Current NRC regulations were developed based on traditional large NPP LWR designs. Current requirements related to the human role in the plant deal primarily with avoiding human error and improving human reliability in normal and abnormal operational conditions. This includes requirements for control room staffing, criteria for evaluation of HSIs, and conducting HFE activities in the power plant. Some provision is currently made for new designs, for example for minimum staffing of the NPP, as described in the Code of Federal Regulations, 10 CFR 50.54. However, most new reactor designs, and particularly SMR designs, differ substantially from traditional designs in a number of aspects, including size and number of reactors, inherent passive safety systems, fuel type and coolant type. These differences present unique issues in terms of licensing and regulation.

Although current NRC guidance provides a general framework for conducting design-specific reviews, the review of control room and HSI designs as well as staffing plans and potential exemption requests is expected to be challenging for SMR designs. This is because of the differences between the new reactor designs and previously licensed reactor designs and also because of a lack of research and design data to provide an adequate technical basis for decisions. Initial evaluation conducted by the NRC has identified a number of differences between SMR and other advanced reactor design and operating philosophies and the designs for larger reactors currently licensed or being evaluated for licensing. These differences include the following:

• SMRs may require different operator tasks. The task requirements will include operating multiple units in different modes of operation. A major challenge will be to identify tasks that may be omitted and those that could substantially affect operator workload.

• Very limited operational experience will be available to use as a resource, especially if these designs are FOAK. The use and observation of simulator activities will be important to verify the task analyses and staffing plans. Parallels in other industries may be useful, if they exist.

• Integration challenges exist in defining not only tasks required for operating the unit, but also for interacting with other on-site maintenance and support organisations for multiple units.

• The skill set for control room operators, especially those required to manage more than one product, may require a different distribution of qualifications (e. g. more senior reactor operators fewer reactor operators, or even a new class of personnel).

• For some advanced SMR designs, operators will face the challenge of supervising the operation of additional units as they are placed on-line. As the number of modules increases, the demands on the operators will change, and potentially the number of operators required for safe operation (that is, multiple staffing plans may be needed to address the addition of more units during the construction period or subsequent operating periods).

These challenges indicate that HSIs and their selection and deployment cannot be considered in isolation of the tasks and environments associated with them. It has to be an integral part of the HFE process, which, in turn, has to be integrated with the rest of the design organisation’s engineering processes.

The licensing issues related to SMRs are covered elsewhere in this Handbook, but for the human factors engineer it is essential to find an early resolution of regulatory issues regarding the use of new HSIs. Early resolution will enable designers to incorporate appropriate changes during the development of their concepts of operation, designs, task analyse, and staffing plans before submitting a design review or licence application. It will also support the NRC staff’s review of the design and license applications.

In the PRA-guided design or safety-driven design, PRA is instituted from the very inception of the concept development, and is used continually and iteratively to evaluate and inform the design and identify safety-beneficial design changes. Obviously, any reactor development can benefit from the PRA-guided design, but there are specific technical and practical considerations that make SMRs more likely to benefit from this approach than evolutionary advanced large LWRs. Essentially, SMR designs tend to start with fewer well-defined details and preconceived solutions and are therefore more open when considering even significant modifications. At the same time, they are more conducive to novel design solutions (such as the integral reactor vessel) needed to address the identified safety weak spots.

The PRA-guided design as implemented for the IRIS reactor is illustrated in Figure 8.4 (Kling et al., 2005; Petrovic et al., 2012). The process is conceptually straightforward, but in practice involved tens of redesign iterations.

By consistently applying PRA to enhance and extend its safety-by-design approach, IRIS has lowered the predicted CDF to below 10-7 events/reactor-year and LERF to below 10-9 events/reactor-year. PRA analyses and design modifications are indicated in Figure 8.5.

