- 1 Introduction
- 2 In the Beginning
- 3 Whiteshell Reactor #1 – A Manitoba Milestone
- 4 Experimental Loops
- 5 November 1, 1965 – Manitoba History
- 6 Nuclear Safety Research – Another Contribution to Nuclear Engineering
- 7 Nuclear Waste Disposal Research and the Underground Research Laboratory
- 8 Other Contributions to Nuclear Engineering
- 9 Important People for the Development of Whiteshell Laboratories
- 10 References
Atomic Energy of Canada Limited (AECL) established a role in Manitoba in 1963 when the Whiteshell Nuclear Research Establishment (WNRE; later renamed Whiteshell Laboratories) first took shape. WNRE was Canada’s second nuclear science research and development laboratory and the first facility of its kind in western Canada. Through its years of operation, the people of the WNRE made significant contributions to the science and engineering knowledge of Canada’s nuclear industry. This article highlights just a few of these many accomplishments.
Whiteshell Laboratories contribution to nuclear engineering progress in Canada has been impressive. Although this article does not include all of the contributions that have been made over the past 50 years, it does illustrate the depth and breathe of the engineering accomplishments that have made Manitoba an important part of the Canada’s nuclear engineering family. As the industry moves forward, these contributions are expected to continue, particularly in the areas of facility decommissioning and waste storage technologies.
In the Beginning
In the late 1950s AECL’s managers thought Chalk River Laboratories (CRL) was nearing the saturation point. A quick survey indicated that three provinces were lacking federal research facilities: Newfoundland, Alberta and Manitoba. Newfoundland, it was felt was not an option at the time, having joined Canada less than ten years previously in 1949. Alberta had no need of atomic energy, blessed as it was with abundant oil and gas. So it would be Manitoba.
AECL wanted this new research laboratory to develop the organically cooled reactor. A preliminary survey went forward under the supervision of Shawinigan Engineering. AECL president J.L. Gray journeyed to Manitoba to meet with premier Duff Roblin. In November, 1959, he reported progress to the board: a probable site near the Seven Sisters Falls on the Winnipeg River; and an opinion by the federal government's housing agency that a new town would be developed.
Negotiations with Manitoba were complicated. The new research centre would not be costless for the province. It would have to look after some of the infrastructure, such as roads and a bridge across the Winnipeg River, as well as housekeeping details. With help from the federal government, an agreement was approved by cabinet on July 21, 1960 and the Whiteshell Nuclear Research Establishment (WNRE) was born.
Final agreement was reached on joint facilities, between AECL and Manitoba, just as the company was finalizing their plans for demonstrating organic cooled reactors. It was called simply OTR for Organic Test Reactor. A design would be ready for the start of the construction season in April 1962.
Whiteshell Reactor #1 – A Manitoba Milestone
The Whiteshell Reactor #1 (WR-1), WNRE’s signature facility, was built starting in 1962. The 60 Megawatt reactor was designed and built by Canadian General Electric for $14.5 million in only three years. By June 1965, WR-1 was substantially complete.
WR-1 was built to test the concept of using an organic fluid to cool the reactor. The expected advantage was that they they can operate hotter and at lower pressures than water-cooled reactors. This was because the organic coolant had a lower vapour pressure than water. Higher operating temperatures increase the thermal efficiency of the attached turbine system (the amount of electricity produced divided by the amount of heat produced in the reactor core). Lower pressures reduce maintenance costs and pressure vessel design requirements. It allowed WR-1 designers to use thinner-walled pressure tubes, which reduced the number of neutrons absorbed in the tubes, giving the reactor a high neutron flux.
The reactor had vertical fuel channels. Neutrons were moderated by heavy water in a large calandria vessel surrounding the fuel channels. This calandria was a stainless steel tank approximately 5 m high and 2.75 m in diameter. Fifty-four aluminum tubes penetrate the calandria vessel. Pressure tubes, which contained the fuel and circulating organic coolant, were located inside these calandria tubes. The fuel was compacted and sintered uranium dioxide, slightly enriched to provide a useful neutron flux (2.4% U-235 in natural uranium, clad in zirconium-2.5% niobium alloy).
