Nuclear criticality safety is a field of nuclear engineering dedicated to the prevention of nuclear and radiation accidents resulting from an inadvertent, self-sustaining nuclear chain reaction. Additionally, nuclear criticality safety is concerned with mitigating the consequences of a nuclear criticality accident. A nuclear criticality accident occurs from operations that involve fissile material and results in a tremendous and potentially lethal release of radiation. Nuclear criticality safety practitioners attempt to minimize the probability of a nuclear criticality accident by analyzing normal and abnormal fissile material operations and providing controls on the processing of fissile materials. A common practice is to apply a double contingency analysis to the operation in which two or more independent, concurrent and unlikely changes in process conditions must occur before a nuclear criticality accident can occur. For example, the first change in conditions may be complete or partial flooding and the second change a re-arrangement of the fissile material. Controls (requirements) on process parameters (e.g., fissile material mass, equipment) result from this analysis. These controls, either passive (physical), active (mechanical), or administrative (human), are implemented by inherently safe or fault-tolerant plant designs, or, if such designs are not practicable, by administrative controls such as operating procedures, job instructions and other means to minimize the potential for significant process changes that could lead to a nuclear criticality accident.

Seven factors influence a criticality system.

1. Geometry or shape of the fissile material: If neutrons escape (leak from) the fissile system they are not available to interact with the fissile material to cause a fission event. Therefore the shape of the fissile material affects the probability of occurrence of fission events. A large surface area such as a thin slab has lots of leakage and is safer than the same amount of fissile material in a small, compact shape such as a cube or a sphere.
2. Interaction of units: Neutrons leaking from one unit can enter another. Two units, which by themselves are sub-critical, could interact with each other to form a critical system. The distance separating the units and any material between them influences the effect.
3. Reflection: When neutrons collide with other atomic particles (primarily nuclei) and are not absorbed, they change direction. If the change in direction is large enough, the neutron may travel back into the system, increasing the likelihood of interaction (fission). This is called ‘reflection’. Good reflectors include hydrogen, beryllium, carbon, lead, uranium, water, polyethylene, concrete, Tungsten carbide and steel.
4. Moderation: Neutrons resulting from fission are typically fast (high energy). These fast neutrons do not cause fission as readily as slower (less energetic) ones. Neutrons are slowed down (moderated) by collision with atomic nuclei. The most effective moderating nuclei are hydrogen, deuterium, beryllium and carbon. Hence hydrogenous materials including oil, polyethylene, water, wood, paraffin, and the human body are good moderators. Note that moderation comes from collisions; therefore most moderators are also good reflectors.
5. Absorption: Absorption removes neutrons from the system. Large amounts of absorbers are used to control or reduce the probability of a criticality. Good absorbers are boron, cadmium, gadolinium, silver, and indium.
6. Enrichment: The probability of a neutron reacting with a fissile nucleus is influenced by the relative numbers of fissile and non-fissile nuclei in a system. The process of increasing the relative number of fissile nuclei in a system is called enrichment. Typically, low enrichment means less likelihood of a criticality and high enrichment means a greater likelihood.
7. Mass: The probability of fission increases as the total number of fissile nuclei increases. The relationship is not linear. There is a threshold below which criticality can not occur. This threshold is called the critical mass.

To determine whether a system containing fissile material is safe, calculations are performed using computer programmes. The analyst describes the geometry of the system and the materials, usually with conservative or pessimistic assumptions. The density and size of any neutron absorbers is minimised while the amount of fissile material is maximised. As some moderators are also absorbers, the analyst must be careful when modelling these to be pessimistic. Computer programmes allow analysts to describe a three dimensional system with boundary conditions. These boundary conditions can represent real boundaries such as concrete walls or the surface of a pond, or can be used to represent an artificial infinite system using a periodic boundary condition. These are useful when representing a large system consisting of many repeated units.

Computer codes used for criticality safety analyses include MONK(UK), KENO(USA), MCNP(USA) and CRISTAL(France).

Traditional criticality analyses assume that the fissile material is in its most reactive condition, which is usually at maximum enrichment, with no irradiation. For spent nuclear fuel storage and transport, burnup credit may be used to allow fuel to be more closely packed, reducing space and allowing more fuel to be handled safely. In order to implement burnup credit, fuel is modeled as irradiated using pessimistic conditions which produce an isotopic composition representative of all irradiated fuel. Fuel irradiation produces actinides consisting of both neutron absorbers and fissionable isotopes as well as fission products which absorb neutrons.

