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extended control rod article in english by Kay Uwe Böhm

PWR control rod assembly, above fuel element

A control rod is a rod used in nuclear reactors to control the rate of fission of uranium and plutonium. They are made of chemical elements capable of absorbing many neutrons without fissioning themselves, such as boron, silver, indium and cadmium. Because these elements have different capture cross sections for neutrons of varying energies, the compositions of the control rods must be designed for the neutron spectrum of the reactor it is supposed to control. Light water reactors (BWR, PWR) and heavy water reactors (HWR) operate with "thermal" neutrons, whereas breeder reactors operate with "fast" neutrons.

Operation principle

Control rods are usually combined into control rod assemblies — typically 20 rods for a commercial Pressurized Water Reactor (PWR) assembly — and inserted into guide tubes within a fuel element. A control rod is removed from or inserted into the central core of a nuclear reactor in order to control the neutron flux — to increase or decrease the number of neutrons which will split further uranium atoms. This in turn affects the thermal power of the reactor, the amount of steam produced, and hence the electricity generated.

Control rods often stand vertically within the core. In pressurised water reactors (PWR), they are inserted from above, the control rod drive mechanisms being mounted on the reactor pressure vessel head. Due to the necessity of a steam dryer above the core of a boiling water reactor (BWR) this design requires insertion of the control rods from underneath the core. The control rods are partially removed from the core to allow a chain reaction to occur. The number of control rods inserted and the distance by which they are inserted can be varied to control the reactivity of the reactor. Typically shutdown times for modern reactors like European Pressurized Reactor or Advanced CANDU reactor are <2s for >90% reduction limited by decay heat.

Materials used

틀:Copy edit Chemical elements with a sufficiently high capture cross section for neutrons include silver, indium and cadmium. Other elements that can be used include boron, cobalt, hafnium, samarium, europium, gadolinium, terbium, dysprosium, holmium, erbium, thulium, ytterbium, and lutetium or their alloys and compounds, e.g. high-boron steel (just in research reactors because swelling from helium and lithium rise after neutron absorption of boron in a (n, alpha) reaction not same like with just (n, gamma) capturing neutron absorbers), silver-indium-cadmium alloy, boron carbide, zirconium diboride, titanium diboride, hafnium diboride, gadolinium nitrate (injected into D₂O moderator of Advanced CANDU reactor), gadolinium titanate, and dysprosium titanate. The choice of materials is influenced by the energy of neutrons in the reactor, their resistance to neutron-induced swelling, and the required mechanical and lifetime properties. The rods may have the form of tubes filled with neutron absorbing pellets or powder etc. out of stainless steel or other neutron window materials like zirconium, chromium, silicium carbide or c¹¹B¹⁵N cubic boron nitride with very less capturing, low density, unburnable, insoluble, not rising any H₂ like zirconium with water and heat, stable and dense up to 2800°C, melting 2973°C with very high thermal conductivity and moderation like graphite usable also for fuel rods with fuel just backed inside alpha boron nitride. The swelling of the material in the neutron flux can cause deformation of the rod, leading to its premature replacement. The burn up of the absorbing isotopes is another limiting lifetime factor reduced by long capturing isotope rows of the same element or just by not using neutron absorbers for trimming over controlling the nuclear fuel amount at run time for example in pebble bed reactors or in possible new type ⁷lithium moderated and cooled reactors using fuel and absorber pebbles over using less fuel pebbles or empty placebo pebbles or over direct fuel & extraction in and out of ⁷lithium circle using continiously centrifuges etc. for taking out fission products like inside from fission produced samarium etc. reducing also the decay heat after shut down (starting normally with about 6-7% and quickly falling) and maximal possible contamination in accident cases (likely <0.2% compared with about 2% Fukushima and 20% Chernobyl of radioactive inventory set free). The rare earth elements that are excellent neutron absorbers are partly less rare than silver (reserves about 500 000t) like ytterbium (reserves about 1 mio. t) and 400 times more common yttrium with middle capturing values and can be found and used together without separation inside minerals like xenotime(Yb) (Yb0,40Y0,27Lu0,12Er0,12Dy0,05Tm0,04Ho0,01)PO4₄, or Keiviit(Yb) (Yb1,43Lu0,23Er0,17Tm0,08Y0,05Dy0,03Ho0,02)2Si2O7 for cheap price. Also xenon is a strong neutron absorber as gas and can be used for controlling and (emergency) stopping of helium cooled reactors but not working in case of pressure loss or as burning protection gas together with argon around vessel part special in case of core catching reactors or if filled up with sodium or lithium etc. Xenon produced from fission inside reactor can be used after waiting enough time until caesium was falling out with practically no radioactivity left. ⁵⁹Cobalt is used also as absorber for winning of ⁶⁰cobalt for x/ray. Control rods can be constructed also as thick turnable rods with tungsten reflector and absorber side turned to stopp by a spring in less than 1s used around the core of pebble bed reactors etc.

Silver-indium-cadmium alloys, generally 80% Ag, 15% In, and 5% Cd, are a common control rod material for pressurized water reactors. The somewhat different energy absorption regions of the materials make the alloy an excellent neutron absorber. It has good mechanical strength and can be easily fabricated. It has to be encased in stainless steel to prevent corrosion in hot water. Also indium is less rare than silver it is in praxis more expensive.

Boron is another common neutron absorber. Due to different cross sections of ¹⁰B and ¹¹B, boron containing materials enriched in ¹⁰B by isotopic separation are frequently used. The wide absorption spectrum of boron makes it suitable also as a neutron shield. Mechanical properties of boron in its elementary form are unfavourable, therefore alloys or compounds have to be used instead. Common choices are high-boron steel and boron carbide. Boron carbide is used as a control rod material in both pressurized water reactors and boiling water reactors. ¹⁰B/¹¹B separation is done commercially with gas centrifuges over BF₃ but can be done also over BH₃ from borane production or directly with an energy optimized melting centrifuge using the heat of fresh separated boron for preheating.

Hafnium has excellent properties for reactors using water for both moderation and cooling. It has good mechanical strength, can be easily fabricated, and is resistant to corrosion in hot water. Hafnium can be alloyed with small amounts of other elements; e.g. tin and oxygen to increase tensile and creep strength, iron, chromium and niobium for corrosion resistance, and molybdenum for wear resistance, hardness, and machineability. Some such alloys are designated as Hafaloy, Hafaloy-M, Hafaloy-N, and Hafaloy-NM. Its high cost and low availability limit its use in civilian reactors, though it is used in some US Navy reactors. Hafnium carbide can be used also as an insoluble material with a very high melting point of 3890°C and density higher than of uranium dioxide for sinking unmolten through a corium.

Dysprosium titanate is a new material currently undergoing evaluation for pressurized water control rods. Dysprosium titanate is a promising replacement for Ag-In-Cd alloys because it has a much higher melting point, does not tend to react with cladding materials, is easy to produce, does not produce radioactive waste, does not swell, and does not outgas. It was developed in Russia, and is recommended by some for VVER and RBMK reactors. Disadvantage is less absorption of titanium and oxide and also other neutron absorbing elements are not reacting with the right already high melting point cladding materials and with just using the unseparated content also with dysprosium inside of minerals like Keiviit Yb inside chromium, SiC or c11B15N tubes are beating price and absorption also without swelling and gasing out and for highest melting point best using HfC.

Hafnium diboride is another such new material. It can be used standalone or prepared in a sintered mixture of hafnium and boron carbide powders.

Possible are also much more compounds of rare earth elements, samarium etc. with boron like europium and samarium boride already used inside colour industry or compounds of boron with less absorbing like titanium but cheap molybdenum as Mo₂B₅ etc. but all swelling with boron means for practice other compounds are better like carbides etc. or compounds with two or more neutron absorber elements together.

Important is that tungsten likely also other elements like tantalum do have about the same high capturing like hafnium but with opposite effect and that is not explainable only by neutron reflection so the only left explaination are resonance gamma rays increasing the fission and breeding ratio over causing more capturing of uranium etc. over metastabil conditions like for isotope ^235mU with a half time of about 26 min.

Additional means of reactivity regulation

Usually there are also other means of controlling reactivity: In the PWR design a soluble neutron absorber (boric acid) is added to the reactor coolant allowing the complete extraction of the control rods during stationary power operation ensuring an even power and flux distribution over the entire core. This chemical shim, along with the use of burnable neutron poisons within the fuel pellets, is used to assist regulation of the long term reactivity of the core, while the control rods are used for rapid changes to the reactor power (e.g. shutdown and start up). Operators of BWRs use the coolant flow through the core to control reactivity by varying the speed of the reactor recirculation pumps (an increase in coolant flow through the core improves the removal of steam bubbles, thus increasing the density of the coolant/moderator with the result of increasing power).

Safety

In most reactor designs, as a safety measure, control rods are attached to the lifting machinery by electromagnets, rather than direct mechanical linkage. This means that automatically in the event of power failure, or if manually invoked due to failure of the lifting machinery, the control rods will fall, under gravity, fully into the pile to stop the reaction. A notable exception to this fail-safe mode of operation is the BWR which requires the hydraulical insertion of control rods in the event of an emergency shut-down, using water from a special tank that is under high nitrogen pressure. Quickly shutting down a reactor in this way is called scramming the reactor.

Criticality accident prevention

Mismanagement or control rod failure was often the cause or aggravating factor for nuclear accidents, including the SL-1 explosion and the Chernobyl disaster.