The vessel was divided into an upper and lower section. The upper section contained the fuel and, when the reactor was operating, the heavy water moderator. The lower section contained helium gas and collected the moderator spillage from the upper section. The reactor control system maintained the moderator level in the upper section by varying the differential helium pressure between the two reactor sections. When the reactor tripped, the helium gas pressure in the lower section was equalized with the upper section allowing the lower section to rapidly receive the moderator from the upper section and drain to the moderator dump tank.
The annuli between the fuel channels and the calandria tubes were purged with CO2 gas to insulate the hot fuel channels from the moderator. Sampling of the CO2 gas provided a means of detecting moderator or organic coolant leaks between the fuel channel and calandria tube.
The reactor was surrounded by heavy concrete shielding (> 2 m thick), which formed the reactor vault walls. Heavy concrete (density of 3,500 kg/m3) was also used in the vicinity of the upper and lower access rooms and the shutdown shields. Stepped pipe chases through the concrete provided access for heavy water and helium lines and for the reactor vault exhaust duct. There were also three penetrations for the ion chambers. The inner surfaces of the concrete walls were cooled by embedded cooling coils.
The top deck plates provide an operational shield between the upper access space and the reactor hall. The deck plates also supported the fuel transfer flask and provided the necessary radiation shielding during fuelling operations. It consisted of two rotating plates and an outer stationary ring. The plates were comprised of cast steel (0.45 m thick) topped by wood fibre hardboard (Masonite; 9 cm thick) and a steel cover plate (0.5 cm thick). The inner (small) rotating plate was supported by the large rotating plate on large ball bearings. The large plate was similarly supported on the stationary outer ring, which, in turn, was supported by the shielding walls of the upper access space. The rotating plates were driven by pinion gears located on the stationary ring and on the large rotating plate, which meshed with gears located at the outer periphery of the large and small rotating plates, respectively. The small rotating plate had two holes for fuelling operations and periscope viewing in the upper access space.
The Primary Heat Transport System (PHT) was designed to remove the heat produced in the reactor core. The system was divided into three circuits. The removed heat was dissipated to the Winnipeg River through three conventional tube-and-shell heat exchangers. River water was used as the secondary coolant. The PHT system had three similar circuits to achieve flexibility for experimental research.
To the outside world the most noticeable feature of the WR-1 reactor was the ventilation stack. The stack was known as the "stank" - a combination emergency coolant tank and ventilation stack.
The WR-1 reactor was housed in a building that had 7 floors, 5 of which were below grade. The building was divided into two areas: the lower 4 levels (with restricted access) contained shutdown reactor components, while the upper 3 floors provided office space or laboratory space for experimental programs. The reliability of the WR-1 safety systems was achieved by means of instrument triplication, parameter duplication and frequent testing. Each trip parameter was monitored by three independent sets of instrumentation. Used fuel, irradiated fuel channels and equipment could be safely transferred from the reactor to water-filled storage facilities. After the fuel or equipment had been cleaned or decayed sufficiently, it would be transferred to long-term site storage.
A unique feature of WR-1 was its four experimental loops and one out-of-reactor hydraulic test loop. Each in-reactor loop consisted of a fuelled test section in a reactor lattice position and piping equipment and instrumentation in an adjacent loop room to maintain required operating conditions of flow, pressure and temperature in the test section. A fuel position was converted to a loop by disconnecting the inlet and outlet feeders from the PHT and connecting the feeders to the loop inlet and outlet piping. The out-of-reactor hydraulic test facility was capable of handling full-sized fuel channels and fuel assemblies. The loop consisted of a circulation pump, a pressurizing pump, three test sections, three electric heaters, a make-up tank/degassifier, a condenser circuit, a purification circuit, a loop cooler, piping and instrumentation.
November 1, 1965 – Manitoba History
WR-1 commissioning proceeded smoothly. As might be expected with a new reactor design, some modifications were required in the pre-critical phase to make the various systems function as intended. For example, to prevent gas-locking of the moderator pumps during a moderator dump, it was necessary to extend the discharge pipe across the dump tank and install baffles to keep the entrained helium from the pump inlet. It was also necessary to re-route the ion-chamber cables to eliminate false signals due to "cross-talk" from the crane control circuitry.
WR-1 went critical on November 1, 1965. The start-up was smooth and uneventful; the low-power commissioning continued throughout the month, with WR-1 operating almost continuously at 0.01% of full power. This allowed for measurements of nuclear coefficients, neutron fluxes, coolant dosimetry, and regulating-system performance. The main problems encountered during high-power commissioning of WR-1 were difficulties in obtaining satisfactory response from the automatic temperature control system, and in obtaining satisfactory performance from the thermal power control system. The use of standard- modules in a relatively complex control circuit proved to be unsatisfactory, and many were replaced by higher quality versions.
