Plant control and monitoring functions are performed in two main areas: MCR and LCSs.

7.7.1.1 MCR

A control room is generally understood to be the nerve centre of the plant and it often forms part of a control centre that could also house some of the operational domains shown in Figure 7.2, for example, the TSC and the outage control centre. The MCR in older nuclear power plants is dedicated to the control of a single unit; new plant designs, and especially SMRs with their compact footprint, are more likely to have a single MCR for multiple modules (see ‘Multi-module control rooms’ below). Some emerging designs envisage a single control room for up to 12 modules. This kind of control room will be larger than today’s control rooms for a single unit, but owing to the level of integration and automation, they may actually reduce the complexity of the overall instrumentation and control architecture by allowing common systems to share a single operator console (O’Hara et al., 2008).

It is normally assumed that a central control room is necessary as part of a strategy to rationalise plant operations, to minimise duplication of equipment and to optimise the capability of automation systems. Central control rooms for modern plants are also considered to enhance communication between units, enable better coordination of plant-wide operations and maintenance, and provide a more effective response to upsets.

As indicated earlier, an important consideration for new designs would be the location of the control centre. Traditionally, the MCR is located somewhere on the ‘nuclear island’, which normally consists of the containment, including reactor, steam generator and primary coolant circuits. Because the nuclear island is seismically qualified and provides back-up systems like electrical supply and heating, ventilation and air conditioning (HVAC), this was typically the choice for the location of the control room. In fact, this is still one of the strictest criteria for control room habitability described in NUREG-0800 (2007); the NRC’s Standard Review Plan.

As mentioned before, it is assumed that new reactors will use more passive safety designs like negative temperature coefficient of reactivity, natural circulation of coolant, or less need for active controls and fewer active protection systems like forced cooling circuits. Designers should now determine if requirements for seismic qualified control systems and HSIs will change and if this means that the control room need not be on the nuclear island. Designers should also determine if the availability and reliability of wireless technology and fibre optics is sufficient justification for having the MCR remote from the reactors. Other important considerations would be if the need for operator response to certain events would still force location of the control room to be near the reactor. Due to the current strict NRC regulations (see NUREG-0696, 1981), proving these new concepts of operation is likely to be an importance challenge for designers.

Issues related to seismic events would warrant a separate chapter. Instead, we point to only one specific SMR feature. Because of their compact design, it is technically and economically feasible for most SMRs to place the nuclear island on seismic isolators and thus limit the impact of seismic events, and significantly improve the PRA indicators. This in turn could further support the reduced EPZ.

When a fleet of multiple NPP units are considered, some competitive factors intervene, to reduce the incidence of capital cost on the electricity generated. These factors are enabled and provide their best effect by the deployment of successive NPP of the same type on the same site. These factors are introduced in this chapter, despite not being specific to the SMR plant category, because they are expected to play a relevant role in the SMR economic competitiveness paradigm.

The ‘flowline principle’ is a process that could be used for the manufacture of small reactors. The flowline approach fits in between batch and assembly line manufacturing as shown in Figure 3. The history of nuclear manufacturing has evolved from a base of bespoke component manufacturing. This manufacturing base cannot support the volumes projected for small reactors. A realistic alternative is the development of a hybrid manufacturing method that sits between batch production and production line. This hybrid is termed flowline and has been successfully applied in a number of high-technology, high-integrity, intermediate-volume businesses.

The flowline is an approach where the unit in work is moved in a number of discrete steps. Each step is called a station and each station has a number of predefined steps to be performed at each point. The tooling for each activity is held at each station, eliminating the need for changes in set-up. The work piece is moved from station to station when the work is complete. The work performed at each station is planned and timed so that each station completes at the same time and then the line can increment along. For the small reactor this technique is ideal. The flowline can be scaled up and down to meet the market growth. Figure 12.4 shows how the activities can be grouped at the different stations to optimise the step time across all four stations.

This growth scaling can be achieved by changing the number of steps that are

Increasing facility cost

Increasing production volumes

performed in each workstation. If the number of stations is one then the flowline is operating as a bespoke manufacturing cell, at the other extreme if the number of stations were infinite then the flowline would be operating like a production line.

