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The Whiteshell Reactor No. 1, or WR-1, was a Canadian research reactor located at AECL's Whiteshell Laboratories (WNRL) in Manitoba. It was built to test the concept of a CANDU-type reactor that replaced the heavy water coolant with an oil substance. This had a number of potential advantages in terms of cost and efficiency.

Research reactor nuclear reactors that serve primarily as a neutron source

Research reactors are nuclear reactors that serve primarily as a neutron source. They are also called non-power reactors, in contrast to power reactors that are used for electricity production, heat generation, or maritime propulsion.

Whiteshell Laboratories

The Whiteshell Laboratories, originally known as the Whiteshell Nuclear Research Establishment, or WNRE for short, was an Atomic Energy of Canada (AECL) laboratory in Manitoba, northeast of Winnipeg. It was originally built as a home for the experimental WR-1 reactor, but over time came to host a variety of experimental systems, including a SLOWPOKE reactor and the Underground Research Laboratory to study nuclear waste disposal. Employment peaked in the early 1970s at about 1,300, but during the 1980s the experiments began to wind down, and in 2003 the decision was made to close the site. As of 2017 the site is undergoing decommissioning with a planned completion date in 2024.

Manitoba Province of Canada

Manitoba is a province at the longitudinal centre of Canada. It is often considered one of the three prairie provinces and is Canada's fifth-most populous province with its estimated 1.369 million people. Manitoba covers 649,950 square kilometres (250,900 sq mi) with a widely varied landscape, stretching from the northern oceanic coastline to the southern border with the United States. The province is bordered by the provinces of Ontario to the east and Saskatchewan to the west, the territories of Nunavut to the north, and Northwest Territories to the northwest, and the U.S. states of North Dakota and Minnesota to the south.


The 60 MWth reactor was designed and built by Canadian General Electric for a cost of $14.5 million CAD. [1] It achieved criticality on 1 November 1965 [1] and full power in December 1965. An effort to commercialized the design began in 1971 but ended in 1973 when the heavy water cooled units became the standard. From then on WR-1 operated at reduced power limits for irradiation experiments and heating the WNRE site.

General Electric American industrial company

General Electric Company (GE) is an American multinational conglomerate incorporated in New York City and headquartered in Boston. As of 2018, the company operates through the following segments: aviation, healthcare, power, renewable energy, digital industry, additive manufacturing, venture capital and finance, lighting, and oil and gas. GE has a subsidiary in Bermuda.

Critical mass smallest amount of fissile material needed for a sustained nuclear chain reaction

A critical mass is the smallest amount of fissile material needed for a sustained nuclear chain reaction. The critical mass of a fissionable material depends upon its nuclear properties, its density, its shape, its enrichment, its purity, its temperature, and its surroundings. The concept is important in nuclear weapon design.

Irradiation is the process by which an object is exposed to radiation. The exposure can originate from various sources, including natural sources. Most frequently the term refers to ionizing radiation, and to a level of radiation that will serve a specific purpose, rather than radiation exposure to normal levels of background radiation. The term irradiation usually excludes the exposure to non-ionizing radiation, such as infrared, visible light, microwaves from cellular phones or electromagnetic waves emitted by radio and TV receivers and power supplies.

WR-1 was shut down for the last time in 1985, was defuelled, and as of 2013 is undergoing decommissioning scheduled to be completed in 2023.

Nuclear decommissioning process whereby a nuclear power plant site is dismantled

Nuclear decommissioning is the process whereby a nuclear facility is dismantled to the point that it no longer requires measures for radiation protection. The presence of radioactive material necessitates processes that are potentially occupationally hazardous, expensive, time-intensive, and present environmental risks that must be addressed to ensure radioactive materials are either transported elsewhere for storage or stored on-site in a safe manner. The challenge in nuclear decommissioning is not just technical, but also economical and social.


Basic fission

Natural uranium consists of a mix of isotopes, mostly 238U and a much smaller amount of 235U. Both of these isotopes can undergo fission when struck by a neutron of sufficient energy, and as part of this process, they will give off more neutrons. However, only 235U can undergo fission when struck by neutrons from other uranium atoms, allowing it to maintain a chain reaction. 238U is insensitive to these lower-energy neutrons and it thus not fissile like 235U. 235U is even more sensitive to neutrons if they are slowed from their original relativistic speeds to much lower energies, the so-called thermal neutron velocities.

