor is described as one of the important possibilities to realize HTR systems. Graphite is the main structural material for the fuel ele-ments and for the re.ectors. The coolant is helium under high pressure (Fig. 1.3), which .ows through the core from top to bottom and is heated up from around 250每750 ∼C for plants using steam generators. For gas turbine and high-temperature process heat application the core outlet temperature rises up to 900每950 ∼C or 1000 ∼C in the future. The graphite structure consists of the side re.ector and from top and bottom re.ector. The core bottom has a suited inclination to allow the out.ow of the fuel elements to a discharge tube. All graphite structures are included in a metallic core barrel, the thermal shield is covering all internals. The fuel consists of very small coated particles, which are embedded into the graphite matrix of the spherical fuel elements. The neutron spectrum is thermal. Because the parasitic absorption of graphite is low, a good neutron economy can be realized. The burnup is high and allows values between 80,000 and 100,000 MWd/t of heavy metal corresponding to an enrichment of 8每10%. Today the low enriched uranium cycle is preferred; however, in the future the use of thorium is possible too. Some further details of TRISO-coated particles can be explained on behalf of Fig. 1.3. The fuel is UO2 in the form thermal spectrum helium as coolant graphite as structural material and moderator coated particle fuel (TRISO) high burnup of fuel high helium outlet temperature fuel element form (hexagonal, spherical) safety characteristics: self acting decay heat removal with limitation of maximal temperature during accidents different applications (steam turbine, gas turbine, combined cycles, high temperature process heat) 1.3 Principle Characteristics of HTR 5 Fig. 1.3 Coated particles, fuel elements, and core of a pebble-bed HTR (a) coated particles (TRISO particle) 1: loading of fuel elements, 2: reflector, 3: core, 4: discharge of fuel elements, 5: control and shutdown elements (b) fuel element (pebble -bed fuel element) (c) core and reflectors of very small kernels of 500 lm diameter. The .rst buffer layer around the kernel (around 50每90 lm) is porous and can store .ssion products which leave the fuel zone. The three following layers (C/SiC/C) with small thicknesses (40 lm/35 lm/40 lm) act as a very reliable pressure vessel system and retain the .ssion products in normal operation almost completely and in severe accidents till a fuel tem-perature of around 1600 ∼C almost completely too. Release values of 10.5 of the inventory of .ssion products during operation are state of the art of technology today. The TRISO particles are the basis for all further devel-opments of HTR concepts worldwide today. In the past, BISO particles (without a SiC layer) have been used in AVR and THTR successfully too. However, the .ssion product retention of TRISO particles is better, as will be explained later in Chap. 10. A number of 10,000每20,000 of these small coated particles are arranged in the pressed matrix of a spherical fuel element. The fuel zone is surrounded by a graphite shell, which is free from fuel and which normally has a thickness of 5 mm. The graphite shell consists of the same material as the fuel matrix and is tightly connected to it. A fresh fuel element contains 7每10 g of heavy metal in oxide form, dependent on the core layout. This concept of fuel elements has the thermodynamic advantage that there exists a nearly homogenous production of heat in the fuel element and that the good heat conductivity of the graphite determines the temperature pro.les inside the fuel elements. Therefore, it is possible to generate very high coolant tem-peratures with relatively low fuel temperatures compared to other types of reactors. The spherical fuel elements with an outer diameter of 60 mm are stochastically arranged in the reactor core. The graphite re.ector forms the container for the core. The bottom has a conical form to allow the out.ow of balls from the core. The fuel element described here allows high burnup and helium circuits which have relatively low contamina-tion, if suited dimensioning of the reactor is applied. The pebble-bed core of modular HTR is cooled by helium .owing from the top to the bottom of the core. The balls are moving very slowly in parallel .ow to the cooling gas. The fuel balls are given in at the top of the reactor and .ow with an average speed of around 1 mm/h from top to the bottom of the core. The power density in the core is small compared to that of other reactor systems. The values are in the orders of 2每4 MW/m3, mainly for safety reasons. The power density in the UO2 kernels of the coated particles however is similar to the values used in the UO2 pellets of the LWR today. Caused by these values and because the heat conductivity in the graphite matrix is very high, heat conduction is the main step during heat transport in the fuel element and the temperatures in the fuel are relatively low. Principally the helium cycle contains the core and the reactor structures, which allow the passage of the helium .ow entering and leaving the core. These are elements of the top and bottom re.ectors. In case of