Rosen M.A.,University of Ontario Institute of Technology |
Naterer G.F.,University of Ontario Institute of Technology |
Chukwu C.C.,University of Ontario Institute of Technology |
Sadhankar R.,Atomic Energy of Canada Ltd. |
Suppiah S.,Atomic Energy of Canada Ltd.
International Journal of Energy Research | Year: 2012
Issues related to equipment scale-up and process simulation are described for a thermochemical cycle driven by nuclear heat from Canada's proposed Generation IV reactor (Super-Critical Water-Cooled Reactor; SCWR), which is a CANDU derivative using supercritical water cooling. The copper-chlorine (Cu-Cl) cycle has been identified by Atomic Energy of Canada Limited as the most promising cycle for thermochemical hydrogen production with SCWR. Water is decomposed into hydrogen and oxygen through intermediate Cu-Cl compounds. This article outlines the challenges and design issues of hydrogen production with a Cu-Cl cycle coupled to Canada's nuclear reactors. The processes are simulated using the Aspen Plus process simulation code, allowing the cycle efficiency and possible efficiency improvements to be examined. The results are useful to assist the development of a lab-scale cycle demonstration, which is currently being undertaken at the University of Ontario Institute of Technology in collaboration with numerous partners. © 2010 John Wiley & Sons, Ltd.
News Article | September 26, 2016
SNC-Lavalin has signed an agreement with two Chinese nuclear energy firms to develop, market and build an advanced CANDU type nuclear reactor The Montreal, Canada, based engineering and construction giant SNC-Lavalin, which five years ago, bought AECL’s reactor division from the government, has a new joint venture with China National Nuclear Corp. (CNNC) and Shanghai Electric Co. The immediate results of the agreement will be the creation of two nuclear reactor design centers, one in China and the other in Canada. The design centers will collaborate to complete the Advanced Fuel CANDU Reactor (AFCR). It is expected that the first two units will be then built in China and then the reactor will offered via export to global markets. “’The market potential for AFCR technology in China is considerable. Each AFCR can use recycled-fuel from four light-water reactors (LWRs) to generate six million megawatt-hours (MWh) of additional carbon-free electricity without needing any new natural uranium fuel. This would be enough new electricity to power four million Chinese homes, and also displace six million tonnes of carbon emissions per year vs. coal, the equivalent of removing one million cars from the road. China has more than 33 LWR nuclear power reactors in operation and another 23 LWRs under construction.” The agreement occurred during an official four-day visit to Canada by Chinese Premier Li Keqiang. Canadian PM Justin Trudeau promoted the visit as a thaw in relations between the two nations following a decade of chilly diplomacy under the Conservative government of PM Stephen Harper. According to news coverage in the Toronto Globe & Mail for 9/22/16, John Luxat, a professor of nuclear safety analysis at McMaster University, told the newspaper the new reactor technology has “high potential for use in China because of the large number of light water reactors” who spent fuel could be used by CANDU designs. However, AltgaCorp investment analyst Chris Murray told the newspaper he sees the design and marketing effort to be a slow, drawn out effort and does not expect there to be any near-term financial impact. CANDU stands for CANada Deuterium Uranium, because it was invented in Canada, uses deuterium oxide (also known as heavy water) as a moderator, and uranium as a fuel. CANDU reactors are unique in that they use natural, unenriched uranium as a fuel; with some modification, they can also use enriched uranium, mixed fuels, and even thorium. Thus, CANDU reactors are ideally suited for using spent fuel from light water nuclear reactors, or downblended uranium from decommissioned nuclear weapons, as fuel, helping to reduce global arsenals. CANDU technical description and schematic courtesy of AECL and the Canadian Nuclear Association CANDU reactors can be refueled while operating at full power, while other light water designs, including PWRs and BWRs, must be shut down for refueling. Moreover, because natural uranium does not require enrichment, fuel costs for CANDU reactors are very low. Canada is one of the world’s leading