December 5, 2019
By: Donald Jones, retired nuclear industry engineer, 2019 December 04
The announcement on 2019 December 01 by the premiers of Saskatchewan, Ontario and New Brunswick about cooperation in the development and deployment of Small Modular Reactors (SMRs) (Ref. 1) should not have come as a total surprise. Ontario Power Generation (OPG) has been working with a Micro Modular Reactor (MMR- a smaller version of a SMR) vendor to assist in getting its design through the pre-licensing vendor design reviews of the Canadian Nuclear Safety Commission (CNSC). Bruce Power and New Brunswick Power have also been working with SMR vendors. There are many (Ref.2) SMR vendors at different stages in the review pipeline of the CNSC with no two reactors being the same. In some cases the design is an improved version of small reactors that have operated successfully in the past but now meeting current design codes and safety regulations in a modular configuration. Small reactors of various capacities and capable of rapid power manoeuvring have been used for many years to power submarines, aircraft carriers (U.S.A) and heavy duty ice breakers (Russia). Small reactors of many different designs are not new but the concept of designing them for serial construction and collectively to comprise a large nuclear power plant is new. Read the rest of this entry »
October 15, 2019
By: Donald Jones, retired nuclear industry engineer, 2019 October 15.
Combined Cycle Gas Turbine (CCGT) power plants produce electricity very efficiently with steady state baseload and intermediate load overall efficiencies of over 60 percent for newer units. Most presently operating units would have somewhat lower efficiencies. Unfortunately on most power grids with significant wind and/or solar generation they rarely operate at steady state (reference 1).
Operating CCGT plants at part load and stopping and starting the gas turbine(s) leads to lower plant efficiency and increased emissions of greenhouse gases per megawatt-hour generated as well as wear and tear on the units. Every time a CCGT is warming up gas is being burned while the gas turbine is slowly increasing power and warming up the heat recovery steam generators (HRSG), and the steam turbine when steam becomes available. This increases the heat rate giving higher kg GHG/MWh. If a CCGT had a bypass stack the gas turbine could be delivering power very quickly but at high heat rate (low overall efficiency) since it would be operating simple cycle and giving higher kg GHG/MWh. When increasing power (manoeuvring/ramping) in the operating range the HRSG and steam turbine metal have to be warmed up which takes away gas for no useful power output, increasing kg GHG/MWh.
Ontario has a very low carbon intensity electricity grid averaging 40 g CO2e/kWh. In 2018 more than 93 percent of the electricity generated in Ontario came from non-GHG emitting resources, predominantly nuclear and hydro with some wind and solar. Nuclear provided 61 percent of generation in 2018. There is about 10000 MW of gas-fired and oil-fired generation connected to the transmission grid mostly CCGTs burning natural gas and just under 5000 MW (nameplate capacity) of wind and solar. CCGTs are used to meet peak load demands and provide operational flexibility. Read the rest of this entry »
April 3, 2018
By: Donald Jones, retired nuclear industry engineer, 2018 March 29
The raw performance data for 2017 are taken from the Power Reactor Information System (PRIS) database of the International Atomic Energy Agency (IAEA). Note that the Load Factor term used in the PRIS database has the same meaning as Capacity Factor (CF). CFs are based on the (net) Reference Unit Power and on the (net) Electricity Supplied, as defined in the PRIS database. For Ontario, at least, the Energy Availability Factor in the PRIS database can be read as the Unit Capability Factor, see later. More information and performance data for 2016 are in references 1 and 2.
The performance of some of Ontario’s nuclear generating stations is affected by the surplus baseload generation (SBG) in the province. The surplus usually arises because of unreliable intermittent wind generation coming in at times of low demand and wind generation is expected to increase even more over the next several years. Some nuclear units saw electricity output reductions during periods of surplus baseload generation (SBG). This means the CFs are not a true performance indicator for those units (reference 3). A better metric of performance in these cases would be the Unit Capability Factor (UCF – used by Ontario Power Generation and by Bruce Power). The Energy Availability Factor (EAF) is another performance indicator and is shown in the PRIS database. The EAF adjusts the available energy generation for energy losses attributed to plant management, planned and unplanned, and for external energy losses beyond the control of plant management while the UCF only includes energy losses attributed to plant management and excludes the external losses beyond control of plant management like load cycling/load following, grid failures, earthquakes, cooling water temperature higher than reference temperature, floods, lightning strikes, labour disputes outside the plant etc.