WR-l operators concluded that the commissioning and operating of an organic cooled reactor was more trouble-free and straightforward than that of a pressurized-water-cooled system. They attributed this to two characteristics of the organic system: high coolant-outlet temperature was attainable with relatively low operating pressure in the primary system, and the radiation fields near the primary piping, feeders and headers were very low, minimizing the problems of access to these areas for normal operation.
With WR-1 operating well, there was considerable incentive to make improvements that would demonstrate the feasibility of a CANDU-based Organically Cooled Reactor (OCR), provide better conditions for experimenters and lower the fueling costs. Towards this goal a number of changes were made in 1975, including:
- the stainless-steel pressure tubes (relatively high absorbers of neutrons) were all replaced by zirconium-alloy tubes,
- a third organic primary coolant circuit was added to service seventeen of the remaining sites, which were then commissioned as fuel sites.
WR-1 operated with uranium dioxide driver fuel from start-up in 1965 until 1973, when uranium dioxide was gradually phased out in favour of uranium carbide fuel. The overall average string burn-up for all the uranium-dioxide fuel irradiated in WR-1 (about 1,100 bundles) was 128 MWh/kgU, which satisfied the original target of 120 MWh/kgU.
Experimental irradiations of uranium-carbide fuel started in 1966, and in 1973 the irradiation of uranium carbide as the reactor-driver fuel, in quantities sufficient to obtain statistical information, commenced. Conversion of WR-1 to uranium-carbide driver fuel was completed by the end of 1977.
The performance of the uranium-carbide fuel was excellent both from the viewpoint of burn-up achieved and failure rate. Average burn-up of the first 125 bundles retired was 253 MWh/ kgU, (original target burn-up was 240 MWh/ kgU). Not only was the failure frequency low, but the consequences of failure were not serious. It was found that while activity releases were high enough to detect failure, they remained low for a long enough period (six weeks or more) to continue operation of the failed fuel until the next scheduled shutdown before removing it. One identified problem was hydrogen migration to the bundle end plates that caused their embrittlement, and six bundles had to be retired due to end-plate breakage during shuffling operations. The problem was solved by using larger hydrogen-sink volumes in the end-plate region.
WR-1 was a most useful research facility, testing experimental fuels, reactor materials, and other coolants for 20 years. The reactor was a busy place, usually working around the clock. It had an availability of 85% over its lifetime, which was exceptionally high for a research reactor.
WR1 was shut down for the last time on May 17, 1985, its place in history secured as the world's only operating heavy water-moderated reactor cooled by an organic fluid.
Nuclear Safety Research – Another Contribution to Nuclear Engineering
The safety of nuclear reactors has been a matter of major interest from the beginning of Canada's nuclear program. CANDU power plant design and the related safety philosophy have evolved in parallel and continue to evolve today.
In the early days (the late fifties and early sixties) of the nuclear power program, safety analysis dealt mainly with the consequences of the failure of individual critical process systems and/or pieces of equipment, as identified during the design process. However, as the program developed and the plants became larger, more attention was paid to dual failures, situations that could occur much less frequently but could have much greater consequences.
Reactor safety R&D programs all over the world expanded greatly in the seventies, and concentrated more and more on investigating the consequences of low-frequency, worst-case accident scenarios. This thrust for more detailed safety analysis tended to be self-ratcheting; i.e., the more detailed the model, the more questions that are raised, requiring more R&D to answer them.
From the outset, the objectives of the CANDU reactor safety R&D program was to develop a thorough understanding of the phenomena that might occur during reactor accidents, and to develop and verify the mathematical simulations used in the plant safety analyses. Whiteshell staff focused on:
Reactor Corrosion and Activity Transport
Perhaps the most important concerns related to safety in any reactor are the fraction of the fission products that will escape from the fuel during an accident, and where they will go. Thus, from the beginning, an important part of the safety R&D program has been to investigate how fission products escape from the uranium oxide fuel matrix and how they migrate throughout the plant.
