Status report 69 - Advanced Candu Reactor 1000 (ACR-1000)

31 Pages·2012·4.82 MB·English

Status report 69 - Advanced Candu Reactor 1000 (ACR-1000)
Overview
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Description
The Advanced CANDU Reactor®‑1000 (ACR‑1000®[1]) design is Atomic Energy of Canada Limited’s (AECL®) evolutionary, Gen III+,
1200 MWe class pressure tube reactor. The designer’s objectives have been to meet the industry and public expectations for safe, reliable,
environmentally friendly, low-cost nuclear power generation. It has been designed to be licensable internationally by ensuring its
compliance with Canadian nuclear regulations and the fundamental safety objectives of the IAEA's safety standards.
The ACR-1000 is a light water cooled, heavy water moderated pressure tube reactor, using low enriched uranium (LEU) fuel with
on-power fuelling.
1.1 Historical basis
AECL has established a successful, internationally recognized line of CANDU®[2] pressure tube reactors that use a heavy water moderator,
in particular, the medium‑sized CANDU 6 reactor. AECL has consistently adopted an evolutionary approach to the enhancement of
CANDU nuclear power plant designs over the last 30 years.
The ACR‑1000 design has evolved from AECL’s in‑depth knowledge of CANDU structures, systems, components (SSCs) and materials,
as well as from the experience and feedback received from operating CANDU plants. The ACR‑1000 incorporates major improvements in
economics, inherent safety characteristics, and performance, while retaining the proven features of the CANDU family of nuclear power
plants.
The CANDU system is ideally suited to this evolutionary design approach since the modular fuel channel reactor design can be modified
extensively, through a series of incremental changes in the reactor core design, to adjust the power output and improve the overall safety,
economics, and performance. A summary of the CANDU plants in‑service is provided in Table 1‑1.
1.2 Features
The ACR-1000 design retains many essential features of the CANDU plant design, including horizontal fuel channel core, a low
temperature heavy-water moderator, a water filled reactor vault, two independent safety shutdown systems, a highly automated control
system, on-power fuelling and a reactor building that is accessible for on-power maintenance and testing. It is designed with a focus on
operations and maintenance (O&M), drawing on AECL’s experience in the design, construction and operation of existing CANDU plants
for utilities around the world, as well as on valuable customer input.
The following key differences from the traditional CANDU design are incorporated into the design of the ACR-1000:
The use of LEU fuel contained in CANFLEX-ACR®[3] fuel bundles,
The use of light water instead of heavy water as the reactor coolant,
Lower moderator volume to fuel ratio.
These features together with a number of other evolutionary changes lead to the following benefits for the ACR-1000 design:
A more compact core design, which further reduces heavy water inventory;
An increased burn-up as a result of the fuel enrichment;
Increased safety margins (e.g., improved critical heat flux performance),
Improved overall turbine cycle efficiency through the use of higher pressures and higher temperatures in the coolant and steam
supply systems;
Reduced emissions, through the elimination of tritium production in the coolant and other environmental protection
improvements;
Enhanced severe accident management by providing backup heat sinks;
Improved performance through the use of advanced operational and maintenance information systems;
Improved separation of redundant SSCs important to safety through the use of a four quadrant plant layout.
These advancements as well as improvements to project engineering, manufacturing, and construction technologies, including the use of
state-of-the-art tools, result in significantly reduced capital cost and construction schedules while enhancing the inherent safety of the
ACR-1000 design.
1.3 Safety philosophy
The ACR-1000 plant is designed based on the “defence-in-depth” safety philosophy applied to all CANDU plants with enhancements to
further improve the overall safety of the plant. This includes the core design incorporating a negative power coefficient of reactivity,
improved performance of safety systems, a robust containment design to meet Canadian and international practice for new nuclear power
plants, and ensuring a measure of inherent safety by incorporating a set of reactivity coefficients that result in a self-limiting reactivity
transient even in the event of a large loss-of-coolant accident (LOCA) in which the shutdown systems do not operate.