The absorption cross section for ¹⁰B (top) and ¹¹B (bottom) as a function of energy

Homogeneous neutron absorbers have often been used to manage criticality accidents which involve aqueous solutions of fissile metals. In several such accidents, either borax (sodium borate) or a cadmium compound has been added to the system. The cadmium can be added as a metal to nitric acid solutions of fissile material; the corrosion of the cadmium in the acid will then generate cadmium nitrate in situ.

In carbon dioxide-cooled reactors such as the AGR, if the solid control rods were to fail to arrest the nuclear reaction, nitrogen gas can be injected into the primary coolant cycle. This is because nitrogen has a larger absorption cross-section for neutrons than carbon or oxygen; hence, the core would then become less reactive.

As the neutron energy increases, the neutron cross section of most isotopes decreases. The boron isotope ¹⁰B is responsible for the majority of the neutron absorption. Boron-containing materials can be used as neutron shields to reduce the activation of objects close to a reactor core.

See also

• Nuclear power
• Nuclear reactor
• Nuclear safety

External links

• Boiling Water Reactor Plant, BWR Technology Program

extended pebble bed reactor article extension in english by Kay Uwe Böhm

Sketch of a pebble-bed reactor.

Graphite pebble for reactor

The pebble-bed reactor (PBR) is a gas-cooled bed of normally graphite or other moderating pebbles like unburnable and insoluble near diamond hard c¹¹B¹⁵N cubic boron nitride pebbles but also fast breeding pebbles are possible out of silicium carbide outside, a thin layer of (depleted) uranium (carbide) and silicium inside in combination with a tungsten side reflector all with a negative temperature coefficient over increased capturing of the mass of graphite on increased temperature with increased collisions not only from increased capturing of ²³⁸uranium or ²³²thorium or with more effect silicium also addable as silicium carbide as burning protection layer around and inside as thick TRISO layer. Possible are also BeO beryllium oxide pebbles but beryllium is too rare and expensive for the needed mass and pure ¹¹B is also addable inside C¹¹B¹⁵N for more moderation like SiC for less moderation and increased capturing for decreased maximum temperature. The high masses are slowing down also temperature changes and are increasing the internal to external surfaces for also passive heat radiation and are decreasing the power density so that there is near no decay heat problem. It is a type of very-high-temperature reactor (VHTR), also one of the six classes of nuclear reactors in the Generation IV initiative. The basic design of pebble-bed reactors features spherical fuel elements called pebbles. These tennis ball-sized pebbles are made of graphite and pyrolytic graphite (which acts as the moderator), and they contain thousands of micro-fuel particles called TRISO or new QUADRISO particles. These TRISO fuel particles consist of a fissile material (such as ²³⁵U) and maybe breedable thorium surrounded by a coated ceramic layer of silicon carbide for structural integrity and fission product containment. In the PBR, thousands of pebbles are amassed to create a reactor core, and are cooled by a gas, such as helium (density 0.1786 g/l), specific heat capacity 5193 J/(kg · K), molar 20.786 J/mol·K), ¹⁵nitrogen₂ (without ¹⁴N(n, p)¹⁴C reaction, molar heat capacity 29.124 J·mol−1·K−1, chemical not inert), neon (density 0.9g/l, chemical inert and not going radioactive with same specific molar heat capacity like helium as single atom gas) or carbon dioxide(but with high pressure good for turbines in case of direct gas turbines but not for security in overheat cases), which does not react chemically with the fuel elements.

This type of reactor is claimed to be passively safe; that is, it removes the need for redundant, active safety systems. Because the reactor is designed to handle high temperatures, it can cool by natural circulation and still survive in accident scenarios, which may raise the temperature of the reactor to 1,600 °C. Because of its design, its high temperatures allow higher thermal efficiencies than possible in traditional nuclear power plants (up to 50%) and has the additional feature that the gases do not dissolve contaminants or absorb neutrons as water does, so the core has less in the way of radioactive fluids.

The concept was first suggested by Farrington Daniels in the 1940s, but commercial development did not take place until the 1960s in the German AVR reactor by Rudolf Schulten. but this system was plagued with problems and political and economic decisions were made to abandon the technology. The AVR design was licensed to South Africa as the PBMR and China as the HTR-10, the latter currently the only such design operational with 18 HTR-PM in construction. Also in development Areva Antares in cooporation with Generation IV initiative with indirect over IHX intermediate heat exchange direct gas turbine and added water steam turbines. In various forms, other designs are under development by MIT, University of California at Berkeley, General Atomics (U.S.), the Dutch company Romawa B.V., Adams Atomic Engines, and Idaho National Laboratory.

Pebble-bed design

A pebble-bed power plant combines a gas-cooled core and a novel packaging of the fuel that dramatically reduces complexity while improving safety.

The uranium, thorium or plutonium nuclear fuels are in the form of a ceramic (usually oxides or carbides) contained within spherical pebbles a little smaller than the size of a tennis ball and made of pyrolytic graphite, which acts as the primary neutron moderator. The pebble design is relatively simple, with each sphere consisting of the nuclear fuel, fission product barrier, and moderator (which in a traditional water reactor would all be different parts). Simply piling enough pebbles together in a critical geometry will allow for criticality.

The pebbles are held in a vessel, and an inert gas (such as helium, nitrogen or carbon dioxide) circulates through the spaces between the fuel pebbles to carry heat away from the reactor. If helium is used, because it is lighter than air, air can displace the helium if the reactor wall is breached combined with pressure and cooling loss leading to increased temperature with example 2300°C for THTR-300 after calculations from Forschungszentrum Jülich. Pebble-bed reactors need fire-prevention features to keep the graphite of the pebbles from burning in the presence of air although the flammability of the pebbles is disputed. Ideally, the heated gas is run directly through a turbine but that means a direct way for air coming inside causing burning and pressure loss also from the turbine if getting any hole and with no in principle possible higher efficiency and more already reachable using classical overcritical water steam turbines with the high temperatures from a VHTR transfering the heat over preheated lithium divide circles with heat exchange all around the pebble bed over double melting point increased (W/Mo) steel walls with ribs and lithium flowing inside and cBN plates with ribs for surface increasement to helium and/or cBN tubes inclosed with thick low melting concrete (for stabilization, isolation and as melting buffer) over additives like in glass production and again steel with ribs outside round shaped and black with no further containment needed also for turbines and advantage of high boiling point 1340°C, high critical point 2950°C, melting 180°C also not causing overpressure in case of inflow like water, high specific heat capacity over a huge usable temperature difference, no expansion on solidification like relative rare bismuth, el. magn. pumpable because paramagnetic, relative cheap with about 5 mio. $ for 100 000l and 28 mio. t reserves. Again increasable with new Kay turbines. Additionally not addable with direct gas trubine to two stopp systems activated <1s also passive over melting and mercury swap switches inside working without electricity a stoppsand system with normally cold still exhaustable sand (SiO₂ melting 1715°C boiling 2200°C and neutron absorbers) el.magn. holded from flaps upon the pebble bed as reserve also in case of pressure loss, strong quakes, airplane crash, tank shooting or bunker blaster etc. However, if the gas from the primary coolant can be made radioactive by the neutrons in the reactor, or a fuel defect could still contaminate the power production equipment, it may be brought instead to a heat exchanger where it heats another gas or produces steam. The exhaust of the turbine is quite warm and may be used to warm buildings or chemical plants, or even run another heat engine.

Much of the cost of a conventional, water-cooled nuclear power plant is due to cooling system complexity. These are part of the safety of the overall design, and thus require extensive safety systems and redundant backups with the need of keeping relative cool because else causing overpressure without the possibilty of passive heat radiation like possible with some VHTR. A water-cooled reactor is generally dwarfed by the cooling systems attached to it. Additional issues are that the core irradiates the water with neutrons causing the water and impurities dissolved in it to become radioactive and that the high-pressure piping in the primary side becomes embrittled and requires continual inspection and eventual replacement.

In contrast, a pebble-bed reactor is gas-cooled, sometimes at low pressures of about 10bar for AVR Jülich, 40bar for THTR or 90bar for Escom PBMR. The spaces between the pebbles form the "piping" in the core. Since there is no piping in the core and the coolant contains no hydrogen, embrittlement is not a failure concern. The preferred gas, helium, does not absorb neutrons or impurities. Therefore, compared to water, helium is both more efficient as coolant, chemical totally inert not even reacting with fluorine and not becoming itself radioactive but maybe there are outgoing fission products inside helium depending also on pebble density and cleaning of the inner helium streaming of diffusing tritium (also cehmical bindable with Li etc.) and single atom gases xenon, krypton etc. always to be taken out by density differences inside slow flow side chamber to helium or chemical binding of tritium with lithium etc. like in water reactors. Usefull are also helium chimneys as tubes with holes between the pebble bed for better thermal streaming through with addable blowers and more equal temperatures usable also as placeholder for inner control or stopp rods falling down spring accelerated holded with a tungsten rope before el. magn. holded instead pressed through the pebble bed like for THTR causing problems additionally to control rods placed around normally inside the graphite side reflector or thick turnable rods with tungsten reflector and absorber side turned to stopp <1s by spring also in combination with a tungsten side reflector more thin than with graphite and unburnable decreasing the size of the construction like also more dense c¹¹B¹⁵N pebbles saving costs.

Safety features

An advantage of the pebble-bed reactor over a conventional light-water reactor is in operating at higher temperatures. A technical advantage is that some designs are throttled by temperature, not by control rods. The reactor can be simpler because it does not need to operate well at the varying neutron profiles caused by partially withdrawn control rods.

