Operational requirements and objectives for HSIs are typically derived from the need for cost-effective power production, nuclear safety, technical requirements imposed by plant processes, the level of automation, and the need to protect workers, the public and the environment. HFE work during the early phases of plant design should focus on how systems will be operated by the users, including interfaces and their interoperability with other systems. These requirements will establish how well and under what conditions the system must perform and will thus form part of the basis for HSI design and technology selection.

SMRs may provide additional features, levels of safety, or barriers in the DID

approach. Some SMRs have implemented one or more of the following DID-

supporting features:

• Containment with higher design pressure which may provide an additional barrier or at least extend the time to radioactivity release. This is facilitated by more compact reactor and containment design.

• Passive cooling of the containment, avoiding its failure in certain loss of offsite power (LOOP) scenarios.

• Fully underground placement, potentially ameliorating a release of radioactivity.

• Significantly increased grace period may also be viewed as DID since it may allow a more systematic and better organized evacuation in case all previous barriers fail.

SMRs may represent a viable option to decrease the average capital at risk in the nuclear business, with respect to LR projects. Financial risk is related to the amount of invested capital. Banks usually apply credit risk control through loans portfolio diversification. The same applies to the shareholder investor (e. g. a utility). Very high capital exposure in a single project represents a stress on the balance sheet and a relevant financial and industrial risk exposure, so that a nuclear generation project could be viewed as a ‘bet the farm’ endeavour for a shareholder utility, due to the size of the investment and the length of time needed to commission a nuclear power facility.

A model has been proposed for relating the risk premium to the risk size (Goldberg and Rosner, 2011). The assumption is that the risk premium associated with a project is a function of the wealth of the sponsoring entity, as might be measured by, for example, NPV and debt to equity ratio. The mathematical expression for this relationship shows risk premium rising at an exponential rate as the size of the project approaches the size of the investor-firm.

If the investment size in different base load technologies is compared with the average annual revenues of a utility (Figure 10.5), it becomes evident that SMRs should be viewed more favourably by the investor community and bear lower risk premium than very large reactors. (Examples of current annual revenues for some US utilities: Exelon — $23.5 billion, Duke Energy — $19.6 billion.)

A rating methodology reported in Table 10.5 shows that business diversification in low versus high risk (i. e. nuclear) businesses is among the risk metrics considered in the evaluation of the merit of credit of a company. In this sense, SMRs allow a better industrial risk diversification, on account of a limited investment on total capital budget. In case of small-sized market or reduced capital budget availability, by including SMR in a portfolio mix it is possible to grant a business diversification,

that would be pre-empted by a large plant, reducing the investment risk (Locatelli and Mancini, 2011a).

There are two key limitations to increased sizing of large reactors:

Handbook of Small Modular Nuclear Reactors. http://dx. doi. Org/10.1533/9780857098535.3.293

Copyright © 2015 Elsevier Ltd. All rights reserved.

600 MWe 1200 MWe

Power output

Figure 12.2 Illustrative plant cost vs. power output.

• Transmission and distributions networks limit the options for deploying large amounts of additional capacity from a single location.

• Manufacturing processes limit the size of key vessel and components.

Existing grid capacity has grown from fossil-fired generation sources and has defined the grid structures. Increasing the size of single nuclear generating capacity past this point will incur transmission infrastructure upgrade costs or will potentially entail running the distribution network with increased line losses.

Current manufacturing capacity and capabilities are challenged to deliver ever larger vessels. The global manufacturing supply chain is already at the edge of the envelope to produce components of the physical dimensions required by plants today (e. g. AP1000, CPR1000, EPR and VVER).

With these two constraints the nuclear power generation portfolio has previously
been unable to challenge the economies of scale that the increasingly larger installations provide. It is against this backdrop that the small reactor can play to a different set of strengths, flipping the economics of scale and delivering on economics of volume. With the opportunity to deliver physically smaller components at volume the supply chain paradigm changes. For example, if the supply chain can be restructured with elements that closer resembled other engineered commodities from the aerospace, automotive, or even white goods sectors it could deliver nuclear power at a significantly lower cost. This is the key attribute that has to be understood before supply chain and manufacturing options can be established.

