In nuclear plants, as in almost all industrial plants, there is always a lot of materials handling that requires various common and also sophisticated tools. HSIs for materials handling include simple overhead cranes, forklifts and even programmable robotic systems. Advanced HSIs will play an increasingly important role in fuel and material handling systems, especially for handling hazardous materials like low — and high — level radioactive waste. Several technologies are becoming prominent in new means to improve reliability and safety in hazardous applications. These include graphic visualisation for remote controls that will take traditional methods of materials handling to a new level. Devices envisaged include robots or remote-controlled machines, augmented reality and eye — and position-tracking systems for tasks such as planning for surveillance and maintenance in plant areas where environmental conditions such as radioactivity, heat, cold, dust or toxic materials would prevent or limit human access. In addition to visual monitoring and material handling systems, HSIs would have the ability to notify operators immediately if there is a problem in the system, including jams, misaligned sensors, worn bearings or related issues that cause system degradation. These kinds of technologies will thus offer significant benefits for system as well as human performance.

Most SMRs (including in particular iPWR SMRs) claim some level of enhanced

security compared to current large loop LWRs. This is justified by the following

features that are favorable or easier to implement for SMRs:

• Protection from external physical threats to the plant, such as an airplane crash, by full or partial underground placement of a plant combined with reinforced outer structure(s). This is easier done for a small than for a large plant, and typically includes placing at least the reactor and used fuel pool below grade.

• Access control and prevention of unauthorized access and intrusion by full or partial underground placement, limiting the number of access points and potential intrusion points (potentially to a single entry point). At the same time this reduces the necessary surveillance and protection force and thus has a positive economic impact.

• More difficult access to safety-relevant equipment, again due to underground placement.

• Inherent increased resilience of passive safety systems to sabotage or intentional mal — operation. Clearly, systems with fewer components, functioning on forces of nature, are more difficult to perturb and disable.

• Inherent safety features — accidents that cannot occur in the first place cannot be malevolently initiated either.

Potential synergistic use of the safety and security characteristics is illustrated in

Table 8.6.

8.2 Future trends

Future trends for iPWRs are expected to focus on safety and economics, i. e., those features that simultaneously enhance safety as well as economics. In this author’s view, the most important design trends and features include:

• addressing safety from the very start, through safety-driven design (Safety-by-Design);

• using specific iPWR SMR characteristics to turn them into safety advantages;

• implementing inherent safety features;

• reliance on (exclusively) passive safety systems;

• introduction of additional features, levels of safety, and barriers to promote DID;

• partial or full below grade placement for security reasons;

• licensing with eliminated or reduced off-site EPZ;

• extending the post-accident grace period, with a larger, easier replenishable, ultimate heat sink;

• seismic isolation;

• advanced instrument and controls (I&C) to support safety in operation and status monitoring in off-normal conditions;

• advanced I&C for diagnostics/prognostics;

• fuel with enhanced accident tolerance.

Additionally, low-power-level systems will likely consider in their design:

• natural circulation for heat removal in normal operation;

• soluble boron-free operation (economically preferred for lower power levels);

• very long core life, or so-called ‘battery’ approach;

• near zero self-regulating excess reactivity, eliminating the possibility of prompt criticality, but usually limited to (very) low-power systems (such as ELENA, 3.3 MWth system envisioned for district heating (IAEA, 2012a)).

Related needs and challenges to enable or support desired safety features in the area of improved analytic capabilities and licensing include the following:

• Improving our understanding of passive systems and their vulnerabilities and failure probabilities needed for PRA. Currently, while the PRA approach is well established, specific probabilities and their uncertainties are not yet well quantified for iPWR SMRs.

• Better understanding and addressing the potential for common mode failure and potential for negative mutual impact of multiple units at the same site. As demonstrated by the Fukushima Daiichi accident, this is not a mere theoretical possibility. It requires developing reliable approaches to avoid such occurrences; otherwise, SMRs may not be able to fully claim (e. g. in licensing) the benefits of the smaller source term per unit.

• Risk-informed licensing. This will provide a framework for reducing the emergency planning zone based on adequate risk estimate rather than prescriptive values (Carelli et al., 2008).