Pebble-bed reactors are also capable of using fuel pebbles made from different fuels in the same basic design of reactor (though perhaps not at the same time). Proponents claim that some kinds of pebble-bed reactors should be able to use thorium(prooved by THTR-300 allowing also very long pebble run time up to 30 years like also possible using fast breeding pebbles or fast breeders like Energy Multiplier Module from General Atomics without pebble bed security or Lead-bismuth cooled fast reactor instead just 3 years and 10 times through but did not proove the full thorium fuel cycle also maybe possible using thorium instead boron trimming pebbles like Shippingport Reactor, Molten Salt Reactor and possible future types like ⁷lithium cooled and moderated reactor), plutonium and natural unenriched uranium, as well as the customary enriched uranium. There is a project in progress to develop pebbles and reactors that use MOX fuel, that mixes uranium with plutonium from either reprocessed fuel rods or decommissioned nuclear weapons.

In most stationary pebble-bed reactor designs, fuel replacement is continuous. Instead of shutting down for weeks to replace fuel rods, pebbles are placed in a bin-shaped reactor. A pebble is recycled from the bottom to the top about ten times over a few years, and tested each time it is removed. When it is expended, it is removed to the nuclear waste area, and a new pebble inserted. Possible is also trimming over empty placebo pebbles, absorber pebbles, depleted uranium or thorium pebbles and adding of moderation pebbles in case of a fast breeding SiC-(U)-Si pebble bed.

When the nuclear fuel increases in temperature, the rapid motion of the atoms in the fuel causes an effect known as Doppler broadening. The fuel then sees a wider range of relative neutron speeds. Uranium-238, which forms the bulk of the uranium or thorium in case of the THTR-300 in the reactor, is much more likely to absorb fast or epithermal neutrons at higher temperatures. This reduces the number of neutrons available to cause fission, and reduces the power of the reactor. Doppler broadening therefore creates a negative feedback because as fuel temperature increases, reactor power decreases. All reactors have reactivity feedback mechanisms, but the pebble-bed reactor is designed so that this effect is very strong and does not depend on any kind of machinery or moving parts. Because of this, its passive cooling, and because the pebble-bed reactor is designed for higher temperatures, the pebble-bed reactor can passively reduce to a safe power level in an accident scenario. This is the main passive safety feature of the pebble-bed reactor, and it makes the pebble-bed design (as well as other very-high-temperature reactors) unique from conventional light water reactors which require active safety controls. Additionally also the mass of the graphite moderator is capturing neutrons on increased temperatures and that happens also the barns values for capturing of graphite ¹²C are relative low over the very high collision numbers with already 114 collisions needed for reaching thermal speed with first from top to down in neutron energy reaching a breeding surplus area before fission is increasing but still less than breeding with a clear fission surplus reached first with very less neutron energy and that effect is working also with ¹¹B inside c¹¹B¹¹N and stronger over silicium addable also as SiC burning protection layer or inside pebbles as more Triso or Quadriso SiC. If the fission would be just reduced by uranium-238 the pebble bed would reach the same temperature like without graphite at an enrichment of about 8-10% but in every case there is also a reduction of uranium-238 else it would be an atomic bomb. With fast breeding SiC-(U238)-SI pebbles combined with a tungsten reflector the neutrons are on every collison with silicium less slowed down staying longer time inside the breeding surplus area additionally the number of collisions with uranium are increased over reduction of the silicium amount compared to the uranium amount and that is also needed for keeping the reactor running together with more uranium-235 compared to uranium-238 but there is also less silicium needed than graphite for the same reduction effect on increased temperatures because of much higher barns capturing values for silicium than graphite at very less neutron energy with disadvanatge also an increased capturing area before but also increased in capturing with increased temperature. The same mass of graphite inside a pebble bed than uranium has already more collisions because of more nucleoses for the same mass and it`s collected nucleoses projection surfaces with substraction of increased gravity effects of more huge uranium nucleoses and additionally there is about 20 times the mass of graphite than uranium inside a pebble bed causing normally staying near an over moderation with near too much collisions and most fission happening in the very less neutron energy area with a clear fission surplus there over breeding if not captured already by the graphite and that makes also the increased fission on less high temperatures forming the negative temperature coefficient not explainable only over capturing of <sup>238</sup>U. <sup>232</sup>Th is also strong capturing like <sup>238</sup>U but starting with increased capturing a little later than <sup>238</sup>U from top to down and reaching a little higher barns values at very less neutron speed so a thorium breeding increasement is reachable also over slow neutrons and much adding of thorium possible over high moderation like with <sup>7</sup>FLiBe or <sup>7</sup>Li. The <sup>238</sup>U or <sup>232</sup>Th is also added as a reflector breeding blanket around in fast or slow breeders with a fission area between with never neutrons lost there always causing breeding or after breeding fissions of <sup>233</sup>U or plutonium cooled with sodium inside sodium breeders with nearly exact the same capturing than silicium causing neutron loss and making also an increased capturing on increased temperature but not enough with not enough sodium and enough temperature limited by the boiling point if inserted inside pebbles of 883°C instead 3265°C for silicium. Silicium is also not melting at normal running temperatures of a pebble bed with about 900°C with a melting point of 1414°C. Silicium has also less density than graphite increasing the volume an internal to external surfaces keeping itself unmolten nuclear fuel enclosed, if molten the nuclear fuel is just sinking down inside the pebbles and if the SiC shell is opend to air it reacts exotherm to not spreading Si<sup>3</sup>N<sup>4</sup> or slowly to not spreading sand SiO<sup>2</sup> at 850-1200°C but building out also passivation layers. A thin depleted uranium layer inside SiC-Si pebbles is working partly like a sodium fast breeder reflector blanket but is letting most neutrons through to the other pebbles so that the pebbles are still needing neutrons from neighbor pebbles and all effects are forwarded through the pebble bed including slow down from side neutron absorbers rods. SiC and cBN are also normally not bursting if very quickly cooled down maybe from water steam coming inside and both are insoluble in water, both with very high thermal conductivity with best values for cBN with 740 W/(m*K), both unburnable and with very high melting points SiC 2730°C or after newest datas 3070°C and cBN 2973°C with remarkable thermal stability of cBN up to 2800°C stable and dense inside inert gas and harder than diamond if both are hot. Graphite is plastical formable >2500°C, porous, burnable >600°C in air, not abrasion proof and not very hard and pyrolitic graphite is just somewhat more dense after heating up near to it`s melting point building up that way more chemical connections between the graphite layers but never as dense like SiC or cBN. Graphite is going also radioactive adding up neutrons to ¹⁴carbon not happening much with boron-11 changing to ¹²carbon after capturing or ¹⁵nitrogen with very less capturing changing to ¹⁶oxygen. c¹¹B¹⁵N is ideal also for fuel rods with nuclear fuel just backed inside enough without SiC fuel pellets coating or graphite expansion room like inside QUADRISO needed and never H² rise as with zircalloy and also usable for control rods with neutron absorbers placed inside tubes as excellent thin layer neutron window material with moderation about graphite. cBN price is about 100$/kg and about 10 mio. t boron world reserve. For a big pebble bed with about 1GWth power, about 1 mio. 6cm 200g pebbles delivering about 1kW at run time and with starting decay heat, after stopping about <70W down to 50W after 10s down to 40W after 1min down to 20W after 10min, 200t moderator mass needed showing also the security together with the internal to external surfaces of the pebble bed for passive heat removal about for a cube 6*6*6 = 216m² and with 1m every side for helium flow around behind a tungsten alloy reflector 8*8*8 = 512m² wall surfaces for passive and active cooling with lithium divide circles etc. means that high melting point pebbles can`t melt after stopping from decay heat and also normally not from maximum heat without stopping but limitation from negative temperature coefficient of an intact pebble bed, not intact if burned down with addiotionally sinking down fuel. Costs for 200t cBN are about 20 mio. $ additionally to boron and nitrogen separation costs but also graphite pebbles and fuel rods are not cheap in praxis and run time with thorium is very long and the pebbles alone are making secure already but boron should be long time recycled. Recommended is normally not to use 93% uranium enrichment for substitution of ²³⁸U with ²³²Th like done with THTR-300 because that means weapon grade uranium that can be taken out and cheaper about 20% enrichment is enough with about per pebble 1g ²³⁵U 4g ²³⁸ 6g ²³²Th burning up also more than three times of U238 than a normal uranium pebble bed because of extended runtime over Th. ¹⁰B separation from ¹¹B called also depleted boron is done commercially with gas centrifuges over BF₃ but can be done also over BH₃ from borane production maybe usable directly for cBN production like BCl₃ or directly with an energy optimized melting centrifuge using the heat of fresh separated boron for preheating. ¹¹B is also usable for U¹¹B₂ fuel changing over Np¹¹B₂ to Pu¹¹B₂ as harmless secure chemical state used also for long time storage without chemical and radioactive toxicity like for dangerous pure Pu. ¹⁵N₂ free out of air is easy separatable in gas centrifuge with >3% density difference compared with <1% for UF₆ or maybe available already separated where all parts of air are separated for commercial selling like with Linde process.

The pebble bed design should take care that pebbles are not damaged mechanically and not blocked anyway like happend with THTR-300 boron trimming pebbles with a testing tube before etc. Pebble inlet at top and outlet at bottom should be with 2 gate steps and at top inlet taken care also for equal spreading and bottom outlet could be with a screw moving the pebbles sidewards out later computer controlled automatically measured out and reinserted or substituted etc. Always if the construction is needing a construction hole there sould no gas go out causing a pressure loss etc. also if it can manage secure a pressure loss and no significant radioactivity is going out also fully opend in case of dense not burning pebbles.