steam-generating plants, the steam generators, helium circulators, and connecting ducts are further com-ponents of the primary circuit as well. Figure 1.4 shows the principle of the primary cycle of this characteristic applica-tion of HTR. The helium is heated up in the core from 250 to 750 ∼C corresponding to the characteristics of the axial distribution of the power density in the core. In the steam generator, the temperature span given before is used to preheat the feed water, to evaporate it, and to deliver superheated steam. The helium temperature of 750 ∼C allows the production of hot steam with a temperature of 530每600 ∼C and steam pressures of 10每30 MPa. These are conditions of modern steam cycle plants with total net ef.ciencies of 40每44%. Figure 1.5 contains an overview of the primary circuit of a modular HTR (e.g., HTR-Module, power 200 MWth). The primary circuit of a modular HTR is integrated into a reactor pressure vessel, the steam generator vessel, and a connecting vessel. Reactor pressure vessel and steam generator vessel are arranged side by side. Control and shutdown of modular pebble-bed reactors are carried out by absorber elements arranged in the side re.ector. In addition to the active shutdown system, there is a very important safety feature of modular HTR inherently available. The temperature coef.cient of reactivity for this reactor type is strongly negative in all different situations of normal operation and accidents. At rising temperatures, HTR plants always would be shut down by negative reactivity coef.cients. The charging of fresh fuel elements and the discharging of spent fuel elements are carried out continuously during full-power operation. Therefore excess reactivity for burnup is avoided in this type of reactors. This is a very important aspect of reactor safety (see Chap. 10). The total primary system of a modular HTR is integrated into an inner concrete cell. This cell (see Fig. 1.6) has the function to carry the load of all components of the primary circuit and to act as a .rst buffer volume in case of depressurization accidents. Fur-thermore, it acts as a shielding system for the radiation .eld around the reactor pressure vessel. The inner surface of the cavity for the reactor pressure vessel is covered with a sur-face cooler, which can remove the decay heat after a time of around 1 day. If this cooling system would fail too, the decay heat would be stored and transported by the concrete structure. The inner concrete cell could tolerate an over-pressure of around 0.2 MPa for a short time, and it is con-nected via a .lter system to the stack. The outer reactor containment building has mainly the function to protect the reactor against impacts from the outside (see Chap. 10). It is not necessary to be totally dense. There are different cycles of fuel burnup. One is the MEDUL cycle (MEhrfach DUrchLauf = several passes through the core), and the other is the OTTO cycle (One Through Then Out) with only one pass-through the core (Fig. 1.7). The OTTO cycle shows big attractions to realize very high helium temperatures at relatively low fuel tem-peratures, as will be explained in Chaps. 3 and 15. In the MEDUL cycle, the fuel elements pass the reactor core 6每15 times. After each pass, the burnup is measured. The MEDUL cycle allows to realize relatively small peak factors of power density, compared to the OTTO cycle. In all cycles, the spent fuel elements are directly stored in intermediate storage vessels. The ※compact§ storage, which in the light-water reactor is a water pool, is not necessary for high-temperature reactors. The spent fuel elements can directly be stored in air-cooled (self-acting) vessels for intermediate storage (see Chap. 11). The danger of damage of fuel elements like in Fukushima Accident (Japan 2011) in those storages is avoided. No accidents from internal reasons or external reasons can be de.ned, which could cause large release of .ssion products from this storage system. There-fore, an intermediate storage over many decades will be possible. Because of the geometric shape of the fuel element and because of continuous loading and discharge of fuel ele-ments, the fuel cycle allows a high .exibility. The cycle, i.e., the type of fuel elements, can be changed relatively easily during the operation life. Uranium and Thorium loadings as well as very different coated particle concepts have been tested in the past time. Reprocessing of the fuel is possible. The low power density in the core and the large masses of graphite in the core and of the re.ectors give the reactor an extremely inertial behavior related to thermal transients. Failures in the cooling system or in the decay heat removal 1.3 Principle Characteristics of HTR 7 Fig. 1.5 Primary enclosure of a modular HTR [7] 1: core, 2: core internals, 3: reactor pressure vessel, 4: steam generator, 5: helium circulator, 6: hot gas duct, 7: outer surface cooler, 8: control and shutdown systems, 9: concrete structures of inner concrete cell (a) vertical section (b) horizontal section Fig. 1.6 Reactor containment building of a modular HTR (e.g., HTR-Module, 200MWth) [7] system do not cause failure of fuel or structures, if the the reactor pressure vessel just by heat