sources of uranium with rich deposits in Saskatchewan and other provinces. It has no uranium enrichment capabilities. The safety systems of CANDU reactors are independent from the rest of the plant, and each key safety component has three backups. This redundancy increase the overall safety of the system, and it also makes it possible to test the safety system while the reactor is operating under full power. There are 19 CANDU reactors in Canada and 31 globally including two in China, two in Argentina, and two in Romania. While all three countries are potential markets for the new SNC-Lavalin / CNNC design, only China has committed, in principle to building the new ACFR. It is unclear to what extent the new AFCR benefits from a design heritage with the now suspended work on the ACR-1000 which was proposed in 2007 and 2008 for Canadian and UK power markets. The ACR-1000, a 1200 MW CANDU type reactor design, was proposed to be built in the tar sands region of Alberta for power and process heat customers and at Point Lepreau in New Brunswick for electric power customers. Neither projects ever made it off the drawing boards. Efforts to license the 1200 MW unit with the Canadian Nuclear Safety Commission ended in Spring 2008 when AECL also withdrew the design from consideration in the UK generic design assessment. AECL CEO Hugh MacDiarmid was quoted at the time as saying, “We believe very strongly that our best course of action to ensure the ACR-1000 is successful in the global market place is to focus first and foremost on establishing it here at home.” But there were no sales at home due to Bruce Power declining to consider the 1200 MW reactor. In June 2011 SANC-Lavalin bought the reactor division of AECL for the bargain basement price of $15 million which included all of AECL’s intellectual property related to CANDU reactor designs. The advanced CANDU reactor (ACR), in its current design status, frozen in 2008, is a Generation III+ nuclear reactor design and is a further development of existing CANDU reactors designed by Atomic Energy of Canada Limited (AECL). The ACR is a light-water-cooled reactor that incorporates features of both pressurized heavy water reactors (PHWR) and advanced pressurized water reactors (APWR) technologies. It uses a similar design concept to the steam-generating heavy water reactor (SGHWR). The difference between heritage CANDUs and the ACR is that it uses low enriched uranium (LEU) fuel, (3-5% U235), ordinary (light) water coolant, and a separate heavy water moderator. The ACR also incorporates characteristics of the CANDU design, including on-power refueling with the CANFLEX fuel; two fast, totally independent, safety shutdown systems; and an emergency core cooling system. The relatively small reactor core reduces core size by half for the same power output over the older CANDU design. The ACR fuel bundle is a variant of the 43-element CANFLEX design (CANFLEX-ACR). The use of LEU fuel would result in higher burn-up operation than traditional CANDU designs. None of these features were found to be compelling by potential customers and AECL shelved the entire effort to develop the ACR. About the New AFCR According to SNC_Lavalin the Advanced Fuel CANDU reactor (AFCR) (fact sheet) is a 700MW Class Generation III reactor based on the successful CANDU 6 and Enhanced CANDU 6 (EC6) reactors with a number of adaptations to meet the latest Canadian and international standards. This is 300 MW less in power than the ACR and also differs technically from the ACR in that it uses only heavy water as a moderator. Its fuel flexibility allows it to use recycled uranium or thorium as fuel. SNC-Lavalin calls such materials “natural uranium equivalent” fuels, It uses a heavy water moderator and heavy-water coolant in a pressure tube design. CANDU reactors can be refuelled on power. The firm claims it will have “one of the highest lifetime capacity factors among the world’s reactors.” The development of the AFCR was first reported by World Nuclear News in November 2014. That report also provided insights into the place in China’s nuclear fuel cycle that would be the niche for the reactor. WNN noted in its report that the used fuel from four conventional PWR reactors can completely supply one AFCR unit (as well as providing recycled plutonium for MOX). This process significantly reduces the task of managing used fuel and disposing of high-level wastes. The R&D effort also explored the use of thorium as a