As with any large project, attention must be given to society values, such as an appetite for renewable energy, concerns for nuclear materials and waste management, proliferation concerns, and general safety issues. Public acceptance is influenced by perceived positive (or negative) impact of a new facility, industry, or system on the local community. This impact could be quantified in terms of the number of jobs created and sustained by the proposed project, or the annual gross domestic product increase attributed to the project. These and other community economic considerations have an important effect on public acceptance and support for a proposed project, related political decisions, zoning, and facility siting. SMRs offer the possibility to be located closer to population centers and the final customer due to the reduced exclusion zone that derives from the smaller plant footprint. The significantly enhanced safety and reduced source term for integral PWR designs relative to existing gigawatt-scale LWR plants are also expected to allow siting closer to the end user. By producing non-electricity products (heat, chemicals, etc.) near the point of use, the economical attractiveness of the planned facility is increased and the market size is enlarged (particularly as aging coal plants require replacement).

NHES present a unique acceptance challenge in regulatory space. Co-location of nuclear and chemical subsystems that are subject to different licensing and regulatory bodies is new territory for regulators. Siting and operation of a nuclear plant require adherence to a cadre of regulations that define the necessary plant exclusion zone, the minimum number of operators per reactor, etc. Chemical plants are similarly regulated with regard to exclusion zone, emissions, etc. If implemented in their current form, the independent nuclear and chemical plant regulations would prohibit siting the plants within a single energy park facility. Joint permitting is essential to a tightly coupled hybrid facility, but it could also be a significant factor in gaining public and political approval. These challenges may entail development of a new licensing process with the Nuclear Regulatory Commission (NRC) and the appropriate chemical industry regulatory bodies to enact appropriate risk assessments that have not previously been considered. An interim solution could be independent siting of each subsystem with transmission lines crossing the site boundaries. However, this configuration would not offer significant improvement over the current hybrid grid configuration, as it would not allow for tight coupling and integrated control of the various subsystems.

The NuScale design (45 MWe per module) employs natural circulation in its primary system where the core heats the water, causing it to rise through a central hot leg where it then enters the upper plenum and is cooled by a helical coil steam generator. Upon cooling it then proceeds down a cold leg downcomer where it is returned to the core. Because of this unique design, NuScale has developed an integral systems test facility, the NuScale integral scaled test (NIST) facility [14, 15] to obtain data on its natural circulation design feature. Figure 14.8 is an illustration of the NIST facility.

The NIST was built on the Oregon State University (OSU) campus and originally developed and constructed in the 2000-2003 timeframe by OSU, Idaho National Laboratory, and NEXANT-Bechtel. NuScale has leased the facility since 2008 and has made a number of proprietary modifications to the NIST configuration to match the current NuScale design. The objective of the testing at NIST to generate well-scaled thermal-hydraulic data for system characterization and safety code validation while also providing a platform for the design and test of plant operational procedures, control logic schemes and to inform safety methodology development. Additional testing is to include LOCA, non-LOCA, flow stability, and startup testing.

D. F Delmastro

Centro Atomico Bariloche (CNEA), San Carlos de Bariloche, Argentina

16.1 Introduction

About 50 years ago Argentina decided to use a natural uranium fuel cycle for electricity production using nuclear power plants (NPPs). This decision was aimed to allow control of the whole fuel cycle. Small and medium reactors based on heavy-water reactors (HWRs) were installed and many facilities were constructed for uranium conversion, special alloy and Zircaloy tube production, fuel element production and heavy-water production.

Since the 1950s Argentina has developed and constructed research reactors. This activity was very successful and research reactors were exported to Peru, Algeria, Egypt and Australia.

In the early 1980s Argentina developed uranium enrichment capabilities using the gas diffusion method. This achievement facilitates the use of enriched uranium in local NPPs and the Central Argentina de Elementos Modulores (CAREM) project started. This project consists of the development, design and construction of small NPPs based on integrated pressurized water reactors (iPWRs). First, a prototype of an electrical output of about 27 MW, CAREM 25, will be constructed in order to validate the innovation of the CAREM concept and then develop a commercial version. After several years of development the CAREM project had reached such a level that the Argentine government decided to construct a CAREM prototype. Several activities have been performed, allowing the start of the CAREM prototype construction in February 2014. Currently the CAREM project is the main activity in small modular nuclear reactor research in Argentina. These activities are briefly described in this chapter.

