Critical heat flux (CHF) tests using SMART fuel bundles have been conducted to provide database for the development of a CHF correlation which is essential for the evaluation of the thermal design criteria and safety analysis of SMART. A series of experiments was carried out to investigate the thermal mixing and CHF in 5 X 5 test bundles with uniform and non-uniform axial power shapes. A high-pressure water test loop included pressure housing with 5 X 5 SMART fuel simulators, high-head pump, pressurizer with non-condensable gas, heat exchangers, preheater and mixers. The CHF experiments varied the pressure from 2.5 to 17 MPa and the mass flux from 200 to 2500 kg/m2 s which covers the normal operating conditions as well as the AOO (anticipated operational occurrences) for SMART. Several hundreds of CHF data items obtained from the water loop CHF experiments were employed for the development and validation of a CHF correlation which was applied to the thermal design and safety analysis of SMART.

Current large LWR coolant pumps function to provide forced primary coolant flow to remove the heat generated by the fission process. Existing large PWRs utilize two to four reactor coolant pumps to provide the forced coolant flow through the primary. Natural circulation flow in current large PWRs will not provide sufficient flow to remove the heat being generated during power operation. Each 6000 to 10 000 horsepower pump provides flow of approximately 100 000 (US) gallons per minute (378,500 litres/min) to remove the heat generated by the reactor fuel assemblies for delivery to the steam generators. A driving head of approximately 90 psi (0.6 MPa) is generated by the reactor coolant pumps. A loss of flow from one or more reactor coolant pumps in a current large PWR results in a reactor trip (NRC, 2006).

These large pumps have seals to limit primary leakage and the seals require cooling. This provides a scenario for a small leak path as well as a scenario for an inter-system leakage path. Newer Generation III+ PWRs like the AP1000 design use a canned reactor coolant pump design to eliminate this possibility.

The integral nature of the iPWR pressure vessel makes the utilization of reactor coolant pumps more challenging. The two larger iPWR designs of the four active iPWR designs currently under development in the United States, the Generation mPowerTM design and the Westinghouse SMR design, plan to incorporate multiple reactor coolant pumps. Because of space limitations in the primary flow path, it is not possible to incorporate the large reactor coolant pumps used by the current PWRs. Instead, smaller reactor coolant pumps in these designs will be required to deliver the necessary flow. While specific reactor coolant pump design detail is unavailable, the resulting driving head will likely be smaller as well. Presumably, a canned type pump will be incorporated. The pumps may be located in the hot leg at the top of the steam generators or in the cold leg beneath the steam generators, depending on the design. The Generation mPowerTM design calls for 12 pumps (Kim, 2010) and the Westinghouse design calls for eight pumps (Memmott et al., 2012). In addition, the Korean SMART iPWR design also plans to use four canned reactor coolant pumps (Lee, 2010). The IRIS design planned to use eight spool pumps; one pump per steam generator (Carelli et al., 2004). The remaining two active SMR designs currently under development in the United States, the NuScale iPWR design and the Holtec SMR-160 design (a PWR-based SMR), do not plan to use reactor coolant pumps (IAEA, 2011). These designs will incorporate natural circulation cooling. In addition, the Argentine CAREM iPWR design and the Japanese IMR iPWR design plan to incorporate natural circulation coolant flow for normal operation (IAEA, 2011). All the iPWR designs will be capable of removing reactor decay heat following reactor shutdown using natural circulation cooling. Therefore, a dedicated AC backup is not required for the designs using reactor coolant pumps.

The non-safety instrumentation systems are wide open for technological advancements. Software controlled non-safety digital systems are attractive to the utility owner and do not hold as much concern with the regulator as safety systems. (US Regulator action is trending to more demanding requirements for NSSS digital control systems, as common mode failures in NSSS control systems have implications on the safety system response of the plant.)

The new high-tech devices such as fiber-based sensors, and ultrasonic sensors have the attractiveness for use throughout the NSSS and BOP systems. The small scale of many iPWRs provides an opportunity to use more state-of-the-art devices with better accuracy, better ease of installation, lower maintenance, better availability, etc.

