BASIC PROFESSIONAL TRAINING COURSE Module XVII Fuel cycle
BASIC PROFESSIONAL TRAINING COURSE Module XVII Fuel cycle, spent fuel management and transport of radioactive materials Version 1. 1, September 2015 This material was prepared by the IAEA and co-funded by the European Union.
2 NUCLEAR FUEL CYCLE Learning objectives After completing this chapter, the trainee will be able to: 1. Broadly describe the nuclear fuel cycle; 2. List the main phases of the nuclear fuel cycle; 3. Recognize the difference between open and closed fuel cycles; 4. Describe basic principles of each phase of the fuel cycle; 5. Describe basic features of PWR and BWR fuel elements. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
3 Introduction • The Nuclear Fuel Cycle (NFC) includes: the set of processes and operations needed to manufacture nuclear fuel; irradiation in nuclear power reactors; irradiated fuel storage; reprocessing and recycling or disposal. • Several nuclear fuel cycles may be considered, depending on: − The type of reactor, − The type of fuel used, and − Whether or not the irradiated fuel is reprocessed and recycled. • All NFCs: − Start with mining of uranium, − End with disposal of spent fuel and/or other radioactive waste. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
4 Introduction • The first stage of the NFC is uranium production (mining and milling) − Uranium ore is extracted from ground and processed to the final product, “yellowcake”, a powder form of uranium oxides (mainly U 3 O 8). • The second stage of the NFC is conversion − Yellowcake is purified and converted to uranium hexafluoride (UF 6), − In NFC for certain types of reactors, the product of this step is uranium dioxide (UO 2), which can be directly used for uranium fuel fabrication. • The third stage of the NFC is enrichment − The fissile isotope 235 U content of UF 6 is increased in comparison with the content of 238 U • The fourth stage of the NFC is fuel fabrication − UF 6 is converted to UO 2 , which is converted to ceramic pellets and loaded into fuel rods. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
5 Introduction • The fifth stage of the NFC is irradiation/nuclear reactor operation − Finished fuel is loaded into nuclear reactors and irradiated, − This step is the purpose of the whole NFC • The final step of the NFC is spent fuel reprocessing and recycling − Spent fuel is first transferred to spent fuel storage. − Spent fuel can be transferred to interim storage (wet storage in water pools, or dry storage in casks). − Spent fuel can be conditioned for longer term interim storage – spent fuel conditioning or transferred to reprocessing facilities for recycling. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
6 Introduction Nuclear Fuel Cycle Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
7 Introduction • The NFC also includes some other processes and operations: − Uranium ore exploration, i. e. activities related to the finding and development of the uranium ores, − Heavy water production, which is necessary to run certain type of reactors, − Zirconium and nuclear grade stainless steel metal and tubing production, − Management of high level and other wastes, − And finally, transportation activities associated with moving materials between operations. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
8 Non-Proliferation • The Nuclear Non-Proliferation Treaty (NPT or NNPT) is an international agreement aimed at preventing the spread of nuclear weapons and promoting cooperation in the commercial uses of nuclear energy and disarmament. − The NPT established a system of safeguards under the IAEA. − Created in 1968 and signed by 190 States. − Permits ownership of nuclear weapons only by the five countries that possessed them at the treaty’s inception: China, France, Russia, the United Kingdom and the United States. − These five countries pledged not to transfer nuclear weapons technology to other states and to reduce their weapons stockpiles. − To combat the threat of proliferation, the international nuclear energy community has adopted robust controls to ensure that it can secure and fully account for nuclear materials manufactured for the production of electricity, and their by-products. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
9 Uranium production • Uranium occurs with an abundance of 2. 7 ppm by weight in the Earth's crust. − It is 600 times more abundant than gold and about as abundant as tin which makes it a rather common metal. • Uranium ores usually contain 0. 1% to 0. 5% of uranium. − Although higher grades (up to several per cent) have been found in some cases. • Current production of uranium comes from about twenty countries − Historically almost forty countries have produced uranium. • The production of uranium in 2012 was 58, 816 t. U, with Kazakhstan producing more than one third of world production. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
10 Uranium production Country shares in 2012 uranium production Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
11 Uranium production Recent world uranium production. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
12 Uranium production • In 2012, 45% of uranium production was by in situ leaching (ISL). • Conventional methods: − Underground mining provided 26%, and − Open-pit mining provided 20% of production. • Other methods: − Co-product or by-product recovery from copper, gold and phosphate operations provided 6. 6%, − Heap-leaching (leaching from the heaps of ore) 1. 7% and − Other methods provided 0. 7% of uranium production. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
13 Uranium production Uranium mining and milling • Uranium mining is similar to other kinds of mining unless the ore is very high grade. − Special mining techniques (dust suppression, and remote handling techniques) are employed to limit worker radiation exposure and to ensure the safety of the environment and general public. • Deposits can be identified and mapped from the air due to the radiation signature of uranium’s decay products. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
14 Uranium production Uranium mining and milling • Mining and milling activities associated with uranium recovery involve two distinct extraction phases: − In mining uranium ore is extracted from the Earth − Deep underground shafts or shallow open pits, − In situ method, without removing ore from the ground. − In milling: − The mined ore is crushed, and − A second extraction process chemically leaches the uranium from the ore – concentrates and precipitates it to produce yellowcake. − Yellowcake is a mixture of uranium oxides, with U 3 O 8 representing ~ 85% of them. It varies in colour, but is usually brown or grey. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
15 Uranium production Uranium mining and milling Uranium mine Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
16 Uranium production Uranium mining and milling • There are three primary milling methods: − Conventional milling, − Heap leaching, − In situ recovery/leaching (ISL method). • The last method (in situ recovery or ISL) is in fact a combination of mining and milling in one operation. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
17 Uranium production Uranium mining and milling Crush, Leach & Filter Mine Heap Leach Concentrate & purify Ion exchange (IX) Solvent extraction (SX) ISL Strip, precipitate & dry (YELLOWCAKE) Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
18 Uranium production Uranium mining and milling • Conventional milling: − Uranium ore is crushed into smaller particles, − Which are subsequently extracted or leached by using sulphuric acid or alkaline solutions. − Extracted uranium is then concentrated to produce yellowcake. − Mills are typically located in areas of low population density, processing ore from mines within a geographic radius of few tens of kilometres. − Mill tailings are: − Fine-grained, sandy waste − Deposited in an impoundment or "mill tailings pile, " - carefully regulated, monitored, and controlled Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
19 Uranium production Uranium mining and milling Uranium ore crusher Conventional uranium mill in the United States. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
20 Uranium production Uranium mining and milling • Heap leach operations involve the following processes: − Small pieces of uncrushed ore are placed in a "heap" on an impervious pad of plastic, clay, or asphalt, with perforated pipes under the heap. − An acidic solution (sometime alkali – depends on rock chemistry) is then sprayed over the ore to dissolve the uranium it contains. − The uranium-rich solution drains into the perforated pipes, where it is collected and transferred to an ion-exchange system. − A solvent extraction or ion-exchange system extracts and concentrates the uranium to produce yellowcake. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
21 Uranium production Uranium mining and milling Heap leach recovery process. Heap Leach Pad project at Trekkopje uranium mine. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
22 Uranium production Uranium mining and milling • The in situ recovery (ISL) process: − A solution called a lixiviant (typically containing water mixed with oxygen and/or hydrogen peroxide, as well as sodium carbonate or carbon dioxide) is injected through a series of wells into the ore body to dissolve the uranium. − The lixiviant solution is then collected in a series of recovery wells, from which it is pumped to a processing plant, where the uranium is extracted from the solution through an ion-exchange process. − The uranium extract is then further purified, concentrated, and dried to produce yellowcake. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
23 Uranium production Uranium mining and milling Hobson ISL processing in Karnes County, USA In situ recovery Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
24 Uranium production Feature Recovery Method Surface Features Conventional Uranium Mill Physical and chemical process to extract uranium from mined ore. Mill building(s), process tanks, tailings impoundment, and evaporation ponds. Heap Leach Facility ISL Physical and chemical process to extract uranium from mined ore that has been piled in a heap. Chemical process to extract uranium from underground deposits. Process buildings, heap pile consisting of crushed ore. Well fields, header houses, pipes, processing facility, storage or evaporation pond. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
25 Uranium production Feature Waste Generated Decommissioning Status at End of Use Conventional Uranium Mill Heap Leach Facility ISL Mill tailings, pipes, pumps, and other process equipment that cannot be decontaminated. Heap pile remains in place after processing; pipes, pumps, and other process equipment that cannot be decontaminated. Liquid waste (disposed of in injection wells or through an evaporation system), pipes, pumps, and other process equipment that cannot be decontaminated. Demolition of mill and site buildings, final cover system installed over tailings pile, groundwater monitoring. Site permanently transferred to relevant authority for long-term care; annual inspections. Demolition of site buildings, final cover system installed over heap pile, groundwater monitoring. Restoration of groundwater, decommissioning of injection wells, removal of pipes and processing building. Site permanently transferred to relevant authority for long-term care; annual inspections. Site released for unrestricted use when cleanup criteria are met. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