conduction, heat parameters of the core are chosen adequately. It is possible radiation, and free convection of air. to realize modular reactors, which never can melt and which Many experiments have been carried out to measure the will not be overheated in an unallowable way, if the active .ssion product release from irradiated spherical fuel decay heat removal totally fails. In this case, the decay heat elements. is transported inside the reactor system and given off from Fig. 1.7 Principles of MEDUL (a) (b) (a) and of OTTO cycle (b) 1: core; 2: discharge system; 3: burnup measurement; 4: recycling system; 5: intermediate storage of spent fuel elements In modular HTR with cylindrical core up to a thermal power of 250 MW, this heat transport works self-acting and a temperature of fuel higher than 1600 ∼C can never occur. As will be explained later in more detail at this temperature, the release of important .ssion products from the coated particle fuel (TRISO in pebbles) is still very limited, typi-cally to less than 10.4 of the inventory. Furthermore, this release is a very slow process and solid .ssion products will be retained in the primary system. Release rate of less than 10.5 from the primary system into the inner concrete cell is a consequence of this fuel behavior during accidents. This has as a consequence that sensible radiological consequences outside the nuclear power plants are avoided. Larger thermal power with the same safety standard can be realized with an annular core structure. The paths for heat transport to the outside have to be short enough, and this limits the thermal power to around 450 MWth, if self-acting decay heat removal with maximal fuel temperature of 1600 ∼ C is required as a safety principle in this case too. These limits for the thermal power correspond to the today avail-able forged steel vessel. If other types of vessels with larger diameter, as an example prestressed vessels, are used, the thermal power can be further raised up. Even in case of very extreme impacts from the outside, the limitation of the maximal fuel temperature to values of around 1600 ∼C would be valid. The reactor containment building and the concrete inner cell have to withstand the usual impacts from the outside, like gas cloud explosions, airplane crash, and earthquakes. The inner concrete cell together with an outer storage and .lter system and a closure system has the function to avoid overly large amounts of air ingress into the primary circuit after a depressurization accident. The HTR also needs some auxiliary systems. A gas puri.cation plant cleans the helium and removes impurities of H2, CO, H2O, O2,CO2, and CH4, and some dust from the primary coolant. Additionally, there is a helium storage system, which especially has large importance in case of gas turbine application. Following today requirements of waste handling, there will be an interim storage for spent fuel elements directly on the site of the power plants. For the reactor containment building, air-conditioning systems and special cooling sys-tems to ful.ll operational requirements are installed. Naturally, the reactor possesses the usual measurement installations to measure temperatures, pressures, mass .ows, neutron .uxes, and deviations of these parameters in time. A reactor protection system, which uses all the information and initiates safety actions, will also be part of a HTR plant. The electrical systems necessary for the operation of the plant and the installations to deliver the produced electricity to the electrical grid correspond to that, which are well known in power plant technology. 1.4 Application of Modular HTR in the Energy Economy Modular HTR systems can be applied in many .elds of energy economy (Fig. 1.8). They can be combined to steam cycles, to gas turbine processes, and to combined cycles for the production of electrical energy [8每14]. Furthermore, the helium outlet temperature from the core has to be raised up til 900每1000 ∼C to be able to carry out high-temperature processes. Modular heat sources can be used to operate cogeneration plants to deliver electrical energy and process heat or process steam for many industrial applications. In the future, it will be possible to supply high-temperature pro-cesses like hydrogen production, coal gasi.cation, or steel production with high-temperature process heat from modular HTR too [15每17]. Figure 1.9 shows the principle .ow sheet of a modular HTR combined with a steam cycle. In this case, only hot live steam would be produced and expanded in the turbine. Depending on the status of live steam (temperature: 530每 600 ∼C, pressure: 10每32 MPa), subcritical, supercritical, and ultra-supercritical steam turbine which use the state-of-the-art technology in coal-.red power plants can be realized. Total plant ef.ciencies are between 40 and 46%. Several modular units can supply one turbine, which allows the realization of power plants of 200 MWe, 300 MWe and higher values of electrical power. These types 1.4 Application of Modular HTR in the Energy Economy 9 1: reactor, 2: helium circulator, 3: steam generator, 4: turbo machinery,cooling 5: condenser with systems, 6: feed water preheating, 7: feed water storage vessel, 8: feed water pump of