The I&C designer’s dilemma for instrumentation for the iPWR non-safety systems, is whether to use new technology with less of a nuclear track record or to use the traditional nuclear instrumentation with years of nuclear experience and pedigree. Some iPWR BOP systems and some iPWR NSSS systems could utilize the same instrumentation as a large traditional nuclear plant. The degree of similarity of the iPWR BOP systems to the existing large PWR BOP systems would determine the potential to use traditional instrumentation; however, the new and unique nature of the iPWR’s design would create the opportunity for new instrumentation applications. It is likely that new iPWR non-safety designs will utilize some new technology already on the market, but not currently used at the existing large PWRs.

A range of advanced sensors embedded in HSIs allow operators to expand their ability to sense the state of the environment and the behaviour of artefacts within the environment by means of haptic devices or ‘tangible interfaces’. Such devices take advantage of the sense of touch to convey a range of information by applying forces such as vibration, force feedback and sensing location and motion. This tactile stimulation can be used to assist in the detection of changing conditions or orientation of objects that the operator cannot handle manually due to hazards such as heat or radiation. In more advanced devices it can create the illusion of virtual objects and the ability to control them in a computer simulation, to control such virtual objects, and also to enhance the remote control of machines and devices (telerobotics). Again, the reader is referred to Buxton’s work (Buxton, 2011, chapters 7, 8, 9, 13, 14).

A common example of haptic interaction in the form of vibratory feedback is found in the Sony, Xbox and Nintendo game controllers mentioned before. Haptic devices may also incorporate tactile sensors that measure forces exerted by the user on the interface. (Jones and Sarter, 2008). It is easy to imagine how operators would be able to use a device like this to ‘feel’ the bearing vibration of a turbine while monitoring the spin-up process!

7.8.5.2 Brain interaction

Recent state-of-the-art developments promise to offer interaction possibilities considered impossible just a few years ago. For example, direct brain-machine interaction has long been considered science fiction (think of the 1984 novel Neuromancer by William Gibson or the 1999 movie The Matrix!), but it is fast becoming a reality. Consumer-oriented devices like Emotiv Systems’ Insight neuroheadset already demonstrate impressive capabilities to control devices and software. With devices like this designers can dramatically enhance interactivity and the level of immersion in the application by, for example, enabling the system to respond to a user’s facial expressions and adjusting the application’s behaviour dynamically in response to user emotions such as frustration or excitement, and enabling users to manipulate objects in an application or even turn them on or off or change their state by simply using the power of their thoughts. This is reality and no longer science fiction; it is not too hard to imagine that these devices will find their way into certain applications in industry within 20 years…

The GIF evaluation process includes nine specific steps that are organized into four main activities:

• D — Defining the work

• M — Managing the process

• P — Performing the work

• R — Reporting the work.

Each step is primarily associated with one of these activities. The nine steps are thoroughly explained in Figure 9.4 and the accompanying text below. Clearly, some level of management is associated with each step. Reporting cannot all be done at the end, but must be generated as the work progresses; thus, the process is iterative, and sometimes the steps are concurrent. Note that the steps in the process are numbered in the order they are first performed (as shown in Figure 9.4), but grouped for discussion under the four main activity areas described above.

New SMR LWR designs offer significantly enhanced safety, security and simplicity of design as describer earlier in this Handbook. Both the industry and the NRC recognize that many of the existing technical requirements would be applicable to these new designs. Traditionally, LWRs were licensed using deterministic engineering judgment and analysis to prove the safety case and establish the licensing basis. However, with the significant improvements in safety design, the NRC is permitting greater emphasis on the use of probabilistic risk assessment (PRA) techniques and risk insights to establish the licensing basis for SMRs. All US SMRs will develop and use design-specific PRAs to support their licensing basis.

While the NRC has indicated that it will permit greater use of PRAs to support and establish a licensing basis, the use of the PRA would be commensurate with the quality and completeness of the design and attendant PRA presented with the application. Depending on the quality and completeness, the NRC might use the PRA and risk-insights to complement a deterministic analysis to establish a licensing basis (including the selection of licensing basis events). Or, it might rely more heavily on the PRA and use a deterministic engineering judgment and analysis to complement the PRA. In the post-Fukushima environment, the quality of a design-specific PRA coupled with the use of deterministic engineering judgment and analysis becomes crucial for the licensing of SMRs that either seek relief from traditional LWR safety requirements, or are used to support revised licensing requirements. The use of PRA risk information for licensing of SMR LWR designs is particularly appropriate because the quality and completeness of the PRA is bolstered by the maturity of the LWR design, and the extent and richness of the operating history information.