This attribute of scalable production capability is important in delivering the mature nth-of-a-kind (NOAK) costs. For a small reactor the first-of-a-kind (FOAK) to NOAK cost transition is important, where customer commitment will be based against perceived costings. It is important that there is an achievable path to reduce
the costs between the initial and mature production units. The flowline can be seen to deliver these cost reduction opportunities in a scalable manner, as shown in Figure 12.5.

To achieve an increase in production volumes the number of stations on the flowline can be increased and the number of activities performed at each station can be reduced. This is a unique attribute of the flowline that enables it to be scaled up as the market demand grows. For example, say that the initial units are built on an eight-station flowline with 40 operations at each station. If the number of stations was increased to ten the number of operations at each station can be reduced to 32. The underlying assumption in this hypothetical example is that each station already has the maximum number of activities at each station.

The deployment of the flowline concept has been shown to be a well-established process, proven to be capable of replication in different geographical locations and industrial segments. It is suited for complex assemblies with an intermediate production volume. The flowline concept adopts a single-track flow system to the assembly of a product, using individual workstations or process steps to progress the assembly procedure. The step flow moves down the line are defined by the term Takt (cycle) times. Each activity within each station is mapped in terms of its Takt time. To ensure the greatest level of repeatability and predictability from the final product at each step minimum inspection is preferred. With the incorporation of ‘design for assembly’ processes a lot of inspection can be eliminated. This is part of the product simplification and standardisation activities. There is a bias toward automated online validation and testing of components or assemblies during the assembly process.

Certain key features are essential to the success of any flowline system. Single product family, no rework, stable supply of parts/assemblies/kits and a high reliability of workstation equipment, tools, etc. With a single product family there is a consistent approach to the work that allows for standardised work instructions to be created and scheduled for specific stations. This is turn requires a stable parts supply. Units

are assembled in a uniform and predictable manner. Kits of parts can be ‘kitted’ away from the flowline, even at an alternative location. These kits will contain all the parts for a specific assembly operation. For example for the fitting of a sensor, the sensor, the gland and seals will all be supplied as a kit along with the correct fasteners. This enables the assembly team to verify that they have all the materials to perform the installation of that component prior to commencing each step. The kitting and delivery of components onto this flowline is therefore also stable and becomes an assured process. With this stability comes to opportunity to pre-kit and collate parts in the build sequence prior to incorporation in the final module build. The final aspect is tooling stability. With known and repeatable assembly operations being performed consistently at a specific work station there is a reduction in the opportunity for errors from misplaced or misused tooling and fixtures.

Design of any flowline is heavily reliant upon the product design and its suitability to conform to flowline criteria. The application and use of process and design failure mode and effect analysis (FMEA) tools during the conceptual design of the product is fundamental to delivering a robust flowline system. As an example these manufacturing ideas have been shown to work well, incorporating simple things like baulking features in the design to ensure that parts are orientated correctly to each other. As an alternative example assembly operations can be grouped such that all fastenings of a specific size are fitted at one assembly station. This reduces the potential assembly errors. The adoption of poka-yoke (fail-safeing) of parts, the utilisation of common tools will also contribute to the success of any flowline system.

Consideration of the facility layout is of paramount importance, while employing the use of lean tools to design a fully efficient flowline capability. This capability will take into account the on-time delivery of sub-modules from suppliers and vendors into the assembly facility.

The storage of inventory in such a facility is not a feature of the efficient flowline philosophy. Therefore delivery of parts, sub-modules and kits direct to the assembly facility workstations is considered essential to support the pure implementation of this philosophy. However, industrial experience shows that an initial ‘buffer’ stock of critical parts is required as a new flowline is established. One option is to pre-kit parts for assembly away from the flowline and sequence them for assembly. The level of buffer inventory held may be reduced and optimized as the process finds its natural rhythm. The rhythm of the flowline is referred to as its drumbeat. The drumbeat does not change. If the volume through the line needs to increase, one of the options is to further sub-divide the operations performed at each station.