26 Conversion • Conversion includes the process steps to purify and convert the yellowcake to uranium hexafluoride (UF 6). • UF 6 is suitable as a feed for enrichment because − It is in gaseous form at slightly elevated temperatures (above ~60°C), − Fluorine has only one natural isotope – the enrichment mechanism is based on differences in molecular form. • There are two principal methods of converting uranium oxide to UF 6: − Wet chemical process, − Dry fluoride volatility process. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
27 Conversion • In the wet process: 1. Uranium concentrate is dissolved in nitric acid, 2. The solution is purified and then calcined (heated strongly) to produce UO 3 powder. 3. UO 3 powder is hydrofluorinated with hydrofluoric acid, which converts it into UF 4 (a green salt). 4. In the last stage the UF 4 is converted into uranium hexafluoride (UF 6) through fluorination. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
28 Conversion • In the dry process: 1. Uranium concentrate is pelletized and directly reduced with hydrogen to UO 2 in a fluidized bed reactor. 2. UO 2 product is reacted with hydrogen fluoride (HF) to form uranium tetrafluoride (UF 4). 3. UF 4 is fed, with gaseous fluorine, into a production unit consisting of a flame-reactor and a fluidized bed reactor, to produce uranium hexafluoride, UF 6, gas. 4. UF 6 is purified in a two stage pressure distillation process. − This distillation process is necessary, because, in contrast with the wet process, no purification is carried out in earlier stages. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
29 Conversion • UF 6 sublimes (changes from a solid to a gas) at atmospheric pressure and 56. 5°C, making it suitable for use in the enrichment process. • At lower temperatures and under moderate pressure, the UF 6 can be liquefied. • As it cools, the liquid UF 6 becomes a white crystalline solid − It is shipped in this form in specially designed, thick walled, mild steel cylinders. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
30 Conversion • Conversion plants exist in Argentina, Brazil, Canada and Iran as well as in states that have produced nuclear weapons. • The health and safety risks involved are more related to the use of highly reactive chemicals (fluorine and its compounds) than to nuclear or radiological risks. • Sensitive and secure measurements under international supervision (IAEA Safeguards) are used; − To verify the mass and enrichment of UF 6 transferred in and out of the facilities with the aim of preventing the diversion of nuclear materials and limiting the potential for nuclear weapons proliferation. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
31 Enrichment • Natural uranium is comprised of three isotopes: • − 238 U (99. 28% by mass), − 235 U (0. 71% by mass) and − 234 U (0. 005% by mass). 235 U is a fissile nuclide and is the only naturally occurring nuclide which can be used as nuclear fuel in thermal reactors. • Increasing the 235 U isotope above its natural concentration (0. 71%) is termed uranium enrichment. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
32 Enrichment • Heavy water reactors and early gas-cooled reactors can run on natural uranium. • Light water reactors require enrichments from 2% to 5% 235 U. • Research reactors use fuel ranging from natural uranium to enrichment greater than 90% 235 U, but most of them have enrichments approaching 20%. • Enrichment less than 20% 235 U – low-enriched uranium (LEU), • Enrichment to 20% 235 U or greater – highly enriched uranium (HEU). • Uranium with a 235 U content less than natural uranium is called depleted uranium (DU). Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
33 Enrichment • The isotopic separation of uranium is a physical process, based on small difference in the mass of isotopes. Enrichment can also be achieved by blending highly enriched uranium into a larger quantity of low enriched, natural or depleted uranium. • Isotopic separation processes are made up of many stages – both in series and parallel (it is usual to speak of separation factors per stage of the process). − Each process stage has only a small separation factor, − Many stages in series are needed to get the desired enrichment level. − Each stage has only a limited throughput, − Many stages are needed in parallel to get the required production rate. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
34 Enrichment • The basic component of an enrichment plant is the separation element (SE). • The SE is a device that separates the incoming feed stream into two outgoing streams: − An enriched stream, in which the process material is enriched to some degree in the desired isotope, and − A depleted stream that is correspondingly depleted in this isotope. • Both these streams are in the chemical form of UF 6 , which is in gaseous form. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
35 Enrichment • Most important features of a separation element (SE) are the separation factor and the throughput. • The separation factor (α) is the degree of separation achieved in a given separation element or stage. − α is approximately the ratio of the concentration of the enriched stream to the concentration of the depleted stream. • Throughput is measured by the mass that an SE can process in unit time. − Multiple stages are used to achieve the required enrichment. − Elements in different stages may need to differ in physical characteristics. − Parallel identical elements are used to achieve the necessary throughput (rate of production). Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
36 Enrichment Separative Work Unit (SWU) • The SWU is the best thought of as related to the amount of energy required to take 1 kg of material from one enrichment level to another. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
37 Enrichment Separative Work Unit (SWU) • For instance, to produce one kilogram of uranium enriched to 5% 235 U requires: − 7. 9 SWU if the plant is operated at a tails assay 0. 25%, or − 8. 9 SWU if the tails assay is 0. 20% (thereby requiring only 9. 4 kg instead of 10. 4 kg of natural U feed). • There is always a trade-off between the cost of enrichment SWU and the cost of uranium. • Techniques include gaseous diffusion, gas centrifugation and laser isotope separation. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
38 Enrichment Gaseous diffusion • Mixture of gas molecules (235 UF 6 and 238 UF 6) is confined in a vessel and is in thermal equilibrium with its surroundings. • Average thermal velocity of the lighter 235 UF 6 molecules is slightly greater than that of the heavier 238 UF 6 molecules. − 235 UF 238 UF 6 molecules strike the porous membrane more frequently than the 6 molecules. • The membrane separating the two product streams has pores large enough to permit the escape of individual molecules, but sufficiently small so that bulk flow of the gas is prevented. The lighter 235 UF 6 molecules pass through the membrane more readily than the heavier 238 UF 6 ones. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
39 Enrichment Gaseous diffusion • The escaped gas is then enriched with respect to the lighter component of the mixture, and the remaining gas is depleted. • Compressed UF 6 feed gas is made to flow inside a porous membrane or barrier tube. • Approximately one-half of the gas passes through the barrier into a region of lower pressure – enriched gas and is sent to the next higher stage of the cascade. • Remaining depleted gas, is sent back to the previous stage. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
40 Enrichment Gaseous diffusion • By a use of a large cascade of many stages, high separations can be achieved. • The first process that was capable of producing enriched uranium in industrially useful quantities. • The gas diffusion process is energy intensive. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
41 Enrichment Gas centrifugation • The process uses the principle of centrifugal force to create a density gradient in a gas containing components of different molecular weights. • The gas centrifuge a hollow, vertical cylinder (i. e. rotor) that is spun about its axis at a high angular velocity inside an evacuated casing. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
42 Enrichment Gas centrifugation • Gaseous UF 6 is fed into the rotor and accelerated. • The higher centrifugal force on heavier 238 UF 6 molecules increases their concentration near outer wall of the cylinder more than the concentration of lighter 235 UF 6 molecules. • Gas near the axis is enriched with lighter molecules containing 235 U. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
43 Enrichment Laser Isotope Separation (LIS) • LIS is based on the fact that electron energy states of atom are very precisely defined and depend on the mass of the nucleus. • Different isotopes of the same element, have different electronic energies and absorb different colours of laser light. • If an atom absorbs light with energy exactly corresponding to one of its electronic states, it can become ionized, i. e. obtains a positive electric charge. • Such charged ions can then be easily separated from neutral atoms in an electric field. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
44 Enrichment Laser Isotope Separation (LIS) • In LIS enrichment, uranium metal is vaporized a vacuum chamber. • The vapour stream is then illuminated with laser light tuned precisely to a colour at which 235 U absorbs energy. • Ionized 235 U atoms are collected on negatively charged surfaces inside the separator unit. • The product material is condensed as liquid on these surfaces and then flows to a charged surface where it solidifies as metal nuggets. • The atoms of 238 U, which were unaffected by the laser beam, pass through the product collector, condense on the tailings collector, and are removed. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
45 Enrichment Deconversion of depleted uranium • Deconversion is a process used to chemically convert the depleted UF 6 to a more stable uranium compound. − The inventory should be managed safely and efficiently in a way that protects the health and safety of workers and the public, and protects the environment. • The depleted UF 6 Management Program involves three primary activities: − Cylinder surveillance and maintenance, − Conversion of depleted UF 6 to a more stable chemical form for use or disposal, and − Development of beneficial uses for depleted uranium. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
46 Enrichment Deconversion of depleted uranium • The cylinder surveillance and maintenance typically includes: − Regular inspections and maintenance of cylinders and storage yards; − Restacking and respacing the cylinders to improve drainage and to allow for more thorough inspections; − Repainting ends of skirted cylinders and repainting cylinder bodies as needed to arrest corrosion; and − Constructing new cylinder storage yards and reconditioning existing yards improve storage conditions. • Chemical processes to convert depleted uranium to a more stable uranium compound are similar to those used in the production of fuel. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
47 Fuel fabrication • Nuclear fuel can be made of uranium oxide, uranium carbide, metallic uranium and various other chemical compounds or alloys, including mixed plutonium and uranium oxides. • The majority of power reactors utilize fuel made of uranium dioxide (UO 2), in the form of ceramic pellets. − Pellets are milled to a very precise size and shape, and − loaded into long metal tubes (cladding tubes) to form fuel rods. − Many such fuel rods make up a fuel assembly. • A typical fuel fabrication process may be divided into three stages which are: − Conversion of UF 6 to UO 2 and pelletizing; − Fuel rod manufacturing process; and − Fuel assembly manufacturing process. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