4.4.1 Fuel designs in the smaller cores

From a nuclear design perspective, the first iPWR design specific that needs to be considered is the compactness of the core compared with large PWRs, both in terms of the number of fuel assemblies, and in some cases, the height of the fuel assemblies. Table 4.1 provides a summary of the key nuclear design parameters for the four US iPWR designs submitted in 2011 to the US Department of Energy (DOE) first solicitation, along with a modern large PWR, the Westinghouse AP1000.2 It is important to note that although the fuel heights, and overall loadings are different in each of the iPWRs, the 17 X 17 array of fuel pins is consistently the same throughout (Figure 4.6). This is primarily driven by the excellent operational performance of these fuel types in large PWRs operating today, and in particular it builds on the extensive development work that has evolved fuel designs towards a larger number of thinner fuel pins in order to improve thermal margins and allow higher rod powers. These are equally important drivers in the deployment of iPWRs, hence the choice of 17 X 17 fuels. The lower linear heat ratings in the case of the NuScale and mPower iPWRs in particular will reduce fuel and cladding temperatures during operations and accidents, as well as allow a higher relative power for the lead fuel rods during normal operations and therefore the ability to tolerate more power peaking in the core.

As shown in Figure 4.4, a large PWR has anywhere from 157 to 193 fuel assemblies compared with iPWRs which typically have between 37 (SMR-160)3 and 89 (Westinghouse SMR).4 It may appear that having fewer fuel assemblies make the design of the core and the loading pattern more straightforward. But in fact, fewer assemblies means fewer degrees of freedom to optimize the loading pattern and smooth out the reactivity and resulting power distribution across the core. The degrees of freedom are not just in terms of the fuel design and BP loadings of the fresh fuel, but once beyond the first cycle of operation, the batches of previously irradiated fuel will have a variety of burnups and hence resulting reactivities that can be distributed throughout the core to smooth the power distribution to achieve the power peaking limits.

The smaller core size also results in more neutron leakage radially (and axially, as described later in this section), and results in a more significant variation in assembly power across the core from the center where there is more fuel and hence more power, to the outside of the core where lower power is generated. The radial

Table 4.1 Summary of key nuclear design parameters for a modern, large PWR (AP1000) and a range of iPWRs iPWR name (vendor) AP10002 (Westinghouse) Westinghouse-SMR4,5 (Westinghouse) mPower6 (Babcock and Wilcox) SMR-16037 (Holtec) NuScale15 (NuScale Power) Power Thermal (MWth) 3400 800 530 500 160 Electrical (MWe) 1150 225 180 160 45 Reactivity control Control rods Control rods Control rods Control rods Soluble boron Soluble boron Control rods Soluble boron Fuel in core No. of assemblies 157 89 69 37 37 Array 17 X 17 17 X 17 17 X 17 17 X 17 17 X 17 Active fuel height 4.3 m/14 ft 2.4 m/8 ft 2.4 m/8 ft 3.7 m/12 ft 2.0 m/6.5 ft Mass in core (MTHM) 85 27 20 15 9 Fissile enrichment <5 wt% U235 <5 wt% U235 <5 wt% U235 <5 wt% U235 <5 wt% U235 Cycle length Month 18 24 48 48 24 Fuel demand MTHM per reload 36.6 20 14.7 MTHM per GW year 21.2 Unknown 29.3 33.8 Unknown Linear rating kW/m 19 14 12 14 8 kW/ft 5.8 4.2 3.6 4.2 2.4 *All values are approximate.

leakage can be reduced by ensuring only the fuel with lower reactivities (generally higher burnups) are loaded on the periphery of the core, or by developing and using radial reflectors in the outer core such as using stainless steel, rather than water, as used in the large PWRs. Although non-standard in large PWRs, there have been designs (for example, the AREVA EPR), where stainless-steel reflectors have been developed and demonstrated to be effective. Additional development work on radial reflectors for iPWRs is likely to yield notable benefits, not only in the economics of the fuel, but also by raising the power in the outer assemblies and hence improve the radial power variations across the core.