The level of site preparation can be greater with the commoditised module delivery. This can have the effect of moving the nuclear island off the critical path for site construction activities. This potential impact of this is typified by looking at the key dimensions of the typical steam turbine generators, as in Table 12.3.

It can be seen that the larger turbine generators quickly become significant units and sit outside of conventional road transportation envelopes. There are regional variations on transportation sizes. Packages around 5 m wide and tall can usually be accommodated as oversize loads. Moving beyond 5 m moves away from a turbine generator set that can be road transported as a single module. A turbine generator that is not modular can put the rotating machine on the critical path for the site build programme. This is substantiated with information from combined-cycle gas turbine construction programmes. From the time the last generator casings are delivered to site it is a further year until the turbine is running on its turning gear, and a further three months before the set is ready to be synchronised for the first time. It is easily conceivable in a small reactor power plant with larger turbine generators, which are not capable of road transportation as a single unit, that the turbine can sit close to the on-site programme critical path.

12.5.1 Deployment model: in service

The opportunity to deploy a small reactor to compete against in any energy portfolio is challenged by its relatively high $/kW overnight capital cost. Some benefit is derived from a factory build approach of plant modules, but this alone does not achieve the desired cost target. Similarly a position on volume production with an overhead amortised over a larger number of modules will make further inroads into the $/kW. The final element in reduction of the competitive position for a small reactor lies in the in-service deployment.

In March 2012 Jack Bailey — VP Nuclear Generation Development for TVA — offered a view on the FOAK plant in-service model. He commented that initial goals for centralised regional services could not be realised for early deployed small reactor plants. The authors would challenge that view, asserting that it is this operating model that needs to be incorporated from the outset. In the same way as a production line needs to be designed at the outset for a level of product delivery, the in-service support package needs to be designed from the outset. For example, if the small reactor is to secure a level of plant sales comparable to current closed cycle gas turbine plants then the training burden alone needs a level of innovative thinking to support the volume of operating staff that need to be qualified on the plant. Early statements about centralised engineering support need to be developed to maintain support for the plants coming on-line. Part of the market appeal for a small reactor plant is the lower capital outlay. The reduced capital outlay opens up the small reactor to a new population of nuclear owners and operators. These ‘new to nuclear’ owners will not have mature engineering upkeep organisations to support their small reactor plants. As a consequence the third aspect to achieve a competitive cost base for a small reactor is the deployment of a fleet model that provides a centralised support service over a number of sites.

12.3 Conclusion

It can be seen that the supply and deployment of the small reactor offers an opportunity for the nuclear industry to adopt a revolutionary supply chain model, and maybe a new group of suppliers. There are opportunities to build an alternative supply chain established around manufacturing techniques that have been validated in other sectors.

With greater volumes of units flowing through the factories the incorporation of advanced manufacturing techniques can be substantiated. These techniques in turn offer efficiencies in manufacture that are reflected in lower manufacturing costs. The factory assembly of modules can be supported in a manner that offers incremental capacity increases reflecting market certainty.

The incorporation of contemporary automation for factory build techniques delivers a greater level of automatic part tracking from standardised product assembly. This in turn opens up options for fleet management across multiple sites managed to a common template. The deployment model for a small reactor is a revolutionary step for the nuclear industry.

Reference

Ford H and Crowther S, 1922, My Life and Work, Kessinger Publishing.

A tightly coupled integrated system will benefit substantially from demonstration via integrated test facility. Such a facility would allow for both physical and virtual representation of subsystem components, requiring either physical or data linkage between the subsystems. A proposed integral test facility could support component testing, partially integrated system testing, and/or fully integrated system testing including interfacing with the electrical grid.

Significant scaled, non-nuclear, integrated system testing has been conducted for space nuclear power and propulsion systems [28-30], which in some cases are essentially scaled versions of the proposed terrestrial hybrid energy systems. Low — capacity fission-power systems are currently being developed for space application via both computational modeling and experimental efforts. The National Aeronautics and Space Administration (NASA) and the Department of Energy Office of Nuclear Energy (DOE-NE) have adopted a hardware-based approach to system development intended for early identification of challenges to the system design, fabrication, assembly, and operation and to assist in design optimization. To minimize cost and development time, a non-nuclear test approach is used to demonstrate integrated system operation during the system design and development stage, offering opportunity for system reconfiguration without the radiological hazards associated with a fueled system.