power plants containing several reactors, which are combined to one turbine, are proven technology in the .eld of Magnox reactors and in conventional technology. For special sites even dry air-cooling towers can be combined to this plant. In the THTR, an air-cooling tower suited for an amount of waste heat of more than 500 MWth was tested successfully. This allows to build nuclear power plants even in regions with lack of cooling water. The high helium outlet temperature of the reactor allows the coupling with a helium turbine cycle too. Dependent on maximal helium temperature, helium pressure, pressure ratio in the gas turbine, and the relative pressure losses in the cycle, the net ef.ciency of these pro-cesses is between 42 and 46%. Naturally, this process is suited very well to use a dry air-cooling tower for the waste heat removal too. Alternatively an intermediate heat exchanger (IHX) can be applied to decouple the primary and secondary cycles. This can have advantages for the design and operation of turbo-machines, as far as contaminations are considered. The reduction of the cycle ef.ciency by the IHX cycle can be in the order of around 2%. A further improvement of ef.ciencies will be gained by introduction of combined cycles. This solution could include a circuit with or without an intermediate heat exchanger too. Behind the gas turbine, the temperature of the helium is high enough to operate a steam generator, which delivers hot steam of 450每500 ∼C. This allows to operate a conventional steam cycle with high ef.ciencies. One big advantage of the combined process is that the inlet temperature into the reactor has the usual value of 250每 300 ∼C. This corresponds then to the normal design of core internals as used in the steam cycle plant. Dependent on the choice of parameters, combined cycles allow ef.ciencies between 45 and 48%. Besides the production of electrical energy, in future nuclear energy can be applied in the heat market too. It is possible to operate cogeneration plants and to deliver elec-tricity and heat for different purposes. Interesting applica-tions are delivery of district heat, process steam for re.neries and chemical plants and heat for the seawater desalination. For the enhanced oil recovery, which will play a very important role in the next decades, huge amounts of super-heated steam are necessary. Figure 1.10 indicates some important processes and average process temperatures. Cogeneration can be carried out with help of steam turbine cycles and gas turbine systems. High helium temperatures above 750每950 ∼C and higher values allow to carry out special high-temperature processes. Fig. 1.10 Aspects of cogeneration using modular HTR and some applications The conversion of fossil fuels to light hydrocarbons, syn-thesis gas, or hydrogen becomes possible. Methane can be reformed together with steam in an endothermal reaction at temperatures between 650 and 800 ∼C to produce synthesis gas or hydrogen. This steam reforming process requires heat from the helium circuit for the reaction, for process steam production, and for the total energy demand of gas puri.-cation and compression. Figure 1.11 shows an overview on some applications of high-temperature heat, produced by a modular HTR. The possibility to realize modular HTR already in rela-tively small units under economic conditions opens the energy market for this new technology. Especially the new safety concepts of modular HTR allow the coupling with processes mentioned before and make the site election for this new application of nuclear heat easier. Overall, it can be stated that all the processes mentioned before become very attractive, if the requirement to reduce CO2 emissions in the worldwide energy economy will be really ful.lled. Additionally rising costs of fossil fuel will favor the application of nuclear heat in the future. 1.5 Safety Aspects of Modular HTR For each nuclear reactor, a safe enclosure of the radioactive isotopes inside the reactor system is necessary. This is required for normal operation and for accidents [18], to retain the radioactivity inside the fuel, as mentioned before. process temperature range 1. agrio-industrial application 2. seawater desalination 3. district heating 4. process steam (chemical application) 5. steam for refineries 6. steam for tertiary oil production 7. oil shale and oil sand processing 0 200 400 600 Fig. 1.11 Aspects of use of high-temperature heat from a modular HTR process top level temperatures in process chemical equations retorting of oil shale and oil sand Oil fine production steam reforming of methane steam gasification of lignite steam gasification of hard coal thermochemical water splitting 0 200 400 600 800 1000 CnHm C2H4+ CH4+H2O CO+3H2 C+H2O CO+H2 C+H2O CO+H2 H2O 1/2O2+H2 (multistage process) 1.5 Safety Aspects of Modular HTR 11 Additionally a reactor protection system is available. Pro-tection of the investment for the plant is necessary too. Many guidelines, developed for light-water reactors, are used for HTR plants similarly. Especially the barrier system for .s-sion product retention can be realized in such a way that no core melting can occur even in case of extreme accidents in this type of reactors. Following this