48 Fuel fabrication Conversion and pelletizing process • Enriched UF 6 is converted to UO 2 powder through a chemical process: − Dry process is IDR (Integrated Dry Route Powder Process), − Wet processes are ADU (ammonium diuranate) and AUC (ammonium uranyl carbonate). • The powder slug is ball-milled into a fine powder. • In the pelletizing process, green pellets of about 60% theoretical density are produced; − By compaction and then sintered in the hydrogen sintering furnace at a temperature >1700 °C. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
49 Fuel fabrication Conversion and pelletizing process • The sintered pellets are ground to make the pellet diameter within the specification range. • Before loading the cladding tube, pellets are inspected; − For diameter, length, perpendicularity, cracks and chips, uranium content, O/U ratio, enrichment, moisture and impurities contents. • As far as possible pellet grindings and reject pellets are recycled in the process. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
50 Fuel fabrication Fuel rod manufacturing process • The fuel rod consists of: − Metal tube (cladding), − Top and bottom end plugs, − Fuel pellets and − Plenum spring. • Cladding is usually made of zirconium alloy • The pellets are loaded into the cladding tube which is bottom end plug welded. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
51 Fuel fabrication Fuel rod manufacturing process • Plenum spring compensates for thermal expansion and structural changes of fuel pellets during power operation. • Fuel rod is pressurized with helium and top end plug welded; − The initial pressure of helium in fresh fuel is approximately equal to the primary coolant pressure at operating temperatures (~ 155 bar). • Plenum (upper empty space of the fuel rod) accommodates released fission gases. − So that the pressure in the fuel rod does not increase over design limit. • All fuel rods are inspected (by radiography, visual inspection, helium leak testing, and for discoloration at the welds, plenum length, full length and straightness). Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
52 Fuel fabrication Fuel assembly manufacturing process • The fuel rods are loaded into a ‘skeleton’ consisting of: − Top and bottom nozzles, − Spacer grids, − Instrumentation tubes and − Guide thimbles. • After the fuel rods are inserted into the skeleton, top and bottom nozzles are attached to the skeleton. • Various inspections confirm that the distance between the fuel rods, fuel assembly torsion, length and other dimensions are correct. • Different types of reactor require different fuel assemblies. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
53 Fuel fabrication PWR fuel assembly manufacturing process Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
54 Fuel fabrication PWR fuel • Fuel for western PWRs is built with a square lattice arrangement, • Assemblies are characterized by the number of rods they contain, typically, 17× 17 in current designs. • A PWR fuel assembly stands between 4 and 5 metres high, is about 20 cm across and weighs about half a tonne. • The assembly has vacant rod positions – space left for the vertical insertion of a control rod. • Some assembly positions may be designated as a "guide thimble“; − Neutron source rod, specific instrumentation, or a test fuel segment can be added. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
55 Fuel fabrication PWR fuel • Russian PWR reactors are usually known by the Russian acronym VVER. • Fuel assemblies are characterized by their hexagonal arrangement, − but are otherwise of similar length and structure to western PWR fuel assemblies. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
56 Fuel fabrication BWR fuel • BWRs also use fuel rods comprising zirconium-clad uranium oxide ceramic pellets. • Their arrangement into assemblies is again based on a square lattice, typically with pin geometries ranging from 6 x 6 to 10 x 10. • BWR fuel assemblies operate more as individual units, and different designs may be mixed in any core load, giving flexibility to the utility in fuel purchases. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
57 Fuel fabrication BWR fuel • BWR fuel is fundamentally different from PWR fuel in certain ways: − Four fuel assemblies and a cruciform shaped control blade form a 'fuel module', − Each assembly is isolated from its neighbours by a water-filled zone in which the cruciform control rod blades travel (they are inserted from the bottom of the reactor), − Each BWR fuel assembly is enclosed in a zirconium alloy sheath which directs the flow of coolant water through the assembly and during this passage it reaches boiling point, and − BWR assemblies contain larger diameter water channels – flexibly designed to provide appropriate neutron moderation in the assembly. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
58 Fuel fabrication BWR fuel Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
59 Fuel during power generation • The fuel assemblies are loaded into the reactor core. • The reactor core is housed in the reactor pressure vessel (RPV). − It consists of a heavy-walled reactor vessel with all its necessary support and coolant flow guiding structures. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
60 Fuel during power generation • Arrangement of fuel assemblies in the core is dictated by three goals: 1. Ensuring uniform power level over the core, 2. Maintaining the integrity of the fuel elements, 3. Optimising fuel burn-up within safe operating limits. • Nuclear fuel operates in a harsh environment; − High temperature, chemical corrosion, radiation damage and physical stresses. • Fuel assemblies are designed so that at their projected maximum burn-up level their risk of failure is still low. − Fuel ‘failure’ – cladding has been breached and radioactive material leaks into the reactor coolant water. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
61 Fuel during power generation • Fuel leaks have a big impact on reactor operations; − Increase operator doses − Not a major risk to plant safety. • Primary coolant water is continuously monitored. • The permissible level of released radioactivity is strictly regulated against specifications. − Specifications take into account the continuing safe operation of the fuel. • Depending on its severity a fuel leak will require different levels of operator intervention. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
62 Fuel during power generation • Very minor leak: no change to operations – the faulty fuel assembly with leaking rod(s) is removed at next refuelling, inspected, and possibly re-loaded. • Small leak: allowable thermal transients for the reactor are restricted. This might prevent the reactors from being able to operate in a “load-follow” mode and require careful monitoring of reactor physics. The faulty fuel assembly with leaking rod is generally removed and evaluated at the next scheduled refuelling. • Significant leak: the reactor is shut down and the faulty assembly located and removed. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
63 Fuel during power generation • Usually a replacement assembly is needed if there is a leaking fuel rod: − Replacement fuel is one cost component associated with failed fuel. − There is also the cost penalty and/or replacement power from having to operate at reduced power or having an unscheduled shutdown. − There may also be higher operation and maintenance costs associated with mitigating increased radiation levels in coolant decontamination. • Fuel management is a balance between the economic imperative to burn fuel for longer and the need to keep well within failure-risk limits. − Improving fuel reliability extends these limits, and therefore is a critical factor in providing margin to improve fuel burn-up. • Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
64 Fuel during power generation • There is a limit on how far fuel burnup can be extended; − Due to the strict criticality safety limitation imposed on fuel fabrication facilities such that a maximum uranium enrichment level of ~5% can be handled. • Higher burnup does not necessarily mean better energy economics. − Utilities must carefully balance the benefits of greater cycle length against higher front-end fuel costs (uranium, enrichment). − Refuelling outage costs may also be higher, depending on length, frequency and the core re-load fraction. • An equally important trend in nuclear fuel engineering is to be able to increase the power rating for fuels. − Limited by the material properties of the cladding. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
65 Spent fuel storage • Spent nuclear fuel is generated from the operation of nuclear reactors of all types. • Spent fuel needs to be safely managed may be considered waste. • Management options may involve: − Direct disposal (as part of what is generally known as the “once through fuel cycle” or “open cycle”) or − Reprocessing (as part of what is generally known as the “closed fuel cycle”). • Either management option necessarily includes storage of the spent fuel for some period of time. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
66 Spent fuel storage • The time period for storage can differ from a few months to several decades. − The time period for storage will be a significant factor in determining the storage arrangements adopted. • Storage options include: − Wet storage in some form of storage pool or − Dry storage in a facility or storage casks built for this purpose. • Storage casks can be located in a designated area on a site or in a designated storage building. • Irrespective of the consideration of spent fuel (either waste or an energy resource), the safety aspects for storage remain the same as those for radioactive waste, which are established in the IAEA GSR Part 5. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
67 Spent fuel storage Pool storage • Reliable radiation shielding and cooling is necessary. − Radionuclides in spent fuel still generate a significant amount of heat and are an extremely high source of radiation. • Cooling and shielding are accomplished by a deep pool of water adjacent to the reactor. − Spent fuel is transferred in it after discharge from the reactor. • Damaged fuel elements are inserted in special racks. − To prevent radioactive contamination of cooling water − To maintain safe geometry, avoiding criticality. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
68 Spent fuel storage Pool storage • Water cooling and shielding is necessary for the first few of years after discharge. − During this period, loss of coolant could result in fuel overheating. • If the spent fuel pool water evaporates, exposing the fuel cladding to air, the fuel temperatures could be high enough for fuel cladding to oxidise in air and lose its integrity. − Resulting in a release of volatile fission products. • The accident in Fukushima exposed the vulnerability of spent fuel pools. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
69 Spent fuel storage Pool storage • Cooling ponds at reactors were, originally designed to hold only a few years’ discharges. − Expectation was that, within a few years, the spent fuel would be shipped to a reprocessing plant. • Operators responded first by increasing the storage density of the spent fuel. − Each fuel assembly is usually enclosed in a rack, usually lined with neutron-absorbing plates to assure that the arrangement is sub-critical. • Spent fuel pools at heavy water reactors are designed with about ten years of storage and do not need to have racks with additional neutron-absorbing plates to ensure sub-criticality. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