The reactor is cooled by an inert, fireproof gas, so it cannot have a steam explosion as a light-water reactor can. The coolant has no phase transitions—it starts as a gas and remains a gas. Similarly, the moderator is solid carbon; it does not act as a coolant, move, or have phase transitions (i.e., between liquid and gas) as the light water in conventional reactors does.

A pebble-bed reactor thus can have all of its supporting machinery fail, and the reactor will not crack, melt, explode or spew hazardous wastes. It simply goes up to a designed "idle" temperature, and stays there. In that state, the reactor vessel radiates heat, but the vessel and fuel spheres remain intact and undamaged. The machinery can be repaired or the fuel can be removed. These safety features were tested (and filmed) with the German AVR reactor. All the control rods were removed, and the coolant flow was halted. Afterward, the fuel balls were sampled and examined for damage and there was none.

PBRs are intentionally operated above the 250 °C annealing temperature of graphite, so that Wigner energy is not accumulated. This solves a problem discovered in an infamous accident, the Windscale fire. One of the reactors at the Windscale site in England (not a PBR) caught fire because of the release of energy stored as crystalline dislocations (Wigner energy) in the graphite. The dislocations are caused by neutron passage through the graphite. At Windscale, a program of regular annealing was put in place to release accumulated Wigner energy, but since the effect was not anticipated during the construction of the reactor, and since the reactor was cooled by ordinary air in an open cycle, the process could not be reliably controlled, and led to a fire. The 2nd generation of UK gas-cooled reactors, the AGRs, also operate above the annealing temperature of graphite.

Berkeley professor Richard A. Muller has called pebble-bed reactors "in every way ... safer than the present nuclear reactors".

Containment

Most pebble-bed reactors contain many reinforcing levels of containment to prevent contact between the radioactive materials and the biosphere.

1. Most reactor systems are enclosed in a containment building designed to resist aircraft crashes and earthquakes.
2. The reactor itself is usually in a two-meter-thick-walled room with doors that can be closed, and cooling plenums that can be filled from any water source.
3. The reactor vessel is usually sealed.
4. Each pebble, within the vessel, is a 60 밀리미터 (2.4 in) hollow sphere of pyrolytic graphite.
5. A wrapping of fireproof silicon carbide
6. Low density porous pyrolytic carbon, high density nonporous pyrolytic carbon
7. The fission fuel is in the form of metal oxides or carbides

Pyrolytic graphite is the main structural material in these pebbles. It sublimates at 4000 °C, more than twice the design temperature of most reactors. It slows neutrons very effectively, is strong, inexpensive, and has a long history of use in reactors. Its strength and hardness come from anisotropic crystals of carbon. Pyrolytic graphite is also used, unreinforced, to construct missile reentry nose-cones and large solid rocket nozzles.

Pyrolytic carbon starts to burn in air like anthracite coal >600°C (diamond ignition point about 720-800°C in oxygen and 850–1,000°C in air) and temperature increased much over already high normal running temperature about 900°C because normally if air is coming inside with additional pressure loss. Infamous examples include the accidents at Windscale and Chernobyl—both graphite-moderated reactors. Some engineers insist that pyrolytic carbon cannot burn in air, and cite engineering studies of high-density pyrolytic carbon in which water is excluded from the test. However, all pebble-bed reactors are cooled by inert gases to prevent fire. No existing pebble design did have already any burning protection layer that should be also on, if used, the graphite reflector etc. or/and any fire erasing possibility or other unburnable materials should be used like c¹¹B¹⁵N and tungsten reflector like used inside not already build THOR reactor design from Kay Uwe Böhm. Infamous examples include the accidents at Windscale and Chernobyl—both graphite-moderated reactors. However, all pebble-bed reactors are cooled by inert gases to prevent fire. No existing pebble design did have already any burning protection layer only extreme thin SiC layers against outgoing of fission products instead on pebble surfaces and if used the graphite reflector etc. or/and any fire erasing possibility or using other unburnable materials like c¹¹B¹⁵N and tungsten reflector like used inside not already build THOR reactor design from Kay Uwe Böhm together with other improvements added into this article for reaching real Full Zero Risk all cases at lowest price possible. If the outer containment is anyway broken the pebbles out of c¹¹B¹⁵N can still hold all fission products secure inside because not melting, not burning, not soluble, near diamond hard not breaking and if one pebble is broken from about one million pebbles just a little can come out on the breaking surface because all is backed inside and cannot be spreaded with helium and sand so also tank shooting and bunker blasters cannot cause a radioactive catastrophe. A pebble bed outside is normally going out because more flat and without side neutrons back without alpha and beta rays coming out but gamma rays and against gamma rays just enough distance is to be kept already done by the place around because the distance is decreasing in square with fresh pebbles undangerous. together with other improvements added into this article for reaching real Full Zero Risk all cases at lowest price possible. If the outer containment is anyway broken the pebbles out of c¹¹B¹⁵N can still hold all fission products secure inside because not melting, not burning, not soluble, near diamond hard not breaking and if one pebble is broken from about one million pebbles just a little can come out on the breaking surface because all is backed inside and cannot be spreaded with helium and sand so also tank shooting and bunker blasters cannot cause a radioactive catastrophe. A pebble bed outside is normally going out because more flat and without side neutrons back without alpha and beta rays coming out but gamma rays and against gamma rays just enough distance is to be kept already done by the place around because the distance is decreasing in square with fresh pebbles undangerous.

Production of fuel

Most authorities agree (2002) that German fuel-pebbles release about three orders of magnitude (1000 times) less radioactive gas than the U.S. equivalents. All kernels are precipitated from a sol-gel, then washed, dried and calcined. U.S. kernels use uranium carbide, while German (AVR) kernels use uranium dioxide.

Criticisms of the reactor design

Combustible graphite

The most common criticism of pebble-bed reactors is that encasing the fuel in combustible graphite poses a hazard. When the graphite burns, fuel material could potentially be carried away in smoke from the fire. Since burning graphite requires oxygen, the fuel pebbles should have a burning protection layer for example SiC on the pebble surfaces not only an extreme thin layer out of SiC not sure building out a Si_2> passivation layer or pyrophoric ZrC (Areva Antares) inside on about 15 000 0.9mm TRISO or QUADRISO sub particles for a 6 cm pebble, and the reaction vessel should stay purged of oxygen. While silicon carbide is strong in abrasion and compression applications, it does not have the same strength against expansion and shear forces. Some fission products such as xenon-133 have a limited absorbance in carbon, and some fuel kernels could accumulate enough gas to rupture the silicon carbide layer. Even a cracked pebble will not burn without oxygen, but the fuel pebble may not be rotated out and inspected for months, leaving a window of vulnerability.

Containment building

Some designs for pebble-bed reactors lack a containment building, potentially making such reactors more vulnerable to outside attack and allowing radioactive material to spread in the case of an explosion. However, the current emphasis on reactor safety means that any new design will likely have a strong reinforced concrete containment structure. Also, any explosion would most likely be caused by an external factor, as the design does not suffer from the steam explosion-vulnerability of some water-cooled reactors.

Waste handling

Since the fuel is contained in graphite pebbles, the volume of radioactive waste is much greater, but contains about the same radioactivity when measured in becquerels per kilowatt-hour. In principle the fission products can be taken out of the pebbles for same processing like with water reactor waste reducing the volume of the very radioactive waste and always easy pebble dry storage possible also long time about 30m deep far outside in deserts and the pebble materials can be recycled for reusing in pebble beds. The waste tends to be less hazardous and simpler to handle. Current US legislation requires all waste to be safely contained, therefore pebble-bed reactors would increase existing storage problems. Defects in the production of pebbles may also cause problems. The radioactive waste must either be safely stored for many human generations, typically in a deep geological repository, reprocessed, transmuted in a different type of reactor, or disposed of by some other alternative method yet to be devised. The graphite pebbles are more difficult to reprocess due to their construction, which is not true of the fuel from other types of reactors. Proponents point out that this is a plus, as it is difficult to re-use pebble-bed reactor waste for nuclear weapons.

1986 accident

Critics also often point out an accident in Germany in 1986, which involved a jammed pebble damaged by the reactor operators when they were attempting to dislodge it from a feeder tube (see THTR-300 section). This accident released radiation into the surrounding area, and probably was one reason for the shutdown of the research program by the West German government.

2008 report

In 2008, a report about safety aspects of the AVR reactor in Germany and some general features of pebble-bed reactors have drawn attention. The claims are under contention. Main points of discussion are

• No possibility to place standard measurement equipment in the pebble-bed core, i.e. pebble bed = black box
• Contamination of the cooling circuit with metallic fission products (Sr-90, Cs-137) due to the insufficient retention capabilities of fuel pebbles for metallic fission products. Even modern fuel elements do not sufficiently retain strontium and cesium.
• improper temperatures in the core (more than 200 °C above calculated values)
• necessity of a pressure retaining containment
• unresolved problems with dust formation by pebble friction (dust acts as a mobile fission product carrier)

Rainer Moormann, author of the report, requests for safety reasons a limitation of average hot Helium temperatures to 800 °C minus the uncertainty of the core temperatures (which is at present at about 200 °C).

The pebble-bed reactor has an advantage over traditional reactors in that the gases do not dissolve contaminants or absorb neutrons as water does, so the core has less in the way of radioactive fluids. However, the pebbles generate graphite particulates that can blow through the coolant loop and will absorb fission products if fission products escape the TRISO particles.