concept, future reactors must be realized in such a way, that ※no need for evacuation or relocation measures outside the plant§ exist. The goal for a well designed modular HTR is to ful.ll this requirement of the future. Some major generally aspects of safety of modular HTR are summed up below [19每26], details on accidents are explained in Chap. 10: . Because of low power density, dispersed arrangement of fuel and good heat conductivity of the graphite matrix the fuel temperatures in normal operation are relatively low (<1000 ∼C) even for the case of relatively high coolant outlet temperature (750 ∼C). The TRISO-coated particles retain the .ssion products very effectively (release rates <10.5) in normal operation. Therefore the activity of the helium cycle can be relatively low. . The TRISO-coated particles retain important .ssion products very effectively (release rates <10.5) in severe accidents (max. fuel temperature <1600 ∼C). Even higher temperatures might be tolerable at incidents in future using advanced coatings. . Generally because of the large masses of graphite in the core, the reactor behavior is very inert in all transients in normal operation and in accidents up to very extreme situations. . The helium-graphite system is chemically inert and free from corrosion. Only some small impurity contents (ppm of H2O, H2,CO2, CO) in the primary circuit can cause very limited corrosion during normal operation. A gas puri.cation system limits these effects. . The coolant helium does not change phase; therefore, constant conditions for the cooling of core are given. . Helium plays no role in the neutron balance, because the interaction with neutrons is totally negligible. This is very important for the reactivity balance and possible disturbance. . Because of the continuous loading of fuel in the pebble-bed HTR no excess reactivity for burnup com-pensation is necessary. This limits the excess reactivity to relatively small values, depending on the requirements to operate single modules at part load. . Because the core contains large amounts of 238Uor 232Th, there is a very strong negative temperature coef-.cient of the reactivity, which guarantees an inherent safe stabilization of the reactor power in case of power excursions. . In case of hypothetical accidents, characterized by a total loss of coolant and active cooling, the core of HTR behaves very inert. If a low power density (around 3 MW/m3) of the core is chosen, and if the core dimensions are well suited, the core can never melt. The maximal fuel temperature can never exceed 1600 ∼C in modular HTR following the principles mentioned above. . The decay heat is transported from the core to the envi-ronment just by radiation and conduction of heat and free convection of air. These are inherent features of the system which never can fail. Just very few fuel elements get high temperatures in severe accidents (Chap. 11), and therefore, the release of radioactivity is very limited. Details of the concept regarding air ingress and water ingress into the core are discussed in Chap. 11. It should be mentioned already here that solutions based on self-acting principles are possible and allow to limit the consequences of these accidents to relatively small effects. Generally, many safety studies delivered the following results for modular HTR plants: if just less than 10.5 of the total radioactive inventory of a modular HTR can be released to the environment, the radiological consequences outside the nuclear power plants are very small. There will be neither cases of early fatality, nor late fatalities, which can be cor-related to the nuclear accident, no need for evacuation or relocation of people. 1.6 Fuel Cycles of Modular HTR For the HTR different fuel cycles are possible and have been developed to a large extend [27每34]. Table 1.3 gives an overview on the variants. The simplest form of fuel path for HTR contains princi-pally the enrichment of uranium and the fabrication of coated particles and fuel elements, the burnup in the reactor, and after that the intermediate storage of spent fuel elements. Later then conditioning and .nal storage in geological deposits will follow. In this so-called open cycle after a long intermediate storage phase〞more than 50 years appear to be optimal today〞a conditioning of the spent fuel elements will happen and after that the .nal storage in geological deposits. The conditioning can contain further mantling with ceramics and the arrangement of the pebbles in containers. As .nal deposits today salt mines, granite structures, tuff or clay formations are thought to be suited candidates. Ceramic materials offer very good preconditions for .nal storage, due to their excellent corrosion behavior. More ※Open§ cycles Th/U (93%) (HEU) Thorium cycle with uranium (93% enriched) as fuel ThO2/UO2 (93%) Th/U (20%) (MEU) Medium enriched (*20%) Thorium每Uranium cycle ThO2/UO2 (20%) U (8%) (LEU) Low enriched (.8%) UO2 (8%) Closed cycles Th/U Thorium cycle ThO2, 93% enriched UO2 from reprocessed fuel Th/dent. U Denatured cycle ThO2, 20% enriched UO2, U from reprocessing, enriched 15% PB Prebreeder ThO2, 93% enriched UO2 in separate fuel elements NB Net breeder ThO2,UO2 of reprocessed fuel, mainly 233U details on