A program was started at Whiteshell in the mid-1970s to study iodine and cesium compounds and their stability. The kinetics of the relevant thermodynamic relationships were derived and incorporated into a computer code. The subsequent analysis showed that under accident conditions iodine and cesium combine to form cesium iodide and cesium hydroxide. Since these salts dissolve in water, and water-cooled reactors (such as AECL’s CANDU) had plenty of water present during an accident, only a small fraction of these fission products should be released to the atmosphere. The results of this work explained the fission-product behaviour observed during the Three Mile Island accident. It is now believed that even in a worst-case accident the majority of the fission products that would escape from the fuel would be dissolved in water and that any release to the environment would be extremely small. In addition, all CANDU plants have well-engineered, highly reliable containment systems.
In the early years of nuclear power development, little was known about the behaviour of power reactor fuel of the CANDU design at any level of irradiation. At that time, the research programs were directed at obtaining a basic understanding of fuel behaviour under normal operating conditions and establishing limits for safe operation of the fuel. Again, the question was asked how fuels would react in accident conditions. In series of tests at Whiteshell in the 1970s, CANDU fuel bundles were heated in vacuum to up to 1600oC to study their mechanical behaviour at elevated temperatures. In these tests the bundles slumped into contact with the pressure tube and bundle elements slumped into contact with their neighbours. However, there were no sheath failures and the bundle end-plate’s end-cap welds remained intact.
One of the most important safety-related issues was the rate of coolant loss from the Primary Heat Transfer (PHT) system in the event of a major pipe break. A unique feature of the CANDU design was the flow-distribution headers. Flow stratification in the headers could affect the distribution to the feeders during emergency coolant injection. To investigate this effect, a full-scale header facility representative of the reactor in Pickering, Ontario, was constructed at Whiteshell in 1974. In this facility, which had 110 feeder pipes, the header temperature could be brought to the desired value using electrical heaters, and water injection was initiated using a quick-acting valve. During coolant injection the liquid level and pressure were measured at several positions along the header, and feeder flow rates were monitored, for a range of conditions.
To test the model used in the overall plant simulation, a series of experiments were performed at Whiteshell in the early 1970s. These tests explored:
- steady-state operating parameters with both single-phase (water) and two--phase (steam plus water) flow conditions at the primary side inlet;
- boiler water level;
- secondary-side steam/water ratio;
- recirculation flow rate, in response to changes in the primary and secondary circuit conditions; and
- the characteristics of the flow oscillations within individual boiler tubes operating with two-phase inlet conditions.
All of these separate-effects tests greatly improved understanding of what might occur during an accident. However, to check the complete analysis code, integrated tests were required. The first such test was performed in the RD-4 facility at Whiteshell, built by 1974. This was a small-scale recirculating water loop containing two pumps, two tubular heated sections (to simulate fuel channels), and two heat exchangers.
The RD-4 facility was succeeded in 1977 by the much larger RD-12 facility, which had heated sections four metres long with electrically heated elements to simulate the fuel. This facility also had recirculating U-tube type boilers, with an interacting secondary circuit and a much wider range of operation than RD-4.
A pressurized cold-water injection system was also provided, to supply water simultaneously to the four headers when the loop pressure fell below a preset value, thereby simulating a loss of coolant accident (LOCA) with emergency coolant injection. The experimental program included tests with various pipe-break sizes, located at different points in the loop circuit, various cold-water injection pressures, and several modes of boiler cooling. In all of these experiments detailed measurements were made to determine coolant flow rate, pressure, temperature and density distribution throughout the loop, as well as fuel-element surface temperatures, and differential pressures across various loop components.
The RD-12 facility was followed in 1983 by an even larger facility, called RD-14, which was a model of a primary coolant loop with the various components arranged to reproduce the gravitational effects in a CANDU plant. It consisted of two full-scale, full-power (6 MW each) fuel channels, each containing a full-length 37-element, electrically-heated bundle to simulate the fuel, plus full-size feeders and two full-height steam generators, all arranged in the CANDU figure-eight configuration. The steam generators had full-size U-tubes, but the number of tubes was reduced in proportion to the number of heated channels, to give the correct heat transfer area per fuel channel. The loop was designed so that fluid mass-flow rate, transit times and pressure/enthalpy terms in the primary system of the loop were the same as those in a typical CANDU under both forced- and natural-circulation conditions.