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March 18, 2016
By: Donald Jones, P.Eng., retired nuclear industry engineer, 2016 March 19
At the end of 2015 Darlington had a four unit average lifetime Capacity Factor (CF) of 83.6 percent and an average annual CF of 76.1 percent. Bruce A had a four unit average lifetime CF of 69 percent and an average annual CF of 86.1 percent. Bruce B had a four unit average lifetime CF of 83.5 percent and an average annual CF of 84.4 percent. The six unit Pickering station had a six unit average lifetime CF of 72.4 percent and an average annual CF of 78.6 percent. Performance data for 2014 are discussed in reference 1.
The raw performance data for 2015 are taken from the Power Reactor Information System (PRIS) database of the International Atomic Energy Agency (IAEA). Note that the Load Factor term used in the PRIS database has the same meaning as CF. CFs are based on the (net) Reference Unit Power and on the (net) Electricity Supplied, as defined in the PRIS database.
The performance of some of Ontario’s nuclear generating stations is affected by the surplus of generation in the province. The surplus usually arises because of unreliable intermittent wind generation coming in at times of low demand and wind generation is expected to increase even more over the next several years. Some nuclear units saw electricity output reductions during periods of surplus baseload generation (SBG). This means the CFs are not a true performance indicator for those units (reference 2). A better metric of performance in these cases would be the Unit Capability Factor (UCF – used by Ontario Power Generation and by Bruce Power) or the Energy Availability Factors (EAF) that are shown in the PRIS database. The EAF adjusts the available energy generation for energy losses attributed to plant management and for external energy losses beyond the control of plant management while the UCF only includes energy losses attributed to plant management and excludes the external losses attributed to grid related unavailability and other things. This means that on unreliable grids, for example, UCF will be significantly higher than EAF but for Ontario there will be no significant difference. The UCF and the EAF take into account reductions in plant output due to load cycling and load following. For units that load cycle and/or load follow the CF will be significantly lower than the EAF. For example, Bruce B unit 5 has a 2015 annual CF of 86.4 percent and an EAF of 91 percent. The UCF and the EAF are based on reference ambient conditions so, unlike the CF, they cannot exceed 100 percent. In some cases the CF is more than the EAF because of lower than design cooling water temperatures that increase the electrical output of the unit. The only reason for using the EAF here (see later for Bruce units) instead of the UCF is that EAFs are now available in PRIS and UCFs are not presently available (well, the author could not find them).
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June 23, 2015
By: Donald Jones, P.Eng., retired nuclear industry engineer – 2015 June 22
How far can we extend the continuous on-line power operation (breaker-to-breaker runs) of the world’s commercial Generation II and Generation III nuclear power plants. Is 1,000 days possible?
To date the world record for PWR (Pressurized Water Reactor) continuous on-line operation is the 705 day run by Three Mile Island unit 1 an 819 MWe (net) unit in the U.S. that went into commercial operation in 1974 September. The run ended in 2009 October when the unit went into a planned refuelling outage. This run broke the previous world record of 692 days of another PWR, Calvert Cliffs unit 2, an 850 MWe (net) unit in the U.S. that was put into commercial operation in 1977 April. This run ended in 2009 February with a refuelling outage.
LaSalle unit 1, a 1137 MWe (net) unit in the U.S., that was put into commercial operation 1984 January, holds the world record for a BWR (Boiling Water Reactor) with 739 days when it came off-line in 2006 February. As it happens its twin, LaSalle unit 2, became the second place world record holder when it completed a run of 711 days on 2007 February. LaSalle unit 2 went into commercial operation in 1984 October. LaSalle units now hold first and second places in the world for a continuous run of any LWR (Light Water Reactor).
The world record for any type of reactor is held by a CANDU. This is Pickering unit 7, a 516 MWe (net) unit in Ontario, Canada, with a continuous run of 894 days when it came off-line for maintenance in 1994 October. This unit was put into commercial operation in 1985 January. CANDU is a PHWR (Pressurized Heavy Water Reactor). Rajasthan unit 5, a 202 MWe (net) PHWR in India, put into commercial operation in 2010 February, holds second place to Pickering unit 7 in world ranking after completing a 765 day continuous run and going into its planned biennial maintenance outage in 2014 September. Besides these record breaking runs there have been many runs of over 400 days by the different types of reactors.