70 Spent fuel storage Dry storage • Spent fuel can be stored in a dry storage − Generation of heat and radiation reduces with time. − After several years, air cooling is sufficient but significant shielding is still required for radiation protection. • The intense gamma radiation emitted by spent fuel requires that fuel is loaded into the casks for dry storage under water or remotely behind shielding. • Compared to spent fuel pools, casks for dry storage are passive, and resistant to aircraft crash and earthquakes. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
71 Spent fuel storage Dry storage • Dry storage casks may consist of steel a canister to hold the spent fuel, surrounded by a heavy shell of reinforced concrete for protection and radiation shielding. • Cooling is provided by natural convection of air. • A more compact design has the canisters inserted horizontally or vertically into a concrete monolith (vault), sized to hold six or more canisters with channels for convective air cooling. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
72 Spent fuel storage Dry storage • The area density of dry cask storage is about 0. 1 ton/m 2. • The lifetime output of a 1 GWe LWR, about 1200 tons of spent fuel discharged during a 60 -year lifetime, could therefore be stored on an area about 100 m 2. • Such an area is easily available within the exclusion zone associated with most nuclear power plants. • Dry storage can be in the open or inside a building. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
73 Geological disposal • Disposal and isolation from the biosphere for hundreds of thousand years could be accomplished by burying the spent fuel or HLW in a mined repository some hundreds of meters underground. • Some radionuclides in spent fuel are very long-lived. • After about 100 years, the radiation level has significantly reduced since the most radioactive, short-lived radionuclides have decayed. • The remaining radioactivity is dominated by plutonium and americium isotopes. − Alpha emitters, − Reduced risk of external exposure, − Significant health risk if ingested or inhaled. − Their radioactivity also has very specific impacts on the strategy for geological disposal. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
74 Geological disposal • The geochemistry of these and other long-lived radioisotopes in a geological medium is important to the science of geological disposal. • The geological conditions of the repository should minimize release from the waste form. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
75 Spent fuel reprocessing • Reprocessing or recycling is a mechanical and chemical process in which spent fuel is separated into different materials (uranium plutonium, minor actinides, fission products, structural materials). • The main purpose is to extract the remaining fissile material, in particular plutonium, from the spent fuel. • Reprocessing was developed before power reactors, to support weapon programmes. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
76 Spent fuel reprocessing Reasons for reprocessing • In power reactors used nuclear fuel is reprocessed to extract fissile materials: − For recycling and to supply Fast Breeder Reactors − To reduce the volume of high-level wastes − To allow plutonium destruction by further irradiation in a reactor. • New reprocessing technologies are being developed to be deployed in conjunction with fast neutron reactors which will burn all longlived actinides. − A significant amount of plutonium recovered from used fuel is currently recycled into MOX fuel. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
77 Spent fuel reprocessing Reasons for reprocessing • Over the last 50 years the main reasons for reprocessing were: − To recover unused uranium and plutonium in the used fuel elements and thereby close the fuel cycle, gaining some 25% to 30% more energy from the original uranium. − To reduce the volume of material to be disposed of as high-level waste to about one fifth. • In the last decade interest has grown in recovering all long-lived actinides together so as to recycle them in fast reactors so that they are converted to short-lived fission products. − Reducing the long-term radioactivity in high-level wastes, and − Reducing the possibility of plutonium being diverted from civil use. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
78 Spent fuel reprocessing Products of reprocessing • The composition of reprocessed uranium (Rep. U) depends on: − Initial enrichment and − Time the fuel has been in the reactor. • It is mostly 238 U. Normally will have about 0. 5% 235 U and smaller amounts of 232 U and 236 U created in the reactor. • Plutonium from reprocessing will have an isotopic concentration determined by the fuel burn-up level. − Higher burn-up levels increase the proportion of non-fissile, neutron absorbing isotopes. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
79 Spent fuel reprocessing Mixed Oxide (MOX) fuel • Mixed oxide (MOX) fuel is produced by mixing uranium dioxide (UO 2) and plutonium dioxide (Pu. O 2). • It provides about 2% of the new nuclear fuel used today. • It is manufactured from plutonium recovered from spent reactor fuel. • The amount of separated uranium and plutonium which may be recycled, including from ex-military sources is equivalent to about three years' supply of natural uranium from world mines. • In addition, there about 1. 6 million tonnes of enrichment tails, with recoverable fissile uranium. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
80 Spent fuel reprocessing Mixed Oxide (MOX) fuel • Today MOX is used in: − Europe (about 40 reactors use it), and − Japan (about 10 reactors). • These reactors generally use MOX fuel as about 1/3 of their core (some will up to 50%). • Advanced light water reactors such as the EPR or AP 1000 are able to accept complete fuel loadings of MOX if required. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
81 Spent fuel reprocessing Mixed Oxide (MOX) fuel • Use of up to 50% MOX − No changes in operating characteristics of a reactor, − Designed or adapted slightly – additional control rods are needed. • Use of more than 50% MOX − Significant changes are necessary, − Accordingly designed reactor. • Burn-up of MOX fuel is about the same as that for UO 2 fuel. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
82 Spent fuel reprocessing Mixed Oxide (MOX) fuel • Reprocessing and MOX becomes more economic as uranium prices rise, but fuel fabrication has to cope with highly radiotoxic plutonium. • MOX becomes more attractive as the need to reduce the volume of spent fuel increases. − 7 UO 2 fuel assemblies are used to produce one MOX, resulting in only about 35% of the volume and mass for final disposal. • MOX is also used in fast neutron reactors in several countries, particularly France and Russia. − It was first developed for this purpose. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
83 TRANSPORT OF NUCLEAR MATERIALS Learning objectives After completing this chapter, the trainee will be able to: 1. To state major Safety standards related to safe transport of radioactive materials; 2. Summarise objectives of regulations for safe transport; 3. Describe graded approach and safety principles in regulations; 4. Describe different forms of radioactive materials; 5. Explain the importance of A 1 and A 2 values; 6. Describe different packages; 7. Describe the usage of Transport index and Criticality safety index; 8. Describe the importance of labelling; 9. State what type of packaging is used for transport of LILW waste; 10. Discuss requirements for packages for transportation of spent fuel. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
84 Regulatory framework • Since 1961 IAEA has published regulations for the safe transport of radioactive material. • Requirements based on the IAEA regulations have been adopted worldwide by IAEA Member States and international and regional organisations − The most important international organisations: International Maritime Organization (IMO), International Civil Aviation Organization (ICAO), International Air Transport Association (IATA), and Universal Postal Union (UPU) − Some regional agreements: The Regulations Concerning the International Carriage of Dangerous Goods by Rail (RID), The European Agreement concerning the International Carriage of Dangerous Goods by Road (ADR), The European Agreement concerning the International Carriage of Dangerous Goods on Inland Waterways (ADN), and The Regulations for the Transport of Dangerous Goods on the Rhine (ADNR). Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
85 IAEA transport regulations • Regular revisions • Latest revision: Regulations for the Safe Transport of Radioactive Material, Safety Requirements SSR-6, Vienna (2012), − Regulations are based on Fundamental Safety Principles, Safety Fundamentals No. SF-1 and International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, Safety Series No. 115 − The Regulations are supplemented by Safety Guides: − TS-G-1. 1: Advisory Material for the IAEA Regulations for the Safe Transport of − − Radioactive Material. TS-G-1. 2 (ST-3): Planning and Preparing for Emergency Response to Transport Accidents Involving Radioactive Material. TS-G-1. 3: Radiation Protection Programmes for the Transport of Radioactive Material. TS-G-1. 4: The Management System for the Safe Transport of Radioactive Material. TS-G-1. 5: Compliance Assurance for the Safe Transport of Radioactive Material. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
86 IAEA transport regulations (Cont. ) • Regulations apply to the transport of radioactive material by all modes on land, water, or in the air. • Fundamental Safety Principles: The prime responsibility for safety must rest with the person or organization responsible for facilities and activities that give rise to radiation risks. − For transport of radioactive material: the consignor should be responsible for safety during transport. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
87 IAEA transport regulations (Cont. ) • Objective of regulations: To establish requirements that must be satisfied to ensure safety and to protect people, property and the environment from the effects of radiation during the transport of radioactive material. • Protection is achieved by: − containment of radioactive contents; − control of external radiation levels; − prevention of criticality; and − prevention of damage caused by heat. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
88 IAEA transport regulations (Cont. ) • The requirements are satisfied by: − graded approach to contents limits for packages and conveyances, − graded approach to performance standards applied to package designs depending on the hazard of the radioactive contents, − imposing conditions on the design and operation of packages and on the maintenance of packagings, including consideration of the nature of the radioactive contents, and − requiring administrative controls, including, where appropriate, approval by competent authorities. • Confidence in compliance with regulations is achieved through management system and compliance assurance programmes. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
89 IAEA transport regulations (Cont. ) • Application of the requirements relies on the principles of inherent safety, passive safety and active safety controls. • These principles are incorporated in regulations through − limiting the nature and activity of the radioactive material which may be transported in a package of a given design; − specifying design criteria for each type of package; − providing information on hazards by labels, marking, placards, and shipping papers; − applying simple rules of handling and stowage of the packages during transport and in-transit storage. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