Natural uranium refers to uranium with the same isotopic ratio as found in nature. It contains 0.711% uranium-235, 99.284% uranium-238, and a trace of uranium-234 by weight (0.0055%). Approximately 2.2% of its radioactivity comes from uranium-235, 48.6% from uranium-238, and 49.2% from uranium-234.

Isotope nuclides having the same atomic number but different mass numbers

Isotopes are variants of a particular chemical element which differ in neutron number, and consequently in nucleon number. All isotopes of a given element have the same number of protons but different numbers of neutrons in each atom.

Nuclear fission nuclear reaction or a radioactive decay process is also known as nuclear fission.

In nuclear physics and nuclear chemistry, nuclear fission is a nuclear reaction or a radioactive decay process in which the nucleus of an atom splits into 2 smaller, lighter nuclei. The fission process often produces gamma photons, and releases a very large amount of energy even by the energetic standards of radioactive decay.

In a mass of pure natural uranium, the number and energy of the neutrons being released through natural decay are too low to cause appreciable fission events in the few 235U atoms present. In order to increase the rate of neutron capture to the point where a chain reaction can occur, known as criticality, the system has to be modified. In most cases, the fuel mass is separated into a large number of smaller fuel pellets and then surrounded by some form of neutron moderator that will slow the neutrons, thereby increasing the chance that the neutrons will cause fission in 235U in other pellets. Often the simplest moderator to use is normal water; when a neutron collides with a water molecule it transfers some of its energy to it, increasing the temperature of the water and slowing the neutron.

Neutron moderator medium that reduces the speed of fast neutrons, turning them into thermal neutrons that can sustain nuclear chain reactions; e.g. water, graphite, heavy water, beryllium

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

The main problem with using normal water as a moderator is that it also absorbs some of the neutrons. The neutron balance in the natural isotopic mix is so close that even a small number being absorbed in this fashion means there are too few to maintain criticality. In most reactor designs this is addressed by slightly increasing the amount of 235U relative to 238U, a process known as enrichment. The resulting fuel typically contains between 3 and 5% 235U, up from the natural value of just under 1%. The leftover material, now containing almost no 235U and consisting of almost pure 238U, is known as depleted uranium.

Enriched uranium is a type of uranium in which the percent composition of uranium-235 has been increased through the process of isotope separation. Natural uranium is 99.284% 238U isotope, with 235U only constituting about 0.711% of its mass. 235U is the only nuclide existing in nature that is fissile with thermal neutrons.

Depleted uranium uranium with a lower content of the fissile isotope U-235 than natural uranium

Depleted uranium is uranium with a lower content of the fissile isotope U-235 than natural uranium. Natural uranium contains about 0.72% U-235, while the DU used by the U.S. Department of Defense contains 0.3% U-235 or less. Uses of DU take advantage of its very high density of 19.1 g/cm3. The less radioactive and non-fissile uranium-238 constitutes the main component of depleted uranium.

Conventional CANDU

The CANDU design addresses the moderation problem by replacing the normal water with heavy water. Heavy water has an extra neutron, so the chance that the original fission neutron will be absorbed during moderation is greatly reduced. Additionally, it is subject to other reactions that further increase the number of neutrons released during operation. The neutron economy is improved to the point where even unenriched natural uranium will maintain criticality, which greatly reduces the complexity and cost of fueling the reactor, and also allows it to use a number of alternative fuel cycles that mix in even less reactive elements. The downside to this approach is that the 235U is spread out through a larger fuel mass, which makes the reactor core larger for any given power level. This can lead to higher capital costs for building the reactor core.