Limited practical experience

There is significantly less experience with production-scale pebble-bed reactors than light water reactors. As such, claims made by both proponents and detractors are more theory-based than based on practical experience.

History

The first suggestion for this type of reactor came in 1947 from Prof. Dr. Farrington Daniels at Oak Ridge, who also created the name "pebble-bed reactor". The concept of a very simple, very safe reactor, with a commoditized nuclear fuel was developed by Professor Dr. Rudolf Schulten in the 1950s. The crucial breakthrough was the idea of combining fuel, structure, containment, and neutron moderator in a small, strong sphere. The concept was enabled by the realization that engineered forms of silicon carbide and pyrolytic carbon were quite strong, even at temperatures as high as 2000 °C (3600 °F). The natural geometry of close-packed spheres then provides the ducting (the spaces between the spheres) and spacing for the reactor core. To make the safety simple, the core has a low power density, about 1/30 the power density of a light water reactor.

Germany

AVR

 이 부분의 본문은 AVR reactor입니다.

AVR in Germany.

A 15 MW_e demonstration reactor, Arbeitsgemeinschaft Versuchsreaktor (AVR translates to experimental reactor consortium), was built at the Jülich Research Centre in Jülich, West Germany. The goal was to gain operational experience with a high-temperature gas-cooled reactor. The unit's first criticality was on August 26, 1966. The facility ran successfully for 21 years, and was decommissioned on December 1, 1988, in the wake of the Chernobyl disaster and operational problems. During removal of the fuel elements it became obvious that the neutron reflector under the pebble-bed core had cracked during operation. Some hundred fuel elements remained stuck in the crack. During this examination it became also obvious that the AVR is the most heavily beta-contaminated (strontium-90) nuclear installation worldwide and that this contamination is present in the worst form, as dust. In 1978 the AVR suffered from a water/steam ingress accident of 30 metric tons, which led to contamination of soil and groundwater by strontium-90 and by tritium. The leak in the steam generator, leading to this accident, was probably caused by too high core temperatures (see criticism section). A re-examination of this accident was announced by the local government in July, 2010.

The AVR was originally designed to breed uranium-233 from thorium-232. Thorium-232 is about 400 times as abundant in the Earth's crust as uranium-235, and an effective thorium breeder reactor is therefore considered valuable technology. However, the fuel design of the AVR contained the fuel so well that the transmuted fuels were uneconomic to extract—it was cheaper to simply use natural uranium isotopes.

The AVR used helium coolant. Helium has a low neutron cross-section. Since few neutrons are absorbed, the coolant remains less radioactive. In fact, it is practical to route the primary coolant directly to power generation turbines. Even though the power generation used primary coolant, it is reported that the AVR exposed its personnel to less than 1/5 as much radiation as a typical light water reactor.

The localized fuel temperature instabilities mentioned above in the criticism section resulted in a heavy contamination of the whole vessel by Cs-137 and Sr-90. Some contamination was also found in soil/groundwater under the reactor, as the German government confirmed in January, 2010. Thus the reactor vessel was filled with light concrete in order to fix the radioactive dust and in 2012 the reactor vessel of 2100 metric tons will be airlifted to an intermediate storage. There exists currently no dismantling method for the AVR vessel, but it is planned to develop some procedure during the next 60 years and to start with vessel dismantling at the end of the century. In the meantime, after transport of the AVR vessel into the intermediate storage, the reactor buildings will be dismantled and soil and groundwater will be decontaminated. AVR dismantling costs will exceed its construction costs by far. In August 2010 the German government published a new cost estimate for AVR dismantling, however without consideration of the vessel dismantling: An amount of 600 million € ( $750 million) is now expected (200 million € more than in an estimate of 2006), which corresponds to 0.4 € ($0.55) per kWh of electricity generated by the AVR. Consideration of the unresolved problem of vessel dismantling is supposed to increase the total dismantling costs to more than 1 bn €. Construction costs of AVR were 115 million Deutschmark (1966), corresponding to a 2010 value of 180 million €. A separate containment was erected for dismantling purposes, as seen in the AVR-picture.

Thorium high-temperature reactor

 이 부분의 본문은 THTR-300입니다.

Following the experience with AVR, a full scale power station (the thorium high-temperature reactor or THTR-300 rated at 300 MW) was constructed, dedicated to using thorium as the fuel. THTR-300 suffered a number of technical difficulties, and owing to these and political events in Germany, was closed after only four years of operation. One cause for the closing was an accident on 4 May 1986 with a limited release of the radioactive inventory into the environment. Although the radiological impact of this accident remained small, it is of major relevance for PBR history. The release of radioactive dust was caused by a human error during a blockage of pebbles in a pipe. Trying to restart the pebbles' movement by increasing gas flow led to stirring up of dust, always present in PBRs, which was then released, radioactive and unfiltered, into the environment due to an erroneously open valve.

In spite of the limited amount of radioactivity released (0.1 GBq ⁶⁰Co, ¹³⁷Cs, ²³³Pa), the THTR management tried to hide the accident, possibly because this accident pointed to some specific problems with pebble-bed reactors, mostly pebble flow and radioactive dust. The management might have thought that the emission would not be detectable due to the Chernobyl fallout happening at the same time. They continued to blame the Chernobyl fallout for all of the contamination found in the surroundings, until the presence of Pa-233 in the vicinity was detected. The radioactivity in the vicinity of the THTR-300 was finally found to result 25% from Chernobyl and 75% from THTR-300. The handling of this minor accident severely damaged the credibility of the German pebble-bed community, and pebble-bed reactors lost a lot of support in Germany.

The reactor also suffered from the unplanned high destruction rate of pebbles during normal operation, and the resulting higher contamination of the containment structure, and problems with compact pebble allocations, which caused deformations in the control rods and of the side reflector arrangement. Ammonia, which was added to helium as lubricant for core rods moving in the pebble bed, was found to cause unacceptable corrosion on metallic components. Pebble debris and graphite dust blocked some of the coolant channels in the bottom reflector, as was discovered during fuel removal some years after final shut-down. A failure of insulation required frequent reactor shut-downs for inspection, because the insulation could not be repaired. Further metallic components in the hot gas duct failed in September 1988, probably due to thermal fatigue induced by unexpected hot gas currents. This failure led to a long-term shut-down for inspections. In August, 1989 the THTR company almost went bankrupt, but was financially rescued by the government. Because of the unexpected high costs of THTR operation, and this accident, there was no longer any interest in THTR reactors. The government decided to terminate the THTR operation at the end of September, 1989.

Different designs

China

China has licensed the German technology and is actively developing a pebble-bed reactor for power generation. The 10 megawatt prototype is called the HTR-10. It is a conventional helium-cooled, helium-turbine design. The program is at Tsinghua University in Beijing. The first 250-MW plant is scheduled to begin construction in 2009 and commissioning in 2013. There are firm plans for thirty such plants by 2020 (6 gigawatts). By 2050, China plans to deploy as much as 300 gigawatts of reactors of which PBMRs will be a major component. If PBMRs are successful, there may be a substantial number of reactors deployed. This may be the largest planned nuclear power deployment in history.

Tsinghua's program for nuclear and new energy technology also plans in 2006 to begin developing a system to use the high temperature gas of a pebble-bed reactor to crack steam to produce hydrogen. The hydrogen could serve as fuel for hydrogen vehicles, reducing China's dependence on imported oil. Hydrogen can also be stored, and distribution by pipelines may be more efficient than conventional power lines. See hydrogen economy.

South Africa

 이 부분의 본문은 PBMR입니다.

In June 2004, it was announced that a new PBMR would be built at Koeberg, South Africa by Eskom, the government-owned electrical utility. There is opposition to the PBMR from groups such as Koeberg Alert and Earthlife Africa, the latter of which has sued Eskom to stop development of the project. In September 2009 the demonstration power plant was postponed indefinitely. In February 2010 the South African government stopped funding of the PBMR because of a lack of customers and investors. PBMR Ltd started retrenchment procedures and stated the company intends to reduce staff by 75%.

On the September 17, 2010 the South African Minister of Public Enterprises announced the closure of the PBMR. The PMBR testing facility will likely be decommissioned and placed in a "care and maintenance mode" to protect the IP and the assets.

Adams Atomic Engines

AAE went out of business in December 2010. Their basic design was self-contained so it could be adapted to extreme environments such as space, polar and underwater environments. Their design was for a nitrogen coolant passing directly though a conventional low-pressure gas turbine. US 5309492, [|Adams, Rodney M.], "Control for a closed cycle gas turbine system", published 1994-05-03, issued 1993.Expired on May 3, 2006 due to failure to pay maintenance fees.  and due to the rapid ability of the turbine to change speeds, it can be used in applications where instead of the turbine's output being converted to electricity, the turbine itself could directly drive a mechanical device, for instance, a propeller aboard a ship.

Like all high temperature designs, the AAE engine would have been inherently safe, as the engine naturally shuts down due to Doppler broadening, stopping heat generation if the fuel in the engine gets too hot in the event of a loss of coolant or a loss of coolant flow.