Delivery of parts from the pre-kitting can be arranged to use reusable trays or standardised trolleys. These trays and trolleys act as ‘shadow boards’, allowing the flowline personnel easily and quickly to validate that all the parts for their operations are present at the outset of their build sequence for their flowline workstation. Similarly they can confirm visually that all the parts for that stage of the build sequence have been incorporated. These techniques are not specific to a nuclear build and have been validated in other industrial sectors. With the historically lower build rates from large nuclear plants there has been little need or incentive to incorporate these techniques. For the first time in the nuclear sector the small reactor presents an opportunity to adopt and implement these lean manufacturing techniques. There are significant consequences for a production capability built on this philosophy. The rigor of supply of sub-assemblies and modules flows down to the sub-tier vendors. The supply of parts onto the flow line is important. This discipline cascades down the supply chain. In this respect the small reactor may not be a revival for existing nuclear suppliers, but may be demanding that the industry searches for new suppliers with alternative manufacturing capacity.

With a higher level of process automation incorporated in the assembly activity, attention is also needed to understand the opportunities afforded for automation in the manufacturing control. These technologies have not been deployed in a nuclear context previously with the smaller volumes of manufacture. With the larger volumes of product the investment to support the deployment of these techniques can be considered. The use of RFID (radio-frequency identification) tags provides a semi-automated, real-time monitoring capability of the location of parts during the assembly process. This technique is already used within the aerospace industry where the incorporation of parts are tracked, and can readily be applied in the small reactor product. There are opportunities to include additional build control techniques from other industries too; For example the use of a high-resolution local global positioning system (GPS) network within the factory can be combined with handheld scanners and some direct part marking will allow components to be scanned as they are incorporated onto the assembly. The build control database can record the incorporation of a specific part and serial numbered component. This concept can be extended to include the tooling. For example a torque wrench could contain a GPS device that automatically records the torque setting applied to a specific bolt at a specific location. Techniques such as these are not specific to nuclear applications, it is the repeatable build sequence of a volume of units that makes off-the-shelf techniques like this attractive to small reactor assembly.

In other manufacturing sectors around the globe there are already facilities employing the flowline approach, history can identify equivalent systems applied to large assemblies, not dissimilar in complexity and physical size to the small reactor product: For example, during the period of 1941-1945, the USA employed an optimised assembly process to produce 2751 Liberty ships at 18 different shipyards — each ship was of common design (prefabricated sections) and helped the allies win the war. The first few vessels took around 230 days to build but, based on experience and the application of continuous improvement philosophy, the average eventually dropped to 42 days.

The introduction of the flowline system has realised output efficiency improvements, including:

• reduced inventory holding;

• clear responsibilities and accountabilities are defined for each activity station;

• the provision of control and sequencing of activities;

• a defined level of inherent quality assurance.

Other simple features include:

• materials handling is positioned to avoid obstructing the walking path of the operator;

• flow racks are filled and delivered away from the flowline to avoid interrupting operator work cycles as parts are replenished;

• parts are correctly orientated for easy operation, allowing operators to use both hands

simultaneously;

• parts containers are sized for ease of use by the operators, not the material handler or the supplying process.

In the design of the flowline each of the stations are timed to a common duration. The process steps and the staffing levels are then optimised around each of these steps. Experience from industry shows that 20 operations or 10 operators in each station is the sustainable limit.

Increasing penetration of renewable sources requires rapid adaptation of the current energy infrastructure. Renewable energy comprises a heterogeneous set of options, but for illustrative purposes, this discussion focuses on wind and solar energy. These resources are considered mature and are being deployed at a significant rate in the US and Europe. The National Renewable Energy Laboratory (NREL) maintains user interactive prospectors for wind and solar energy opportunities [9]. The INL has similarly developed a virtual prospector tool for water resources, supporting potential future hydropower projects [10].

The NREL Wind Prospector Tool [11] can be used to identify regions having significant wind generation potential in the US, with the greatest potential in the Midwest and coastal regions. Similarly, the NREL Solar Prospector Tool [12] can be used to identify regions in the US having the greatest solar energy potential, with the Southwest and Western regions of the US having the highest solar intensity. Some of these regions have already introduced significant wind and solar capacity (see refs [13,14]). Incorporation of large amounts of wind and solar energy into a utility system has required load following or load leveling capability to utilize these variable resources when they are available. Based on the output of the various projects, the average utilization factors for wind and solar are approximately 30 and 20 percent, respectively [15]. While there is some cause and effect relationship between solar availability and demand (at least for hot, sunny days where the peak solar insolation coincides with the daily and summer peaks), there is no similar cause and effect relationship between wind availability and electricity demand.