waste management and disposal are contained in Chap. 11. In the pebble-bed reactor, these materials can be used in mixed or separated form. Dependent on the today economic conditions and regarding the situation and perspectives of fuel supply for the next decades, the ※open§ cycles will have advantages for a time schedule of some decades. With rising costs of uranium ore the closed cycles will become inter-esting in far future. In some countries today closed cycles are already considered attractive for different reasons. The recycling of plutonium and the use of thorium are thought to offer advantages under their conditions. The HTR can be operated with different closed fuel cycles. The fuel consists of the nuclides 233U, 235U, 239Pu, 241Pu 234U, 238U, 240Pu as .ssile material and 232Th, as fertile material. In these closed cycles, the fuel elements would be destroyed in a head end process and the fuel would be solved in normal process steps of the PUREX process in case of low enriched uranium cycles. After separation of .ssion prod-ucts, uranium and actinides the last fractions are used in the refabrication to produce new fuel elements. The .ssion products are embedded in glass, and these glass coquilles will again need an intermediate storage, which is similar to that for spent fuel elements, and the storage time should be again 50 years or longer, to get a drastic reduction of decay heat production. The glass coquilles would be .nally stored in deep geological deposits. In all steps of intermediate storage and in the case of direct .nal storage of spent fuel elements the full ceramic structure of coated particles and of the graphite matrix of the fuel element itself are essential preconditions for a safe storage. The process of reprocessing in the future will allow a better use of uranium by a factor of 10 and more compared to today strategies of open fuel cycles. This high use factor of uranium can be realized by introduction of near-breeder systems with reduced burnup and conversion factors near 1 for HTR fuel cycles. Fuel elements for cycles with different enrichment and different breeding materials have been developed and proved in massive tests in AVR and THTR. Fuel elements for cycles with highly enriched Uranium (93%) and Thorium, the so-called HEU cycles, as well as cycles with 8% enriched 238U Uranium with as fertile material (LEU) are fully developed today. Naturally, there are further possibilities of improvements of these fuel elements. Because of the INFCE regulations cycles with more than 20% enrichment today are not accepted because of non-proliferation requirements. This excludes nowadays the use of HEU cycles. For all new HTR projects therefore today LEU cycles are foreseen with around 8每110% enrichment and a burnup of 80,000每100,000 MWd/t of heavy metal. Especially for near-breeders, the HEU cycle with lower burnup is feasible; however, this is an option for a far future, if the uranium ore costs become much higher. The MEU cycle with medium enrichment (<20%) would be still proliferation-resistant and allows high conversion ratios. The pebble-bed HTR generally allows a high fuel cycle .exibility. In the AVR during normal operation, fuel ele-ments corresponding to the different options mentioned above have been tested without any major changes of plant parameters in a continuous change. Fourteen different types of fuel elements have been tested. More details of fuel fab-rication, storage, aspects of breeding and proliferation are explained later in Chap. 11. The new ideas of nuclear waste management, partitioning and transmutation, could also be applied in the HTR fuel 1.6 Fuel Cycles of Modular HTR 13 technology in a far future, if they offer advantages so far as safety and resources saving are considered. Partitioning must allow a .ne separation of .ssion products and uranium as well as of actinides with very low residual content of plu-tonium and minor actinides in the mixture of .ssion prod-ucts. Then, the radiotoxicity of the glass coquilles can be reduced to a large extent and after around 1000 years the danger of the residual waste in the .nal deposit will be smaller than that in a uranium ore mine. However, the additional risks to carry out the processes of partitioning and transmutation have to be taken into account. Some improvements will be possible, if the glass coquille is sub-stituted by a special glass ceramic mixture, which shows an extremely good leaking resistance. 