See also

• Gas turbine modular helium reactor
• Generation IV reactor
• Next Generation Nuclear Plant
• Nuclear fuel
• Nuclear safety
• Rainer Moormann
• Very high temperature reactor

External links

• IAEA HTGR Knowledge Base
• AVR, experimental high-temperature reactor : 21 years of successful operation for a future energy technology ISBN 3-18-401015-5
• High Temperature Reactor 2006 Conference, Sandton, South Africa
• MIT page on Modular Pebble Bed Reactor
• Research on innovative reactors in Jülich
• Differences in American and German TRISO-coated fuels

Idaho National Laboratory - United States

• Conceptual Design of a Very High Temperature Pebble-Bed Reactor 2003
• NGNP Point Design - Results of the Initial Neutronics and Thermal-Hydraulic Assessments During FY-03, Rev. 1
• New Generation Nuclear Plant (NGNP) Project, Preliminary Point Design 2003
• The Next Generation Nuclear Plant - Insights Gained from the INEEL Point Design Studies 2004
• Computation of Dancoff Factors for Fuel Elements Incorporating Randomly Packed TRISO Particles 2005

South Africa

• Coalition Against Nuclear Energy South Africa
• Eskom
• PBMR (Pty.) Ltd.
• Pebble Bed Modular Reactor - PBMR - Home
• Atomic Energy in South Africa
• Earthlife Africa: Nuclear Energy Costs the Earth campaign
• Steve Thomas (2005), "The Economic Impact of the Proposed Demonstration Plant for the Pebble Bed Modular Reactor Design", PSIRU, University of Greenwich, UK
• NPR (April 17, 2006) NPR: South Africa Invests in Nuclear Power

addableliquid metal cooled reactor section in english by Kay Uwe Böhm

⁷Lithium

Also nether used already ⁷Li is an excellent coolant and excellent secure moderator allowing also full thorium fuel cycle with more than 2-3 times heat capacity than water and moderation better than rare beryllium with very less neutron absorption(45 millibarns)(after capturing just decay to ⁴He over ⁸Be means not going itself radioactive). Lithium is melting 180°C boiling 1340°C means a huge usable temperature difference without pressure with about 3/4 (NaK77 1/4) specific heat capacity/kg than water and high possible temperatures without pressure (-loss and decreased overpressure risk) for efficient and compact turbines, secure high critical point (thermodynamic) 2950°C(causing also decreased specific heat capacity problem with water >374°C), strong negative VOID on bubbles security with less cooling loss, low density about 0.5g/ccm (hot), no expansion on solidification like rare bismuth (eutecticum), el. magn. pumpable over paramagnetism of Li, strong thermal expansion for thermal flow, chemical binding of tritium, iodine, bromine etc., most less reactive earth alkali metal, not spreading radioactivity like water (steam) and graphite smoke, price just about 5 mio. $ for 100 000l, 28 mio. t world reserve for Li with 92.5% ⁷Li inside etc. combinable with also suggested new extreme compact and secure (Full Zero Risk all possible cases at lowest price for GWe and kWh) pebbles pillar design from Kay Uwe Böhm with a long high tungsten reflector core tube (for example 3m diameter 18m high) for high thermal flow up and down around with some isolation concrete through cBN boron nitride(thermal conductivity extreme high 740W/(m*K) and there wanted neutron absorption) heat exchange tubes with normal lithium around flowing upwards and again around downwards then to turbines and back with fuel pebbles (out of SiC or moderating c¹¹B¹⁵N melting 2973°C etc.) inside ⁷Li that can be taken in and out continiously also for regulation and trimming with near no neutron absorbers, controlled and stopped with different size absorber control and stopp pebbles and maybe turnable all around from top to bottom control and stopp rods with tungsten reflector and absorber side and addable independant and secure working stoppsand core catcher system. Possible catcher materials are broad band neutron absorbing hafnium carbide melting 3890°C at spread dome directly under core, thin cubic boron nitride plates melting 2973°C upon thick cheap magnesium oxide melting 2852°C. If the reactor core is opend all should come out quickly and quickly spreaded over a cover holded by (also active electric meltable) clamps then a mass of sand (SiO₂ melting 1715°C boiling 2200°C) maybe with additional neutron absorber placed upon is keeping better cool than water with much boiling bubbles and takes decay heat itself never melting all if enough, not rising much pressure, spread risk and keeping near all down under with still closed quick spreadable high melting point pebbles. Around heat exchange (W/Mo) steel with ribs low melting concrete with additives like in glass production working for stabilization, isolation but in overheat situation as melting buffer and then again steel with ribs all shaped round and black for better radiation. The two different pebble sizes can be taken in and out by sieves at run time. El. magn. wandering field pumping of lithium can be done with not melting tungsten spooling and wiring through cBN etc. Preheating can be done also with cBN for electric isolation heating up a high melting material with enough electric resistance through tungsten etc. cBN has a strong one time expansion after backing usable also for fixing inside other materials tubes. Possibble is likely also direct fuel and extraction using continiously taking out fission products centrifuges etc. against leight weight lithium and replacement of nuclear fuel (ThC-PaC-²³³UC or ThO₂-PaO₂-²³³UO₂) normally more heavy than all (lithium bound or density decreased alloyed) fission products maybe using a separation medium like red lead and maybe additional separation over fluoridization and/or fine seperating melting centrifuge watching outflow for colour, reflection, radioactivity kind and intensity etc. keeping the radioactive inventory always extreme low in all accidents cases (likely <0.2% compared with Fukushima 2% Chernobyl 20% set free), decreased decay heat problem, extraction also of reactor poisons like usable samarium etc. at run time like normally xenon or of breeded ²³³U surplus and continiously controlled nuclear fueling instead trimming with neutron wasting absorbers.

extended moderator (nuclear reactor) article in english by Kay Uwe Böhm

In nuclear engineering, a neutron moderator is a medium that reduces the speed of fast neutrons, thereby turning them into thermal neutrons capable of sustaining a nuclear chain reaction involving uranium-235.

Commonly used moderators include regular (light) water (roughly 75% of the world's reactors), solid graphite (20% of reactors) and heavy water (5% of reactors). [Beryllium]] has also been used in some experimental types but rare and expensive with just 200t/year world production, and burnable hydrocarbons have been suggested as another possibility like CD₄. Other possible moderators are ⁷Lithium also an excellent coolant melting 180°C, boiling 1340°C, specific heat capacity 3482 J/(kg · K) about 3/4 of water, thermal conductivity 85 W/(m · K), critical point 2950°C with excellent moderation and excellent negative VOID security on bubbles, not going radioactive, chemical binding tritium, iodine, bromine etc., no expansion on soldification like bismuth (eutecticum), high thermal expansion for thermal flow and paramagnetic for el. magn. pumping, relative cheap price also usable for divide circles not spreading also burning but most less reactive earth alkali metal; and c¹¹B15N boron nitride with moderation about graphite but unburnable, stable and dense up to 2800°C, melting 2973°C, near diamond hard, thermal conductivity 740W/(m · K), unsoluble in water, price for cBN about 100$/kg and 10 mio. t world boron reserves with 80% 11B ideal for pebble bed reactors and fuel rods with no SiC coating needed both used also in ODIN reactor design from Kay Uwe Böhm. Also ¹¹B and ⁷Li inside contsinment material aloneus are usable. In principle usable also helium or explosive D2 deuterium and T2 Tritium as gas under very high pressure dense enough for moderation (for example density H2 700 bar 75% density liquid H2 for car tanks).

Currently operating nuclear power reactors by moderator

Moderator	Reactors	Design	Country
none (fast)	1	BN-600	Russia (1)
graphite	29	AGR, Magnox, RBMK	United Kingdom (18), Russia (11)
heavy water	29	CANDU	Canada (17), South Korea (4), Romania (2), China (2), India (2), Argentina, Pakistan
light water	359	PWR, BWR	27 countries

Moderation

Neutrons are normally bound into an atomic nucleus, and do not exist free for long in nature. The unbound neutron has a half-life of just under 15 minutes. The release of neutrons from the nucleus requires exceeding the binding energy of the neutron, which is typically 7-9 MeV for most isotopes. Neutron sources generate free neutrons by a variety of nuclear reactions, including nuclear fission and nuclear fusion. Whatever the source of neutrons, they are released with energies of several MeV.

Since the kinetic energy, $E$, can be related to temperature via:

$E={\frac {1}{2}}mv^{2}={\frac {3}{2}}k_{B}T$

the characteristic neutron temperature of a several-MeV neutron is several tens of millions of degrees Celsius.

Moderation is the process of the reduction of the initial high kinetic energy of the free neutron. Since energy is conserved, this reduction of the neutron kinetic energy takes place by transfer of energy to a material known as a moderator. It is also known as neutron slowing down, since along with the reduction of energy comes a reduction in speed.

The probability of scattering of a neutron from a nucleus is given by the scattering cross section. The first couple of collisions with the moderator may be of sufficiently high energy to excite the nucleus of the moderator. Such a collision is inelastic, since some of the kinetic energy is transformed to potential energy by exciting some of the internal degrees of freedom of the nucleus to form an excited state. As the energy of the neutron is lowered, the collisions become predominantly elastic, i.e., the total kinetic energy and momentum of the system (that of the neutron and the nucleus) is conserved.

Given the mathematics of elastic collisions, as neutrons are very light compared to most nuclei, the most efficient way of removing kinetic energy from the neutron is by choosing a moderating nucleus that has near identical mass.

Elastic collision of equal masses

A collision of a neutron, which has mass of 1, with a ¹H nucleus (a proton) could result in the neutron losing virtually all of its energy in a single head-on collision. More generally, it is necessary to take into account both glancing and head-on collisions. The mean logarithmic reduction of neutron energy per collision, $\xi$, depends only on the atomic mass, $A$, of the nucleus and is given by:

$\xi =\ln {\frac {E_{0}}{E}}=1+{\frac {(A-1)^{2}}{2A}}\ln \left({\frac {A-1}{A+1}}\right)$.