The CDF for the initial design, after reviewing dominant cut-sets, was estimated to be ~2 X 10-6, a respectable number, but far from the IRIS target. In the next phase, marked in Fig. 8.5 as Step 1, sensitivity cases on individual significant factors (test intervals, diversity, reassessment) were performed and the design was modified, reducing the CDF to ~5 X 10-7. That was a limit achievable by optimizing single parameters. In the next phase (Step 2) more complex design changes were evaluated to understand and improve coupled processes by simultaneous optimization of several parameters, enabling reduction of CDF to ~1.2 X 10-8. The next Step 3 accounted for higher level of design details, which increased the CDF to ~2 X 10-8. Step 4 evaluated IRIS specific auxiliary systems, anticipated transients without scram (ATWS), human

reliability analysis, and further design details. Some weaknesses were identified that temporarily increased the estimated CDF, and then the design was improved, restoring the low CDF value of ~2 X 10-8. Step 5 indicates initial evaluation of external events. Every point on the graph represents an iteration including a PRA and a design modification. After implementing the PRA-suggested modifications to the reactor system layout, the preliminary PRA level 1 analysis estimated the CDF due to internal events (including ATWS) to be about 2 X 10-8, more than one order of magnitude lower than in current advanced LWRs. Such improvement would not have been possible by mere ‘engineering judgment’, without the benefit of systematic PRA-guided improvements, and without the design characteristics of iPWR SMRs.

Clearly, PRA limited to only internal events would be misleading, and in fact, the external events frequently become the limiting factors in iPWR SMRs, but the approach may be extended to external events as well, as for example in Alzbutas et al. (2005).

Multiple SMRs represent both a ‘modular’ design concept and a ‘modular’ investment model: multiple SMRs may offer the investor a step-by-step entry in the nuclear market. As long as multiple SMRs are deployed with a staggered schedule, the investor has the option to expand, defer or even abandon a nuclear project, to adjust the investment strategy in order to catch early market opportunity or to edge a market unexpected downturn. The investment involves sequential steps with multiple ‘go’ or ‘not-to-go’ decisions that allow management to respond to changes in the market or in the regulation environment, or to adapt to technological breakthroughs. The risk edging capability of a modular investment such as multiple, staggered SMRs is enhanced compared with a monolithic LR. This flexibility against future uncertainty can be measured by the real option analysis and exploited to face the investment risks (Locatelli et al., 2012).

So how should industry rise to this challenge: evolution or revolution? Evolution will drive existing processes harder. Revolutionary change will introduce new techniques. New techniques often imply greater risk. A way through this could be to look at techniques that have been validated in other industrial sectors. The small reactor as a volume product is a revolutionary step for the nuclear sector. There is the opportunity to pick up and emulate the lean manufacturing techniques that have driven efficiency in other manufacturing sectors. The rigor and discipline that accompany lean manufacture will cascade down into sub-tier vendors. The key challenge to the success of the revolution is the culture of the supply base. Can existing supplier bases change their organisational structures and cultures to deliver nuclear components in this manner or will the demand be satisfied by new incumbents who already have the appropriate operating culture? Having acknowledged the organisational cultural change that is required, this chapter will focus on the revolutionary challenge from an engineering perspective.

The two halves of this revolutionary challenge will be explored independently, technologies and techniques around component fabrication, and then a platform for integrating these in a production environment.

12.2 Options for manufacturing

Manufacturing processes are characterised and applied depending on the volume of product they are delivering. Each manufacturing process aims to maximise efficiencies, and minimise production costs. With high volume manufacture the assembly line has been established as offering the best solution. Popular examples can be found in automotive applications. A lot of the early production line thinking emerged in this industrial sector, and is credited to Henry Ford and the application of the production line methodology for the Ford Model T. The production line demands a pace, repeating the same sequence of activities at the same location on each manufactured unit. The small reactor has opportunities for factory build, with vessels and components that are physically smaller than conventional nuclear plant. Standardisation of design similarly aligns with the capabilities of a production line approach — ‘Any customer can have a car painted any colour that he wants so long as it is black’ (Ford and Crowther, 1922). Equally the repetition of sequenced activities fits with the assurance that nuclear manufacturing requires.