1.7 Intermediate and Final Storage Spent fuel elements, which have been discharged from the reactor, need a safe intermediate storage for a long time. Today in some countries, time spans of 50每100 years are discussed as optimal storage time. If the elements stay for a long time in the intermediate storage, the heat input for the .nal storage becomes much smaller and many requirements in the .nal storage are reduced. The application of cast iron or cast steel vessels for the storage is worldwide available technology today. These vessels can contain as example 50,000 spherical fuel elements corresponding to a volume of nearly 10 m3. The order of 10 kW directly after .lling in spent fuel elements. This amount of energy can be easily transported just by conduction, radiation, and free convec-tion inside the arrangement of pebbles, through the wall of the vessel and given off to air at the outside of the vessel. Fig. 1.12 Principal .ow sheet of the HTR-PM plant (2 ℅250 MWth) [39] Applying this heat transport chain the surface temperature of the vessel stays below a temperature in the pebble bed below 250 ∼C. Some more details on this type of intermediate storage are given in later chapters. Looking at the safety analysis for this storage principle one does not .nd any reasons that large release rates of .ssion products could occur. Just for the case of terroristic attack by large airplanes, which carry large masses of kerosine, special considerations on .res over long time are necessary. Inter-ventions after some time would avoid any problem. The .nal storage of spent HTR fuel elements has been considered until now as direct storage in salt mines. In the meantime granite and clay are in discussion too as geological deposits. Reprocessing in a far future is an additional option for HTR fuel, if Thorium would be applied as breeding material. 1.8 Overview on the HTR-PM Project The HTR-PM (High-Temperature Reactor〞Pebble-bed Module) will be built in China as the demonstration plant for the Chinese development of high-temperature reactors [35每41]. The power of the modular system is 2 ℅250 MWth. The heat of the reactor plant shall be used to produce steam (566 ∼C/13.25 MPa) for the operation of a steam turbine cycle with an electrical power of 212 MW. The cooling of the power cycle shall be done with seawater. The modular HTR will use a pebble-bed core with a LEU enri-ched fuel cycle and TRISO-coated particles. Figure 1.12 contains a simpli.ed .ow sheet for the power plant and shows, that the process shall be operated without reheating in the helium cycle. The steam turbine shall be separated into a high pressure, medium pressure and a low pressure section. The preheating of the feed water will contain 4 stages and allows the pre-heating of the feed water till 205 ∼C. The condensation shall take place at a pressure of around 40 kPa, which is possible using seawater as heat sink for the waste heat. The helium will be heated in the core from 250 to 750 ∼ C at a pressure of 7 MPa. The .ow will be downwards in the core. As the arrangement of the primary system (Fig. 1.13) indicates, the steam generator will be arranged side by side to the reactor vessel and the position of the steam generator heating bundle is geodetically below the core. The connection between the reactor and the steam generator is carried out by a connection vessel, which contains a coaxial hot gas duct. The helium circulator is positioned at the top of the steam generator and the housing forms a part of this component. The steam generator con-sists of 19 modular units arranged in the vessel. Each modular unit is equipped with a sampler for the feed water and for the hot steam. Some important design data of the HTR-PM plant are given in Table 1.4. Some details of the reactor are shown in Fig. 1.14. It is a pebble-bed system with 420,000 spherical fuel elements. The concept of these fuel elements was already explained in Sect. 1.1. The LEU cycle with TRISO-coated particles will be applied. In Table 1.5 some data of the core are given. Compared to designs of modular HTR, known until now, the core is relatively high and therefore the pressure drop is large. 1: core; 2: reflector; 3: shutdown system; 4: discharge system; 5: steam generator ＊s (modules); 6: helium circulator; 7: connecting vessel with coaxial duct; 8: feedwater supply; 9: steam outlet; 10: shock dampers; 11: reactor pressure vessel; 12: steam generator pressure vessel 1.8 Overview on the HTR-PM Project 15 Table 1.4 Some design data of HTR-PM plant Fig. 1.14 Arrangement of the reactor of HTR-PM [39] Parameter Dimension Value Thermal power MW 2 ℅250 Electrical power MW 200 Process 每 Steam turbine without reheat Cooling 每 Seawater Helium cycle ∼C 250每750 ∼C Steam conditions ∼C/MPa 566/13.25 Safety concept (max. fuel temp. in accident) ∼C <1600 1) 1storage vessel of absorber ball system 2) fuel element loading pipe 3) neutron source channel 4) absorber ball conveying system 5) reactor pressure vessel 6) guide keys 7) side reflector 8) thermal shield 9) core barrel 10) core 11) fuel element discharge pipe 12) sealing rings 13) support bearings 14) hot gas chamber 15) hoop strap 16) material radiation surveillance channel 17) rod and drive of 1. shutdown system (a) vertical section (b) horizontal section Table 1.5 Some data of the core of the HTR-PM reactor Parameter Dimension Value Thermal power per core MW 250 Average power density MW/m3 3.2 Core height m 11 Core diameter m 3.0 Average helium temp. outlet ∼C 750 Lifetime of structures Year of full power 30 Position of control + shutdown elem. 