To address the cost issue, CANDU uses a unique reactor core layout. Conventional reactor designs consist of a large metal cylinder containing the fuel and moderating water, which is run under high pressure in order to increase the boiling point of the water so that it removes heat more efficiently. At the time CANDU was being designed, Canada lacked the facilities to make such large pressure vessels, especially ones large enough to run on natural uranium. The solution was to enclose the pressurized heavy water within smaller tubes and then insert these into a much larger low-pressure vessel known as the calandria. One major advantage of this layout is that the fuel can be removed from the individual tubes which allow the design to be refuelled while operating, while conventional designs require the entire reactor core to be shut down. A small disadvantage is that tubes absorb some neutrons as well, but not nearly enough to offset the improved neutron economy of the heavy water design.

Organic coolant

A significant problem with using any sort of water as a coolant is that the water tends to dissolve the fuel and other components and ends up becoming highly radioactive. This is mitigated by using particular alloys for the tubes and processing the fuel into a ceramic form. While effective at reducing the rate of dissolution, this adds to the cost of processing the fuel while also requiring materials that are susceptible to neutron embrittlement. More of an issue is the fact that water has a low boiling point, limiting the operating temperatures. A material with a higher boiling point can be run at higher temperatures, increasing the efficiency of the power extraction and allowing the core to be smaller.

This was the basic premise of the Organic Cooled Reactor (OCR) design. In the CANDU layout, the moderator and coolant both used heavy water, but there was no reason for this other than expediency. Since the bulk of the moderation occurred in the calandria mass, replacing the small amount in the fuel tubes with some other coolant was straightforward, unlike conventional light water designs where some other moderator would have to be added. [lower-alpha 1] Using oil meant the issues with corrosion were greatly reduced, allowing more conventional metals to be used while also reducing the amount of dissolved fuel, and in turn, radiation in the cooling system. The organic liquid that was selected, OS-84, is a mixture of terphenyls treated catalytically with hydrogen to produce 40 percent saturated hydrocarbons. The terphenyls are petrochemical derivatives that were readily available and were already in use as heat transfer media.

Additionally, by using a material with a higher boiling point, the reactor could be operated at higher temperatures. This not only reduced the amount of coolant needed to remove a given amount of energy, and thereby reduced the physical size of the core, but also increases the efficiency of the turbines used to extract this energy for electrical generation. WR-1 ran with outlet temperatures up to 425°C, [1] compared to about 310°C in the conventional CANDU. This also meant that there is no need to pressurize the cooling fluid beyond what is needed to force it through the cooling tubes at the required rate. This allowed the fuel tubes to be made thinner, reducing the number of neutrons lost in interactions with the tubing, and further increasing the neutron economy.

The reactor had vertical fuel channels, in contrast with the normal CANDU arrangement where the tubes are horizontal. The reactor did not use conventional control rods, but relied on control of the level of the heavy water moderator to adjust the power output. The reactor could be shut down quickly (SCRAMed) by the rapid dumping of the moderator.

In 1971 AECL initiated design engineering of a 500 MWe CANDU-OCR, based on uranium carbide fuel. Carbide fuels would corrode in water, but in the oil-cooled version, this was not an issue. Carbide fuels were much easier to produce than the more complex ceramics being used in most reactor designs. This design effort was shut down in 1973, but WR-1 tested the concept anyway. Another possibility was to use metallic fuel, which would increase the density of the fuel and offer higher burnup. The metallic fuel conducts heat better so that a higher power core could be used in the same space.


In November 1978 there was a major loss-of-coolant accident. 2,739 liters of coolant oil leaked, most of it into the Winnipeg River. The repair took several weeks for workers to complete. There was another leak in 1980 of 680 liters. [2] [3]


WR1 was shut down for the last time for economic reasons, on May 17, 1985 although it was the youngest of AECL's large research reactors. The reactor is in an interim decommissioning stage, defuelled and largely disassembled. The site will be returned to greenfield status at the end of decommissioning.

See also


  1. As is the case in the UK's Magnox designs, which used graphite as the moderator and carbon dioxide gas as the coolant.

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CANDU reactor Canadian PHWR nuclear reactor design

The CANDU is a Canadian pressurized heavy-water reactor design used to generate electric power. The acronym refers to its deuterium oxide moderator and its use of uranium fuel. CANDU reactors were first developed in the late 1950s and 1960s by a partnership between Atomic Energy of Canada Limited (AECL), the Hydro-Electric Power Commission of Ontario, Canadian General Electric, and other companies.