Additionally, testing and validation will be necessary to address specific unique features:

• Testing and experimental validation of natural circulation phenomena and integral primary configuration. This will provide confidence in current models and support licensing.

• Validation of advanced analytical methods. Some novel or unique features (novel fuel, different flow regime from that in current reactors, novel components, etc.) will require new simulation codes or methodologies, that will need to be developed and validated based on carefully devised testing.

Some of the research needs include the following:

• Developing and improving reliable passive decay heat removal approaches, with indefinite grace period. This addresses the main challenge — ensured decay heat removal in all accident situations.

• Developing effective ‘firewalls’ between the nuclear and non-nuclear portion in co­generation applications. SMRs are more suitable for co-generation applications, which rely on effective separation.

• Developing novel components with improved performance. Some examples include novel fuel forms and designs, internal CRDMs and steam generators, integral pressurizer and fully immersed main coolant pumps.

Furthermore, considering trends in traditional PWRs and in non-LWR SMRs, the

following research trends, even if currently not pursued for iPWR SMRs, could be

expected to impact the next generation of iPWR SMRs:

• Fuel with enhanced accident tolerance, such as coated particle type (e. g., tristructural isotropic type, TRISO), originally developed for a very high-temperature reactor (VHTR), now as fully ceramic microencapsulated (FCM) considered for LWR. This fuel would potentially add margin in certain safety aspects, but it has significant challenges of its own, related to fabrication and cost. The author does not expect to see it applied to the first wave of iPWR SMRs.

• Use of other-than-oxide fuel, such as nitride or silicide. One possible driver is the desire for achieving a longer cycle, which could be justified for specific purposes, but it has its own development cost and challenges.

• Amplified negative feedback. Proposed for several high-temperature reactors and in some very low-power iPWRs (e. g., ELENA; IAEA, 2012a), primarily to be used for self­regulation. However, one needs to consider cooling reactivity insertion, to make sure it is not introducing a more negative than positive safety impact for general applications.

• Low-pressure systems. Inspired by lead or molten-salt-cooled reactors that offer attractive safety characteristics, this is clearly not applicable for mainstream iPWRs. However, specialized iPWR applications may be tempted to consider significantly lowering the operating pressure.

Co-siting economies arise from the cost-sharing of some common structures, systems and services by multiple units built on the same site, decreasing the incidence of some fixed costs and, thus, the penalty of the loss of economy of scale (Boarin et al., 2012).

The current designs of large, new-build reactors have components that require transport to site individually, for subsequent assembly on-site once they arrive. The main containment vessel of a small reactor on the other hand will fit within the specifications of a single large reactor component. An AP1000 steam generator is 22.5 m ~(74 foot) long, 5.6 m ~(18 foot) diameter and weighs 700 t ~(1.5 X 106 lb). The same component in an EPR has similar dimensions but weighs 550 t ~(1.2 X 106 lb). Compare this to a small reactor where a complete integral unit can fit within the above envelope, and weigh less than 500 t ~(1.1 X 106 lb) and it becomes apparent that from a manufacturing perspective the components can be manipulated on the flowline with relative ease. The entire small reactor unit can now be transported as a single unit direct for installation to the site, moving more tasks to the factory to complete assembly in a controlled environment.

Integral reactors often have more numerous, smaller components per plant which lends itself to a flowline approach. For example, a typical large, new-build main coolant pump will weigh 124 t (2.7 X 105 lb), which represents a significant part to lift and negotiate in a manufacturing environment. In a small reactor, however, each pump may weigh only 500 kg (1100 lb). This is of a comparable magnitude to the aero-engine flowline model, so it seems appropriate to adopt this technique to the small reactor product.

A primary motive for nuclear-renewable hybrid energy systems is the efficient alternative use of the generated heat when it is not needed for electric power production due to low net demand conditions. The proposed load-dynamic behavior of a system that incorporates variable renewable power generation introduces system complexities as a result of timing (when the heat is available), timescales (required response rate), and the large amount of excess heat that must be diverted to industrial applications. Industrial processes can potentially be designed to absorb the heat at timescales aligned with heat availability. Operating temperatures for each of the ‘classes’ of proposed SMRs result in different options for coupling to the selected process applications [18]. Table 13.2 provides a brief overview of possible process applications and the corresponding operating temperatures.