Oak Ridge National Laboratory (ORNL) has developed a tool for evaluating siting options for new electrical generation using Geographic Information System (GIS) data sources and spatial modeling capabilities. Referred to as OR-SAGE (Oak Ridge Siting Analysis for Power Generation Expansion), the objective of this tool was to use industry-accepted parameters for screening potential reactor sites and to apply GIS data sources and spatial modeling capabilities to evaluate suitability of those sites [16]. The tool considers various input data, such as proximity of electrical transmission lines, population density, regional seismicity, water sources, protected lands, ground slope, hazards (landslide, flood plain), and proximity of external hazardous operations. The tool then establishes criteria for each parameter to assess the feasibility of siting a large or small reactor in a specific location. This method allows for broad regional siting studies to narrow down to regions of interest, candidate areas, candidate sites and finally to a preferred site. Figure 13.2

shows results from OR-SAGE identifying potential sites for large (~1600 MWe) and small (~350 MWe) reactors (new capacity). When overlaid with renewable resource availability, possible sites for a hybrid nuclear-renewable plant become evident.

KAERI started developing SMART in 1997, aiming to export it to countries with small electric grids and water supply issues. For the last 15 years, SMART-specific design methodologies and a computer code system have been built together through a series of validation experiments and equipment verifications. From 2009 to 2012, the SMART Technology Validation and Standard Design Approval Project was carried out. As a result, a full set of licensing documents, including a standard safety analysis report (SSAR), were submitted to the Nuclear Safety and Security Commission (NSSC) at the end of December 2010. After one and half years of intensive licensing review, the SDA (standard design approval) for SMART was officially issued on July 4, 2012, by the NSSC, in compliance with Article 12 of the Nuclear Safety Act.

This may be the first license for an integral reactor in the world. The development of SMART took over 275 million USD and 1500 R&D man-years.

16.3.1 CAREM 25 design

CAREM 25 is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a higher level of safety (Boado Magan et al., 2011):

• integrated primary cooling system;

• primary cooling by natural circulation;

• self-pressurization;

• safety systems relying on passive features.

The CAREM 25 reactor pressure vessel (RPV) contains the core, the steam generators (SGs), the whole primary coolant and the absorber rods drive mechanisms. The RPV diameter is about 3.2 m and the overall length is about 11 m (Delmastro et al., 2002).

The core of the prototype has 61 hexagonal cross-section fuel assemblies (FA) having about 1.4 m active length. Each fuel assembly contains 108 fuel rods, 18 guide thimbles and an instrumentation thimble. Its components are typical of the PWR fuel assemblies. The fuel is enriched UO2. Core reactivity is controlled by the use of Gd2O3 as burnable poison in specific fuel rods and movable absorbing elements (AEs) belonging to the adjust and control system. Chemical compounds are not used in the water for reactivity control during normal operation. The fuel cycle can be tailored to customer requirements, with a reference design of 390 full-power days and 50% of core replacement.

Each AE consists of a cluster of rods linked by a structural element, so the whole cluster moves as a single unit. Absorber rods fit into the guide thimbles. The absorbent material is the commonly used Ag-In-Cd alloy. The AEs are used for reactivity control during normal operation (adjust and control system), and to produce a sudden interruption of the nuclear chain reaction when required (fast shutdown system).

Twelve identical ‘mini-helical’ vertical SG, of the ‘once-through’ type are placed equally distant from each other along the inner surface of the RPV. They are used to transfer heat from the primary to the secondary circuit, producing dry steam at 47 bar, with a minimum 30 °C of superheating. The location of the SG above the core produces natural circulation in the primary circuit. The secondary system circulates upwards within the tubes, while the primary goes in a counter-current flow. An external shell surrounding the outer coil layer and adequate seal form the flow separation system. It guarantees that the entire stream of the primary system flows through the SG. In order to achieve a rather uniform pressure-loss and superheating on the secondary side, the length of all tubes is equalized by changing the number of tubes per coil layer. Thus, the outer coil layers will hold a larger number of tubes than the inner ones. For safety reasons, SG are designed to withstand the primary pressure without pressure in the secondary side and the whole live steam system is designed to withstand primary pressure up to isolation valves (including the steam outlet/water inlet headers) for the case of an SG tube brake.