The RD-14M Facility was established in 1984. The experiments performed there simulated the behaviour of the entire primary heat transport system of a reactor. RD-14M was the key facility needed in the safety analysis for the licensing of CANDU reactors. The facility was one of the largest of its kind in the world, with an overall height of about 34 m. It was used to investigate many LOCA related phenomena, such as:
- blowdown tests for many different pipe break sizes, including the critical break size, where the pressure difference across the reactor core becomes close to zero, resulting in a stagnant flow condition in the fuel channels for an extended period (tens of seconds); and
- two-phase thermo syphoning tests, to study the situation that would arise from a loss of the coolant pumps, or during small-break LOCAs, where the fuel channels are cooled mainly by natural circulation.
Many useful results were obtained from the tests done with the RD-14 and RD-14M facilities and these were used to refine and verify the computer codes. The results of these tests showed that for all LOCAs, of any break size, with or without power to the coolant pumps, fuel sheath temperature would not exceed about 600oC, hence fuel should not fail, as long as the emergency coolant system was available. RD-14M remains operational today and is supporting both operating plants and new CANDU designs.
Nuclear Waste Disposal Research and the Underground Research Laboratory
In the early 1970s nuclear power stations were operating well in Canada but the social climate had changed. There was significant public opposition to nuclear power and one of the major focuses of this opposition was nuclear wastes. Critics claimed that there was no safe way of disposing of the wastes from the used fuel; the early glass block experiments were deemed insufficient and irrelevant because there had been essentially no accompanying geologic work and there were now no plans to reprocess the used fuel. Accordingly, in 1974, AECL initiated a major program to prove that wastes from the used fuel could be managed safely. It would be necessary to shield the fuel for several hundred years, while the bulk of the gamma emitting radionuclides decayed, and to isolate the fuel for a much longer period, to ensure that long-lived radionuclides did not escape in significant concentrations to drinking water. The challenge was to prove that long-term isolation was possible. The program was centred at Whiteshell and was to occupy a major portion of the laboratory's efforts for the next 30 years.
It was decided that underground disposal was the best option for Canada; it was deemed the method whose safety could most readily be proven "beyond reasonable doubt."
One of AECL’s major achievements was the preparation and public defense of a ten-volume Environmental Impact Statement (EIS) for a conceptual deep geologic repository. Completion of this document on the characterization, construction and performance modelling of a conceptual repository in the granite rock of the Canadian Shield was largely based on work conducted at the Underground Research Laboratory (URL).
The URL was situated in a granite batholith towards the western edge of the Canadian Shield, about 50 km northeast of Whiteshell. AECL constructed the facility to provide a representative geological setting for conducting research activities in support of the Canadian nuclear fuel waste management program.
Excavation of the URL shaft to a depth of 255 m was in 1984. The main shaft was extended to a depth of 443 m in 1988, followed by the excavation of the 420 level and the ventilation shaft over the following three years.
The programs at the URL included experiments on solute transport, grouting, buffers, containers, shaft sealing and in situ stress studies. The URL completed its mission in 2010. The comprehensive and multidisciplinary URL research program contributed to defining a robust conceptual design for an underground repository. Results from research at the URL were used in the assessment of the feasibility and safety of deep geological disposal.
Other Contributions to Nuclear Engineering
Over the 50 years of operation, Whiteshell Laboratories has contributed to a number of other important nuclear engineering programs, including:
Containment Test Facility (CTF)
The CTF was used to determine the conditions under which hydrogen combustion could occur within containment. There are two main sources of hydrogen in a CANDU power plant: the exothermic reaction between the Zircaloy fuel sheathing and steam and radiolysis of the heavy-water coolant and moderator. The CTF had two large-scale vessels that simulated containment structures. These studies showed that it was extremely unlikely that conditions could arise where hydrogen combustion could threaten the integrity of a CANDU containment system, if proper precautions are taken.
Large-Scale Vented Combustion Test Facility (LSVCTF)
To complement the CTF, the LSVCTF was built at Whiteshell in the 1980s to study flame propagation and pressure development during vented combustion and to test Passive Autocatalytic Recombiners (PARs) to control hydrogen in nuclear reactors. The PARS are a commercial product for AECL with world-wide sales in excess of $100 Million. Both the CTF and the LSVCTF are also still operational on a commercial basis.