These long runs are terminated when it is time for the planned maintenance outage and are not extended until safety targets can no longer be met, which would mean shutting down the unit at an inopportune time. The practical limit of continuous operation of PWRs and BWRs is set by the need to replace about a third of the nuclear fuel and do maintenance after about two years (720 days) or less. In the U.S. most light water reactors units operate on a 18 month fuel cycle and have maintenance outages scheduled for the spring and autumn months when electricity demand is low. Since a pressure tube PHWR like CANDU can refuel on-line at power the length of continuous operation is indeterminate but in practice there is a need to come off-line for certain tests, maintenance and inspections and upgrades that cannot be done at power. For a PHWR the run could be terminated by initiating one of the two reactor safety shutdown systems with the other reactor safety shutdown system being tested during the maintenance outage. The Enhanced CANDU 6 (EC6) is designed to operate for about three years (1080 days) before coming off line for a month for maintenance and inspections. Having some testing and maintenance done on-line reduces the inspection load during unit maintenance outage. Read the rest of this entry »
March 29, 2015
By: Donald Jones, P.Eng., retired nuclear industry engineer
Most of India’s nuclear reactors are of the pressurized heavy water reactor (PHWR) type with horizontal pressure tubes, just like the Canadian designed CANDU. In fact the first PHWR in India was the Rajasthan Atomic Power Project (RAPP) unit and was a CANDU designed by Atomic Energy of Canada Limited (AECL) that used the Douglas Point unit in Ontario as reference design but modified to aid localization. RAPP-1 entered commercial operation 1973 December. While RAPP-1 was being constructed the design of RAPP-2 was started (Author’s note: I know because I was part of design team). However the detonation of a nuclear device by India in 1974 curtailed completion of the design by AECL and India was on its own as far as nuclear technology was concerned. The design was completed by India and RAPP-2 eventually entered commercial operation in 1981 April. Since those early days India has developed its own indigenous designs of PHWRs with net electrical outputs of 202 MW, 490 MW, and 630 MW. They bear little to no relation to Douglas Point. All 18 PHWR units operating in 2014 (including RAPP-1 which has been shutdown since 2004) were 202 MW (220 MW gross) except for two 490 MW (540 MW gross) units. There were four 630 MW (700 MW gross) units under construction with none in operation. All nuclear power units, except for RAPP-1, are designed, owned, and operated by Nuclear Power Corporation of India Ltd. Several of the country’s PHWRs have been refurbished for extended life operation. For more detailed information on the Indian nuclear program see, Nuclear Power in India (reference 1). Read the rest of this entry »
March 25, 2015
By: Donald Jones, P.Eng., retired nuclear industry engineer, 2015 March 24
Following on from some early conceptual work by Canadian General Electric (CGE), Atomic Energy of Canada Limited (AECL) based the CANDU 6 design on the four unit Pickering A station (that was brought into service 1971-1973) but as a single unit station with a significant power increase, major equipment simplifications, improvements in shutdown and emergency core cooling systems, extensive use of digital computers for control and safety systems etc. In fact the CANDU 6 is unrecognizable as being based on Pickering except maybe for the fuel channel sizing, even though fewer channels are in CANDU 6, and the two loop primary heat transport system that were retained. Since Ontario Hydro was enamored by multi-unit stations CANDU 6 was intended as a single unit for out of province build including off shore. The two lead CANDU 6 projects were Gentilly 2 in Quebec and Point Lepreau in New Brunswick and these were quickly followed by Embalse in Argentina and Wolsong, now Wolsong 1, in South Korea and all came into service in the early to mid 1980s. These can be regarded as the first tranche of CANDU 6 build.
The second tranche of CANDU 6 units came with Wolsong 2, 3 and 4 in South Korea, Cernavoda 1 and 2 in Romania, and Qinshan 4 and 5 in China (the other units at Qinshan site are not CANDU), all entering service between 1996 to 2007. Each of the second tranche CANDU 6 units incorporate lessons learned from operation of the earlier units with changes to meet latest regulatory codes and standards. All three Wolsong units came in on budget and on schedule and the two Qinshan units came in under budget and ahead of schedule. In fact the total project schedule for the CANDU 6 units at the Qinshan site in China was 81 months from contract effective date to in-service.
Unlike the 2013 CANDU 6 performance figures (reference 1) the Capacity Factors are taken from the Power Reactor Information System (PRIS) database of the International Atomic Energy Agency (IAEA). Note that the Load Factor term used in the PRIS database has the same meaning as Capacity Factor. Capacity Factors are based on the (net) Reference Unit Power and on the (net) Electricity Supplied figures, as defined in the PRIS database.
CANDU 6 Units
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