The natural circulation of the coolant produces different flow rates in the primary system according to the power generated (and removed). Under different power transients a self-correcting response in the flow rate is obtained.

Due to the self-pressurizing of the RPV (steam dome) the system keeps the pressure very close to the saturation pressure. At all the operating conditions this has proved to be sufficient to guarantee a remarkable stability of the RPV pressure response. The control system is capable of keeping the reactor pressure practically at the operating set point through different transients, even in case of power ramps. The negative reactivity feedback coefficients and the large water inventory of the primary circuit combined with the self-pressurization features make this behavior possible with minimum control rod motion. It concludes that the reactor has excellent behavioural responses under operational transients.

Nuclear safety has been incorporated in CAREM 25 from the start of the design. The defense-in-depth concept has specially been considered. Many intrinsic characteristics contribute to the avoidance or mitigation of eventual accidents.

CAREM 25 safety systems are based on passive features and must guarantee no need of active actions to mitigate the accidents during a long period. They are duplicated to fulfill the redundancy criteria. The shutdown system should be diversified to fulfill regulatory requirements.

The First Shutdown System (FSS) is designed to shut down the core when an abnormality or a deviation from normal situations occurs, and to maintain the core sub-critical during all shutdown states. This function is achieved by dropping a total of 25 neutron-absorbing elements into the core by the action of gravity. Each neutron-absorbing element is a cluster composed of a maximum of 18 individual rods which are together in a single unit. Each unit fits well into guide thimbles of each FA.

Internal hydraulic control rod drives (CRDs) avoid the use of mechanical shafts passing through RPV, or the extension of the primary pressure boundary, and thus eliminates any possibilities of big loss-of-coolant accidents (LOCA) since the whole device is located inside the RPV. Their design is an important development in the CAREM concept. Nine out of 25 CRD are the fast shutdown system. During normal operation they are kept in the upper position, where the piston partially closes the outlet orifice and reduces the water flow to a leakage. The CRD of the adjust and control system is a hinged device, controlled in steps fixed in position by pulses over a base flow, designed to guarantee that each pulse will produce only one step.

Both types of device perform the SCRAM function by the same principle: ‘rod drops by gravity when flow is interrupted’, so malfunction of any powered part of the hydraulic circuit (i. e. valve or pump failures) will cause the immediate shutdown of the reactor. CRDs of the fast shutdown system are designed using a large gap between piston and cylinder in order to obtain a minimum dropping time thus taking few seconds to insert absorbing rods completely inside the core. For the adjust and control system CRD manufacturing and assembling allowances are stricter and clearances are narrower, but there is no stringent requirement on dropping time.

The second shutdown system is a gravity-driven injection device of borated water at high pressure. It actuates automatically when the reactor protection system detects the failure of the FSS. The system consists of two tanks located in the upper part of the containment. Each of them is connected to the reactor vessel by two piping lines: one from the steam dome to the upper part of the tank, and the other from a position below the reactor water level to the lower part of the tank. When the system is triggered, the valves open automatically and the borated water drains into the primary system by gravity. The discharge of a single tank produces the complete shutdown of the reactor.

The residual heat removal system has been designed to reduce the pressure on the primary system and to remove the decay heat in case of loss of heat sink. It is a simple and reliable system that uses condensing steam from the primary system in emergency condensers. The emergency condensers are heat exchangers consisting of an arrangement of parallel horizontal U-tubes between two common headers. The top header is connected to the reactor vessel steam dome, while the lower header is connected to the reactor vessel at a position below the reactor water level. The condensers are located in a pool filled with cold water inside the containment building. The inlet valves in the steam line are always open, while the outlet valves are normally closed, therefore the tube bundles are filled with condensate. When the system is triggered, the outlet valves open automatically. The water drains from the tubes and steam from the primary system enters the tube bundles and is condensed on the cold surface of the tubes. The condensate is returned to the reactor vessel, forming a natural circulation circuit. In this way, heat is removed from the reactor coolant. During the condensation process the heat is transferred to the water of the pool by a boiling process. This evaporated water is then condensed in the suppression pool of the containment.