However, the potential application of the conventional production line approach is compromised for the small reactor for two fundamental reasons. The first is volume and profile of sales and the ramp-up of those sales over time. The second is level of activity required to assemble each unit factored against the volume of units flowing down the production line.

The emergence of ‘smart’ electricity grids and small modular reactors may have significant impact on future electricity markets. Significant progress in data acquisition systems and information processing have led to smart grids capable of smoothing load/demand curves and integrating energy storage systems, allowing a baseload system to operate at a stable level for longer periods of time in a predictive fashion (based on anticipated grid demand determined from historical trends and current state estimators) rather than varying production intermittently based on fluctuations in renewable energy availability. These advances are also key to reliable operation of hybrid system architecture.

Smart grids could enable the implementation of smaller input sources, such as SMRs, by balancing the load dynamics at a local scale rather than at the large scale required by traditional, large-scale nuclear plants. In this case, SMRs could be sited based on other subsystem requirements. Siting could be in the vicinity of the process feedstock (e. g. coal, natural gas, biomass), near the end-user (e. g. local community or commercial industry), or near the coupled renewable input source. Such siting would reduce transport distances for both electricity and thermal energy, thereby minimizing transmission losses. Hence, SMRs offer operational flexibility, not to mention the already discussed investment flexibility, by introducing a broad range of production opportunities and simplified coupling to renewable sources, and coupling to many more process applications than large-scale nuclear implementations.

The 2012 OECD NEA report on nuclear energy and renewables [1] makes several recommendations regarding full representation (transparency) of power-generation costs at the system level and the need for low-carbon energy technologies in current and future energy markets. Specifically, this report points to the need for low-carbon technologies to supplement variable renewables as renewables continue to grow to a larger grid share. While the report goes on to note that nuclear energy can provide flexible, low-carbon back-up capacity as the renewable penetration on the grid increases, the introduction of SMRs in a hybridized system takes this recommendation a step further. Direct integration of low-carbon, baseload, dispatchable energy (e. g. nuclear) with low-carbon renewables allows policy goals for reduced GHG emissions to be met while retaining high-capacity factors for the baseload source via load — dynamic operation (operating at steady state power to produce multiple products). The NEA report further recommends establishing flexibility at the system level for future low-carbon systems. Accomplishing this flexibility will require increased load-following capability in dispatchable energy sources, such as nuclear, expanded energy storage systems, and increased responsiveness to demand changes. Again, the proposed hybrid energy system configurations would meet this need at a local level via load-dynamic operation (vs. load-following) while reliably meeting the grid demand.

In July 2013 Westinghouse completed the manufacturing and assembly of two test fuel assemblies for its SMR design (225 MWe per reactor). The fuel design for this LW-SMR is based upon Westinghouse’s Robust Fuel Assembly (RFA) technology used in existing PWRs and for the AP-1000 design. In the past, one of the failure mechanisms of concern for PWR fuel is fretting wear. Since this SMR fuel design

is new, having different axial height and grid locations as compared to other current RFA fuel, the fuel rod vibration characteristics will be unique. Thus, Westinghouse is conducting wear testing to ensure that this design has acceptable fuel rod fretting wear performance. The long-term hydraulic testing is performed in Westinghouse’s Vibration Investigation and Pressure drop Experimental Research (VIPER) test loop at its Columbia Fuel Fabrication Facility in Columbia, South Carolina. The VIPER test loop is designed to hold two full-scale, PWR test assemblies that are placed side-by­side. The maximum operating temperature is 400 °F (204 °C), the maximum pressure is 350 psig (238 MPa), and the peak flow rate is 7000 gpm (442 L/s) [17].