Storage Options for Used Fuel
Whiteshell Laboratories experimented with dry storage of used fuel in reinforced concrete canisters with walls thick enough to provide the required shielding. The attraction of the canisters was that they could be built as needed, avoiding the large up-front cost for new fuel storage bays at existing CANDU reactor sites. The objective of the program was not only to demonstrate that used fuel could be stored in this way, but to develop and verify prediction methods that would allow future designs to be optimized. Tests on the electrically heated canisters showed that both cylindrical and square designs were practical and conservative. These tests confirmed the viability of the concept and it was quickly adopted for the storage of excess WR-1 fuel. Optimized designs have since been used for the storage of Gentilly-l, Douglas Point and NPD fuel when those reactors were shut down.
SLOWPOKE Demostration Reactor (SDR)
The Safe Low-Power Critical Experiment was a low-energy, pool-type nuclear research reactor conceived in 1967 at Whiteshell. In 1970 a prototype unit was built at Chalk River Laboratories. The first commercial version of the SLOWPOKE reactor was started up in 1971 at AECL's Commercial Products Division in Ottawa; and in 1976 a commercial design, named SLOWPOKE-2, was installed at the University of Toronto. Between 1976 and 1984, seven SLOWPOKE-2 reactors with Highly Enriched Uranium (HEU) fuel were commissioned in six Canadian cities and in Kingston, Jamaica. In the early 1980s AECL also designed and built a scaled-up version called SLOWPOKE-3 for district heating at Whiteshell. The economics of a district-heating system based on SLOWPOKE-3 technology were initially competitive with conventional fossil fuels for use in remote communities; however the market interest in the SLOWPOKE heating system eventually dwindled due to the low price of natural gas. The SDR was shut down in 1990.
The Nuclear Battery program began in 1984 as a joint project with the Los Alamos National Laboratory (LANL) to develop a small nuclear power supply for unattended short-range radar stations in the new North Warning System (NWS). In hardware development programs at Whiteshell, a toluene organic Rankine cycle engine was commissioned and operated with a propane heat source. An experiment to study graphite oxidation during an air ingress accident was also performed.
Research into the life sciences focused on the biological effects of radiation, the behaviour of radionuclides in the environment, the use of radionuclides for biological research, and developing instruments for radiation protection. Work at Whiteshell has greatly improved our understanding of how radiation produces biological changes in living organisms and how best to protect nuclear workers and the general public.
AECL's environmental research program has always been concerned with two key issues; how radioactive materials move through the environment, and how radiation interacts with the environment. Whiteshell was a world leader in environmental protection, using the many unique facilities at the laboratory, such as the Field-Irradiator Gamma (FIG) area, the Zoological Environment Under Stress (ZEUS) area, and the Cesium Pond experiment to examine these fundenmaental questions.
Important People for the Development of Whiteshell Laboratories
James Lorne Gray was born in Brandon, Manitoba in 1913. After public school in Winnipeg, he graduated with a Masters in Mechanical Engineering from the University of Saskatchewan in 1938. He joined the Royal Canadian Air Force in 1939.
Mr. Gray’s scientific career began at the National Research Council in 1948. He was assigned to the “Chalk River” project. He advanced to become President of AECL in 1958. For the next 16 years he led the corporation through an impressive growth period that saw Canada become a leader in nuclear engineering and technologies.
Mr. Gray was appointed a Companion of the Order of Canada in 1969 and was awarded the Professional Engineers Gold Medal by the Association of Professional Engineers of Ontario in 1973.
Fred Gilbert was the forefather of AECL in Manitoba. He was the first manager at the WNRE. He was also tasked with creating the town site for the employees. Fred Gilbert was born in 1910 in Winnipeg. He obtained his B.Sc. in Engineering Physics from Queens University in Kingston in 1936. He was at the forefront of nuclear industry when he arrived at Chalk River, Ontario.
He oversaw the construction and criticality of NRX (1947) and NRU (1957) at the Chalk River site, as well as CIR in India in 1960. Gilbert was involved in every facet of WNRE and Pinawa, and was a key member of several of the town’s first organizations, including the Pinawa Club.
Born in 1922 in Hamilton, Ontario, Ara Mooradian gained his early training as an engineer and scientist at the University of Saskatchewan and the University of Missouri. His career began at the Consolidated Mining & Smelting Company before joining the staff at the Chalk River Nuclear Laboratories.