Nuclear reactor device to initiate and control a sustained nuclear chain reaction

A nuclear reactor, formerly known as an atomic pile, is a device used to initiate and control a self-sustained nuclear chain reaction. Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nuclear fission is passed to a working fluid, which in turn runs through steam turbines. These either drive a ship's propellers or turn electrical generators' shafts. Nuclear generated steam in principle can be used for industrial process heat or for district heating. Some reactors are used to produce isotopes for medical and industrial use, or for production of weapons-grade plutonium. As of early 2019, the IAEA reports there are 454 nuclear power reactors and 226 nuclear research reactors in operation around the world.

Pressurized water reactor nuclear power plant with a cooling system that operates under high pressure

Pressurized water reactors (PWRs) constitute the large majority of the world's nuclear power plants and are one of three types of light-water reactor (LWR), the other types being boiling water reactors (BWRs) and supercritical water reactors (SCWRs). In a PWR, the primary coolant (water) is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms. The heated water then flows to a steam generator where it transfers its thermal energy to a secondary system where steam is generated and flows to turbines which, in turn, spin an electric generator. In contrast to a boiling water reactor, pressure in the primary coolant loop prevents the water from boiling within the reactor. All LWRs use ordinary water as both coolant and neutron moderator.

Nuclear fuel cycle Process of manufacturing and consuming nuclear fuel

The nuclear fuel cycle, also called nuclear fuel chain, is the progression of nuclear fuel through a series of differing stages. It consists of steps in the front end, which are the preparation of the fuel, steps in the service period in which the fuel is used during reactor operation, and steps in the back end, which are necessary to safely manage, contain, and either reprocess or dispose of spent nuclear fuel. If spent fuel is not reprocessed, the fuel cycle is referred to as an open fuel cycle ; if the spent fuel is reprocessed, it is referred to as a closed fuel cycle.

NRX was a heavy-water-moderated, light-water-cooled, nuclear research reactor at the Canadian Chalk River Laboratories, which came into operation in 1947 at a design power rating of 10 MW (thermal), increasing to 42 MW by 1954. At the time of its construction, it was Canada's most expensive science facility and the world's most powerful nuclear research reactor. NRX was remarkable both in terms of its heat output and the number of free neutrons it generated. When a nuclear reactor is operating its nuclear chain reaction generates many free neutrons, and in the late 1940s NRX was the most intense neutron source in the world.

A thermal-neutron reactor is a nuclear reactor that uses slow or thermal neutrons.

Fast-neutron reactor nuclear reactor in which the fission chain reaction is sustained by fast neutrons

A fast-neutron reactor (FNR) or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons, as opposed to thermal neutrons used in thermal-neutron reactors. Such a reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor.

Loss-of-coolant accident mode of failure for a nuclear reactor

A loss-of-coolant accident (LOCA) is a mode of failure for a nuclear reactor; if not managed effectively, the results of a LOCA could result in reactor core damage. Each nuclear plant's emergency core cooling system (ECCS) exists specifically to deal with a LOCA.

In nuclear engineering, the void coefficient is a number that can be used to estimate how much the reactivity of a nuclear reactor changes as voids form in the reactor moderator or coolant. Reactivity, in the nuclear engineering sense, measures the degree of change in neutron multiplication in a reactor core. Reactivity is directly related to the tendency of the reactor core to change power level: if reactivity is positive, the core power tends to increase; if it is negative, the core power tends to decrease; if it is zero, the core power tends to remain stable. The reactivity of the core may be adjusted by the reactor control system in order to obtain a desired power level change. It can be compared to the reaction of an automobile as conditions around it change, and therefore the corresponding counter-measure that the driver applies to maintain road speed or execute a desired maneuver.

Light-water reactor type of nuclear reactor uses normal water

The light-water reactor (LWR) is a type of thermal-neutron reactor that uses normal water, as opposed to heavy water, as both its coolant and neutron moderator – furthermore a solid form of fissile elements is used as fuel. Thermal-neutron reactors are the most common type of nuclear reactor, and light-water reactors are the most common type of thermal-neutron reactor.