The emergency injection system prevents core exposure in case of a LOCA. In the event of such accident, the primary system is depressurised with the help of the emergency condensers to less than 20 bar, with the water level over the top of the core. At 20 bar a low pressure water injection system comes into operation. The system consists of two tanks with borated water connected to the RPV. The tanks are pressurized, thus when during a LOCA the pressure in the reactor vessel reaches 20 bar, the rupture disks break and the flooding of the RPV starts.

Safety relief valves protect the integrity of the reactor pressure vessel against overpressure, in case of strong imbalances between the core power and the power removed from the RPV. Each valve is capable of producing 100% of the necessary relief. The blow-down pipes from the safety valves are routed to the suppression pool.

The primary system, the reactor coolant pressure boundary, safety systems and high-pressure components of the reactor auxiliary systems are enclosed in the primary containment — a cylindrical concrete structure with an embedded steel liner. The primary containment is of pressure-suppression type with two major compartments: a drywell and wetwell. The lower part of wetwell volume is filled with water that works as the condensation pool, and the upper part is a gas compression chamber.

The building surrounding the containment has been designed in several levels. It supports structures with the same seismic classification, allowing the integration of the RPV, the safety and reactor auxiliary systems, the spent fuels pool and other related systems in one block (Figure 16.1). CAREM 25 NPP has a standard steam cycle of simple design.

Technical and economical advantages are obtained with the CAREM 25 design compared with the traditional design:

• Due to the absence of large-diameter piping associated to the primary system, no large LOCA has to be handled by the safety systems. The elimination of large LOCA considerably reduces the needs in emergency core cooling system (ECCS) components, AC supply systems, etc.

• Eliminating primary pumps and pressuriser results in added safety (loss of flow accident elimination), and advantages for maintenance and availability.

• Innovative hydraulic drive control rods avoid rod ejection accidents.

• A large coolant inventory in the primary results in large thermal inertia and long response time in case of transients or accidents.

• Passive safety systems with a long grace period are incorporated.

• Shielding requirements are reduced by the elimination of gamma sources of dispersed primary piping and parts.

• The large water volume between the core and the wall leads to a very low fast neutron dose over the RPV wall.

• The ergonomic design and layout make the maintenance easier. Maintenance activities like SG tube inspection does not compete with refueling activities because it will be carried out through 12 radial plena with individual removable blind flanges.

5.2.1 Pressure vessel and flange

In a current large PWR, the reactor pressure vessel holds the individual fuel assemblies, the control rods and a significant percentage of the reactor coolant, which also acts as the moderator. Functionally, the reactor pressure vessel provides one of several safety barriers to fission product release, provides support to the control rods, and provides support to the reactor vessel internals, which support the reactor fuel and direct coolant flow within the vessel. A PWR reactor pressure vessel is characteristically a cylindrical vessel with a hemispherical bottom head and a removable, flanged and gasketed, hemispherical top head (NRC, 2006). The bottom head is welded to the cylindrical shell while the top head is bolted to the cylindrical shell via the flanges. The vessel is nominally constructed of low-alloy carbon steel clad on the inside with a thin layer of austenitic stainless steel. The reactor pressure vessel cylinder is typically made up of a number of thick-walled ring forgings that are welded together circumferentially. The circumferential weld nearest the fuel region of the vessel is typically referred to as the beltline region. A typical PWR pressure vessel is shown in Figure 5.2.

Current large PWR reactor pressure vessel concerns include vessel embrittlement, pressurized thermal shock, and primary water stress corrosion cracking (NRC, 2003). Neutron embrittlement occurs over time as the vessel is exposed to neutron flux. The area of greatest concern for embrittlement is the vessel beltline weld (IAEA, 2006; Kang and Kupca, 2009). Pressurized thermal shock (PTS) can occur when cold water is introduced while the vessel is pressurized. This creates increased thermal stress in the vessel wall. As the vessel becomes brittle over time, it can become more susceptible to cracking, especially as a result of the added stresses induced by

PTS. Primary water stress corrosion cracking also has the potential to compromise reactor pressure vessel integrity in the vicinity of vessel head penetration nozzles. This could lead to vessel leakage or in more extreme cases an increased potential for control rod ejection. Primary water stress corrosion cracking is aggravated by the presence of boric acid.

