The contribution of learning (Boarin et al., 2012) applies at various levels: a better work organization on the same site, where the personnel have already had experience in the construction and assembling of previous NPP modules; a learning component in factory fabrication of the equipment; a learning component in the utilization of materials and equipment by more skilled workers, etc. A scale-up of the plant output and the attempt to introduce an original French design, i. e. the N4 reactors, towards the end of the program may only partially explain such an occurrence.

Lovins (1986) presented an interesting theoretical framework, referred to as the Bupp-Derian-Komanoff-Taylor hypothesis, that suggests that with increasing application (‘doing’), the complexity of the technology inevitably increases, leading to inherent cost escalation trends that limit or reverse ‘learning’ (cost reduction) possibilities. In other words, the technology scale-up can lead to an inevitable increase in systems complexity that translates into real-cost escalation, or ‘negative learning’. Nevertheless, learning effects have been recorded in technology-advanced industries (Frischtak, 1994); learning effect description was first published by an aeronautical engineer (Wright, 1936).

The learning effect is also visible in the Korean NPP fleet deployment costs: learning accumulation has played an undeniable effect on a progressive cost decrease (Figure 10.10). KHNP, the owner of all 21 of South Korea’s operating nuclear power reactors, has held a licensee relationship with Westinghouse since the late 1980s when the US-based company supplied the 945 MWe System 80 nuclear steam supply for Yonggwang 3 and 4. After that, KHNP was able to develop variants of System 80 for its own requirements under technology transfer terms in the license agreement. After introducing domestic innovations and updating technology over time, KHNP came up with the Korean Standard Nuclear Plant (KNSP), then the OPR-1000. The current APR-1400 technology represents a further evolution of that design. The construction and power generation costs of the APR-1400 are reported to be 10% lower than those of OPR-1000 units.

Korean NPP is the evidence that learning economy may apply to construction costs: in this case, learning effect was achieved through a concentrated construction (Figure 10.11), with the deployment of twin/multiple units on the same site and by avoiding substantial design modification in order to attain PWR plant standardization and control design complexity. (Wolsong NPP are PHWR-CANDU, provided by AECL, as the only exception to the PWR design.)

It may be argued that in principle, learning accumulation is expected to determine a construction cost and time-progressive decrease of successive NPP units, as it was in the Korean NPP fleet. Nevertheless, as far as western countries are considered, in the real world there is often no evidence of cost and time benefits in large NPP deployment programs. That is why simpler and smaller NPPs, with design modularity and high content of factory fabrication, have a higher chance of controlling complexity and exploiting standardization, enabling learning accumulation on both construction and assembling phases. SMRs are expected to benefit from anticipated learning effects, mostly arising from the construction and assembling of multiple units on the same

site. Given the power size of a nuclear site, more SMR units should be fabricated and installed than LRs, with improved chances to learn. General learning accumulation may be recorded at the engineering procurement and construction (EPC) level residing in the human resources knowledge and approach to the project management, and to the organization and procurement issues, such as supplier selection. This learning applies independently of the site location of the new NPP and is therefore indicated as ‘worldwide’ learning in Figure 10.12. In addition, site-level learning accumulation is also applicable on successive NPP units built on the same site, residing in the best, refined practices and actions by local staff. The magnitude of the two effects is comparable (Boarin and Ricotti, 2011a). The learning effect is destined to fade out over the first five to seven units (Carelli et al., 2010). For this reason, in a mature phase of the market, worldwide learning is not a differential factor for SMRs and LRs, while SMRs keep the benefit from the on-site learning accumulation, which applies in case of multiple units built on the same site.

Implicit in the flowline concept is the standardisation of product design. In a volume manufacture situation, a small reactor vendor will not be able to cope with customisation demands without destroying the economic model for the sector. The design will therefore be fed through a sealed manufacturing route. In itself this is no different from many other high-quality, assured manufacturing operations but it raises questions such as the regulator’s approach to manufacture of components that could be fitted into any one of a number of reactors going through the flowline at that point. Although this issue could be seen as implicit in the NRC’s Part 52 licensing regime, it is not obvious that implications such as this have actually been considered in the detail required.