At Chalk River he was Head of the Development Engineering and Fuel Development Branches. In 1964 Ara became the Managing Director of the Whiteshell Nuclear Research Establishment and later the Vice-President.
Ara Mooradian was noted for his contributions to the development of low cost fuel for CANDU nuclear power generating stations. His honours included the Canada Medal, the W.B. Lewis Award and Fellowships of the Royal Society of Canada and the Chemical Institute of Canada.
Dr. Archie Aikin was born in Saskatoon and attended schools in Winnipeg and Montreal. He obtained a B.Sc. (honours chemistry) from McGill University in 1941, served in the Canadian Army, then returned to McGill to gain his PhD. in chemistry.
Dr. Aikin joined the staff at Chalk River Nuclear Laboratories in 1949 and served there in a series of positions that included work in chemical engineering, nuclear fuels and economic evaluations of nuclear power systems. In 1968 he was appointed to Head Office, Ottawa, to set up the nuclear power marketing section as general manager. He was appointed vice-president, Whiteshell nuclear Research Establishment, on January 1, 1971. He moved to be Vice-President, Commercial Products, for AECL in 1974.
Robert (Bob) Hart arrived to Chalk River Laboratories in 1948. After working on various projects including purification of heavy water in a reactor system, reprocessing of nuclear fuels and studying the physical properties of these fuels, he moved to Whiteshell Laboratories in 1965 as head of the Reactor Core Technology Branch.
He was appointed director of the Applied Science Division in 1969, managing director of the Whiteshell Site in 1973 and a vice president of AECL in 1974. In 1978 he became executive vice-president in charge of the Research Company, AECL. Bob was awarded the W.B. Lewis medal by the Canadian Nuclear Association in 1981, with the following citation: "For giving the Whiteshell Nuclear Research Establishment world recognition in such fields as organic heat transport technology, thermalhydraulic technology for nuclear safety technology analysis, radioactive waste management.
After graduating from the University of Birmingham in England and obtaining a M.Sc. in Chemical Engineering, he taught at the University of Toronto and obtained a PhD. He joined AECL's nuclear R&D program at Chalk River Nuclear Laboratories and worked on the chemistry of CANDU coolant and moderator systems.
In 1963 he headed to the Whiteshell Nuclear Research Establishment at Pinawa, Manitoba. In 1985, Dr. Hatcher became President of AECL Research, responsible for Canada's nuclear science and technology R&D programs. He became President and CEO of AECL in 1989 and for three years led the restructuring of the corporation towards its new emphasis on support of CANDU. Dr. Stanley Hatcher was also past President and Chief Executive Officer of AECL.
Ralph Green (B.Sc., M.Sc. (Dalhousie), Ph.D. (McGill)) joined AECL at the Chalk River Laboratories in 1956, working first in reactor physics at Canada's first research reactor ZEEP, then in accelerator physics, and later as head of the Reactor Control Branch.
In 1979 he transferred to AECL's head office in Ottawa as a senior advisor. In 1982 Ralph was appointed vice-president and general manager of the Whiteshell site. In 1986 he was appointed vice-president of Reactor Development, responsible for all reactor-related R&D in AECL.
After retiring in 1991, he contributed to the 1997 book on the history of AECL, “Canada Enters the Nuclear Age. Ralph was a charter member of the Canadian Nuclear Society, having joined in 1980. Since his retirement, Ralph has been active in the Ottawa Branch of CNS.
Terry Rummery’s academic achievements began with an Honours B.Sc. in Engineering Chemistry, Queen’s University, 1961; a Ph.D. in Physical Chemistry, Queen’s University, 1966; and a National Research Council Overseas Post-Doctoral Fellowship, University College, London, UK, 1966-67.
As President of Atomic Energy Canada (Research), Terry led the program to develop the safe disposal of used nuclear fuel. For his work he was awarded an Honorary Doctor of Science degree from Queen’s University in 1993. A year later he was the recipient of the W.B. Lewis Medal presented by the Canadian Nuclear Society for contributions to nuclear science and engineering. His other accomplishments include Chairman of the Board for the Chemical Institute of Canada, 1998, and Fellowships in the Canadian Academy of Engineering.
Original article by Chris Saunders, P. Eng. and Ray Sochaski, P.Eng: http://heritage.apegm.mb.ca/index.php/File:Whiteshell_-_Manitoba%27s_Contribution_to_Nuclear_Energy_-_March_18,_2015.pdf