The design approach for the ACR-1000 plant ensures safety during construction, commissioning, start-up, and O&M by assuring the
following:
Ample thermal and safety margins for safe operation;
Collective radiation exposure to plant personnel and the public is well below regulatory limits;
Feedback from construction, commissioning, start-up, and O&M is incorporated;
The “defence-in-depth” safety philosophy is applied;
Physical and functional separation of safety systems;
Simplified, more reliable systems;
Safety systems perform their required safety functions during design basis accidents without credit for active mitigation by the
process systems;
Human factors engineering principles and criteria are applied in the design of systems, facilities, equipment, and procedures;
Compliance with applicable government regulations related to environmental protection;
Security and robustness of the plant against malevolent acts and external events;
Severe accidents are considered in the plant design.
Table 1‑1
Representative CANDU Reactors
* Electrical equivalent (electricity plus process steam).

The major nuclear systems of each ACR‑1000 unit are located in the Reactor Building and the Reactor Auxiliary Building. These nuclear
systems include the following:
Reactor assembly, consisting of a calandria with 520 channels in a square lattice pitch.
LEU fuel contained in the CANFLEX‑ACR fuel bundle.
Heavy water moderator system.
Heat transport system (HTS) with light water coolant in a two‑loop, figure‑of‑eight configuration with four steam generators, four
heat transport pumps, four reactor outlet headers and four reactor inlet headers.
Fuel handling system, which consists of two fuelling machines (FMs), each mounted on a FM bridge and columns, located at both
faces of the reactor.
Main steam supply system.
Safety systems, consisting of two shutdown systems, emergency core cooling (ECC) system, emergency feedwater (EFW) system -
defined as the emergency heat removal system, and containment system plus associated safety support systems.
Figure 2‑1 ACR-1000 Nuclear Systems Schematic
2.1 Reactor assembly
The ACR-1000 reactor assembly consists of a horizontal, cylindrical calandria and end shield assembly enclosed and supported by a
steel‑lined concrete vault. The calandria contains heavy water moderator; and the vault contains light water, which serves as both a thermal
and a biological shield. The relatively cool moderator is independent of the hot, pressurized, light water HTS coolant in the fuel
channels. The reactor vault roof supports the reactivity mechanisms deck.
Fuel is replaced while the reactor is on power in order to maintain criticality. This feature contributes to high availability factors and
improved outage flexibility since refuelling outages at fixed cycle times are not required.
Figure 2‑2 Reactor Assembly
2.1.1 Reactor core design
The reactor core design is optimized to produce thermal power at a reduced capital cost, with enhanced control and safety features, while
reducing the quantity of spent fuel. It has been specifically designed with the following characteristics:
A more compact size.
A significant reduction in heavy water inventory requirements due to the more compact size and the use of light water as the
coolant.
Small and predominantly negative values of reactivity coefficients important to accident behaviour.
A substantial simplification in reactor control by operating with negative feedbacks in reactor power.
An increase in the reliability and effectiveness of the ECC system due to the simplified interface with the HTS since they both use
light water.
A flat, radial flux profile enables the reactor to operate at a very high radial power form factor. The axial power shape is also relatively
flat. This shape results naturally from the characteristics of the core, coupled with the bidirectional fuelling scheme, and acts to reduce the
capital costs by minimizing the size of the core relative to its power output.
2.1.2 Fuel channel assembly
The ACR‑1000 fuel channel assembly is illustrated in Figure 2‑3. The fuel channel assembly comprises of a zirconium alloy (Zr‑2.5wt%
Nb) pressure tube, two stainless steel end‑fittings (one at each end of the pressure tube), and various internal components. Each pressure
tube is thermally insulated from the low‑temperature, low‑pressure moderator by the carbon dioxide filled gas annulus between the pressure
tube and the calandria tube. Spacers, positioned along the length of the pressure tube, maintain the annular space and prevent contact
between the two tubes. Each end‑fitting holds a liner tube, a fuel support plug, and a channel closure.