The deployment options for a small reactor build on one of the key attributes of their design. The small modular reactor (SMR) factory build aspect creates a number of significant advantages during the deployment phase. Delivery of assured modules to a customer site has been claimed as a banner headline, but it is worth reviewing the aspects behind these individual claims to understand the contribution of their worth to the overall deployment:

• time to build;

• risk — schedule risk and cost over-run;

• validation — off-site pre-delivery testing;

• international aspirations;

• Critical path through site activities with modular build.

12.5.1.1 Modularity: addressing schedule and cost risk

For the large part nuclear construction project cost increases and schedule delays have their root in site uncertainty when compared to original plans and cost estimates. Modularity has been offered as the solution to these over-runs.

The Construction Industry Institute defines modularity as ‘work that represents substantial offsite construction and assembly of components and areas of the finished product.’ There are incremental levels of modularity that need to be considered; Graduations of modular construction are as follows:

• Skid — modular skid assemblies of equipment that are delivered to site fully instrumented and validated, requiring little installation and commissioning activity. Within this definition it would also be anticipated that the skids are capable of being delivered to site using conventional transportation networks.

• Super skid — this represents the next increment with larger portions of plant intended to be installed in a building.

• Prefabricated — this is the final step in the definition of modular and would extend to a fully prefabricated module comprising plant equipment in an architecturally complete housing.

Has a modular build approach been deployed already with large nuclear plants? There are elements of modular construction encompassed within large nuclear plants. The Vogtle plant built by Southern Co. in Georgia has utilised a level of off-site modular manufacture. This facility has been established to build modules for the plant in a manner that has the same overall goal. However, the modules for a large nuclear plant are essentially one-off modules, built off-site in a piece-by-piece approach. Without the repeat volume flowing through such a facility it can be argued that module build for a large nuclear plant is an off-site variation to the on-site stick build approach and falls short on delivering some of the goals for the module-build approach advocated for small reactors.

The incorporation of advanced modular construction centres around completely shop-fabricated modules. This method of manufacture is more demanding in terms of upfront planning and management. With greater emphasis on design integration, procurement, manufacturing and construction activities there is less natural opportunity for schedule variation. This modular method of site deployment marries beautifully with the flowline approach bringing about greater schedule adherence.

This modular deployment model is not without consequence. The preliminary and final design activities must remain sealed to allow time for the procurement and delivery activities to be completed. Design change at this point erodes the benefits of the advanced modular approach. The mindset of deployable small reactor manufacturing is driving the product line closer and closer towards commodity-based supply chain design.

Having taken a view on the supply and delivery end of the deployment model, there are equivalent considerations that need to be taken into account with the site implementation.

Preliminary, generalized analyses of proposed hybrid system architectures have been conducted by researchers at the INL to provide initial insight into the value of system hybridization (see refs [3], [4], [18-27]). More detailed analyses of tightly coupled HES requires detailed subsystem models accurately to predict dynamic system performance and to identify potential operating procedures to best manage and mitigate variable energy generation sources and to alternate between output products (e. g. electricity and a chemical process). As the tools for dynamic hybrid systems analysis become more sophisticated, and conceptual subsystems and integrated system designs are better defined, more accurate dynamic analyses can be performed to determine optimum system configuration and specific operating parameters, to design integrated system control architecture, and to establish instrumentation requirements for state estimation necessary for control.

13.2 Future trends

Expanded use of nuclear energy for industrial applications raises several technical challenges that can be addressed through conceptual design; system modeling and optimization; component, subsystem, and integrated system testing (non-nuclear and nuclear); and eventual deployment of the advanced energy system. Early work should be focused on identifying key challenges and then addressing those challenges via design solutions. Advanced modeling and simulation tools can provide significant insight to system feasibility and anticipated performance, beginning with simplified steady-state analysis and moving to more complex dynamic analysis with more detailed subsystem models. The more detailed, dynamic system model can then be used as a virtual test bed for control system design prior to hardware implementation. However, many of the significant challenges associated with hybrid systems derive from complex integrated system control and hardware interfaces, including fast­switching smart valves necessary to divert thermal energy between output streams and connections to energy storage systems. These challenges can be addressed through a number of component and subsystem tests for model validation, followed by integrated system testing.