90 The philosophy of the regulations • Packages of radioactive material should be dealt with in the same way as other hazardous goods; • Safety depends primarily upon the package and not on operational controls; • The consignor should be responsible for ensuring safety during transport through proper characterization of the contents, proper packaging of those contents, and properational actions including adequate communications (i. e. shipping papers, marking, placarding and labelling, Transport indexes, Criticality safety indexes, approval certificates, proper shipping names and UN numbers). Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
91 Terminology of the regulations • Package: radioactive content + packaging. • Overpack: an enclosure such as a box, used to consolidate one or • • • more packages so they may be treated as one. Consignment: package(s) or load of radioactive material that is presented for transport. Consignor: the individual or organization that prepares a consignment for transport. Consignee: the corresponding agent that receives the consignment. Carrier: an individual or organization that undertakes the carriage of radioactive material. Conveyance: any means by which the package is transported. Exclusive use: when a single consignor has sole use of the conveyance, such that all loading and unloading is carried out in accordance with the directions of the consignor or consignee. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
92 Forms of radioactive materials • Radioactive materials being transported have elevated radionuclide content. Other characteristics (state, composition, shape, size, chemical properties, etc. ) could vary extensively. • The most important characteristics are those that determine exposure of people during normal transport activities and potential exposures in accident conditions. − radioactive materials are categorised in different classes regarding radioactivity (and consequently radiation levels) and possibility of dispersion of radioactive material. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
93 Forms of radioactive materials (Cont. ) • Materials exempted from regulations: Materials with very low activities or very low specific activities, or approved consumer products with incorporated radioactive materials, present insignificant hazard in any situation. − The basis and numerical values for these exemptions are the same as those in International Basic Safety Standards. • Excepted materials: Materials with low activities, or low specific activities above exemption levels and below certain prescribed levels. If released from package due to accident, these materials present insignificant hazard in foreseeable situations. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
94 Forms of radioactive materials (Cont. ) • Low specific activity (LSA) material: radioactive material that by its nature has low activity per unit mas. − LSA-I: intrinsically radiologically safe due to low specific activity of material, − LSA-II: materials with activity distributed throughout material (it could be solid, gas or liquid) and specific activity below certain level. − LSA-III: similar to LSA-II, but higher activity level and only solid material • Surface contaminated object (SCO): solid object which is not radioactive by itself, but has contaminated surface. − There are two subcategories SCO-I and SCO-II, depending on the level of fixed and non-fixed contamination. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
95 Forms of radioactive materials (Cont. ) • Special form: either indispersible solid radioactive material or sealed capsule containing radioactive material. − Material which we usually describe as “sealed source”. − It can give rise to high radiation levels, but it is unlikely that it will cause any contamination hazard. − Since this property must be valid also in transport accident, stringent qualification is required. − For materials not qualified in this way contamination must be always considered as a probable consequence of an accident • Fissile material: material containing 233 U, 235 U, 239 Pu, 241 Pu, or any combination of these radionuclides that has the capability of undergoing nuclear fission. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
96 A 1 and A 2 values • Hazard assessment is used to determine radioactive material activities which could be transported in packages with certain properties and ability to withstand particular conditions of transport. • A 1 and A 2 are activities, which are result of dose assessment (this includes external effective and committed effective dose due to material released from package) to a person involved in transport accident, where radioactive material was transported in package designed to withstand normal conditions of transport (but not accidents!) • Starting from assessed total effective dose 50 m. Sv, values A 1 and A 2 are calculated for special form material (A 1) and other than special form radioactive material (A 2). Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
97 A 1 and A 2 values (Cont. ) • Values A 1 and A 2 are listed in regulations and serve to determine the type of packaging necessary for a particular radioactive material − For example, activity limits for excepted packages for most of the radionuclides in solid form are 10 -3 A 1 for special form material and 10 -3 A 2 for other forms. − For liquids, this activity limit is 10 -4 A 2, and for instruments with radioactive materials 10 -2 A 1 and 10 -2 A 2. • Type of package for which A 1 and A 2 where calculated as limiting values has been designated as Type A package. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
98 Classification of packages • Graded approach to packaging: more hazardous material (i. e. higher activity of the same radionuclide) requires more “resistive” packaging to different conditions of transport • In regulations three categories of conditions are used: − Routine conditions (incident free, only conventional stresses and strains resulting from transport and handling); − Normal conditions (with minor mishaps, like typical incidents as being rained upon, being dropped, having other packages stacked on top); − Accident condition. • Packages for radioactive materials with activities below A 1 for special form (or A 2 for other forms) are designed to withstand normal conditions (or even only routine conditions), while packages for materials with higher activities must survive accident conditions Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
99 Classification of packages (Cont. ) • Excepted packages are intended for transport of small quantities of radioactive materials (excepted materials or limited number of instruments containing radioactive material). • Should withstand routine conditions of transport and there are no special test requirements. • It must be assumed that package will fail in any form of accident. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
100 Classification of packages (Cont. ) • Industrial packages are used for transport of LSA and SCO material. • Three types of industrial packages: Type IP-1 (designed for routine conditions of transport), Type IP-2 and Type IP-3. • They must withstand normal conditions of transport, which is verified through testing. • Although special packaging could be acquired, many packages from industry, such as steel drums and bins could pass the testing. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
101 Classification of packages (Cont. ) • Type A packages are intended for safe and economical transport of radioactive materials with activities up to A 1 (special form) or A 2 (other forms). • They must maintain integrity under normal conditions of transport (these also include falling from the vehicle, being exposed to water for limited time, being stuck by sharp object, or having other objects stacked on top). − Special tests were developed. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
102 Classification of packages (Cont. ) • Type B packages should withstand most accident conditions without failure of containment or decrease of shielding ability. − They must pass series of mechanical and thermal tests where effects are cumulative. • The limit to the content is not imposed by regulations, but with approval certificate. • Type B(U) and Type B(M) packages • Type B packages are used for variety of sources, from sources for industrial radiography, gauges, to used fuel (“casks”) and vitrified waste. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
103 Classification of packages (Cont. ) • Type C packages are designed to survive the most demanding condition of severe accident. − They are intended for air transport and are tested to conditions that may be encountered in a severe air accident. • Fissile packages are designed to fulfil requirements for the radioactive material and also to ensure criticality safety under a variety of postulated conditions. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
104 Transport index (TI) • Transport index (TI) is a number that is assigned to a package (or overpack, freight container, or conveyance), which is used to provide control over groups of packages for the purposes of minimizing radiation exposure risks. • Transport index (TI) is measured (or calculated for uranium and thorium ores or concentrates) as maximum dose rate at 1 m from the external surface of package in m. Sv/h multiplied by 100. − Example: if the highest measured dose rate at 1 m from the surface of package is 0. 1 m. Sv/h, then the TI is 10. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
105 Transport index (TI) (Cont. ) • TI=10 is the limiting value for a package (or an overpack), which is not a part of consignment under exclusive use − The highest value of dose rate on the external surface of the package is limited to 2 m. Sv/h. • TI for conveyance is sum of TIs of packages or overpacks. − TIs of conveyances, which are not under exclusive use, are limited to 50 (or 200 for cargo planes, or even without limit for large freight containers on seagoing vessels) Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
106 Criticality safety index (CSI) • Criticality safety index (CSI) is number that is used to provide control over the accumulation of packages, overpacks or freight containers containing fissile material. • CSI is result of calculation which takes into account the maximum number of identical packages under consideration which is still subcritical under conditions that provide the maximum neutron multiplication, considering proper safety margin. • CSI for package or overpack in consignment, which is not under exclusive use, is limited to 50. − Sum of CSIs for conveyance, which is not under exclusive use, must be below 50. In case of exclusive use, sum of CSI must be below 100, or even without limit for large freight containers on seagoing vessels Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
Marking, labelling and placarding and shipping papers 107 • Marking, labelling and placarding are part of active safety controls • They provide permanent information on hazards to people involved in normal transport operations, and also to first responders involved in emergency response in case of an accident • All packages with radioactive material must have on the outer surface marking “UN” followed by the UN number. − UN number uniquely identifies the package content as a specific category of dangerous goods. • All packages but excepted must have marked also proper shipping name and type of package approval (e. g. TYPE IP-1, TYPE A, or TYPE B(M)). − Packages TYPE B(U), TYPE B(M) and TYPE C also must have an identification mark allocated to that design by the competent authority. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