For a nuclear hybrid energy system, initial hardware demonstration could similarly integrate a physical reactor simulator that uses electric heaters to mimic the heat generated by nuclear fuel pins. Operation of the reactor simulator would rely on a virtual component derived from system modeling to computationally simulate the neutronic response that would be observed in a fueled system using measured system temperatures as state estimators for the heater control logic. Once all the relevant feedback mechanisms are understood for a particular reactor design, appropriate instrumentation and measurement points can be selected for the non-nuclear test hardware such that the virtual reactivity feedback may be applied appropriately. Other system components could be represented by physical hardware or simulated computationally depending on the stage of facility development, the availability of a specified component or subsystem, or the relative experience with that subsystem that impacts the availability of validated computational models. Integrated system demonstration with hardware-in-the-loop allows the researcher to evaluate integration challenges and to characterize system response time and response characteristics.

‘Virtual’ reactor kinetics applied to the control system architecture can generate significant understanding at a system level, but in a manner that allows for system reconfiguration, control system design and demonstration, operator training, test to failure, etc. without the added complexity and security inherent to a nuclear system.

Successful non-nuclear demonstration of a fully integrated NHES will provide a strong foundation for future build of a nuclear prototype. Several test facilities have been built and operated by potential SMR vendors for specific, focused purposes, including testing to characterize the performance of individual components and materials, to evaluate subsystem thermal-hydraulics and to identify and evaluate integral system effects. In many cases these facilities are designed around a specific reactor concept and are intended for testing of traditional LWR designs, but others may be more generally applicable. Potentially relevant non-nuclear test facilities could include the various non-nuclear test facilities constructed to refine the design of the CAREM (Central Argentina de Elementos Modulares) reactor in Argentina, which began construction of a 25 MWe prototype nuclear unit in February 2014 [31]; Korean Atomic Energy Research Institute (KAERI) has built several facilities to verify and refine the design of the System-integrated Modular Advanced Reactor (SMART) plant [32, 33]; Babcock and Wilcox (B&W) has constructed the Integrated System Test (1ST) facility to verify the mPower reactor design and safety performance in support of NRC licensing activities [34]; and NuScale Power has designed and built a one-third scale, electrically heated prototype facility (the NuScale Integral System Test [NIST] facility) to demonstrate concept viability and stability [35].

Study of the design of these test facilities may offer insight into design of an SMR hybrid energy system test bed. Some of the existing test facilities might be leveraged to benchmark component and subsystem models that could be incorporated in hybrid system simulation. Additional facilities may be necessary to verify the performance or refine the design of specific interface components for hybrid systems, such as fast-switch valves that will be necessary to achieve the desired load-dynamic behavior in a tightly integrated system, and buffer components, such as thermal energy and electrical energy storage technologies. Integral system testing with these components installed can then be conducted to evaluate the subsystem interactions across valves and energy storage devices to verify transient system behavior given both anticipated operational occurrences and transients associated with accident conditions.

13.3 Sources of further information

M. F. Ruth, O. R. Zinaman, M. Antkowiak, R. D. Boardman, R. S. Cherry, and M. D. Bazilian, ‘Nuclear-Renewable Hybrid Energy Systems: Opportunities, Interconnections and Needs,’ Energy Conversion and Management, 78 (2014), 684-694.

W. Phoenix, Personal communications with various nuclear plant engineers, 2011.

AspenTech, Aspen Plus Version 2006 (Build 20.0.3.4127), 2006.

Microsoft Corporation, Microsoft Excel 2007, Version 12.0, Redmond, Washington, 2007.

Acknowledgements

Significant contribution has been made to the topics covered in this chapter by fellow researchers at the Idaho National Laboratory. Contributors include Richard Boardman, Michael McKellar, Richard Wood, Humberto Garcia, Piyush Sabharwall, and Cristian Rabiti. Opinions expressed in this chapter are that of the author and are not attributable to the US Department of Energy or the Idaho National Laboratory.

15.4.4.1 Steam generator cassette

SMART has eight identical SGs which are located on the annulus formed by the RPV and the core support barrel (CSB) (see Figure 15.7). Each SG is of a once-through

design with a number of helically coiled tubes. The primary reactor coolant flows downward in the shell side of the SG tubes, while the secondary feedwater flows upward through the inside of the tube. The secondary feedwater is evaporated in the tube and exits the SG cassette nozzle header at a 30 °C superheated steam condition of 5.2 MPa. In the case of an abnormal shutdown of the reactor, the SG is used as the heat exchanger for the PRHRS, which permits an independent operation of the PRHRS from the hydraulic condition of the primary system. The design temperature and design pressure of the SG cassette are 360 °C and 17 MPa, respectively.