The fuel channel is designed to achieve a target 30‑year operating life with a greater than 90% lifetime capacity factor. The fuel channel
design will accommodate the predicted sag and the axial and diametric growth due to irradiation of the pressure tube over its operating life.
Figure 2‑3 Fuel Channel Assembly
2.1.3 Reactor control
The primary reactivity control elements in the ACR‑1000 design are the mechanical zone control units. Each unit is divided into two
independently movable absorber elements. Day‑to‑day reactivity control is performed by on‑power refuelling and zone control action.
The reactor regulating system also includes mechanical control absorber units that can be used if larger power reductions are
required. Control absorber rods can either be inserted slowly under motor control or dropped into the core by gravity.
All reactivity mechanisms operate in the water filled reactor vault or the low‑temperature, low‑pressure moderator environment and are
simple, rugged, highly reliable, and require little maintenance.
2.2 Fuel
Figure 2-4 CANFLEX ACR Fuel Bundle
The 43‑element CANFLEX‑ACR fuel bundle is used for the ACR‑1000, shown in Figure 2‑4. The outer three rings of fuel elements
contain enriched uranium pellets while the central fuel element contains burnable neutron absorber in a stabilized zirconia matrix. A small
amount of burnable poison is incorporated into the LEU fuel elements for control of initial reactivity. A thin layer of CANLUB (graphite
coating) covers the inside surface of the fuel element sheaths to enhance the fuel elements’ ability to withstand power ramp.
The fuel sheath, endcaps, endplates and appendages are made of Zircaloy‑4 for its nuclear characteristics of low neutron absorption, good
corrosion resistance, and low hydrogen pickup.
The same fuel bundle design is used for all stages of the life of the core, from the initial start-up core to the reference equilibrium core. The
level of enrichment of the uranium dioxide in the fuel pellets for the start-up core is lower than that for the reference fuel. With a much
higher operating margin for initial years of reactor operation, the amount of burnable neutron absorber in the central element can be reduced
to achieve an increase in the exit burnup.
Major features include the use of enriched uranium and neutron absorbers (dysprosium and gadolinium) in a zirconium oxide matrix in the
centre element, and uniform diameter of all elements except the centre element.
One of the features of the ACR-1000 design is its capability to efficiently burn a variety of fuel cycles including thorium and mixed oxide
(MOX). This will enhance economics, extend nuclear fuel resources and assure energy security well into the future, while reducing waste
and enhancing proliferation resistance.
2.3 Heat transport and auxiliary systems
Figure 2-5 Heat Transport System Layout
The HTS consists of two figure‑of‑eight loops circulating pressurized light water coolant through the reactor fuel channels to remove heat
produced by nuclear fission in the core. The fission heat is carried by the HTS coolant to the steam generators to produce steam, on the
secondary side, which subsequently drives the turbine generator.
The major components of the HTS are the fuel channels, four steam generators, four electrically driven heat transport pumps, four reactor
inlet headers, four reactor outlet headers, and the interconnecting piping. Light water coolant is fed to the fuel channels from the inlet
headers at each end of the reactor and returned to the outlet headers at the opposite end of the reactor (see Figure 2-5).
2.3.1 Heat transport pumps
The heat transport pumps are similar to the CANDU 6 and Darlington double‑discharge design. Each pump is driven by a vertical, totally
enclosed, air‑water‑cooled, squirrel cage induction motor. The motors have a built-in inertia arrangement to eliminate the possibility of
missile generation. The arrangement has sufficient rotational inertia to prolong pump run down for a few minutes after loss of Class IV
electrical power. The pump shaft sealing arrangement consists of three main mechanical face seals and one back‑up seal in series. Each
mechanical seal is designed to withstand the full differential pressure. In the unlikely event that all three main seals fail, the back-up seal
prevents significant leakage of coolant to containment during depressurization of the HTS. The three mechanical seals and back-up seal are
constructed as a single seal package assembly in a removable cartridge for easy replacement.