每 In re.ector Fuel elements 每 Pebbles Cycle 每 MEDUL The core diameter is chosen at such height to allow the self-acting transport of decay heat to the environment. The value is around 3 m. The limitation of the highest fuel temperature to values <1600 ∼C in the accident ※total loss of active cooling§ can be ful.lled in this reactor, if the thermal power is limited to 250 MW. Furthermore, all control and shutdown elements for the .rst and second system can be arranged in this case in the side re.ector. The helium .owing through the reactor is heated up from 250 ∼C to a value of 750 ∼C as an average. The pressure will be 7 MPa and this helps to limit the impact of the pressure drop in the relatively high core on the total pressure of the helium circuit. For the .rst shutdown system, control rods with electrical drives are foreseen. The second or reserved shutdown sys-tem consists of small absorber balls, which fall under in.uence of gravity into borings in the re.ector. In addition to this system, which must not be a fast acting one from safety reasons, in AVR and in THTR it was possible to make the reactor subcritical over a longtime period by removing fuel elements from the core. The reactor furthermore has a strong negative temperature coef.cient of reactivity. Table 1.6 Some important data of spherical fuel elements for the HTR-PM plant The discharging of the reactor is carried out continuously under full pressure and at normal operation. For this pur-pose, a special discharge system is arranged at the bottom of the reactor pressure vessel. The loading is done as usually carried out in pebble-bed reactors, at the top of the re.ector. The reloading of the core uses a machine consisting of a rotating disk with a hole of 65 mm diameter. A scrap sep-arator and a burnup measurement system additionally are installed to handle the discharged fuel elements. After measurement of the burnup the fuel elements are either recycled into the core or taken out to the interim storage for spent fuel elements. The recycling rate for the HTR-PM is foreseen as 15 times before they reach their speci.ed burnup. This procedure allows additional statistical tests on the status of the fuel elements. Some important parameters of the spherical fuel elements are listed in Table 1.6. They are oriented on values, which have been proven in AVR operation and in the development programs for the HTR until now. The total primary system is arranged inside an inner concrete cell, which is designed as a dense room for an average pressure of 0.25 MPa. The .nal concept of decay Parameter Dimension Value Form of fuel 每 Pebbles (6 cm diameter) Coated particles 每 TRISO (C/SiC/C) Matrix 每 A3/27 graphite Enrichment % 8.5 Burnup MWd/t HM 90,000 Fuel 每 UO2 Cycle 每 MEDUL Max. power per ball kW/ball 3.5 Max. neutron dose n/cm2 <4 ℅1021 Max. temp. of fuel in operation ∼C <1000 1.8 Overview on the HTR-PM Project 17 heat removal (in case of total loss of all active cooling) uses the principal of self-acting transport of heat from the core to the outside of the reactor pressure vessel. From here the decay heat is given to an outer surface cooling system sur-rounding the reactor vessel. All steps of the transport chain for the decay heat use just conduction and radiation of heat and natural convection of air or air/helium mixtures. The inner concrete cell is positioned inside the reactor contain-ment building, which is special for the modular HTR. It is not as dense as in case of LWR, but it offers safety against strong impacts from the outside. There are penetrations through the inner concrete cell for feed water and steam pipes and for the gas puri.cation, the fuel handling and the helium storage system. The helium-containing auxiliary systems are arranged in rooms, which are connected to the inner concrete cell. Figure 1.15 shows the arrangement of two HTR-Modules explaining on behalf of a horizontal section. The vertical section of the building of the HTR-PM is included in Fig. 1.16. The steam generator is arranged side by side to the reactor pressure vessel as this is usually planned for modular HTR system. The helium circulator is placed at the top of the steam generator, and therefore, there is good access to this component for maintenance, repair, and possibly exchange. The feed water and live steam pipes are connected to the steam generator at the side of the vessel, and from there, the transport of the media to the steam tur-bine and from the feed water preheating system occurs. The inner concrete cell for the primary system has thick-walled structures and is practically dense. The outer containment building offers protection against some impact from the outside. The two modules are connected to one machine house containing the steam turbine and the generator, the con-denser, feed water pumps, preheaters, and the feed water storage vessel. The intermediate storage is designed to contain all spent fuel elements of 40 years of operation of the plant. It will be a building containing dry air-cooled storage vessels. The intermediate storage building offers protection against strong impacts from outside. An isometric picture of the total plant, which gives an impression from the total arrangement of the plant, is shown in Fig. 1.17. On the site of the HTR-PM, the deployment of the .rst concrete was .nished on December 12, 2012, and the .rst reactor pressure vessel was installed on March 20, 2016 (Fig. 1.18 and 1.19). 1: reactor; 2: steam generator; 3: connecting vessel; 4: inner concrete cell; 5: outer reactor building; 6: discharge system; 7: shutdown system; 8: crane 1.8 Overview on the HTR-PM Project 19 Fig. 1.18 HTR-PM .rst concrete deployment