Enhanced CANDU 6 and NuScale SMR have capability to easily integrate wind and solar

August 17, 2016

by: Donald Jones, P.Eng., retired nuclear industry engineer, 2016 August 17.

Nuclear power plants do not like to operate at anything less than 100 percent full power. The main reason is that capital costs for nuclear are high and fuel costs are low so fuel cost savings are negligible at reduced power while revenue losses are appreciable. Another reason is that when reactor power is reduced relatively quickly there is an increase of Xenon-135 in the fuel, a fission product, that tends to reduce reactivity and sets a limit on the rate and depth of any power reduction that can be achieved before the reactor shuts itself down, the so called “poison out”. On a CANDU this is about a 40 percent reactor power reduction to a reactor power of 60 percent after a fast power reduction. Xenon also slows the return to full reactor power. The xenon transient means that frequent power changes, down and up, in support of load following dispatches, would be difficult. Indeed CANDU was not designed to load follow although it was designed to load cycle, that is, reduce reactor power overnight and return to full power in the morning, without bypassing steam around the turbine to the condenser. Light water reactors use enriched fuel so are better able to respond to the xenon transient, at least with a fresh core.

In the past some domestic units and off-shore units (CANDU 6) did accumulate considerable good experience with load cycling, with some deep reactor power reductions, but not on a continuous daily basis. For example back in the 1980s several of the Bruce B units experienced nine months of load-cycling including deep (down to 60 percent full power, or lower) and shallow power reductions. All done without steam bypass. Analytical studies based on results of in-reactor testing at the Chalk River Laboratories showed that the reactor fuel could withstand daily and weekly load-cycling. However this load cycling capability has been configured out of the Ontario CANDUs and they presently operate continuously at 100 percent reactor power. Note that the eight units at the Bruce Nuclear Power Station load cycle when required to do so by bypassing steam around the turbine to the condenser but the reactor remains at full 100 percent power. With certain restrictions station electrical output can be reduced to around 60 percent of the full electrical output (reference 1).
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CANDU cousins in India – Performance in 2015

March 28, 2016

By: Donald Jones, P.Eng., retired nuclear industry engineer, 2016 March 28

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 (not the first nuclear reactor) 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 resemblance to Douglas Point. All 17 PHWR units operating in 2015 (excludes 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 PHWR 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).

The performance data 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, so capacities referenced in this article are net electrical MW output. The lifetime, or cumulative, CF is based on the date of commercial operation and will include the outage time if the unit has been refurbished. Only the performance of India’s PHWRs is reviewed in detail but India’s three operating non-PHWR units are mentioned. Read the rest of this entry »


CANDU 6 Performance in 2015

March 28, 2016

By: Donald Jones, P.Eng., retired nuclear industry engineer, 2016 March 27

History

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 3-1 and 3-2 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.

More information on the CANDU 6 projects can be found in, CANDU 6 Performance in 2014 (reference 1). Note that the Power Reactor Information System (PRIS) database of the International Atomic Energy Agency (IAEA) identifies the Qinshan units in China as Qinshan 3-1 and Qinshan 3-2, that is, units 1 and 2 of Phase 3 of the Qinshan Nuclear Power Project. These were identified previously (reference 1) as Qinshan 4 and 5.

Capacity Factor

The Capacity Factors are taken from the PRIS database. 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. The annual Energy Availability Factor (reference 2) will only be given in this article if it is significantly different from the unit Capacity Factor.

CANDU 6 Units

Point Lepreau, New Brunswick, Canada. At the end of 2015, just over three years after refurbishment, the “refurbished lifetime” Capacity Factor was 75 percent and the annual Capacity Factor for 2015 was 74.0 percent. The lifetime Capacity Factor since start of commercial operation in 1983 was 69.7 percent, including the refurbishment outage. Read the rest of this entry »


Performance of Ontario’s CANDU nuclear generating stations in 2015

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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1,000 day breaker-to-breaker run is possible with CANDU

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 »


CANDU cousins in India – Performance in 2014

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 »


CANDU 6 Performance in 2014

March 25, 2015

By: Donald Jones, P.Eng., retired nuclear industry engineer, 2015 March 24

History

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.

Capacity Factor

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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