This can be reasonably approximated to the very simple form $\xi \simeq {\frac {2}{A+1}}$. From this one can deduce $n$, the expected number of collisions of the neutron with nuclei of a given type that is required to reduce the kinetic energy of a neutron from $E_{0}$ to $E$:$n={\frac {1}{\xi }}(\ln E_{0}-\ln E)$.

In a system at thermal equilibrium, neutrons (red) are elastically scattered by a hypothetical moderator of free hydrogen nuclei (blue), undergoing thermally activated motion. Kinetic energy is transferred between particles. As the neutrons have essentially the same mass as protons and there is no absorption, the velocity distributions of both particles types would be well-described by a single Maxwell–Boltzmann distribution.

Choice of moderator materials

Some nuclei have larger absorption cross sections than others, which removes free neutrons from the flux. Therefore, a further criterion for an efficient moderator is one for which this parameter is small. The moderating efficiency gives the ratio of the macroscopic cross sections of scattering, $\Sigma _{s}$, weighted by $\xi$ divided by that of absorption, $\Sigma _{a}$: i.e., ${\frac {\xi \Sigma _{s}}{\Sigma _{a}}}$. For a compound moderator composed of more than one element, such as light or heavy water, it is necessary to take into account the moderating and absorbing effect of both the hydrogen isotope and oxygen atom to calculate $\xi$. To bring a neutron from the fission energy of $E_{0}$ 2 MeV to an $E$ of 1 eV takes an expected $n$ of 16 and 29 collisions for H₂O and D₂O, respectively. Therefore, neutrons are more rapidly moderated by light water, as H has a far higher $\Sigma _{s}$. However, it also has a far higher $\Sigma _{a}$, so that the moderating efficiency is nearly 80 times higher for heavy water than for light water.

The ideal moderator is of low mass, high scattering cross section, and low absorption cross section.

	Hydrogen	Deuterium	Beryllium	Carbon	Oxygen	Uranium
Mass of kernels u	1	2	9	12	16	238
Energy decrement $\xi$	1	0,7261	0,2078	0,1589	0,1209	0,0084
Number of Collisions	18	25	86	114	150	2172

Distribution of neutron velocities once moderated

After sufficient impacts, the speed of the neutron will be comparable to the speed of the nuclei given by thermal motion; this neutron is then called a thermal neutron, and the process may also be termed thermalization. Once at equilibrium at a given temperature the distribution of speeds (energies) expected of rigid spheres scattering elastically is given by the Maxwell–Boltzmann distribution. This is only slightly modified in a real moderator due to the speed (energy) dependence of the absorption cross-section of most materials, so that low-speed neutrons are preferentially absorbed, so that the true neutron velocity distribution in the core would be slightly hotter than predicted.

Reactor moderators

In a thermal nuclear reactor, the nucleus of a heavy fuel element such as uranium absorbs a slow-moving free neutron, becomes unstable, and then splits ("fissions") into two smaller atoms ("fission products"). The fission process for ²³⁵U nuclei yields two fission products: two to three fast-moving free neutrons, plus an amount of energy primarily manifested in the kinetic energy of the recoiling fission products. The free neutrons are emitted with a kinetic energy of ~2 MeV each. Because more free neutrons are released from a uranium fission event than thermal neutrons are required to initiate the event, the reaction can become self-sustaining — a chain reaction — under controlled conditions, thus liberating a tremendous amount of energy (see article nuclear fission).

Fission cross section, measured in barns (a unit equal to 10⁻²⁸ m²), is a function of the energy (so-called excitation function) of the neutron colliding with a ²³⁵U nucleus. Fission probability decreases as neutron energy (and speed) increases. This explains why most reactors fueled with ²³⁵U need a moderator to sustain a chain reaction and why removing a moderator can shut down a reactor.

The probability of further fission events is determined by the fission cross section, which is dependent upon the speed (energy) of the incident neutrons. For thermal reactors, high-energy neutrons in the MeV-range are much less likely to cause further fission. (Note: It is not impossible for fast neutrons to cause fission, just much less likely.) The newly released fast neutrons, moving at roughly 10% of the speed of light, must be slowed down or "moderated," typically to speeds of a few kilometres per second, if they are to be likely to cause further fission in neighbouring ²³⁵U nuclei and hence continue the chain reaction. This speed happens to be equivalent to temperatures in the few hundred celsius range.

In all moderated reactors, some neutrons of all energy levels will produce fission, including fast neutrons. Some reactors are more fully thermalised than others; for example, in a CANDU reactor nearly all fission reactions are produced by thermal neutrons, while in a pressurized water reactor (PWR) a considerable portion of the fissions are produced by higher-energy neutrons. In the proposed water-cooled supercritical water reactor (SCWR), the proportion of fast fissions may exceed 50%, making it technically a fast neutron reactor.

A fast reactor uses no moderator, but relies on fission produced by unmoderated fast neutrons to sustain the chain reaction. In some fast reactor designs, up to 20% of fissions can come from direct fast neutron fission of uranium-238, an isotope which is not fissile at all with thermal neutrons.

Moderators are also used in non-reactor neutron sources, such as plutonium-beryllium and spallation sources.

Form and location

The form and location of the moderator can greatly influence the cost and safety of a reactor. Classically, moderators were precision-machined blocks of high purity graphite with embedded ducting to carry away heat. They were in the hottest part of the reactor, and therefore subject to corrosion and ablation. In some materials, including graphite, the impact of the neutrons with the moderator can cause the moderator to accumulate dangerous amounts of Wigner energy. This problem led to the infamous Windscale fire at the Windscale Piles, a nuclear reactor complex in the United Kingdom, in 1957.

Some pebble-bed reactors' moderators are not only simple, but also inexpensive: the nuclear fuel is embedded in spheres of reactor-grade pyrolytic carbon, roughly of the size of tennis balls. The spaces between the balls serve as ducting. The reactor is operated above the Wigner annealing temperature so that the graphite does not accumulate dangerous amounts of Wigner energy.

In CANDU and PWR reactors, the moderator is liquid water (heavy water for CANDU, light water for PWR). In the event of a loss-of-coolant accident in a PWR, the moderator is also lost and the reaction will stop. This negative void coefficient is an important safety feature of these reactors. In CANDU the moderator is located in a separate heavy-water circuit, surrounding the pressurized heavy-water coolant channels. This design gives CANDU reactors a positive void coefficient, although the slower neutron kinetics of heavy-water moderated systems compensates for this, leading to comparable safety with PWRs."

Moderator impurities

Good moderators are also free of neutron-absorbing impurities such as boron. In commercial nuclear power plants the moderator typically contains dissolved boron. The boron concentration of the reactor coolant can be changed by the operators by adding boric acid or by diluting with water to manipulate reactor power. The German World War II nuclear program suffered a substantial setback when its inexpensive graphite moderators failed to work. At that time, most graphites were deposited on boron electrodes, and the German commercial graphite contained too much boron. Since the war-time German program never discovered this problem, they were forced to use far more expensive heavy water moderators. In the U.S., Leó Szilárd, a former chemical engineer, discovered the problem.

Non-graphite moderators

Some moderators are quite expensive, for example beryllium, and reactor-grade heavy water. Reactor-grade heavy water must be 99.75% pure to enable reactions with unenriched uranium. This is difficult to prepare because heavy water and regular water form the same chemical bonds in almost the same ways, at only slightly different speeds.

The much cheaper light water moderator (essentially very pure regular water ) absorbs too many neutrons to be used with unenriched natural uranium, and therefore uranium enrichment or nuclear reprocessing becomes necessary to operate such reactors, increasing overall costs. Both enrichment and reprocessing are expensive and technologically challenging processes, and additionally both enrichment and several types of reprocessing can be used to create weapons-usable material, causing proliferation concerns. Reprocessing schemes that are more resistant to proliferation are currently under development.

The CANDU reactor's moderator doubles as a safety feature. A large tank of low-temperature, low-pressure heavy water moderates the neutrons and also acts as a heat sink in extreme loss-of-coolant accident conditions. It is separated from the fuel rods that actually generate the heat. Heavy water is very effective at slowing down (moderating) neutrons, giving CANDU reactors their important and defining characteristic of high "neutron economy."

Nuclear weapon design

 이 부분의 본문은 Uranium hydride bomb입니다.

Early speculation about nuclear weapons assumed that an "atom bomb" would be a large amount of fissile material, moderated by a neutron moderator, similar in structure to a nuclear reactor or "pile". Only the Manhattan project embraced the idea of a chain reaction of fast neutrons in pure metallic uranium or plutonium. Other moderated designs were also considered by the Americans; proposals included using uranium hydride as the fissile material. In 1943 Robert Oppenheimer and Niels Bohr considered the possibility of using a "pile" as a weapon. The motivation was that with a graphite moderator it would be possible to achieve the chain reaction without the use of any isotope separation. In August 1945, when information of the atomic bombing of Hiroshima was relayed to the scientists of the German nuclear program, interned at Farm Hall in England, chief scientist Werner Heisenberg hypothesized that the device must have been "something like a nuclear reactor, with the neutrons slowed by many collisions with a moderator."