13.5.1 General considerations

A number of integration issues must be considered when coupling process heat applications to nuclear reactors. Key considerations include:

• reactor outlet temperature;

• reactor inlet temperature;

• fluid composition;

• pressure of primary coolant, heat transfer loop, and process heat application;

• primary coolant heat capacity;

• tritium migration.

The reactor outlet temperature defines the temperature and heat that can be provided to the process heat application. Low-temperature reactors can still provide process heat to most applications but will require temperature amplification via heat recuperation or high-temperature topping heat from fossil fuels and/or electric resistance heating to couple with higher-temperature process heat applications. Greenhouse gas emissions

can still be reduced relative to traditional generation sources in these cases. Process heat application studies under the NGNP program found that a reactor outlet temperature of 825-850 °C would provide sufficient heat to handle most applications. However, coupling a high-temperature reactor to a low-temperature process heat application can reduce the overall process efficiency due to inefficient use of the high-temperature heat. Optimal coupling of subsystems based on temperature outlet/inlet requirements may result in the most efficient use of the reactor thermal output, but regional needs may sway the selection of a system configuration based on economic factors.

Each generalized reactor type has a specified temperature difference (rise) and pressure difference (drop) across the core. For instance, a helium-cooled HTGR has a nominal temperature rise of 350-400 °C, while the core temperature rise for a fluoride salt-cooled high-temperature reactor is of the order of 100-150 °C. The temperature of the heat transfer fluid returned from the process heat application must be at or below the reactor inlet temperature. If it is at a higher temperature, heat must either be rejected to the environment or used in an additional process (i. e. power production using a bottoming cycle). A smaller core temperature rise corresponds to a higher average temperature process heat than designs having a larger core temperature rise, generally leading to a reduction in the number of reactors needed to provide the high-temperature heat.

The coolant and secondary heat transfer loop fluid composition can affect which process heat applications may be integrated. The melting temperature of molten salts and liquid metals (e. g. sodium) must be considered when applied to process heat applications. Some applications would return the working fluid at temperatures below the coolant solidification temperatures, which could plug heat exchangers and piping, without careful design of the heat transfer interfaces and potentially boosting the temperature of the return fluid. Coolants such as water, steam, carbon dioxide and helium do not have such issues, although condensation must be considered.

Process heat applications require heat not only at specific temperatures, but also specific pressures. Low-pressure coolants, such as liquid metals and molten salts, provide ideal pressure conditions at the core, but large pressure differences may occur at the process heat exchangers where high temperatures and pressure differences must be considered in the heat-exchanger design. High-pressure coolants, such as helium and pressurized water, may have the same effect if used for low-pressure process heat applications.

Process heat applications generally require a secondary heat transfer loop to isolate the primary core coolant loop from the process heat application. A primary purpose of this secondary loop is to reduce the migration of tritium from the reactor core to the process application. Additional heat transfer loops may be added to reduce tritium migration, but each loop reduces the process heat temperature based on the temperature difference required to transfer heat across the additional heat exchangers.

Some process heat applications use the heat rejected from the power conversion cycle of the reactor. Low-temperature applications, such as seawater desalination and district heating, may be able to use the heat rejected from the condenser of a Rankine cycle. The temperatures needed for these applications are higher than the usual rejection heat temperature such that some temperature amplification may be required. Heat may alternately be extracted from the low-pressure turbines at a higher temperature; this reduces the electric power generated, but the overall system thermal efficiency is increased due to the additional heat utilization.

The existing proven PWR technologies are basically utilized for the SMART design. However, it also adopts new and innovative design features and technologies that must be proven through tests, experiments, analyses and/or the verification of design methods.

