Engineering of systems using nuclear reactions or managing radioactive materials, including reactor energy conversion, materials performance and waste management.
Attenuation of an ionising radiation field by intervening material. A shield can reduce external exposure without preventing contamination from a released radioactive substance.
A factor accounting for scattered radiation that contributes to a specified response beyond the uncollided exponential attenuation. It depends on material, geometry, spectrum and the response being counted; is a stipulated simple model, not a universal law.
Prevention of unwanted dispersal of radioactive material by barriers, enclosures and controlled transfers. It differs from shielding, which attenuates a radiation field without necessarily retaining the radioactive source.
The coupled analysis of heat generation, heat transfer and coolant fluid mechanics in a nuclear reactor. A change in fuel geometry can affect neutron moderation, hydraulic resistance and thermal margins together, so holding one quantity fixed does not generally hold the others fixed.
An accident involving loss of reactor coolant through a breach of its pressure boundary. Depressurization, loss of inventory and changes in heat removal require a transient assessment; stopping the fission chain reaction does not remove decay heat.
A system intended to supply emergency cooling to limit fuel damage and maintain heat removal following a coolant-loss event. Its performance depends on the accident conditions and flow paths as well as nominal injection capability.
Thermal power generated per unit axial fuel-rod length, measured in . At fixed rod radius and material state, it controls the conductive fuel-to-surface temperature rise; coolant conditions determine the surface temperature from which that rise begins.
Thermal power per active-core volume under a stated volume convention. Coolant channels can be included in that core volume even when excluded from fuel volume, so core and fuel power densities need not change together.
Recording and reconciliation of nuclear-material inventories and transfers. This is a management concept, not a procedure for producing or separating fissile material.
Unwanted radioactive material on or in an object, substance or environment. It is distinct from activation of the object itself.
Administrative and technical actions bringing a permanently shut-down nuclear facility to a demonstrated end state permitting removal of some or all regulatory controls. Waste management and site-release evidence are integral to completion.
The specified condition at completion of nuclear decommissioning, supported by final characterisation and records. It can permit unrestricted reuse or retain defined restrictions.
Removal of materials or objects from further radiological regulatory control under specified criteria. Absence of an obvious radiation signal is not sufficient evidence of clearance.
Encasement of radioactive structures in a long-lived material. Historical classifications included it as a third strategy; IAEA GSR Part 6 excludes it for planned permanent shutdown, allowing consideration only in exceptional circumstances.
A strategy maintaining a facility safely for a period before final dismantling. radioactive decay can reduce later dose, while surveillance, deterioration and knowledge retention remain obligations.
A strategy beginning dismantling and decontamination soon after permanent shutdown, making early use of operating knowledge and enabling earlier site reuse.
Control of radioactive wastes from generation through treatment, conditioning, storage and disposal. Concentrating a radionuclide in a smaller stream changes its location, not its radioactive decay law.
A liquid or gaseous outflow containing radionuclides. Its measured inventory and treatment history are distinct from whether release is authorised under the applicable discharge controls.
Removal of contamination from a surface or material. The cleaning residues and wash solutions become part of the waste inventory.
Prevention of an unintended self-sustaining nuclear fission chain reaction in waste management. neutron absorbers, fissile-isotope dilution and their long-term geochemical retention are relevant; a cross-section comparison is not a criticality assessment.
Storage that allows short-lived radioactive constituents to radioactive decay before subsequent handling or discharge. A clearance or discharge decision still requires the applicable measurement and acceptance criteria.
Temporary retention of radioactive gas so that nuclear radioactive decay lowers the radioactive activity released. A guaranteed minimum residence time, an average residence time and a tank filling time are different design quantities.
Separation or transformation of contaminants in liquid waste. evaporation, ion exchange and chemical precipitation concentrate radionuclides into residues that need further management.
Transfer of dissolved contaminants to separable solids by precipitation, coprecipitation or surface uptake. The resulting sludge requires separation and conditioning; residual dissolved radioactive activity must be measured.