fuel for the new reactor. In June 1998, construction started on a CANDU 6 reactor in Qinshan China of the Qinshan Nuclear Power Plant, as Phase III (units 4 and 5) of the planned 11 unit facility. Commercial operation began in December 2002 and July 2003, respectively. These are the first heavy water reactors in China. In 2015 China signed agreements in principle with Romania and Argentina to supply CANDU reactors. In a World Nuclear News report in November 2015 report details were revealed that China and Argentina had in 2014 signed a new high-level agreement towards construction of a third CANDU type pressurized heavy water reactor (PHWR) at the Atucha plant in Argentina. Under the agreement, CNNC will be providing goods and services and long-term financing. The utility in Argentina will be designer, architect-engineer, builder and operator of the new PHWR (Atucha 3). Under the agreement, over 70% of the components to be used in the plant will be supplied by Argentine companies. CNNC is now expected to advance the negotiations with Chinese financial institutions to conclude project financing. Atucha 3 will be a part Canadian-developed Candu reactor running on natural uranium fuel, like the 648 MWe Embalse Candu reactor in Córdoba province. Because of the localization strategy for major components, and the history of the supply chain in Argentina with the other CANDU reactors, it is unlikely that Atucha 3 could be based on the new AFCR design. Atucha 3 is expected to cost almost $6 billion and to take eight years to build at the Atucha Nuclear Power Plant Complex in Buenos Aires province, where the 335 MWe Atucha I and 745 MWe Atucha 2 currently operate. Also in November 2015 World Nuclear News reported Romania’s Nuclearelectrica signed a memorandum of understanding (MOU) with China General Nuclear (CGN) for the development, construction, operation and decommissioning of units 3 and 4 of the Cernavoda nuclear power plant. The Romanian national nuclear company said a joint venture project company is to be established, with CGN owning at least 51% of the share capital. That company will oversee construction of the units, which will be 700 MWe Candu 6 reactors. Two Candu units already operate at the Cernavoda site. Romania and China signed a letter of intent in November 2013 during a visit to Bucharest by Chinese premier Li Keqiang. Cernavoda is home to two operating Candu 6 pressurized heavy water reactors (PHWRs) supplied by Candu Energy’s predecessor, Atomic Energy of Canada Ltd (AECL), and built by a Canadian-Italian consortium of AECL and Ansaldo. Unit 1 started up in 1996, but work was suspended on a further four units in 1991. Unit 2 was subsequently completed and has been in operation since 2007. Given Romania’s history with CANDU reactors, and its intent to apply its operating experience with them to Units 3 & 4, it is unlikely that country would be a market for the new AFCR model. Romania will supply the fuel for all four reactors. According to the same World Nuclear News report, the new conventional CANDU units will have an operating life of 30 years with the possibility of extension by an additional 25 years. With Argentina and Romania committed to conventional CANDU, off-the-shelf, technology, it is unclear what the commercial prospects will be for the new AFCR CANDU design. The design intent to use spent nuclear fuel in the reactor would make it attractive to many countries. China will build and operate the first two units to prove to potential customers that the design is safe, affordable, and will have a long and cost-competitive service life. Assuming the units can be built in China for $3,000 to $4,000 per Kw, a 700 MW unit will cost approximately $2.1 billion to $2.8 billion which is far less than the cost in the U.S. for a 1000 MW Westinghouse AP1000. Similar cost comparisons would be expected for new nuclear reactors in the UK. However, China is proposing its new PWR design, the Hualong One, for the UK market. Once China has proven the technical and financial viability of the AFCR CANDU, it will face the uncertain prospects of design safety reviews for first-of-a-kind units by nuclear regulatory agencies in countries where it wants to sell the reactors. By leveraging the well-known CANDU technology, SNC-Lavalin and CNNC are placing a bet that they will find willing buyers of their new nuclear reactor.
Pietralik J.M.,Atomic Energy of Canada Ltd. |
Schefski C.S.,Atomic Energy of Canada Ltd.