15.2.2.1 Thermal-hydraulic test

A series of fundamental tests and experiments have been carried out throughout the SMART development phases to examine the physical phenomena related with the specific SMART design concepts. The main purpose of these experiments was twofold: to understand the thermal hydraulic behavior of the specific design concepts and to obtain fundamental data to be used, in turn, for further feedback to the optimization of design. Among the experiments conducted, specific SMART design-related experiments are as follows:

• boiling heat transfer characteristics in the helically coiled steam generator tube;

• experiment for natural circulation in the integral arrangement of the reactor system,

• two-phase critical flow tests with non-condensable gases to investigate the thermal hydraulic phenomena of critical flow in the presence of non-condensable gases;

• critical heat flux measurement for 5 X 5 test bundles simulating SMART fuel assembly and out-of-pile mechanical/hydraulic tests for full-scale SMART fuel assemblies;

— Design certificate — Standard design — Experiments	Potable water Electricity
Figure 15.1 SMART development program.

• water chemistry and corrosion tests at a loop facility to examine the corrosion behavior and characteristics of fuel cladding, internal structural materials and steam generator tube materials at reactor operating conditions;

• experiments on wet thermal insulation to determine the insulating effects for the internal pressurizer (PZR) design and to derive a heat transfer coefficient for the design;

• experiments on phenomena and characteristics of heat transfer through the condensing mechanism of the heat exchanger inside PRHRS tanks.

Brief introductions to the typical tests and experiments implemented to verify the

SMART design characteristic are given below.

On 17 December 2009, the National Law 26566/2009 declared of national interest, inter alia, the design, construction and start-up of CAREM prototype, establishing a special regime. CNEA was entrusted to complete these tasks. The Preliminary Safety Analysis Report and the Quality Manual were presented to the Argentinean Regulatory Authority in 2009.

In 2010 the ARN issued a new procedure for the licensing of prototype NPPs. Information is being provided to and analyzed by the regulatory body according to this new procedure. The Universidad Tecnologica National — Facultad Regional Avellaneda has performed the Environmental Impact Study of CAREM reactor prototype and it was presented to the authorities of Province of Buenos Aires in December 2012.

Site activities such as soil studies and environmental analyses have been performed. The construction of a high-pressure and high-temperature rig for testing the innovative HCRD mechanism has been finished. This rig can also be adapted for testing the structural behavior of the FA. In 2013 the ARN and the province of Buenos Aires issued the necessary permits allowing the start of prototype construction. The construction of CAREM prototype started in February 2014. Contracts and agreements are being taking with different Argentinean stakeholders to perform detail engineering. In 2013 the construction of the RPV was granted to a local supplier.

16.2 Future trends

The CAREM project is the main R&D project in Argentina and in the future it will lead the activities related to SMRs in the country. The first step of this project is the construction of the prototype of about 27 MWe (CAREM 25). The following steps consider the prototype operation and the commercial modules’ development.

For the commercial modules’ development, the economic aspects are very important. CAREM power was initially fixed at 15 MWe but the prototype power was increased to achieve a better economic performance. In general the economy of size can be use to improve the competitiveness of a given reactor configuration. But there are technical and economic constraints that limit the maximum economical size of a given concept. In integrated primary system reactors the size of the reactor pressure vessel is a relevant one. The CAREM concept economy was analyzed using a very advanced tool and two alternatives were evaluated for the primary system flow rate (natural circulation and assisted circulation). Below 150 MWe, the natural convection option is preferred. But over that power level, the size and cost of the RPV are outside the acceptable range so the forced convection option is preferable. The maximum power achievable using pumps is about 300 MWe.

The development of a natural circulation commercial module of about 100 MWe is foreseen as the next step. This module will better benefit from the experience developed with CAREM prototype design, engineering, licensing, construction and operation while incorporating the economic improvements arising from the economy of size. The construction of the first CAREM commercial unit at the province of Formosa is under consideration. Sitting and module optimal size studies are ongoing.

A further step considers the development of a forced convection module of larger power. Different options could be considered to assist flow circulation in CAREM concept. CAREM configuration has many similarities with BWRs and the use of jet pumps was preliminary considered (Delmastro et al., 2006). In integrated primary system reactors an annular space is available in the down-comer below the steam generators and the jet pumps could be located there. Another option is the use of internal centrifugal canned pumps located in the lower plenum of the reactor pressure vessel. But as primary system natural circulation precludes the loss of flow accident, the use of forced circulation must take into account the potential new requirements over the thermal margins evaluation and the safety systems.