Marking, labelling and placarding and shipping papers – Labels • Each package, overpack and freight container shall bear one of the labels with trefoil symbol • The name of radionuclide, activity of the radioactive material, and Transport index should be written on the label. • Number 7 on labels denotes Class 7 according to UN classification of dangerous goods • Packages with fissile material should also have additional label with inscription “FISSILE” and Criticality safety index Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 108
Marking, labelling and placarding and shipping papers – Labels (Cont. ) • The category of label depends on the TI: − for TI = 0 (in fact for TI < 0. 05) category I-WHITE is chosen, − for 0 < TI < 1 category II-YELLOW is chosen, and − for 1 < TI < 10 category III-YELLOW is chosen. − For TI > 10 category III-YELLOW should be used for consignment under exclusive use. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 109
Marking, labelling and placarding and shipping papers – Placards • Large freight containers, rail and road vehicles carrying packages, overpacks or freight containers labelled with any of the labels should bear placards with trefoil symbol − Placard is similar to labels for packages, except for the size (it is larger) and there is no data printed on placard. − The orange placard must contain the UN number of the radioactive material. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 110
Marking, labelling and placarding and shipping papers – Shipping papers 111 • For each consignment, the consignor shall deliver transport documents with: 1. Basic data regarding radioactive material and packages (UN number, proper shipping name, primary hazard Class number 7, subsidiary hazard class division number, name of radionuclide, maximum activity, category of package, TI, etc. ); 2. A statement regarding actions, if any, that are required to be taken by the carrier; 3. Supplementary requirements for loading, stowage, carriage, handling and unloading of the package, overpack or freight container; 4. Restrictions on the mode of transport or conveyance and any necessary routeing instructions; and 5. Emergency arrangements appropriate to the consignment. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
112 Transport of LLW and ILW Low-level and intermediate-level wastes (LLW and ILW) are generated throughout the nuclear fuel cycle and from the production of radioisotopes used in medicine, industry and other areas. • Low-level radioactive wastes: − low levels of radiation, slightly above normal background; − Most often solid materials (clothing, tools, contaminated soil, etc. ) − Transported from its origin to waste treatment sites, or to an intermediate or final storage facility. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
113 Transport of LLW and ILW Low-level wastes: • The packaging used for the transport of low-level waste does not require special shielding. • Typically, transported in 200 litre drums (TYPE IP packages), often compacted • 36 standard, 200 litre drums go into a 6 -metre transport container. • Low-level waste is moved by road, rail, and by sea. Most low-level waste is only transported within the country where it is produced. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
114 Transport of LLW and ILW Much ILW comes from NPPs and reprocessing facilities: • Broad composition; • Shielding required. ILW are taken from their source to an interim storage site, a final storage site (as in Sweden), or a waste treatment facility. They are transported by road, rail and sea. The radioactivity level of ILW is higher than LLW: • The classification of ILW is decided for disposal purposes, not on transport grounds; • The transport of ILW takes into account any specific properties of the material. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
115 Transport of spent fuel In some occasions spent fuel has to be transported off site: • Sea transport has been used from Japan to Europe and from the continent to the UK. • Most of the transport within continental Europe and Russia is by train. • Smaller casks, containing 0. 5 to 2 tons of spent fuel, are transported by truck. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
116 Transport of spent fuel Transport casks are TYPE B(U) packages: • Thick metal walls; • Inner layer of lead for gamma-ray absorption; • Outer layers may contain polymer (including hydrogen in to slow neutrons) and boron to absorb the slowed neutrons. The basis for safety requirements is the IAEA Safety Requirement, SSR-6. The application for a transport license is made by Safety Analysis Report (SAR). The cask must satisfy the requirements for routine, normal and accidental conditions identified within the IAEA regulations. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
117 Transport of spent fuel The contents of a SAR include: • General – package identification, use, purpose, description, contents, principal design criteria, compliance regulations, requirements and acceptance criteria: − Structural − Thermal − Containment − Shielding − Criticality • Operating procedures • Maintenance programme Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
Transport of spent fuel Acceptance tests, maintenance programme and monitoring Prior to first use, it is necessary to demonstrate that the cask conforms to the safety requirements outlined in the SAR. These tests are performed during manufacturing, commissioning and before the cask’s first shipment. Acceptance tests to be considered are: 1. Manufacturing tests, including those applicable by construction codes within a specific country, such as: − Acceptance inspections, including dimensional checks, visual inspections for defects, and weld and fabrication examinations; − Lifting points overload tests for trunnions and lifting features; Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 118
Transport of spent fuel Acceptance tests, maintenance programme and monitoring − Containment testing, including leak tightness tests for seals and hydraulic structural tests on the cask cavity, if applicable; − Thermal tests to confirm that the cask operates in normal use as predicted by analysis; − Functional tests to confirm the operation of cask components, such as valve and orifice function, component fit-ups and general operations; − Impact limiter qualification testing. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 119
Transport of spent fuel Acceptance tests, maintenance programme and monitoring 2. Commissioning tests: − Handling trials; − Plant interfaces; 3. Performance tests before the first shipment: − Containment system – leak-tightness of seals, if applicable; − Radiation shielding; − Thermal shielding; − Heat dissipation characteristics - normal conditions; − Confinement system; − Presence of neutron poisons. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 120
Transport of spent fuel Acceptance tests, maintenance programme and monitoring Maintenance programme • The purpose: to maintain the integrity of the cask (compliant with the SAR and the licence conditions). The SAR is required to outline the maintenance programme for the cask once the cask is in operation. • For transport casks, this is based on either the number of transport cycles completed or by periodic maintenance based on the cask’s time in operation. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 121
Transport of spent fuel Acceptance tests, maintenance programme and monitoring A typical maintenance programme for a transport cask should consider periodic inspection and testing that covers the following: • General condition for damage and deterioration in particular parts and components for containment and lifting features; • Structural and pressure tests, including cavity overpressure testing (hydraulic tests, if applicable) and lifting point load tests; • Leak tightness tests, confirming the continued acceptance of seals if applicable; • Functional tests of the components; • Presence of neutron absorbers with fuel baskets or containers; • Confirmation of thermal and shielding performance. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 122
Transport of spent fuel Acceptance tests, maintenance programme and monitoring Monitoring: to ensure that the operation of the cask is in accordance with the license conditions prior to cask loading, during transport and following cask unloading. Operational checks before loading include: • Turnaround maintenance checks and approvals (these include inspection, tests and repair / replacement and maintenance); • Visual checks; • Functional tests, to confirm that the operation of cask components is acceptable, in particular orifices or valves; • Containment tests, such as pressure testing the seal interspaces to confirm that leak tightness is in accordance with the SAR. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 123
Transport of spent fuel Acceptance tests, maintenance programme and monitoring Operational checks in preparation for transport, during loading and after loading, include: • Cavity dryness and pressure/inert atmosphere (dry cask); • Cavity ullage and inert atmosphere (wet cask); • Seal tests, to confirm that the leak tightness of the containment boundary; • Radiation dose measurements, ; • Contamination check; • Temperature measurements, which includes the measurement of external surfaces temperatures. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 124
Transport of spent fuel Acceptance tests, maintenance programme and monitoring Physical protection • Physical protection measures include designed features, security measures, and various administrative controls. For the case of spent fuel transport, these may include the attachment of IAEA type seals to the cask prior to transport and the confirmation that the seals are intact at receipt of the cask following transport. Integrated Management System (IMS) • All operation and maintenance steps must be subject to IMS rules, including unambiguous step-by-step instructions that are easy for the personnel to follow. IMS programmes are required to cover the design, manufacture, testing, operation and maintenance of the cask. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 125
126 Transport of plutonium • Plutonium: − Radioactive transuranic element, − Normally produced in nuclear reactor as a by-product, − 15 isotopes, − Masses from 232 to 246, − Half-lives from 20 minutes to 76 million years. • Principal form of radiation is alpha particles; • − Relatively harmless outside the body, − Primary hazard from inhalation. 239 Pu and 241 Pu are fissile isotopes – transportation must be designed in the way that critical mass cannot be formed. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
127 Transport of plutonium • Usually transported as a metal, as an oxide or as MOX fuel; − Plutonium oxide is a chemically unreactive ceramic material, − Plutonium metal is relatively insoluble, but dissolves in acid or strong carbonate solution. • Plutonium is a strategic material; − Any amount transported (or released) must be under effective stewardship and carefully accounted for. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
128 Transport of plutonium • Plutonium is transported in different types of sealed packages, each can contain several kilograms of material. • Criticality is prevented by: − The design of the package, − Limitations on the amount of material contained within the package, and − The number of packages carried on a transport vessel. − Special physical protection measures apply to plutonium consignments. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