14.3 Future trends

The factors that will determine the ultimate deployment of SMRs, and in particular A-SMRs, will focus on the ability of these concepts to compete economically with large LWRs offering lower costs in terms of both construction and operation, improved performance including conversion efficiencies perhaps with advanced PCS technologies, demonstrated enhancements in safety, and where possible generating less waste. The A-SMRs typically involve innovative designs where new fuels and materials are introduced. Several of these designs are for high-temperature applications. For new non-LWR reactor concepts, the emphasis of future R&D will likely focus on the development, demonstration, and qualification of these new materials and fuels.

R&D for new materials will include the development of materials that will need to be compatible with the several coolant types (liquid metals, gas, and liquid salts) at elevated operating temperatures as well as core structural materials such as graphite for HTGR and FHR applications. Advanced steels will be required for fast reactor concepts. R&D efforts are ongoing in the US as discussed earlier in this chapter on the development and qualification of new materials for advanced reactors including A-SMRs under both the ART and ARC programs at DOE-NE.

Work in the US is expected to continue on further development and qualification of TRISO fuels for use in HTGRs and FHRs as well as fast reactor fuels (oxide and metallic). These new fuels and cladding materials will need to be capable of withstanding irradiation at higher fuel burnup as well.

The demonstration of these new materials and fuels will require both modeling/ simulation and experimental capabilities. Given the significant resources required for new facilities, additional emphasis will be placed on modeling and simulation tools for the integration of reactor physics, thermal hydraulic, and structural mechanics models to evaluate the expected performance of these new concepts. Also, the re­purposing or refurbishing of experimental facilities used in the R&D performed earlier the US will likely be evaluated as options to building new facilities to conduct verification and validation of the models.

Additional R&D will continue in looking at how to integrate digital instrumentation and control (I&C) systems and advanced control architectures to enable integration of control, diagnostics, and decision making for highly automated multi-unit plant operations as well as devising alternate concepts of operation for multi-unit SMR designs. Successfully integrating diagnostics capability into new A-SMR design will assist in providing a sound technical basis for extended operation beyond the initial licensed time frame.

Integrating SMRs in general with renewable electric power sources as hybrid systems as a way to balance power on the electric grid due to the intermittency of power produced by such renewable sources as wind and solar and utilize the excess thermal power from the reactor for process heat applications is garnering more attention today. With some large LWRs being shut down due to economic factors resulting from the lower cost of natural gas and decreased electrical demand, these hybrid systems potentially offer a new application for SMRs in general. To effectively evaluate such systems, R&D will be needed in the way of modeling to understand the interactions and control systems for coupling the renewable and nuclear plants. Also, in these coupled systems the reactor may be used more in a load following manner, which will require R&D for controls based upon process heat application and ensuring the safe operation when cycling the reactor through increasing and decreasing power operations.

As some of these A-SMR concepts mature and evolve as more realistic options for future deployment, one may see two trends emerge. Within the US, the first will be focused on potential industry — government partnerships somewhat different from to the cost-share arrangements between DOE and two of the LW-SMR vendors, m-Power and NuScale. These partnerships are focused on a cost shared arrangement to support a design certification approval from NRC. After having demonstrated to a reasonable degree the favorable attributes of these advanced concepts in terms of economic attractiveness, efficient operations, and enhanced safety via government supported R&D, it will likely be necessary for an industry-government arrangement to support the final demonstration in the form of a test or prototype reactor. Such an arrangement would be more appealing to industry if the government were to identify scenarios for deployment to assist federal agencies including the Department of Defense in meeting clean power goals as established by the current Administration. It is straightforward to identify areas where there is significant power demand by groups of collocated federal agencies. Such a consideration would benefit the government agencies with a dedicated source of reliable power and provide the nuclear industry the opportunity to demonstrate the safe and reliable operation of these advanced concepts for further commercial deployment.