In fuel storage pools using burnup credit, separate regions are designed for storage of fresh and irradiating fuel. In order to store fuel in the irradiating fuel store it must satisfy a loading curve which is dependent on initial enrichment and irradiation.

More than a dozen advanced reactor designs are in various stages of development. Some are evolutionary from the PWR, BWR and PHWR designs above, some are more radical departures. The former include the Advanced Boiling Water Reactor (ABWR), two of which are now operating with others under construction, and the planned passively safe ESBWR and AP1000 units (see Nuclear Power 2010 Program).

• The Integral Fast Reactor (IFR) was built, tested and evaluated during the 1980s and then retired under the Clinton administration in the 1990s due to nuclear non-proliferation policies of the administration. Recycling spent fuel is the core of its design and it therefore produces only a fraction of the waste of current reactors.
• The Pebble Bed Reactor, a High Temperature Gas Cooled Reactor (HTGCR), is designed so high temperatures reduce power output by doppler broadening of the fuel's neutron cross-section. It uses ceramic fuels so its safe operating temperatures exceed the power-reduction temperature range. Most designs are cooled by inert helium. Helium is not subject to steam explosions, resists neutron absorption leading to radioactivity, and does not dissolve contaminants that can become radioactive. Typical designs have more layers (up to 7) of passive containment than light water reactors (usually 3). A unique feature that may aid safety is that the fuel-balls actually form the core's mechanism, and are replaced one-by-one as they age. The design of the fuel makes fuel reprocessing expensive.
• The Small Sealed Transportable Autonomous Reactor (SSTAR) is being primarily researched and developed in the US, intended as a fast breeder reactor that is passively safe and could be remotely shut down in case the suspicion arises that it is being tampered with.
• The Clean And Environmentally Safe Advanced Reactor (CAESAR) is a nuclear reactor concept that uses steam as a moderator — this design is still in development.
• The Hydrogen Moderated Self-regulating Nuclear Power Module (HPM) is a reactor design emanating from the Los Alamos National Laboratory that uses uranium hydride as fuel.
• Subcritical reactors are designed to be safer and more stable, but pose a number of engineering and economic difficulties. One example is the Energy amplifier.
• Thorium based reactors. It is possible to convert Thorium-232 into U-233 in reactors specially designed for the purpose. In this way, thorium, which is more plentiful than uranium, can be used to breed U-233 nuclear fuel. U-233 is also believed to have favourable nuclear properties as compared to traditionally used U-235, including better neutron economy and lower production of long lived transuranic waste.
  • Advanced Heavy Water Reactor (AHWR)— A proposed heavy water moderated nuclear power reactor that will be the next generation design of the PHWR type. Under development in the Bhabha Atomic Research Centre (BARC), India.
  • KAMINI — A unique reactor using Uranium-233 isotope for fuel. Built in India by BARC and Indira Gandhi Center for Atomic Research (IGCAR).
  • India is also planning to build fast breeder reactors using the thorium – Uranium-233 fuel cycle. The FBTR (Fast Breeder Test Reactor) in operation at Kalpakkam (India) uses Plutonium as a fuel and liquid sodium as a coolant.

A generation 2 nuclear reactor is a design classification for a nuclear reactor, and refers to the class of commercial reactors built up to the end of the 1990s. Prototypical generation II reactors include the PWR, CANDU, BWR, AGR, and VVER.

These are contrasted to generation I reactors, which refer to the early prototype and power reactors, such as Shippingport, Magnox, Fermi 1, and Dresden. The nomenclature for reactor designs, describing four 'generations', was proposed by the US Department of Energy when it introduced the concept of generation IV reactors.

The designation generation II+ reactor is sometimes used for modernised generation II designs built post-2000, such as the Chinese CPR-1000, in competition with more expensive generation III reactor designs. Typically the modernisation includes improved safety systems and a 60 year design life.

Generation II reactor designs generally had an original design life of 30 or 40 years. However many generation II reactor are being life-extended to 50 or 60 years, and a second life-extension to 80 years may also be economic in many cases.

Advantages and disadvantages of Generation 4 nuclear reactor:

Relative to current nuclear power plant technology, the claimed benefits for 4th generation reactors include:

• Nuclear waste that lasts a few centuries instead of millennia
• 100-300 times more energy yield from the same amount of nuclear fuel
• The ability to consume existing nuclear waste in the production of electricity
• Improved operating safety

One disadvantage of any new reactor technology is that safety risks may be greater initially as reactor operators have little experience with the new design. Nuclear engineer David Lochbaum has explained that almost all serious nuclear accidents have occurred with what was at the time the most recent technology. He argues that "the problem with new reactors and accidents is twofold: scenarios arise that are impossible to plan for in simulations; and humans make mistakes". As one director of a U.S. research laboratory put it, "fabrication, construction, operation, and maintenance of new reactors will face a steep learning curve: advanced technologies will have a heightened risk of accidents and mistakes. The technology may be proven, but people are not".