129 Transport of plutonium • A typical transport consists of one truck carrying a protected shipping container. − The container holds a number of packages with a total weight varying from 80 to 200 kg of plutonium oxide. • A sea shipment may consist of several containers; − Each of them holding between 80 to 200 kg of plutonium in sealed packages. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
130 Transport of vitrified waste • The highly radioactive wastes (especially fission products) are segregated and recovered during the reprocessing operation. • These wastes are incorporated in a glass matrix by a process known as 'vitrification', which stabilises the radioactive material. • The molten glass is poured into a stainless steel canister; − Where it cools and solidifies. • A lid is welded into place to seal the canister. • The canisters are then placed inside a Type B cask; − Similar to those used for the transport of used fuel. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
131 Transport of vitrified waste • Typically a vitrified waste transport cask contains up to 28 canisters of glass. • Return vitrified waste shipments from Europe to Japan started in 1995: − Up to 28 canisters (total 14 tones) are packed in each 94 -tonne steel transport cask. • In 1995 -2007, twelve shipments were made from France of vitrified HLW comprising 1310 canisters containing almost 700 tones of glass. • Return shipments from the UK have commenced, and there will be about 11 shipments over eight years. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
SAFETY ASPECTS OF THE NUCLEAR FUEL CYCLE Learning objectives After completing this chapter, the trainee will be able to: 1. Describe criticality safety aspects in the nuclear fuel cycle; 2. Describe radiation safety aspects of the NFC; 3. Describe chemical hazards in the NFC; 4. Describe fire and explosion hazards in the NFC; 5. Describe safety aspects of effluents in the NFC; 6. Summarise other safety aspects of NFC installations. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 132
Safety aspects in different phases of the fuel cycle - summary 133 • Mining and milling The primary risk is due to hazards found in any ore mining and milling operation. There is no criticality hazard and little fire or explosive hazard. • Conversion Chemical hazards dominate those from radiation. • Enrichment Criticality safety becomes important. There also chemical and radiological hazards from any UF 6 release. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
Safety aspects in different phases of the fuel cycle - summary 134 • Fuel fabrication Chemical, radiological, and criticality hazards similar to hazards at enrichment plants. Criticality risk, radiation and radiotoxicity are significantly higher in the case of MOX fuel. • Power operation Radioactive material in nuclear fuel can present a significant hazard during reactor operations. Fuel integrity should be assured in normal operational conditions and in most reactoraccident conditions. Coolant chemistry and operational practice should be designed to avoid adverse effects on fuel materials and to reduce hazards arising from the transport of radioactive contamination in the coolant. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
Safety aspects in different phases of the fuel cycle - summary 135 • Spent fuel storage Criticality and radiological hazards for people and environment. • Reprocessing Wide range of significant radiological, chemical and criticality risks. • Transport of spent fuel The primary risk is from the radioactive inventory. Prevention and detection of, and response to, theft, sabotage, unauthorised access and illegal transfer or other malicious acts involving nuclear material and other radioactive substances must be ensured in all stages of the nuclear fuel cycle. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
136 Criticality • Concerned with preventing inadvertent nuclear chain reaction; • Potentially lethal effects of gamma and neutron radiation doses to workers and possible release of fission products; • Important when handling enriched uranium or plutonium. • One of the dominant safety issues for fuel cycle facilities: − Great diversity of technologies and processes; − Fissile Materials (fuel pellets, rods and elements, solutions, slurries, liquids, gases, powders, etc. ) exist in some of these facilities; − This requires operator attention to account for this material throughout the facility. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
137 Criticality • Criticality hazards may be plausible during: − enrichment, − reprocessing, − (enriched) uranium fuel fabrication, − MOX fuel fabrication, − fresh fuel storage (and transportation), − spent fuel storage (and transportation), − waste treatment, and waste disposal facilities need to be evaluated, first by engineering design. • This includes assurance that excessive amounts of fissile material do not accumulate above the specified limits in vessels, transfer pipes and other parts of the facility. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
138 Criticality Process safety limits must not be violated and must have enough margins to preclude criticality during postulated abnormal operating condition. Particular attention should be paid to: • fissile material in liquid, solid and gaseous waste streams, • process changes or modifications that may be inadequate (from the point of view of criticality), • nuclear material accounting and control procedures that may lack the appropriate accuracy to ensure subcriticality, • and controls that are used to prevent the accumulation of nuclear materials outside intended locations. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
139 Criticality Modes of control of criticality safety shall include any one or a combination of the following: − control of the mass of fissile material present in an equipment; − control of the geometry of the process equipment; − control of the concentration of fissile material in solutions; − presence of neutron absorbers; − limitation of moderation. Fuel cycle facilities with credible criticality hazards should have in place a programme to ensure subcriticality. Provision should be made to cope with an accident and to notify the facility personnel should an inadvertent criticality occur. Emergency arrangements must be in place and should be tested regularly. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
140 Criticality accidents Most have happened in the early years of using nuclear power; 22 have been reported in process facilities up to the year 2000 [23]. Tokaimura uranium processing facility accident, Japan, 1999. Three operators were engaged in processes to produce a 18. 8% enriched uranium solution for shipment. The procedure used deviated from the licensed one: − The uranium solution was being placed in a geometrically unfavourable precipitation tank; − A criticality excursion occurred when two workers added an excessive amount of uranium solution to the tank; − The workers observed a blue flash and fled the location; gammaradiation detectors went off prompting all workers to evacuate to a muster area (there was no criticality evacuation alarm system). Two workers subsequently died from radiation exposure. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
141 Radiation safety Radiation protection programme should be in place Potential accident sequences are considered to minimise radiological risk. Fuel cycle facility radiation protection practices must include: • A documented programme to ensure that occupational radiological protection is optimised; • Adequate qualification requirements for the radiation protection personnel; • Approved written procedures; • Radiation protection training for all personnel who have access to radiologically restricted areas; Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
142 Radiation safety Radiation protection programme should be in place, contd. : • Engineering controls (passive and active) for airborne radioactive material, administrative controls (designation of areas by contamination risk etc. ) and, where necessary, respiratory protection; • A radiation survey and monitoring programme that includes requirements for control of radioactive contamination within the facility and monitoring of external and internal radiation exposures and taking appropriate action; • Programmes to maintain records, to report radiation exposures to the regulating authority, and to reinstate an acceptable in-plant radiological environment in the event of an incident. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
143 Radiation safety Design features of nuclear installations to prevent radiation exposure Preventive and protective measures are always taken to reduce the external exposure, and especially to reduce the hazard to workers associated with ingestion or inhalation of radioactive substances. The most important engineered measures to accomplish this are: • Containment (e. g. hot cells, glove boxes, sumps and bunds); • Ventilation to maintain air flows into the most contaminated areas; • Ventilation/off-gas cleaning equipment (e. g. scrubbers, electrostatic precipitators and HEPA filters). Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
144 Chemical hazards • Fuel cycle facilities may also be considered as chemical plants for processing of nuclear materials. This involves the use of strong reagents to dissolve the materials so that the subsequent chemical reactions may take place. • The use of uranium hexafluoride (UF 6) in conversion facilities involves the handling of significant quantities of highly reactive hydrogen fluoride which is chemotoxic and poses a significant hazard to workers. • Other chemicals encountered at fuel cycle facilities in industrial quantities include ammonia, nitric acid, sulphuric acid, phosphoric acid, and hydrazine. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
145 Chemical hazards • Strong acids are used to dissolve uranium, other materials and also the spent fuel during fuel element reprocessing, enabling the separation of the plutonium and uranium from the fission products. • The separated fission products, which comprise 99% of the radioactivity in the spent fuel, pose a significant radiological hazard in what is typically a complex chemical slurry. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
146 Chemical hazards • Unplanned release of chemicals may adversely affect safety controls. For example, hydrogen fluoride could disable an operator whose normal activities may be relied upon to ensure safe processing. • A robust chemical risk control process will include process descriptions to support an understanding of the chemical process risks (including radiological risks caused by or involving chemical accidents) and would allow understanding of protective measures and potential accident sequences. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
147 Chemical hazards • Appropriate methods should be used to predict the toxic footprint of any potential release of hazardous chemicals to the environment. The tolerability of the toxicological consequences should be assessed against international and national standards. • Chemical exposure standards are available from a variety of national and international organisations, e. g. relevant ISO standards. Fuel cycle facilities should be designed and operated in a manner that ensures that the risks of hazardous chemical exposure and contamination are controlled and protection optimised. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