Nuclear fuel material that can be used in nuclear fission or fusion to derive nuclear energy

Nuclear fuel is material used in nuclear power stations to produce heat to power turbines. Heat is created when nuclear fuel undergoes nuclear fission.

The Advanced CANDU reactor (ACR), or ACR-1000, is a Generation III+ nuclear reactor designed by Atomic Energy of Canada Limited (AECL). It combines features of the existing CANDU pressurised heavy water reactors (PHWR) with features of light-water cooled pressurized water reactors (PWR). From CANDU, it takes the heavy water moderator, which gives the design an improved neutron economy that allows it to burn a variety of fuels. However, it replaces the heavy water cooling loop with one containing conventional light water, greatly reducing costs. The name refers to its design power in the 1,000 MWe class, with the baseline around 1,200 MWe.

Pool-type reactor

Pool-type reactors, also called swimming pool reactors, are a type of nuclear reactor that has a core immersed in an open pool of usually water. Some sodium-cooled reactors like the BN-600 have sodium pools instead. The rest of this article will assume that water is being used.

Supercritical water reactor

The supercritical water reactor (SCWR) is a concept Generation IV reactor, mostly designed as light water reactor (LWR) that operates at supercritical pressure. The term critical in this context refers to the critical point of water, and must not be confused with the concept of criticality of the nuclear reactor.

Nuclear reactor core portion of a nuclear reactor containing the nuclear fuel

A nuclear reactor core is the portion of a nuclear reactor containing the nuclear fuel components where the nuclear reactions take place and the heat is generated. Typically, the fuel will be low-enriched uranium contained in thousands of individual fuel pins. The core also contains structural components, the means to both moderate the neutrons and control the reaction, and the means to transfer the heat from the fuel to where it is required, outside the core.

The Clean and Environmentally Safe Advanced Reactor (CAESAR) is a nuclear reactor concept created by Claudio Filippone, the Director of the Center for Advanced Energy Concepts at the University of Maryland, College Park and head of the ongoing CAESAR Project. The concept's key element is the use of steam as a moderator, making it a type of reduced moderation water reactor. Because the density of steam may be controlled very precisely, Filippone claims it can be used to fine-tune neutron fluxes to ensure that neutrons are moving with an optimal energy profile to split 238
nuclei – in other words, cause fission.

Carolinas–Virginia Tube Reactor nuclear reactor

Carolinas–Virginia Tube Reactor (CVTR), also known as Parr Nuclear Station, was an experimental pressurized tube heavy water nuclear power reactor at Parr, South Carolina in Fairfield County. It was built and operated by the Carolinas Virginia Nuclear Power Associates. CVTR was a small test reactor, capable of generating 17 megawatts of electricity. It was officially commissioned in December 1963 and left service in January 1967.

A pressurized heavy-water reactor (PHWR) is a nuclear reactor, commonly using natural uranium as its fuel, that uses heavy water (deuterium oxide D2O) as its coolant and neutron moderator. The heavy water coolant is kept under pressure, allowing it to be heated to higher temperatures without boiling, much as in a pressurized water reactor. While heavy water is significantly more expensive than ordinary light water, it creates greatly enhanced neutron economy, allowing the reactor to operate without fuel-enrichment facilities (offsetting the additional expense of the heavy water) and enhancing the ability of the reactor to make use of alternate fuel cycles. At the beginning of 2001, 31 heavy water cooled and moderated nuclear power plants were in operation, having a total capacity of 16.5 GW(e), representing roughly 7.76% by number and 4.7% by generating capacity of all current operating reactors.

Organic nuclear reactor

An organic nuclear reactor, or organic cooled reactor (OCR), is a type of nuclear reactor that uses some form of organic fluid, typically a hydrocarbon substance like polychlorinated biphenyl (PCB), for cooling and sometimes as a neutron moderator as well.


  1. 1 2 3 "WR-1". Manitoba Branch of the Canadian Nuclear Society. 2005-03-18. Archived from the original on 2005-03-18. Retrieved 2016-11-07.
  2. Taylor, Dave (March 24, 2011). "Manitoba's forgotten nuclear accident".
  3. "Nuclear leak into river Negligible " Winnipeg Free Press. Ritchie Gage July 30, 1981