2.3.2 Steam generators
The ACR-1000 steam generators are similar in design to CANDU 6, except for an increased tube diameter and a larger overall physical
size. The steam generators transfer heat from the light water coolant on the primary side of the steam generator to raise the temperature of,
and boil, feedwater on the secondary side to produce steam. Each steam generator consists of an inverted, vertical U‑tube bundle installed
in a cylindrical shell, and an integrated preheater section. Each is connected to supply headers for emergency feedwater and reserve water
systems. Steam‑separating equipment is provided in the steam drum in the upper part of the shell. The four steam generators are designed
to maintain both primary and secondary pressure boundaries and heat sink requirements during a design basis earthquake (DBE).
2.3.3 Feeder header assembly
The reactor inlet headers and outlet headers at each end of the reactor provide a means of distributing the flow to the inlet and from the
outlet ends of the respective channels. Both the inlet and outlet feeders are low‑carbon type 316N stainless steel pipes to prevent
flow‑assisted corrosion and minimize the susceptibility to stress corrosion cracking. The headers and upper feeders at each end of the
reactor are pre‑assembled as modules (see Figure 2‑6).
Figure 2-6 ACR-1000 Headers/Upper Feeders Modules
2.3.4 Pressure and inventory control system
The ACR‑1000 pressure and inventory control system provides pressure and inventory control and overpressure protection for the
HTS. The principal system components are the pressurizer, degasser condenser, degasser cooler, coolant storage tanks, as well as the heat
transport pressurizing pumps and associated control and safety valves, and instrumentation. The principal functions of the system are to:
Control HTS pressure and inventory over the full range of HTS and reactor operating modes;
Limit HTS pressure variations caused by transients to acceptable values;
Accommodate HTS coolant thermal expansion and contraction;
Provide degassing of the HTS coolant;
The major components of this system are described below:
The pressurizer is connected to the HTS via one of the two reactor outlet headers in each loop. Flow to and from the heat transport
purification system provides pressure control and inventory make‑up to the HTS during reactor transients. Spray flow and heaters
provide pressure control.
The degasser condenser acts as a receiver for relief flow from the HTS, acts as a receiver and condenser for steam relief from the
pressurizer, and is used as a means of degassing the HTS. The degasser condenser pressure is controlled using a cooling spray flow
and electrical heaters.
The degasser cooler heat exchanger cools the water flow at the bottom nozzle of the degasser condenser and returns it to the
pressurizing pump suction.
Pressurizing pumps provide water make‑up to the HTS during normal operation and upset conditions.
The coolant storage tank provides inventory to accommodate the swell and shrinkage for the HTS as required.
Liquid relief valves provide overpressure protection to the HTS by discharging the coolant to the degasser condenser.
2.4 Moderator system
The moderator system is a low‑pressure and low‑temperature system. The system maintains the structural integrity of the calandria
assembly by removing the heat generated in the heavy water by neutron moderation, and the heat transferred to the heavy water from the
calandria structures. The heavy water also serves as a neutron reflector and a medium for dispersing neutron poisons for rapid reactor
shutdown.
The moderator system consists of pumps and heat exchangers that circulate heavy water moderator through the calandria vessel. During
normal operation, the heavy water moderator is drawn from the calandria via four outlet ports located near the top of the calandria
vessel. The heavy water is circulated by one moderator pump and through four heat exchangers to remove the moderator heat load. The
flow returns to the calandria via twelve inlet nozzles located just beneath the outlet ports.
The locations of the inlet and outlet nozzles on the calandria vessel are optimized to ensure uniform moderator temperature distribution
throughout the reactor.