148 Fire hazards Facilities use flammable, combustible, and explosive materials in their process operations, such as: − a tributylphosphate-kerosene mixture for solvent extraction, − bitumen or polymer for conditioning radioactive waste, − hydrogen in sintering furnaces, − chemical reagents for oxide reduction. The design of the facilities should minimise the inventories of combustible materials and ensure control of thermal processes to reduce the potential for fire and explosions. A significant source of hazard in higher radiation facilities is the radiolytic generation of molecular hydrogen. Extreme care must be taken to control accumulation of radiolytic hydrogen. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
149 Fire hazards • Fire can become the driving force for significant releases of radioactive and toxic material from the facilities as well as a threat to containment barriers. Extensive fire detection, suppression, and mitigation controls are necessary. • A fuel cycle facility safety assessment considers the radiological and other consequences from fires and explosion. Safety controls are instituted to protect the workers, the public and the environment. These safety controls are designed to provide the requisite protection during normal operations, anticipated operational occurrences, and credible accidents at a facility. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
150 Fire hazards • The hazards encountered in fuel cycle facilities vary considerably and a special fire hazard analysis should be carried out for each individual installation (can‘t be standardised as in NPPs). • With older installations, analyses of this kind allow safety authorities to determine what improvements are needed to meet current safety standards. In order to carry out these analyses, some member countries have developed special design codes and expert systems that they use in conjunction with existing technical rules and regulatory requirements or guides. Analysis of fire hazards involves a sequential review of the provisions made for preventing, detecting and fighting fires. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
151 Explosion hazards The main possible generators of explosions are: • the use of hydrogen in the sintering furnaces of fuel fabrication; • the explosive combustion of zirconium powder; • the decomposition of hydrazoic acid or ammonium nitrate; • the reaction of solvent or diluent with nitric acid in evaporators, etc; • the production of hydrogen by radiolysis; • the oxidation of U; • the use of reducing agents (hydrazine, etc. ); • the use of solvents and diluents; • the presence of formaldehyde in evaporators; • the presence of nitrites in resins and bitumen. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
152 Effluents Some fuel cycle facilities pose special safety hazards to the environment because they produce large quantities of effluents or highly hazardous effluents. • Effluents in the forms of liquid or gas must be treated in order to optimise the environmental and health impacts from release. • Filtration systems are used to prevent dispersion of aerosol substances within the plant or external release. • A liquid recovery system is used to recycle selected products (filtration, distillation, etc. ) and to minimise the generation of waste. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
153 Effluents • Effluent monitoring programme allows fuel cycle facilities to measure and monitor the concentrations of radioactive materials in airborne and liquid effluents and to establish that the protection of the public and the environment is optimised and in below regulatory limits. • Airborne effluents from routine and non-routine operations are usually continuously sampled. It is also usual to have an “environmental sampling programme” using a combination of fixed and ‘random” sampling locations around the facility to ensure that any releases from monitored or unmonitored sources are adequately accounted for and corrective action taken where necessary. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
154 Other safety issues There are several other important aspects of safety in nuclear fuel cycle facilities that have common grounds for all types of facilities, including nuclear power plants. Examples of these “other” safety aspects include: • Deterministic and/or probabilistic safety assessment (described in more detail in Volumes 6 and 7) • Siting considerations and Environmental Impact Assessment (described in more detail in Volume 9) • Maintenance (described in more detail in Volume 13) • Human performance (described in more detail in Volume 22) • Monitoring of the health of the workforce (including estimation of internal radiation doses) Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
155 IAEA FUEL CYCLE RELATED PROGRAMS Learning objectives After completing this chapter, the trainee will be able to: 1. Describe fuel cycle criticality safety aspects; 2. Appreciate IAEA nuclear fuel cycle safety standards; 3. Describe the Safety evaluation of fuel cycle facilities during operation (SEDO); 4. Describe the Fuel incident notification and analysis system (FINAS); 5. Describe the Integrated nuclear fuel cycle information systems (INFCIS); 6. Describe the fuel bank. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
156 Nuclear fuel cycle safety standards Uranium production IAEA together with the OECD Nuclear Energy Agency (NEA) collects and provides information on uranium resources, production and demand. Publication Uranium - Resources, Production and Demand (Red Book) covers the following topics: • Estimates of uranium resources in several categories of assurance based on existence and economic attractiveness; • Production capability; • Nuclear capacity; • Related reactor requirements. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
157 Nuclear fuel cycle safety standards Fabrication and in-reactor performance of nuclear fuel The IAEA programmes cover the following topics of expertise: • Production of nuclear-grade uranium; • Development, design and engineering; • Fabrication; • Behaviour, analysis and modelling; • Utilization and management; • MOX, alternative fuels, advanced fuel technologies and materials; • Economic and other aspects, e. g. environmental issues; • Quality assurance and control. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
158 Nuclear fuel cycle safety standards IAEA supports Member States through several projects: Management of spent nuclear fuel • Promoting technologies and strategies for spent fuel management; and • Providing technical guidance on good practices for long term storage of spent fuel. Advanced fuel cycles including recycling • Supporting emerging nuclear fuel cycle technologies for advanced and innovative reactors; and • Supporting development of proliferation resistant fuel cycles. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
Safety Evaluation of Fuel Cycle Facilities during Operation (SEDO) 159 The objectives of the SEDO are to: • Provide useful information on opportunities for improving operational safety; • Identify good practices; • Broaden the experience of facility staff through informal exchange of information; • Instruct the facility staff in the use of the SEDO methodology which could be used for conducting future self-assessments. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
Safety Evaluation of Fuel Cycle Facilities during Operation (SEDO) • SEDO is not a regulatory inspection which assesses fuel cycle safety against national regulatory requirements. It does not rank the operational safety performance of different facilities. • SEDO includes assessment of: conversion facilities, enrichment facilities, fuel fabrication facilities, spent fuel storage facilities and reprocessing and associated waste treatment facilities as well as fuel cycle R&D facilities. • SEDO is intended to be a peer review conducted by a team of international experts with experience in the operational and technical areas being evaluated. Advice on the safety performance of the facility are based on IAEA FCF Safety Standards. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 160
Safety Evaluation of Fuel Cycle Facilities during Operation (SEDO) Review areas of a full SEDO mission are: • Management, organization and administration • Training and qualification • Operation • Maintenance and periodic tests • Modifications • Other technical support and services • Criticality safety • Radiation protection • Waste management • Fire, chemical and industrial safety management • Emergency planning and preparedness • Effluent management and environmental protection Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 161
162 Fuel Incident Notification and Analysis System (FINAS) • Main objective: to provide timely feedback on safety related events to help to prevent the occurrence or recurrence of such incidents; • Web-based system for the exchange of lessons learned from operating experience gained in fuel cycle facilities; • Includes the collection, evaluation and dissemination of event reports, and the organization of meetings and workshops; • Based on the voluntary commitment of participating Member States; • Participating Member States designate a FINAS national coordinator. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
Integrated Nuclear Fuel Cycle Information System (INFCIS). The on-line information system includes: • Nuclear Fuel Cycle Information System (NFCIS) • Post Irradiation Examination Facilities Database (PIE) • World Distribution of Uranium Deposits Database (UDEPO) • Nuclear Fuel Cycle Simulation System (NFCSS) • World Thorium Deposits and Resources (Th. DEPO) • Minor Actinide Property Database (MADB) Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials 163
164 Fuel Bank • The IAEA Low Enrichment Uranium bank (owned and managed by the IAEA) will help to assure a supply of LEU for power generation. • In the case that a Member State´s LEU supply to a nuclear power plant is disrupted, and the supply cannot be restored by the commercial market, it may call upon the IAEA LEU bank to secure LEU supplies, without distorting the commercial market. • The LEU will be made available to an eligible IAEA Member State at the market prices prevailing at the time of supply. • The IAEA LEU bank will be sited in Kazakhstan. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
165 Fuel Bank Requirements for supply LEU from the bank will only be supplied to a Member State which fulfils the following eligibility criteria: • The Member State is experiencing a supply disruption of LEU to a nuclear power plant and is unable to secure it from the commercial market, or through State-to-State arrangements; • The IAEA has made a conclusion that there has been no diversion of declared nuclear material and no issues relating to safeguards; • The Member State has brought into force a comprehensive safeguards agreement requiring the application of IAEA safeguards to all its peaceful nuclear activities. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials
166 Fuel Bank Recipient State's obligations • The LEU from the IAEA LEU bank can only be used for fuel fabrication for the generation of energy at an NPP; • The LEU may not be used for any military purpose; • It shall not further enrich, reprocess, retransfer or re-export the LEU unless the IAEA agrees; • It shall apply the applicable IAEA safeguards, safety standards and physical protection measures to the LEU; • It shall take responsibility for all liability for any nuclear damage that may be caused by a nuclear incident associated with the LEU supplied under the Agreement; • It shall take responsibility for spent fuel and wastes. Basic Professional Training Course; Module XVII Fuel cycle, spent fuel management and transport of radioactive materials The views expressed in this document do not necessarily reflect the views of the European Commission.
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