Journal of Engineering for Gas Turbines and Power | Year: 2011
The three groups of parameters that affect flow-accelerated corrosion (FAC) are the flow conditions, water chemistry, and materials. Nuclear power plant (NPP) data and laboratory tests confirm that, under alkaline water chemistry, there is a close relationship between local flow conditions and FAC rates in the piping components. The knowledge of the local flow effects can be useful for developing targeted inspection plans for piping components and predicting the location of the highest FAC rate for a given piping component. A similar evaluation applies also to the FAC in heat transfer equipments such as heat exchangers and steam generators. The objective of this paper is to examine the role of the flow and mass transfer in bends under alkaline FAC conditions. Bends experience increased FAC rates compared with straight pipes, and are the most common components in piping systems. This study presents numerical simulations of the mass transfer of ferrous ions and experimental results of the FAC rate in bends. It also shows correlations for mass transfer coefficients in bends and reviews the most important flow parameters affecting the mass transfer coefficient. The role of bend geometry and, in particular, the short and long radii, surface roughness, wall shear stress, and local turbulence, is discussed. Computational fluid dynamics calculations and plant artifact measurements for short- and long-radius bends are presented. The effect of the close proximity of the two bends on the FAC rate is also examined based on CANDU (CANDU is a registered trademark of the Atomic Energy of Canada Limited) NPP inspection data and compared with literature data.
Nakla M.E.,King Fahd University of Petroleum and Minerals |
Groeneveld D.C.,Atomic Energy of Canada Ltd |
Groeneveld D.C.,University of Ottawa |
Cheng S.C.,University of Ottawa
International Journal of Multiphase Flow | Year: 2011
An experimental investigation of inverted annular film boiling heat transfer has been performed for vertical up-flow in a round tube. The experiments used R-134a coolant and covered a pressure range of 640-2390kPa (water equivalent range: 4000-14,000kPa) and a mass flux range of 500-4000kgm-2s-1 (water equivalent range: 700-5700kgm-2s-1). The inlet qualities of the tests ranged from -0.75 to -0.03. The hot-patch technique was used to obtain the subcooled film boiling measurements. It was found that the heat transfer vs. quality curve can be divided into four different regions, each characterized by a different mechanisms and trends. These regions are dependent on pressure, mass flux and local quality. A detailed examination of the parametric trends of the heat transfer coefficient with respect to mass flux, inlet quality, heat flux and pressure was performed; reasonably good agreement between observed trends and those reported in the literature were noted. © 2010 Elsevier Ltd.
Banath J.P.,Cancer Agency Research Center |
Klokov D.,Cancer Agency Research Center |
Klokov D.,Atomic Energy of Canada Ltd |
MacPhail S.H.,Cancer Agency Research Center |
And 2 more authors.
BMC Cancer | Year: 2010
Background: Evidence suggests that tumor cells exposed to some DNA damaging agents are more likely to die if they retain microscopically visible γH2AX foci that are known to mark sites of double-strand breaks. This appears to be true even after exposure to the alkylating agent MNNG that does not cause direct double-strand breaks but does produce γH2AX foci when damaged DNA undergoes replication.Methods: To examine this predictive ability further, SiHa human cervical carcinoma cells were exposed to 8 DNA damaging drugs (camptothecin, cisplatin, doxorubicin, etoposide, hydrogen peroxide, MNNG, temozolomide, and tirapazamine) and the fraction of cells that retained γH2AX foci 24 hours after a 30 or 60 min treatment was compared with the fraction of cells that lost clonogenicity. To determine if cells with residual repair foci are the cells that die, SiHa cervical cancer cells were stably transfected with a RAD51-GFP construct and live cell analysis was used to follow the fate of irradiated cells with RAD51-GFP foci.Results: For all drugs regardless of their mechanism of interaction with DNA, close to a 1:1 correlation was observed between clonogenic surviving fraction and the fraction of cells that retained γH2AX foci 24 hours after treatment. Initial studies established that the fraction of cells that retained RAD51 foci after irradiation was similar to the fraction of cells that retained γH2AX foci and subsequently lost clonogenicity. Tracking individual irradiated live cells confirmed that SiHa cells with RAD51-GFP foci 24 hours after irradiation were more likely to die.Conclusion: Retention of DNA damage-induced γH2AX foci appears to be indicative of lethal DNA damage so that it may be possible to predict tumor cell killing by a wide variety of DNA damaging agents simply by scoring the fraction of cells that retain γH2AX foci. © 2010 Banáth et al; licensee BioMed Central Ltd.