After the success of the Manhattan project, all major nuclear weapons programs have relied on fast neutrons in their weapons designs. The notable exception is the Ruth and Ray test explosions of Operation Upshot-Knothole. The aim of the University of California Radiation Laboratory design was to produce an explosion powerful enough to ignite a thermonuclear weapon with the minimal amount of fissile material. The core consisted of uranium hydride, with hydrogen, or in the case of Ray, deuterium acting as the neutron moderator. The predicted yield was 1.5 to 3 kt for Ruth and 0.5-1 kt for Ray. The tests produced yields of 200 tons of TNT each; both tests were considered to be fizzles.

The main benefit of using a moderator in a nuclear explosive is that the amount of fissile material needed to reach criticality may be greatly reduced. Slowing of fast neutrons will increase the cross section for neutron absorption, reducing the critical mass. A side effect is however that as the chain reaction progresses, the moderator will be heated, thus losing its ability to cool the neutrons.

Another effect of moderation is that the time between subsequent neutron generations is increased, slowing down the reaction. This makes the containment of the explosion a problem; the inertia that is used to confine implosion type bombs will not be able to confine the reaction. The end result may be a fizzle instead of a bang.

The explosive power of a fully moderated explosion is thus limited, at worst it may be equal to a chemical explosive of similar mass. Again quoting Heisenberg: "One can never make an explosive with slow neutrons, not even with the heavy water machine, as then the neutrons only go with thermal speed, with the result that the reaction is so slow that the thing explodes sooner, before the reaction is complete."

While a nuclear bomb working on thermal neutrons may be impractical, modern weapons designs may still benefit from some level of moderation. A beryllium tamper used as a neutron reflector will also act as a moderator.

Materials used

• Hydrogen, as in ordinary "light water." Because protium also has a significant cross section for neutron capture only limited moderation is possible without losing too many neutrons. The less-moderated neutrons are relatively more likely to be captured by uranium-238 and less likely to fission uranium-235, so light water reactors require enriched uranium to operate.
  • There are also proposals to use the compound formed by the chemical reaction of metallic uranium and hydrogen (uranium hydride—UH₃) as a combination fuel and moderator in a new type of reactor.
  • Hydrogen is also used in the form of cryogenic liquid methane and sometimes liquid hydrogen as a cold neutron source in some research reactors: yielding a Maxwell–Boltzmann distribution for the neutrons whose maximum is shifted to much lower energies.
  • Hydrogen combined with carbon as in paraffin wax was used in some early German experiments.
• Deuterium, in the form of heavy water, in heavy water reactors, e.g. CANDU. Reactors moderated with heavy water can use unenriched natural uranium.
• Carbon, in the form of reactor-grade graphite or pyrolytic carbon, used in e.g. RBMK and pebble-bed reactors, or in compounds, e.g. carbon dioxide [3]. Lower-temperature reactors are susceptible to buildup of Wigner energy in the material. Like deuterium-moderated reactors, some of these reactors can use unenriched natural uranium.
  • Graphite is also deliberately allowed to be heated to around 2000 K or higher in some research reactors to produce a hot neutron source: giving a Maxwell–Boltzmann distribution whose maximum is spread out to generate higher energy neutrons.
• Beryllium, in the form of metal. Beryllium is expensive and toxic, so its use is limited.
• Lithium-7, in the form of a lithium fluoride salt, typically in conjunction with beryllium fluoride salt (FLiBe). This is the most common type of moderator in a Molten Salt Reactor.

Other light-nuclei materials are unsuitable for various reasons. Helium is a gas and it requires special design to achieve sufficient density; lithium-6 and boron-10 absorb neutrons.

References

• 《DOE Fundamentals Handbook: Nuclear Physics and Reactor Theory. Vol. 2 (DOE-HDBK-1019/2-93)》 (PDF). U.S. Department of Energy. January 1993. 2013년 11월 29일에 확인함. 

See also

• Nuclear cross section
• Neutron reflector
• Neutron scattering
• Zircalloy

Coal power plant air pollution etc. by Kay Uwe Böhm

Partly compensating natural and anthropogenic share of temperature change effects showing also sulfate temperature decrease much from coal power plants

Modern coal power plants should have like most done already a flue-gas desulfurization and normally not made already a not too expensive centrifugal smoke filtering system inside chimney channel(maybe because new from Kay Uwe Böhm). A high speed centrifuge standing upwards inside the smoke streaming with blades formed for moving the solids and from high pressure condensing water to centrifuge side and down out with the smoke gas still going upwards. With additionally separation the valuable metals like Hg, Cd, Ni, Pb, As, Cr, U, Th etc. inside smoke could be taken also out for using and selling. Adding of cheap methanol inside spray tower flue-gas desulfurization instead or with chalk can produce much Dimethyl sulfate as a sulfate aerosol for cooling down the atmosphere with much more cooling effect (can cause an ice time with available masses of sulfur compounds and methanol) than the strongest greenhouse gas sulfur hexafluoride warming effect working more local but also global and regulatable watching the weather situation if much is streaming high up through the chimney. Existing coal power plants are releasing already significant amounts of cooling sulfate aerosols coming timely later in existence after chemical change in air. Also much air pollution decreasing is using high efficient turbines like new Kayturbines and supercharging with cBN cubic boron nitride tubes filled with molten liquid lithium (melting >180°C, boiling 1340°C, 3/4 specific heat capacity of water/kg)as excellent cheap coolant heated up also with cBN tubes at hotest point of flame. The carbon dioxide air pollution rise is limited at all by exploration maximums and that way not dangerous in future coming for coal from china with >50% world coal production but less than 30 years reserve and peak coal about 2020-2025, peak oil about 2012-2015, peak gas <2025, with fossil energy maximum about 2020-2025 reached and not near unlimited until 2100 like inside unrealistic IPCC SRES scenarios.

Carbon monoxide and other air pollution of the 9 greatest brown coal power plants in germany (PRTR 2010)

Power plant	CO2 (Tons)	NOx/NO2 (Tons)	SOx/SO2 (Tons)	Fine Dust (Tons)	Hg (kg)	Cd (kg)	Ni (kg)	Pb (kg)	As (kg)	Cr (kg)
Niederaußem	28.100.000	17.900	6.870	386	499	<10	<50	<200	49,9	<100
Jänschwalde*	23.800.000	18.700	21.400	573	592	<10	308	<200	129	<100
Weisweiler	19.900.000	12.700	3.060	456	271	<10	103	<200	67	<100
Neurath	16.900.000	11.700	3.190	251	181	<10	<50	<200	42,2	<100
Boxberg	15.100.000	10.700	7.810	167	226	<10	152	236	<20	<100
Frimmersdorf	14.400.000	9.070	5.620	257	153	<10	<50	<200	35,7	<100
Lippendorf**	12.500.000	8.570	13.800	108	1.160	68	1.960	789	21	466
Schwarze Pumpe	11.200.000	4.610	7.060	<100	243	62,9	<50	369	35,8	224
Schkopau	5.120.000	3.320	4.770	74,6	227	129	<50	<200	<20	<100
Sum without "<"	147.020.000	97.270	73.580	2.273	3.552	260	2.523	1.394	381	690
DE All together 2010	834.511.385	1.328.717	444.035	211.284	9.412	4.723	105.802	193.968	6.120	55.060
Share of all together	18 %	7,3 %	17 %	1,1 %	38 %	5,5 %	2,4 %	0,7 %	6,2 %	1,3 %
* with Fuel surrogate and Waste-to-energy ** with biosolids-Waste-to-energy

Carbon monoxide and other air pollution of the 14 greatest stone coal power plants in germany PRTR 2010

Power plant	CO2 (Tons)	NOx/NO2 (Tons)	SOx/SO2 (Tons)	Fine dust (Tons)	Hg (kg)	Cd (kg)	Ni (kg)	Pb (kg)	As (kg)	Cr (kg)
Scholven	9.390.000	7.090	4.330	244	135	31	86	<200	51	<100
Mannheim	6.510.000	3.550	1.490	148	146	<10	<50	<200	68	<100
Voerde	6.240.000	4.700	2.840	<100	38,3	<10	<50	<200	<20	<100
Staudinger*	4.480.000	2.770	665	69,9	45,6	19,1	131	<200	113	192
Heyden	3.870.000	2.920	1.380	86,7	28,4	<10	<50	<200	<20	<100
Heilbronn	3.240.000	2.160	1.660	<100	34	<10	<50	<200	<20	<100
Werne*	3.140.000	1.900	1.170	<100	11,5	<10	<50	<200	<20	<100
Wilhelmshaven	3.100.000	2.040	1.390	136	29,9	11,7	<50	<200	<20	<100
Bergkamen	3.020.000	2.100	2.040	<100	18,1	<10	<50	<200	<20	<100
Herne	2.480.000	1.790	1.340	<100	30,3	<10	<50	<200	<20	<100
Altbach**	2.220.000	1.350	906	<100	30	<10	<50	<200	<20	<100
Karlsruhe*	2.170.000	1.140	1.080	<100	19	<10	<50	<200	<20	<100
Veltheim**	1.740.000	1.290	400	52,6	10,1	22,4	<50	<200	156	<100
Bexbach	1.300.000	910	746	<100	<10	<10	<50	<200	<20	<100
Sum without "<"	52.900.000	35.710	21.437	737	576	84	217	-	388	192
DE All together 2010	834.511.385	1.328.717	444.035	211.284	9.412	4.723	105.802	193.968	6.120	55.060
Share of all together	6,3 %	2,7 %	4,8 %	0,3 %	6,1 %	1,8 %	0,2 %	-	6,3 %	0,3 %
* with earth gas share, ** with oil- and earth gas share