16.6 Sources of further information and advice

The International Atomic Energy Agency every year produces many technical documents related to nuclear power energy and particularly nuclear reactors and SMRs. Many of them include information related to SMR research and development and deployment in Argentina (IAEA, 1989, 2002, 2004, 2006).

In a PWR, a pressurizer functions to maintain the pressure of the primary coolant system in a range such that no boiling occurs in the primary system under normal and transient operations. In a current large PWR, the pressurizer is a separate cylindrical tank connected to the reactor coolant system piping by a surge line, nominally 10 inches (25 cm) (Datta and Jang, 2007), and a spray line, nominally 4 inches (10 cm) (NRC, 2012a). Pressure is normally controlled using heaters and a spray valve to maintain the pressure range. A balance of water and steam exist in the pressurizer space. The water level in the pressurizer provides an indication of water inventory in the reactor coolant system.

Some large PWR designs also use a power operated relief valve (PORV) connected to a pressurizer relief tank to assist in controlling pressure. The PORVs will open to reduce pressure prior to the system reaching the RCS safety valve relief set point. In large PWR designs that utilize a PORV, a stuck open PORV is a potential small — break LOCA initiating event.

A PWR pressurizer provides the surge volume for the reactor coolant system. If the RCS temperature increases, less dense reactor coolant system water surges into the pressurizer, compressing the steam space and increasing the primary pressure. Steam will naturally begin to condense to lower the pressure, which may be sufficient for a small or slow transient associated with a minor RCS temperature change. More commonly, however, pressure will rise to the spray valve set point. The open spray valve then admits colder primary water from the RCS cold leg into the steam space to quickly condense steam and subsequently reduce pressure. The spray flow is driven by the discharge pressure of the reactor coolant pump. Conversely, if the RCS temperature decreases, the suddenly denser reactor coolant water causes water to flow out of the pressurizer into the reactor coolant system hot leg, expanding the steam space and decreasing the primary pressure. This is controlled by use of pressurizer heaters, which heat the pressurizer water and subsequently increase pressure (NRC, 2006).

An iPWR integrates the pressurizer into the top of the reactor pressure vessel. A baffle plate or head plate with drilled openings separates the pressurizer from the reactor coolant system and acts as the surge line. Because of the integral nature of the pressurizer design, the 10-inch surge line is eliminated. The volume of an iPWR pressurizer is significantly larger than a current large PWR pressurizer relative to reactor thermal power. In the IRIS iPWR design, the pressurizer volume is about five times larger per unit of power than for a current large PWR design (Ingersoll, 2011). This larger pressurizer volume, coupled with larger reactor coolant system water inventory overall relative to reactor thermal power, provides slower pressure transients in general. This provides a number of operational benefits. First, an operator will have more time to analyze changes in plant operating conditions and respond accordingly. Second, the need for a fast acting spray valve is virtually eliminated because in most cases the natural steam condensation following a slower insurge of coolant into the large volume pressurizer will adequately maintain the RCS pressure under normal operations and expected transients. Finally, the integrated location of the pressurizer in an iPWR design provides a more direct indication to the operator of the water level above the top of the reactor fuel at all times (IAEA, 1995).

The ‘normal’ spray valve is eliminated in many iPWR designs because it is not essential to fine tune the system pressure as discussed above and, in designs with no reactor coolant pumps, there is insufficient driving head in the RCS to provide adequate spray flow. Even in iPWR designs employing smaller reactor coolant pumps, the pressure differential or driving head created in the RCS is much lower than that in a current large PWR. Current large PWRs employ an ‘auxiliary’ spray valve to back up the normal spray valve when the connected reactor coolant pump is unavailable or the normal spray valve is inoperable. The auxiliary spray valve in a current large PWR is typically driven by the charging pump discharge. The NuScale iPWR design plans to use this approach to provide for pressure reduction through pressurizer spray actuation (NuScale, 2012). Other iPWR designs will likely use this approach to include a pressurizer spray function as well. As a result, the spray line is not necessarily eliminated in an iPWR design, but it is generally limited to less than the 4-inch size employed in current large PWRs. Pressurizer heaters in an iPWR design function the same as the heaters in current large PWR designs.