The second emerging trend may involve more cooperation between the US and the international community in the development of new and/or shared use of existing experimental facilities, test loops, and perhaps test reactors. Such arrangements would support the government reaching the state of development described in the preceding paragraph at which point the nuclear industry could be engaged to complete final development. Presently, international cooperation is underway on advanced reactor concepts among some 13 countries under the GIF cooperative effort focusing on some six advanced concepts. The costs to construct and operate such test and experimental facilities may be viewed as prohibitive for any one country to undertake. Other opportunities for international collaboration exist via the International Atomic Energy Agency and the Nuclear Energy Agency with the Organization for Economic Cooperation and Development.

In 1965 the government requested the CNEA to prepare a feasibility report on the construction of an NPP in the region were most of the population and industry were concentrated. The CNEA’s report showed the feasibility of installing a nuclear power plant and the Atucha site was selected.

In 1966 the CNEA was authorized to start the bidding procedure and 17 offers were received. By the end of 1967 the decision was taken. The construction of Atucha-1 started in 1968 and the plant was connected to the grid in 1974. The main supplier was Siemens AG, from Germany, and the design was a pressure vessel design, moderated and cooled by heavy water, with natural uranium fuel, based on Siemens-KWU PWRs and the German pressurized heavy-water reactor MZFR. Local industry participation in the project was about 40%. The reactor has operated well, with an overall load factor of about 70%. The thermal reactor power of Atucha-1 is 1179 MW, the generator output is 357 MWe and the net plant power totals 335 MWe.

In 1972 it was decided to construct a second NPP in Embalse, province of Cordoba. The call for bids followed the same trend of the Atucha-1 and a 50% minimum local industry participation was requested. In March 1973, the offer of a 600 MWe natural uranium-HWR with pressure tubes, made jointly by Canadian and Italian firms was selected. A very wide technology transfer agreement was negotiated between Argentina and Canada, and signed in March 1974. Different reasons caused both delays in the construction and only partial fulfillment of the technology transfer agreement. CNEA became in 1979 the main contractor for the assembly works, and Argentine firms were responsible for the assembly of many systems. The plant entered into commercial service in 1984. The reactor has operated well, with an overall load factor of about 84%.

The reactor is of a CANDU600 type. The moderator and coolant are heavy water. The fuel channel are horizontal pressure tubes. The moderator is at low pressure and separated from the coolant. The thermal reactor power is 2.109 MW, generator output is 648 MW(e) and net plant power totals 600 MW(e).

In 1979 a call for offers for the third NPP, to be built next to the Atucha-1, was issued. Argentina was requesting a NPP, a heavy-water production plant, and the constitution of a joint company between CNEA and the supplier, who would become architect-engineer for the construction of the present and future nuclear plants in the country. The German and Swiss offer of a 700 MW(e) version of the Atucha-1 pressure vessel reactor and a heavy-water production plant was selected. In 1981 the architect-engineering company ENACE was formed as a joint venture with Germany, and construction of the Atucha-2 started.

Argentine suffered economic crisis, the project was delayed, and construction was formally stopped in 1994. The Argentine nuclear sector was restructured. A state-owned shareholder company, Nucleoelectrica Argentina SA (NA-SA), was created for the operation of the two nuclear stations and the construction and subsequent operation of the third one. In August 2006 the government decided to re-start construction of Atucha-2 under NA-SA management. The first connection to the grid is expected in 2014. Participation of the local industry was very important.

The Atucha-2 was developed from the Atucha-1 design by Siemens-KWU. Many of the components have a conceptual design identical to those of Atucha-1, while the plant layout and other features are derived from the design of the pre-Konvoi and Konvoi 1300 plants. The thermal reactor power is 2.160 MW, the generator output is 745 MWe and the net plant power totals 692 MW(e).