Generation 4 Nuclear Reactors (Gen IV) are a set of theoretical nuclear reactor designs currently being researched. Most of these designs are generally not expected to be available for commercial construction before 2030, with the exception of a version of the Very High Temperature Reactor (VHTR) called the Next Generation Nuclear Plant (NGNP). The NGNP is to be completed by 2021. Current reactors in operation around the world are generally considered second- or third-generation systems, with most of the first-generation systems having been retired some time ago. Research into these reactor types was officially started by the Generation IV International Forum (GIF) based on eight technology goals, including to improve nuclear safety, improve proliferation resistance, minimize waste and natural resource utilization, and decrease the cost to build and run such plants.

The reactors are intended for use in nuclear power plants to produce nuclear power from nuclear fuel.

Reactor types

Many reactor types were considered initially; however, the list was downsized to focus on the most promising technologies and those that could most likely meet the goals of the Gen IV initiative. Three systems are nominally thermal reactors and three are fast reactors. The VHTR is also being researched for potentially providing high quality process heat for hydrogen production. The fast reactors offer the possibility of burning actinides to further reduce waste and of being able to breed more fuel than they consume. These systems offer significant advances in sustainability, safety and reliability, economics, proliferation resistance and physical protection.

Thermal reactors

Very-high-temperature reactor (VHTR)

The very high temperature reactor concept uses a graphite-moderated core with a once-through uranium fuel cycle, using helium or molten salt as the coolant. This reactor design envisions an outlet temperature of 1,000 °C. The reactor core can be either a prismatic-block or a pebble bed reactor design. The high temperatures enable applications such as process heat or hydrogen production via the thermochemical iodine-sulfur process. It would also be passively safe.

The planned construction of the first VHTR, the South African PBMR (pebble bed modular reactor), lost government funding in February, 2010. A pronounced increase of costs and concerns about possible unexpected technical problems had discouraged potential investors and customers.

Supercritical-water-cooled reactor (SCWR)

The supercritical water reactor (SCWR) is a concept that uses supercritical water as the working fluid. SCWRs are basically light water reactors (LWR) operating at higher pressure and temperatures with a direct, once-through cycle. As most commonly envisioned, it would operate on a direct cycle, much like a Boiling Water Reactor (BWR), but since it uses supercritical water (not to be confused with critical mass) as the working fluid, would have only one phase present, like the Pressurized Water Reactor (PWR). It could operate at much higher temperatures than both current PWRs and BWRs.

Supercritical water-cooled reactors (SCWRs) are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current LWRs) and considerable plant simplification.

The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil fuel fired boilers, a large number of which are also in use around the world. The SCWR concept is being investigated by 32 organizations in 13 countries.

Molten-salt reactor (MSR)

A molten salt reactor is a type of nuclear reactor where the coolant is a molten salt. There have been many designs put forward for this type of reactor and a few prototypes built. The early concepts and many current ones rely on nuclear fuel dissolved in the molten fluoride salt as uranium tetrafluoride (UF₄) or thorium tetrafluoride (ThF₄), the fluid would reach criticality by flowing into a graphite core which would also serve as the moderator. Many current concepts rely on fuel that is dispersed in a graphite matrix with the molten salt providing low pressure, high temperature cooling.

Fast reactors

Gas-cooled fast reactor (GFR)

The gas-cooled fast reactor (GFR) system features a fast-neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The reactor is helium-cooled, with an outlet temperature of 850 °C and using a direct Brayton cycle gas turbine for high thermal efficiency. Several fuel forms are being considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations are being considered based on pin- or plate-based fuel assemblies or prismatic blocks.

Sodium-cooled fast reactor (SFR)

The SFR is a project that builds on two closely related existing projects, the liquid metal fast breeder reactor and the Integral Fast Reactor.

The goals are to increase the efficiency of uranium usage by breeding plutonium and eliminating the need for transuranic isotopes ever to leave the site. The reactor design uses an unmoderated core running on fast neutrons, designed to allow any transuranic isotope to be consumed (and in some cases used as fuel). In addition to the benefits of removing the long half-life transuranics from the waste cycle, the SFR fuel expands when the reactor overheats, and the chain reaction automatically slows down. In this manner, it is passively safe.