PORVs are not incorporated into the iPWR pressurizer designs. A stuck open PORV was the root cause of the Three Mile Island accident, which is eliminated in all iPWR designs. Safety relief valve discharge is directed to containment or another water storage tank inside containment. Therefore, a pressurizer relief tank is not incorporated into the iPWR designs. In addition, piping to provide a nitrogen cover gas on the pressurizer relief tank and the need for a drain system and an associated pump are eliminated in the iPWR designs.

Traditional PWR safety qualified pressure transmitters are devices that are able to supply an electronic signal that is proportional to an absolute pressure measurement or a delta pressure measurement. Nuclear versions of these devices were originally manufactured and qualified in the 1970s and 1980s, but have had various features modernized for recent advancements in electronics and materials. The most important feature of the traditional device is the qualification of that device. A class 1E safety qualification for any transmitter means that the device has been through a series of tests and longevity assessments that prove that it is operational and survivable for nuclear conditions. Specifically it means that the device can meet the radiological, temperature, pressure, and longevity requirements for a nuclear reactor environment. Specifics on the qualification requirements for class 1E equipment is contained in IEEE 323-197419, 198320 and 200321 and IEEE 344-197522, 198723, 200424 and 2013.25

The issue with the traditional transmitters is that they were designed for an environment unlike the one that will be experienced in an iPWR design. In a traditional PWR, the containment environment, where most safety classified traditional pressure transmitter electronics are mounted, is a large, airy environment with relatively low ambient temperature and pressure, and with easy access for maintenance and repair of equipment; whereas most iPWR containments are designed as capsules that closely envelop the reactor vessel. These capsules have limited access with small enclosed environments of air, vacuum, or water that may include high ambient temperature and pressure. The location and mounting of these pressure transmitters will be quite different from the traditional approach, and the limited access for maintenance or repair will need to be factored into the design. In some cases, there will be no traditional pipe mountings for the sensing elements, and the use of sensing lines may not be possible in all cases. These hurdles in instrumentation, going from the traditional PWR to the iPWR, will necessitate the redesign of traditional models and/or the development of new technology.

The quandary in which iPWR instrumentation designers find themselves is that the least risk-averse approach, from a licensing and schedule perspective, is to use the instrumentation that already has the qualification pedigree and has worked in the past with traditional PWRs. However, the many changes in iPWR mechanical and physical design mean that the traditional devices may not function correctly or even fit in this new environment and design. This dilemma forces the I&C designer to look outside the ‘traditional box’ for answers.

New technologies, discussed in other sections, offer some solutions and some problems. Some new technologies offer smaller packaging, submersible options, more reliable manufacturing, flexible-mounting options, remote electrical processing, and fewer maintenance-intensive options. It is easily seen how these features would be welcomed by the iPWR I&C designer. The major drawback is the lack of device qualification and the lack of longevity field data that substantiates the survivability of these devices in the actual environment. As with the generation II reactors that populate many countries today, the iPWR I&C systems will undoubtedly experience the same growing pains as those experienced in the 1970s, as new designs are tried and lessons are learned.

As mentioned above, digital technology is expected to expand greatly in iPWRs. The same concern for common mode and common cause failure potential with software-based digital systems will continue to be an issue for iPWRs as it is with the large PWRs. The FPGA technology (Section 6.6) and technology like it that provide the reliability of digital devices without the need for operational software, has the appeal to take the forefront of safety and NSSS system processing in iPWRs.

7.8.5.1 Gesture interaction

Gesture interaction is a way for computers to interpret purposeful human motions, thereby creating a bridge between machines and humans that allows a richer interaction experience than the primitive input methods of keyboard and mouse. Using gesture recognition, operators can literally point a finger at the computer screen to interact directly with objects, without actually touching the screen. This technology is still in its infancy but several devices and applications are beginning to appear. Devices like the Xbox Kinect and the Leap Motion gesture controller are likely to become mainstream options very quickly. In the short term, this probably will not make conventional input devices such as mice, keyboards and even touch screens redundant, but will be added to the range of HSIs to allow operators more flexibility in interacting with plant systems. For an extensive discussion of gesture interaction, the reader is referred to the work of Bill Buxton, a Microsoft researcher (Buxton, 2011).