Current large PWR reactor pressure vessel size will vary with the thermal power rating of the reactor. However, the ratio of cylindrical height to diameter in a large PWR reactor pressure vessel is nominally in the range of 2 to 2.5. The coolant design

pressure of current large PWRs is normally around 2500 psi (17.2 MPa). Vessel penetrations are typically below the fuel region in a PWR vessel for instrumentation. In addition, the large inlet and outlet nozzles for supplying reactor coolant to and from the vessel are characteristically just above the fuel region in the vessel.

As with the current large PWR designs, an iPWR reactor pressure vessel holds the individual fuel assemblies and the control rods. In addition, the iPWR reactor pressure vessel includes the pressurizer and the steam generator. With the exception of isolable support systems, such as the chemical and volume control system, virtually the entire reactor coolant inventory is contained within the iPWR pressure vessel. Functionally, an iPWR reactor pressure vessel performs the same roles as a current large PWR reactor pressure vessel. An iPWR pressure vessel flange is typically located above the top of the fuel and below the integral steam generator, which is significantly different from the pressure vessel head flange in a current PWR. This facilitates refueling in the tall iPWR pressure vessels and also facilitates steam generator inspection. Exceptions include the SMART iPWR, which does retain the reactor pressure vessel flange at the vessel head (Lee, 2010). The Holtec SMR-160 uses an offset steam generator which also allows the reactor pressure vessel flange to remain at the vessel head (Oneid, 2012). Note that the Holtec design is not a true iPWR because the steam generators are flanged externally to the reactor pressure vessel and are not internal to the vessel.

The fuel in an iPWR is usually half the height of current PWR fuel. As a result, the thick-walled ring forgings that make up the iPWR vessel could conceivably be stacked such that no weld exists adjacent to the fuel, thereby diminishing the vessel embrittlement concern. In addition, as can be seen in Figure 5.1, the riser section needs to be large enough in diameter to accommodate control rod motion. Consequently, the internal position of the steam generators in the iPWR designs forces the reactor pressure vessel wall to be farther away from the reactor fuel than it would be in a conventional PWR. Therefore, there will be added water shielding in the downcomer region of the iPWR reactor pressure vessel relative to a conventional PWR and subsequently lower fluence on the reactor pressure vessel. This would also act to diminish the vessel embrittlement concern. Integral PWR vessel materials will not differ from current PWR designs, so PTS will remain an operational concern for iPWR vessels. However, if the vessel fluence and the beltline weld embrittlement issue can be diminished in an iPWR vessel, then the overall iPWR operating window for PTS can likely be relaxed compared to a current large PWR. The international reactor innovative and secure (IRIS) design, with a reactor vessel height to width ratio on the lower end of the spectrum for iPWRs, has a very large downcomer water channel that virtually eliminates the embrittlement issue and a need for an embrittlement surveillance program over the design life of the reactor vessel (Carelli et al., 2004).

Additionally, some iPWR designs are planning for internal control rod drives and other iPWR designs are planning to exclude the use of boron chemical shim for normal operation. The presence of boric acid aggravates primary water stress corrosion; therefore, designs that do not use boron for normal operations will diminish primary water stress corrosion cracking concerns for iPWR pressure vessels.

An iPWR reactor pressure vessel is smaller in volume than a conventional large PWR reactor pressure vessel. However, the ratio of cylindrical height to diameter in an iPWR pressure vessel is approximately 4 to 7 (Kim, 2010; Memmott et al., 2012; NuScale, 2012) to promote natural circulation for emergency operations and for normal operations in some iPWR designs. Since almost the entire reactor coolant inventory is within the reactor pressure vessel and not distributed in loop piping and other components as it is in a conventional PWR design (Figure 5.3), natural circulation cooling is efficient and effective. The iPWR pressure vessel height to diameter ratio is much higher than the corresponding ratio for current PWRs at 2.0 to 2.5. The increased iPWR pressure vessel height to diameter ratio is achieved by

increasing the vessel height and decreasing the vessel diameter, which also facilitates truck delivery of a factory built iPWR pressure vessel. In addition, because the pressure vessel includes additional components that current large PWR vessels do not, the water volume relative to core thermal power is increased considerably. The coolant design pressure of an iPWR can be equivalent to, or slightly lower than, a current large PWR design.