The Integral Fast Reactor or IFR is a design for a nuclear reactor with a specialized nuclear fuel cycle. A prototype of the reactor was built, but the project was cancelled before it could be copied elsewhere.

The SFR reactor concept is cooled by liquid sodium and fueled by a metallic alloy of uranium and plutonium. The fuel is contained in steel cladding with liquid sodium filling in the space between the clad elements which make up the fuel assembly. One of the design challenges of an SFR is the risks of handling sodium, which reacts explosively if it comes into contact with water. However, the use of liquid metal instead of water as coolant allows the system to work at atmospheric pressure, reducing the risk of leakage.

Lead-cooled fast reactor (LFR)

The lead-cooled fast reactor features a fast-neutron-spectrum lead or lead/bismuth eutectic (LBE) liquid-metal-cooled reactor with a closed fuel cycle. Options include a range of plant ratings, including a "battery" of 50 to 150 MW of electricity that features a very long refueling interval, a modular system rated at 300 to 400 MW, and a large monolithic plant option at 1,200 MW. (The term battery refers to the long-life, factory-fabricated core, not to any provision for electrochemical energy conversion.) The fuel is metal or nitride-based containing fertile uranium and transuranics. The LFR is cooled by natural convection with a reactor outlet coolant temperature of 550 °C, possibly ranging up to 800 °C with advanced materials. The higher temperature enables the production of hydrogen by thermochemical processes.

A generation III nuclear reactor is a development of any of the generation II nuclear reactor designs incorporating evolutionary improvements in design developed during the lifetime of the generation II reactor designs. These include improved fuel technology, superior thermal efficiency, passive safety systems and standardized design for reduced maintenance and capital costs.

Improvements in reactor technology result in a longer operational life (60 years of operation, extendable to 120+ years of operation prior to complete overhaul and reactor pressure vessel replacement) compared with currently used generation II reactors (designed for 40 years of operation, extendable to 80+ years of operation prior to complete overhaul and RPV replacement). Furthermore, core damage frequencies for these reactors are lower than for Generation II reactors — 60 core damage events per 1000 million reactor–year for the EPR; 3 core damage events per 1000 million reactor–year for the ESBWR significantly lower than the 10,000 core damage events per 1000 million reactor–year for BWR/4 generation II reactors.

The first generation III reactors were built in Japan, while several others have been approved for construction in Europe. A Westinghouse AP1000 reactor is scheduled to become operational in Sanmen, China in 2013.

Generation III reactors

• Advanced Boiling Water Reactor (ABWR) — A GE design that first went online in Japan in 1996.
• Advanced Pressurized Water Reactor (APWR) — developed by Mitsubishi Heavy Industries.
• Enhanced CANDU 6 (EC6) — developed by Atomic Energy of Canada Limited.
• VVER-1000/392 (PWR) — in various modifications into AES-91 and AES-92

Designs not adopted

• AP600 — A Westinghouse Electric Company design that received final design approval from the NRC in 1998; the EIA states that "Westinghouse has deemphasized the AP600 in favor of the larger, though potentially even less expensive (on a cost per kilowatt or capacity basis) AP1000 design."
• System 80+ — a Combustion Engineering (now incorporated into Westinghouse) design, which "provides a basis for the APR1400 (Generation III+) design that has been developed in Korea for future deployment and possible export."

Generation III+ reactors

Generation III+ designs offer significant improvements in safety and economics over Generation III advanced reactor designs certified by the NRC in the 1990s.

• Advanced CANDU Reactor (ACR-1000)
• AP1000 — based on the AP600 with increased power output
• European Pressurized Reactor (EPR) — an evolutionary descendant of the Framatome N4 and Siemens Power Generation Division KONVOI reactors.
• Economic Simplified Boiling Water Reactor (ESBWR) — based on the ABWR
• APR-1400 — an advanced PWR design evolved from the U.S. System 80+, which is the basis for the Korean Next Generation Reactor or KNGR
• VVER-1200/392M (PWR) — in design of AES-2006 with mainly passive safety features
• VVER-1200/491 (PWR) — in design of AES-2006 with mainly active safety features, international sold as MIR.1200
• EU-ABWR — based on the ABWR with increased powert output and compliance with EU safety standard.
• Advanced PWR (APWR) — 4th Generation of PWR from Mitsubishi Heavy Industries

Generation III++ reactors

• B&W mPower — an Advanced Light Water Reactor in development by Babcock and Wilcox and Bechtel