15.4.4.2 RCP

SMART has four RCPs horizontally installed on the upper shell of the RPV (see Figure 15.8). Each RCP is an integral unit consisting of a canned asynchronous three phase motor. The rotational speed of the pump rotor, 1800 rpm, is measured by a sensor installed in the upper part of the motor.

Being a canned motor pump which does not require pump seals, its characteristic basically eliminates a SB LOCA associated with a pump seal failure which is one of the design basis events for conventional reactors.

D. Cummins

Rock Creek Technologies, LLC, Loudon, TN, USA

6.1 Introduction

Integral pressurized-water reactors (iPWRs), Small modular reactors (SMRs) of the light pressurized-water kind, are the wave of the future. Other chapters have detailed the advantages of iPWR modular designs in meeting the needs of small metropolitan centers or developing countries where the electrical grid infrastructure is not present, as well as the needs of normal power plants with the modular ability to add units. The ‘smallness’ and the ‘modularity’ of the new iPWR designs present several advantages to the nuclear power community.

Some of the common architecture features of an iPWR are:

• small capsule-like design;

• below grade installation;

• steam generators inside the reactor pressure vessel;

• small on-site crew manning for operations and maintenance (more automation);

• a multi-unit control room;

• modularity and grow-ability (ability to add more modules to increase the power output for a small additional increase in infrastructure);

• more passive cooling techniques;

• less accident or safety cases;

• more built-in fail-safe features;

• more prognostic and diagnostic capabilities;

• low or no maintenance instrumentation for cycle to cycle runs;

• incorporation of lessons learned from large PWR experiences.

With these features come challenges for the instrumentation and controls (I&C) design. iPWRs will have more automation, more redundancy, more cyber security, more fail­safe features, more fault-tolerant designs, more prognostic and diagnostic capability, and different measurement methods. With these design challenges it is important to focus on I&C development early in the overall reactor design/development phase in order that solutions, possibly unique solutions, can be developed and qualified early. Just as the mechanical/electrical design requires I&C compatibility, the I&C design requires mechanical/electrical modifications. This give-and-take between the I&C design and the mechanical/electrical design is a necessity in these new builds

Handbook of Small Modular Nuclear Reactors. http://dx. doi. Org/10.1533/9780857098535.2.123

Copyright © 2015 Elsevier Ltd. All rights reserved.

and requires a high level of communication and coordination between functional divisions.

New technologies offer solutions and abilities that the old 1970s and 1980s technology did not have. The advent of digital designs and wireless communications are technological advances that should be considered in every new design. The extent to which an iPWR incorporates these technological advances will determine the amount of safeguards that may need to be built in to the design and the license application. In the case of microprocessor-based digital technology, a detailed defense of common cause and common mode failures must be developed. In the case of hardwired logic-based digital design, such as field programmable gate arrays (FPGAs), common cause and common mode failures are not as much of an issue, but a detailed diversity and defense in depth must still be developed. In the case of digital or wireless communications, cyber security threats and electromagnetic interference/radio frequency interference (EMI/RFI) must be considered.

Traditional measurement methods may not be applicable in the new designs. New environments, submerged vessels, lack of traditional piping and new geometries all play a part in the need for new measurement devices and methods. The traditional devices offer the safety qualification pedigree that a nuclear plant requires, but the traditional devices are designed for a traditional plant and may not be qualified or designed for the new harsher environment and/or the new smaller geometries involved. There is an emerging need for radiation-hardened, high pressure and temperature qualified, submergible, smaller sensors with remote processing electronics. Although the technology is developed, the application of this technology to the new iPWR environment has not been accomplished in many cases; especially the cases of primary vessel level and flow. New qualification programs will need to be developed for the new sensor technologies.

In summary, the maturation of iPWR design is bringing about a paradigm shift in instrumentation design and methodology. What used to work for large light — water PWRs will not necessarily work in the iPWR environment. The new smaller dimensions, differing geometries, and harsher environments will necessitate new instrumentation solutions. This chapter highlights the requirements, challenges, and potential solutions for this new I&C world.