In the 1980s the Empresa Nuclear Argentina de Centrales Electricas (ENACE) designed the ARGOS PHWR 380. Its flow diagram and main technical characteristics were practically the same as those of Atucha NPPs. It had 60 vertical hydraulically actuated absorver rods. The spent fuel pool was located within the reactor building and different fuel options were considered. Its safety design was mainly based on probabilistic safety criteria. Also in the 1980s the Direction de Centrales Nucleares of CNEA designed the TPA 700/300. These were both based on the CANDU 600 and aimed at allowing important local participation. The fuel of Atucha-1 was originally natural uranium but it was substituted by 0.85% enriched uranium in order to reduce fuel costs. The exit burnup was increased from 6 to 11 MW d/kg U.

CNEA has developed under the Combustible Avanzado para Reactores Argentinos (CARA) project an advanced fuel element concept for HWR, specially designed to fit Argentinean fuel-cycle requirements. Atucha-1 and Embalse have quite different designs for the fuel elements. Both NPPs use on-load refueling, but they differ in the number and length of the refueled elements. In Embalse a CANDU reactor type, with a total of 12 fuel elements, 6 meter long channel, two fuel elements are refueled at a time. In Atucha-1 a pressure vessel design, moderated and cooled by heavy water, each vertical channel has one single fuel bundle of about 5 m in length for its active portion, hung by its upper part. Atucha-1 is fueled on-power by a fueling machine that sits above the reactor.

The main objectives of CARA were:

the fuel element could be used in both reactor types;

• NPP performance enhancement;

• thermal-hydraulic and thermal-mechanic extra margins;

• negative void coefficients;

• cost and spent fuel reduction.

This advanced fuel element for HWR has collapsible cladding, 100 cm length and 52 fuel pins. In 2012 a Canadian patent was issued (Florido et al., 2012).

Another variation in some iPWRs that also affects the BP loadings and the complexity of the fuel design is core operation in a once through/single batch reload, i. e., all of the core is replaced each cycle of operation. mPower and SMR-160 are examples where this design option has been examined. Since the fuel is all fresh for each reload, it simplifies the nuclear design because the same fuels, loaded in the same locations with the same enrichment, BP loadings, etc., can be repeated for every cycle of operation, i. e., once a truly optimized nuclear design has been produced, it is simply duplicated for every cycle and for every reactor deployed.

Fresh fuel also has a distinctive axial power profile, with a distinctive chopped cosine power distribution, peaking at the center. Therefore, if all fresh fuel is loaded into the core, in the absence of either control rods inserted or axially varying burnable poison distribution, the power peaking limits will be violated. Irradiated fuel has a much flatter axial power distribution and in those cores with a mix of fresh and depleted fuel, the flattening influence helps to reduce axial power peaking.

All fresh fuel also means that the designer cannot rely on irradiated fuel to help in smoothing the power distribution across the core, and so only BPs and control rods can be used. This tends to result in higher control rod and BP use in the designs rather than utilizing the lower excess reactivity that arises because some of the fuel has been irradiated. As with fresh cores in large PWRs, the use of asymmetric BPs in certain core locations may be required to achieve the power peaking limits, which also adds to the cost of the fuel.

Special attention also has to be paid to the use of a single batch core in terms of uranium ore and enrichment utilization, by ensuring that the fuels achieve their full potential burnup. Fuel loaded on the periphery of the core will tend to have lower powers and resulting burnups compared with the leading power assemblies towards the center of the core, and since the fuel is loaded for only one cycle of operation, there is no opportunity to reload the fuel to achieve higher burnups and ensure that all of the fuel within a given batch achieve similar discharge burnups. This is an indication of how single-batch cores are not as efficient as multi-batch (see Section 4.3.1).

Some of the single-batch core concepts include a cartridge type fuel unit. This suggests (although the designs are yet to be finalized) that the fuel is loaded into the reactor as one single unit, including control rods already inserted. Although this will speed up the core loading and inspection routine, and assist in the criticality safety case for fresh and spent fuel, the cartridge approach will also make it more difficult to inspect and replace any damaged fuel pins. For example, fuel inspection once at the reactor site prior to startup, and rod replacement in the event of a leaker during operations.