Pilgrim D.,Atomic Energy of Canada Ltd
Topical Meeting Held by the ANS Nuclear Criticality Safety Division, NCSD 2013 - Criticality Safety in the Modern Era: Raising the Bar | Year: 2013
Atomic Energy of Canada Limited (AECL) is Canada's premier nuclear science and technology organization. AECL has two locations where criticality safety is considered, Whiteshell Laboratories (WL) and Chalk River Laboratories (CRL). Between the two sites AECL has approximately 3280 employees. Its workforce is very diverse in that the company employs those with engineering, science, technology, administrative and business backgrounds as well as those in skilled trades. However there is also a contingent of employees who do not have post secondary education. The diversity of skills and educational background of AECL's workforce made it challenging to meet the intent of ANS 8.20 standard "Nuclear Criticality Safety Training" and the Canadian regulatory commitments that require all AECL employees to be trained in criticality safety. Nuclear Criticality Safety Program personnel at AECL decided the best approach to provide training for such a diverse community would be on a graded approach. Four categories were created in which every employee could be placed based on their required duties. This ranged from Category A, whose work could not impact criticality safety, to Category D, whose work could have serious consequences to criticality safety. A guidance document was issued directing the AECL management team to categorize their employees. The training outlined in the guidance document was a combination of pre-existing course material and material that needed to be developed. This material included awareness for all staff that covered a high level introduction to criticality safety, computer based training for emergency responders, and a revision to the nuclear criticality safety course that better aligned with ANS-8.20, Nuclear Criticality Safety Training. A graded approach to training began in 2009 and since then there has been a threefold increase in the number of employees receiving the full day training program and approximately 22 per cent of AECL staff has received basic awareness. With the aid of computer-based learning it is expected that the number of employees that will receive basic awareness will increase significantly over the coming year.
Ismail Y.,University of Montréal |
Ismail Y.,Atomic energy of Canada Ltd |
Mccormick S.,U.S. Department of Agriculture |
Hijri M.,University of Montréal
FEMS Microbiology Letters | Year: 2013
Trichothecenes are an important family of mycotoxins produced by several species of the genus Fusarium. These fungi cause serious disease on infected plants and postharvest storage of crops, and the toxins can cause health problems for humans and animals. Unfortunately, there are few methods for controlling mycotoxin production by fungal pathogens, and most rely on chemicals, creating therefore subsequent problems of chemical resistance. We tested the impact of the symbiotic arbuscular mycorrhizal fungus Glomus irregulare on a trichothecene-producing strain of Fusarium sambucinum isolated from naturally infected potato plants. Using dual in vitro cultures, we showed that G. irregulare inhibited the growth of F. sambucinum and significantly reduced the production of the trichothecene 4, 15-diacetoxyscirpenol (DAS). Furthermore, using G. irregulare-colonized potato plants infected with F. sambucinum, we found that the G. irregulare treatment inhibited the production of DAS in roots and tubers. Thus, in addition to the known beneficial effect of mycorrhizal symbiosis on plant growth, we found that G. irregulare controlled the growth of a virulent fungal pathogen and reduced production of a mycotoxin. This previously undescribed, biological control of Fusarium mycotoxin production by G. irregulare has potential implications for improved potato crop production and food safety. © 2013 Federation of European Microbiological Societies. Published by John Wiley & Sons Ltd. All rights reserved.