Removal of water vapour to concentrate mainly nonvolatile contaminants. Entrainment and volatile radionuclides can contaminate the condensate; a smaller concentrate volume does not guarantee a clean distillate.
Operations producing a stable waste form and suitable package, for example immobilisation in cement, glass or ceramic. Treatment residues and secondary wastes must be included in the waste inventory.
A waste form together with its container and any associated internal barriers prepared for handling, storage or disposal.
Immobilisation of radioactive waste in a glass matrix. glass composition, durability, secondary phases and the complete package affect performance.
Interaction of waste glass with water through ion exchange, network dissolution and alteration-product formation. Dissolved-element release need not equal the total extent of glass alteration.
Formation of new solids during glass alteration. Such phases can retain selected elements or consume dissolved network constituents, changing the apparent release and reaction rate.
A glass-durability test with a finite initially specified leachant inventory. Dissolved products, pH and alteration layers evolve, so the test does not maintain the same conditions as continuous fresh-water flow.
Element release departing from bulk glass proportions. Selective ion exchange, retention in an alteration layer, sorption and secondary precipitation can all produce noncoincident normalised solution releases.
A low continuing rate after alteration products or solution chemistry have slowed the initial reaction. It need not be zero or represent thermodynamic equilibrium of the original glass.
A hydrated, compositionally modified interfacial layer formed during glass corrosion. Its porosity, transport properties and continuing reactions can affect the residual alteration rate.
The dilute-solution dissolution rate at specified temperature, pH and composition, before dissolved reaction products suppress the chemical driving force. It is not a maximum across all temperatures and pH values.
A continuously refreshed leach test used to measure dissolution under controlled dilute conditions. At steady state the elemental normalised rate is . Varying helps identify solution-affinity effects.
Dissolution releasing glass constituents in their bulk proportions. Element-normalised releases then coincide if precipitation, sorption and analytical losses do not selectively remove released constituents.
The solution-released mass of an element divided by its initial glass mass fraction and the exposed glass area: . It has units of glass-equivalent mass per area; it does not alone establish congruent dissolution.
The physical and chemical form in which radioactive waste is immobilised for handling, storage or disposal. Its durability is distinct from that of its container and surrounding engineered barriers.
A waste form designed using durable mineral structures as hosts. A natural analogue motivates a design but does not substitute for tests of the actual radioactive composition.
A titanate host represented by . strontium can replace calcium; other ions require suitable charge balance.
A tunnel-structured titanate host, represented by , used to incorporate large cations such as caesium with appropriate charge compensation in a crystal.
A family of mineral-analogue ceramic waste forms. Different titanate phases accept different waste ions by solid solution; phase assemblage is tailored to the waste inventory.
A multiphase Synroc formulation with zirconolite, hollandite, perovskite and titanium-oxide phases. The mineral hosts distribute different radionuclides rather than requiring one lattice to accept the entire inventory.
Material containing radionuclides for which no further use is foreseen and which is subject to radiological control under the applicable framework.
The linked production, use and subsequent management of nuclear fuel. A closed cycle recovers some material for reuse; it still generates radioactive wastes.
The operating interval between specified refuelling events. It must be distinguished from the full material nuclear fuel cycle and from an individual fuel batch's multi-cycle residence time. Reactivity history, thermal limits and fuel management constrain the achievable interval.
An increase in the fraction of a specified uranium isotope. Fuel-cycle accounting tracks feed, product, enrichment tails and separative work; it does not specify an uranium enrichment plant or its operating conditions.
A technology exploiting mass-dependent redistribution of gas under rotation. It became the dominant commercial uranium-enrichment technology with much lower electricity consumption than gaseous diffusion. No operating conditions or equipment specifications are supplied.
A historical isotope-enrichment technology using the different transport of isotopic gas molecules through a barrier. It was displaced commercially by much less electricity-intensive centrifuge technology. This is a historical principle, not a plant design or operating procedure.
A unit of isotope-separation work used in fuel-cycle accounting. If stream masses are in kilograms and is isotope mass fraction, is in kg-SWU, with . This is a resource-accounting relation, not equipment or cascade design.