The moderator system is also connected to the reserve water system, which provides back-up water (light water) to the calandria vessel
following a loss of moderator during a severe accident. In addition, the moderator system interfaces with other auxiliary systems to purify
the heavy water, remove radiolytic deuterium and oxygen, maintain heavy water inventory, and to disperse chemicals for reactivity and
chemistry control.
2.5 Fuel handling system
The function of the fuel handling system is to refuel the reactor on demand for the purpose of controlling the reactor power distribution and
excess reactivity. The fuel handling system stores and handles fuel, from the arrival of new fuel to the storage of spent fuel. The fuel
handling system is divided into new fuel transfer and storage, fuel changing, and spent fuel transfer and storage.
2.5.1 New fuel transfer and storage
The new fuel transfer and storage system supplies the FMs with sufficient fuel to maintain full‑power operation. Since refuelling is
normally performed on‑power, new fuel is stored, inspected, and loaded from an accessible area in the Reactor Auxiliary Building.
2.5.2 Fuel changing
Refuelling is done remotely using two FMs, one at each reactor face. The two most downstream fuel bundles are discharged into one FM
while two new fuel bundles are inserted at the upstream end by the other FM.
A bridge and carriage system allows the FM to traverse the entire reactor face for access to all fuel channels. The system also allows the
FM access to the service port and the new and spent fuel transfer ports.
2.5.3 Spent fuel transfer and storage
The spent fuel transfer and storage system handles spent fuel from the time it is discharged from the FM to the time it is removed from the
spent fuel bay.
Once the FM discharges spent fuel into the transfer tube, the transfer system uses recirculating water to push fuel through the pipe to the
transfer mechanism located in the spent fuel receiving bays. The water in the pipe also serves to cool the fuel during transit. The transfer
mechanism then unloads the fuel from its magazine into a mechanism that places the fuel into storage baskets.
The storage bay provides storage of spent fuel for typically 10 years of operation.
Spent fuel from the receiving bay is moved to the storage bay through a shielded tunnel that connects the two bays. Within the storage
bay, the spent fuel storage baskets are stacked in storage frames. The baskets are suitable for direct transfer to the longer‑term dry (spent)
fuel storage. Equipment, such as the storage bay bridge and handling tools, is provided to permit manipulation of spent fuel and
containers.

3.1 Overall safety concept and licensing approach
The objectives of safety design are to:
Take all reasonably practicable measures to prevent accidents in the nuclear power plant and to mitigate their consequences should
they occur;
Ensure with a high level of confidence that, for all postulated accidents considered in the design including those of very low
probability, radiological consequences would be below prescribed limits;
Ensure that the likelihood of accidents with serious radiological consequences is extremely low.
To ensure that the ACR‑1000 design is licensable in Canada and to facilitate its licensability in other countries, the following licensing
requirements have been applied in the design process:
Relevant Canadian nuclear safety requirements;
IAEA requirements as specified in the IAEA Safety Standards Series No. NS‑R‑1, “Safety of Nuclear Power Plants: Design - Safety
Requirements” (Reference [1]);
Environmental requirements for siting in Canada.
In addition, the ACR‑1000 design has taken into consideration requirements from other jurisdictions and can accommodate them with
limited changes to the reference design.
3.1.1 Defence-in-depth
The concept of defence-in-depth is applied throughout the ACR-1000 plant design to provide a series of levels of defence, including
measures to prevent accidents and measures to provide protection in the event that prevention fails. Consistent with the overall safety
concept of defence-in-depth, the design of the ACR-1000 plant aims to prevent, as far as practicable, challenges to the integrity of physical
barriers; failure of a barrier when challenged, and failure of a barrier as a consequence of the failure of another barrier. This approach is
structured in five levels, as presented below:
Level 1 – Prevention of abnormal operation and of failures of SSCs by conservative design and high quality in construction;
Level 2 – Detection and control of deviations from normal operation by controlling plant behaviour with both inherent and