Le T.,McMaster University |
Ewing D.,McMaster University |
Schefski C.,Atomic Energy of Canada Ltd. |
Ching C.Y.,McMaster University
Nuclear Engineering and Design | Year: 2014
Flow-Accelerated Corrosion (FAC) is a major degradation mechanism affecting carbon steel piping systems in nuclear power plants (NPPs). Flow and mass transfer conditions determine the local distribution of wall thinning, even though chemistry and materials determine the overall potential for FAC. Different localized thinning rates in back-to-back elbow configurations between the first and second elbows have been noted at NPPs, and this difference depends on the distance between elbows, flow conditions, and the configuration of the back-to-back elbows (S-, C-, or out of plane). This paper will focus on mass transfer measurements for back-to-back elbows arranged in an out of plane configuration for different elbow separation distances under single-phase flow conditions. The mass transfer measurements were performed using a mass dissolution technique of gypsum test sections in water. The experiments were performed at a Reynolds number of 70,000 and a resulting Schmidt number of 1280, which is similar to that for the diffusion of the iron magnetite layer of carbon steel piping in water, providing a mass transfer environment analogous to that in NPPs. Experiments were performed with 0, 1, 2 and 5 pipe diameters in length between the elbows. The mass transfer results show regions of higher mass transfer in the second elbow in comparison to the first elbow. The maximum mass transfer enhancement factor decreased from 2.7 to 2.1 as the separation distance increased from 0 to 5 pipe diameters. Flow streaks on the second elbow surface indicated swirling flow and its strength decreased with increasing separation distances. The relative roughness in the upstream pipe was found to be 0.003-0.004. The roughness level in the second elbow is 1.5 times higher than the upstream pipe and decreases with increasing bend separation distance. © 2014 Elsevier B.V.
Duffey R.B.,Atomic Energy of Canada Ltd |
Sur B.,Atomic Energy of Canada Ltd
Energy Procedia | Year: 2011
We describe the contribution nuclear energy will make to global energy needs based on the sound foundation of existing technology, infrastructure, natural resources and human knowledge, while meeting the requirements of security of supply (energy independence) and growing demand. Currently all reactors internationally operate on an unsustainable once-through nuclear fuel cycle using uranium fuel. Future decisions will be increasingly based on strategic considerations involving the complete nuclear fuel cycle, including requirements related to supply assurances, resource utilization, proliferation resistance and radioactive waste disposal. Pressure tube reactor (PTR) technology using fuel channels is uniquely suited to respond to the future needs because of its inherent technical characteristics and associated fuel cycle flexibility. PTR channel technology concepts have also continued to advance based on 50 years of continuous development and improvement, with strategic considerations involving the complete nuclear fuel cycle related to: • Fuel Availability and Supply Assurances • Uranium, Plutonium and Thorium utilization • Waste Minimization • Proliferation Resistance (Safeguards) • Assured Licensability • Improved Safety • Cost Competitiveness We show how nuclear technology development and global sustainability is determined by R&D progress, with challenging technology goals for nuclear energy systems in the four areas of sustainability, economics, safety and reliability, and proliferation resistance and physical protection, leading naturally to the next phase of PTR channel development, namely the high efficiency Supercritical Water Reactor (SCWR). Aggressive targets have been set for R&D and advanced concepts, complementary to the approaches taken in India, which support enhanced safety, cost reduction, resource sustainability, and economical and efficient operation. © 2011 Published by Elsevie Ltd.
Li H.,Atomic Energy of Canada Ltd. |
Tan G.,Atomic Energy of Canada Ltd. |
Tan G.,Ontario Power Generation |
Zhang W.,Atomic Energy of Canada Ltd. |
Suppiah S.,Atomic Energy of Canada Ltd.
Applied Energy | Year: 2012
The Sulfur-Iodine (S-I) cycle has been considered as one of the efficient and promising thermochemical water-splitting cycles for hydrogen production using nuclear energy. However, the catalytic SO3 decomposition process in the S-I cycle demands high temperature heat (>800°C). Existing nuclear reactors cannot provide such heat for SO3 decomposition. AECL proposed a direct resistive heating concept to compensate for the requirement of high temperature heat. An experimental program was established at AECL to demonstrate the concept and to develop reliable catalyst structures for SO3 decomposition. Due to the high temperature and harsh chemical environment, Hastelloy C-276 was selected as the material for the heating element and reactor. The catalyst was directly applied on the surface of an electrical heating element. SO3 was produced online from H2SO4 in a pre-heated vessel. The SO3 decomposition percentage was determined using the measured O2 concentration in the exit gas stream. The results showed that SO3 decomposition can be successfully achieved with the direct resistive heating method. As much as 90% of the initial SO3 was decomposed under the experimental conditions explored. The Pt-based catalyst performed better than the Fe-based catalyst in the low temperature region (<700°C). The effect of carrier gas flow on SO3 decomposition was also considered. © 2011.