Vessel penetrations in an iPWR pressure vessel are significantly different from the penetrations on a PWR pressure vessel. There are no penetrations below the top of the fuel in any iPWR design — instrumentation access is always above the fuel in the iPWR designs. In addition, because there are no external coolant loops, the large PWR inlet and outlet piping penetrations, typically 27.5 to 29 inches (70-79 cm) (NRC, 2006), are eliminated in the iPWR designs. This effectively removes any opportunity for a large-break loss-of-coolant accident (LOCA) in the iPWR designs. The maximum vessel penetration in the iPWR designs is much smaller, typically 2 inches (5 cm) or less, for iPWR reactor support systems. In addition, the smaller iPWR vessel penetrations are well above the top of the reactor fuel relative to the larger PWR designs. This allows a significant amount of water to continue to be available to cool the reactor following a small-break LOCA on an iPWR design.

Cabling is the glue that connects the I&C signals together. There is cabling to the transmitter, cabling from the transmitter to the rack electronics, and cabling from the rack electronics to the end device. Recent regulatory interest has resulted in a new field of diagnostic measurements to verify cable integrity. Some of these techniques advocate pro-active measures with sensors embedded in the cable itself; other methods require baseline tests for future comparisons, so cable health and monitoring must be considered up front in the design.

Periodic cable testing is likely to be a requirement for iPWRs. Most likely, the iPWRs will take a baseline set of tests and then retest at regular intervals and compare results. Some of the tests that are recommend today for I&C cable testing are:

• time domain reflectometry (TDR);

• reverse TDR;

• frequency domain reflectometry (FDR);

• insulation resistance (IR);

• inductance, capacitance, and impedance measurements;

• visual inspections;

• insulation hardness (indenter modulus);

• partial discharge, for the higher-voltage cables.

From these tests it is possible to determine the location of hard faults, degraded areas, moisture intrusion, insulation degradation, and poor cable connection locations. The location and the identification of degraded areas of cable are key factors in timing the replacement of cable sections in advance of failure.

In addition to cable testing, embedding sensors in the cable insulation to determine insulation aging and degradation is a new technology that may be incorporated in the manufacturing process of cables for the industry in the near future.

US NRC Regulatory Guide 1.21818 summarizes the current regulation on cable testing. This guide requires that all power plants have a plan/program for periodic testing and health/aging management of the plant’s cables. This guide covers I&C cables as well as medium voltage cables that support large pumps and valves. Some sections in NUREG/CR-7000 provide additional guidance on cable health and monitoring. These regulations will apply to iPWRs, and iPWRs should not miss the opportunity to take the appropriate measures to ensure proper cable health and aging management for the future.

Devices associated with cabling must also be considered in designing robust cabling systems. It is likely that cable and wiring connectors, as well as the containment penetrations will need new engineering design and development as the expected environments and geometries will be very different from those traditionally encountered.

Audio-based interaction between human and HSI is not new but it is now developing rapidly to provide more powerful and reliable means of obtaining information or performing control actions. Alarm sounds are the most common use of audio technology and most control rooms use coded and modulated sounds to enable operators to distinguish different conditions.

Speech recognition is also not a new technology and it has been applied in many systems with varying degrees of success. Speech recognition has never been used in control rooms, but this technology is also becoming more accurate and reliable. However, although important advances have been made, especially in the ability of such systems to recognise natural language, it remains one of the least reliable interaction methods. Research has shown that even in the best systems, recognition is typically subject to an error rate of 5-10% and with background noise it is even worse, with error rates of 20-40% (Shneiderman, 2000). This makes this kind of interface slow and unreliable and unlikely to be used in mission-critical applications. However, research continues and this may eventually become an option for ‘busy hands, busy eyes’ applications. Both speaker-independent and speaker-dependent speech recognition with background noise cancellation might become more viable options for certain types of control commands in future, especially for fieldwork and maintenance where hands-free operation is often desirable.

Another auditory device that may become important for voice communication, for example between the control room and field operators in the plant, especially in noisy areas, is bone conduction audio. This technology provides real-time information in ‘busy-hands busy-eyes’ tasks. It maintains sound clarity in very noisy environments because the eardrum is bypassed and sound is passed directly to inner ear. It is especially important where there is a need to enhance the presentation of written or graphical information and to notify users about a particular condition without the need for a display. It is also useful where coded audio signals may convey more information than a single alarm tone, or for operators with some hearing impairment.