Neutron leakage is further exacerbated by the use of short fuel in some iPWRs. Typical large PWR fuel assemblies have between 12 and 14 feet (3.66 and 4.27 m) of fuel in 264 fuel pins, whereas several of the iPWRs have moved to a partial height version of the standard 17 X 17 fuel designs, principally because full fuel heights are not required for the lower-powered reactor designs. In each case, the fuel designs have been developed and optimized for the specific requirements and demands of each iPWR, e. g., rating and natural convection. For example, the Westinghouse SMR and mPower iPWRs has an active height of approximately 8 feet (2.4 m), and the NuScale design is approximately 6.5 feet (2.0 m); see Table 4.1 and Figure 4.6 for a relative comparison of fuel heights.

However, there is extensive experience in large PWRs to reduce the axial leakage by using what are known as ‘axial reflectors’. These are simply sections of the active fuel height (typically a few inches at the top and bottom of the fuel stack) where the usual fissile enrichment is reduced or in some cases, natural (0.71 wt% U-235) or even tails (~0.3 wt% U-235, a by-product of the enrichment process) uranium is used. This not only improves the fuel economy by reducing the neutron leakage from the core, but also reduces the costs of the fuel in terms of enrichment needs in an area of the fuel and core where there is relatively little power generated, and therefore little need to enrich the fuel.

However, axial variation in the fuel stack tends to increase the cost of the fuel, not in terms of enrichment or ore costs, but simply because of the complexity of manufacturing a fuel type with more enrichments, zones and diversity; the manufacture and in particular the quality assurance (QA) and quality control (QC) cost components are increased.

By way of illustration, based on current market prices for uranium fuels, if an increase of 0.25 wt% U-235 was required in a reload of fuel to offset either neutron leakage, or inefficient use of the fuel (such as lower burnups achieved), the increase in fuel price would be of the order of a few million dollars per reload. Cost increases because of additional fuel complexity would be the same order of magntiude.

Flow transmitters for NSSS purposes will follow much the same process as the safety system flow devices, only without the safety related qualification program. The devices will have to be accurate and reliable, but will not have the burden of a class 1E safety qualification.

The types of processes that will need measurement in the NSSS iPWR environment are:

• feedwater flow;

• steam flow;

• boric acid flow;

• reactor make-up water flow;

• reactor coolant pump (RCP) seal flow for models with pumps;

• let-down flow;

• charging flow.

In the NSSS controls environment, accuracy will be most important, especially for feedwater flow. Feedwater flow is an important parameter in the calculation of the calorimetric, a calculation from which the reactor power is assessed. The more accurate the instruments, the more megawatts the plant can produce. Also the accuracy afforded by digital control systems will further emphasize the need for accurate measurement devices to mate with the digital processing.

As some of these flow measurements will need to take place close to the reactor vessel, the devices may have to be more radiation, temperature, and pressure hardy than traditional NSSS sensors. This need may drive designers to newer technology, such as the ceramic-based MEMs sensor mentioned in the safety system discussion. Different variations of transit time methods and temperature-based flow methods may find their foothold here with NSSS-based applications. Further discussion on new technologies for flow measurement can be found in Section 6.2.5.

HSIs that employ vision as interaction modality are the most common devices to present information to the operator. This mode of interaction is uni-directional only (device to observer). New developments use cameras and sensors to detect user gaze and motion and use this to create an interactive dimension for displays. Examples are already found in entertainment-related devices like Microsoft’s Xbox® Kinect, Nintendo® Wii. Sony PlayStation® Move and the Leap Motion® gesture controllers. Other devices use gaze detection to determine where the user is looking and use this data to display contextual information or to enable users to navigate through the system by gaze only.

While standard desktop flat screen displays will continue to be the most common means to display information for everyday use, a variety of advanced visual devices are becoming attractive options for textual, graphical and video information. Three types can be identified, ranging from large screen displays that are already available, to more sophisticated technologies that will be available in consumer markets within the next few years.