The depleted material stream in isotope uranium enrichment. enrichment tails retain uranium, so gross feed demand exceeds the enriched product mass.
The stated mass or atom fraction of a selected fuel isotope. An isotope balance requires all streams to use the same convention; a mass fraction is not interchangeable with an atom fraction without conversion.
A concentrated uranium-bearing product of ore processing. Resource accounts must distinguish mass of uranium from gross mass of its oxide concentrate.
A fuel cycle returning recovered constituents to fuel manufacture. Recovery losses, additional waste streams and recycled-fuel specifications affect the resource and cost balance.
A once-through fuel cycle in which spent nuclear fuel is managed without recovering its constituents for a further fuel irradiation.
Chemical treatment of spent nuclear fuel to partition reusable constituents from wastes. Recovery does not eliminate the need for waste conditioning and disposal. This article concerns concepts and material accounting, not operational fissile-separation procedures.
uranium recovered from spent nuclear fuel. Its isotopic composition differs from natural uranium; uranium-236 can impose an additional fuel-performance penalty.
Irradiated nuclear fuel removed from a reactor. Residual actinides coexist with fission products and activation products; radioactivity and decay heat continue after shutdown.
Historical background on the replacement of gaseous diffusion by gas centrifuges and diffusion-plant closures. The later webpage is not used to assign present-day plant counts to 2019.
npj Materials Degradation 4, article 41, published 11 December 2020. Solution and solid depth-profile measurements show that boron can be retained in alteration gels.
npj Materials Degradation 3, article 6, published 7 February 2019. Atomistic simulations examine restructuring and transport in hydrated gels; this is not a universal measured leach law.
Transactions of the Japan Institute of Metals 29, 970–976. The model and comparison with experimental specimens distinguish grain size, aspect ratio and dislocation density.
ORNL-1386, August 1952. Primary thermal-cycling investigation; the opening discussion describes dimensional instability, surface distortion and preferred orientation.
BNWL-SA-5333; CONF-750915-21. Primary comparison of alpha irradiation, self-damage and annealing; the alpha contribution is small relative to recoil damage but is not zero.
Journal of Nuclear Materials 306, 61–72. Simulation study using empirical potentials; a disordered cascade core and surrounding point defects are distinguished.
Chemical Geology 291, 55–68. Primary natural-sample study and stopping/damage simulations; Table 5 compares alpha and recoil displacement yields. Open institutional copy: apo.ansto.gov.au/server/api/core/bitstreams/12f621b9-8bb2-4e4c-84a0-8dc31a548213/content
Module HT-03 covers fluid mechanics, hydraulic losses and centrifugal pumps. Printed pp.51–53 explain the system characteristic curve and pump operating point. The date in the current hosting path does not change the handbook's original publication year.
An archival reproduction of the original Department of Energy handbook. Module NP-03, pp.24–27, explains moderation and temperature feedback. The repository is independently operated by Doxcelerate, rather than a current government web service.
Waste-form characterisation, process development and hot isostatic pressing; current status is not used as a 2021 historical claim.
GSR Part 6; paragraphs 1.9–1.10 distinguish planned strategies from exceptional entombment of a nuclear facility.
Waste-form development paper; section on long-term uranium-daughter criticality and neutron absorbers.
Primary conference record; lead author S. Muraoka, with H. Mitamura, S. Matsumoto, E. R. Vance and K. P. Hart.
MRS Proceedings 663, 259; abstract reports 2M, 4M and pyrochlore transitions.
Key Engineering Materials 48–50, 717–721; ANSTO primary research record.
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Nuclear engineering is a branch of engineering that deals with the processes, systems, and instruments used to harness and utilize nuclear energy and radiation. It encompasses a wide range of applications, including the design, construction, operation, and maintenance of nuclear power plants, the development of nuclear reactors, and the handling of radioactive materials. Key areas of focus in nuclear engineering include: 1. **Nuclear Reactor Design**: Engineering reactors to generate electricity efficiently and safely.