Due to the power output required, and the fuel management of the iPWRs under development today, there is a large variation in the fuel demands per reload (see Table 4.1). These limited examples indicate that the iPWRs tend to require more fuel on a per GW year-electrical basis compared with a large PWR such as AP1000. What has to also be take into account in the overall economics, however, is the outage time — for an iPWR with a three-year cycle length compared with an AP1000 with a one-and-half-year cycle length, the iPWR will have half as many outages for refueling and maintenance, which represents something of the order of 750 more days availability to produce power.

It is likely that each of the iPWR designs will have the potential to operate with a range of fuel management schemes, depending on customer and market needs, and overall economics, and further developments of the nuclear design options can continue during the development and demonstrations phases.

The traditional instrumentation in a nuclear power plant consists of temperature, level, flow, pressure, and power. Almost all actuation and monitoring signals are derived from these five measurements. Recently, however, diagnostic measurements have become increasingly more important. With the advent of new unobtrusive ways to determine the health of plant equipment, the diagnostic measurement field is becoming the fastest growing instrumentation field in nuclear plants. The capability to catch a problem before it leads to a failure is a powerful capability in a nuclear plant. It is expected that iPWRs will take advantage of the latest in diagnostic technology, especially in the field of prognostics.

Diagnostic signals are not likely to become safety related measurements, but they will become more prevalent in future large — and small-scale reactor designs. For large reactors they are important for license extension justification; for iPWRs they are important for containment entrapped systems, as the access to containment (at power) during an operating cycle is not possible in several iPWR designs.

Diagnostic technology has provided a significant advantage for nuclear plants over the past few years. Many decisions about current and future system health have been determined on diagnostic evidence. Currently, one of the most common of diagnostic measurements is temperature, especially with rotating equipment. Thermography ‘guns’ have provided additional insight for electrical components that may be malfunctioning. Oil samples are another precursor to failure in rotating equipment.

Vibration measurements on rotating, oscillating, or even stationary equipment can provide valuable insight on the equipment’s health. While all these measurements are valuable to a large traditional PWR, most of them require a maintenance technician to go out to the equipment and take measurements or samples. The key diagnostic measurement in an iPWR is going to be embedded or in-place sensors with automatic processing and indication, as design and operating constraints will prevent the human interface during power operations.

A manufacturing technique called shape deposition manufacturing (SDM) is a technique for embedding thin film sensors or fiber optic sensors for the continuous measurement of temperature and strain in a metal vessel or casing. Reactor vessels or other critical structures may have these sensors embedded during manufacturing for detection of increased strain or precursors to cracking.13 Another technology which will be considered for some equipment is the use of fiber sensors for strain measurement.14

The field of diagnostic measurement and testing is developing quickly. The advantages for iPWRs are obvious. With the iPWR paradigm for less staff and the fact that between cycle maintenance and testing may not be possible, diagnostic and prognostic tools are necessary. Embedded sensors and the automation of these sensors are important for iPWRs.

Good resources for prognostic information can be found in IEEE publications and IAEA publications. Grant work funded by the DOE has produced several approaches which have value.15

Wearable and head-mounted displays of various types have been prototyped and tested by the military for many years. Devices range from large, heavy, full-immersion, head-mounted virtual reality displays used for specialist training, to lightweight, see-through devices used for augmented reality applications. This technology is now finally becoming a commercial reality also in consumer markets. Devices like Google Glass would offer significant opportunities to simplify common control room tasks, like continually monitoring alarm annunciators while performing other tasks, or having computer-based procedures available with a simple voice command.

Virtual reality has a long history in visualisation of, and interacting with, 3-D environments. This is not only a powerful technology for visualising and verifying designs long before they are built, but when combined with wearable devices like augmented reality headsets that superimpose virtual objects and information on the user’s view of the real world, they enable operators to perform tasks without the need for printed documentation or other support. In this way, information about the user’s surrounding real world also becomes interactive and digitally manipulable. This technology is already being used in some industries to support maintenance and assembly tasks.