The SMART core is composed of 57 fuel assemblies, the design and performance of which are based on the proven 17 X 17 array with UO2 ceramic fuel rods in the commercial PWRs. Each fuel assembly holds 264 fuel rods of 2 m in active height, 24 guide tubes for control rods, and one instrumentation thimble, which are mechanically joined in a square array. A total of eight space grids hold the fuel rods in their specified positions. A specially designed bottom end piece offers an improved resistance to the debris entering the core. Slightly enriched uranium oxide of less than 5 wt% is used as the fuel ingredient, providing a sufficient amount of reactivity required for a 36 month cycle operation. The fuel assembly is designed to accommodate power ramps during the load-following maneuvering normally experienced in a commercial PWR. The maximum burn-up of the SMART fuel rod is estimated as 54 000 MWD/ MTU. In order to reduce the critical boron concentration and to control the radial power peaking, a lumped burnable absorber rod comprised of Gadolinia (Gd2O3) mixed in Urania (UO2) of which the 235-U enrichment is 1.80 wt% is used. Eight to 24 burnable absorber rods are used per fuel assembly.

In a current large PWR, the hot RCS water leaving the top of the reactor fuel is directed into the reactor coolant system hot-leg piping. The hot-leg piping, nominally 29-inch piping (74 cm) (NRC, 2006) connects to a U-tube or a once-through steam generator, where heat is transferred from the primary water to the secondary water to generate steam to drive the turbine generator. Subsequently, colder primary water exits the steam generator into the intermediate leg (cross-over leg) piping, nominally 31-inch piping (79, cm) (NRC, 2006) which connects to the suction side of a reactor coolant pump. The intermediate leg level is typically the largest diameter pipe in the RCS and it is below the level of the top of the fuel assemblies, which can make a leak in this section of piping the most problematic of all large-break LOCAs. The reactor coolant pump discharges into the 27.5 inch (70 cm) (NRC, 2006) RCS cold leg which directs primary water back into the reactor pressure vessel. Current large PWRs employ two, three, or four reactor coolant loops. A typical two-loop large PWR reactor coolant system is shown in Figure 5.3.

Since the pressurizer and steam generator functions are integrated into the iPWR pressure vessel, all the large diameter piping associated with current large PWRs is eliminated. This eliminates any possibility of a large-break LOCA in all the iPWR designs. As a result, active emergency equipment, such as high pressure injection pumps, associated with current large PWRs to mitigate the consequences of a large — break LOCA are also eliminated in the iPWR designs.

The small scale and unique environments of iPWRs open up the instrumentation field to new and innovative solutions. The temptation will be to fall back on what has been used before, and in scenarios where this works, perhaps this will be the path of least resistance. Because of design peculiarities and non-standard physical environments, new instrumentation or redesigned instrumentation will likely be required. In addition to qualifying these new designs and new approaches for new measurement technologies, the iPWR design engineer will need to coordinate with the regulator to accommodate the regulator’s concerns/requirements at the same time.

The sections below identify some of the known challenges and trends in the I&C world for the iPWR designer.

In older control rooms, discrete control input devices (that is, devices that depend upon mechanical motion) are still the most common means for the operator to interact with the plant’s systems. These devices are limited to relatively primitive devices like buttons, switches and levers. As I&C and HSI technology evolves it is becoming possible to control plant components by means of direct manipulation devices like a computer mouse, joystick, keyboard or trackball. As computing power continues to increase, we can expect to see even more sophisticated devices finding their way into the control room and other work areas. In the not too distant future we can expect to see fixed as well as mobile devices that allow not only direct interaction through touch and force feedback, but also indirect interaction through gesture, speech and gaze.

• Probability of adversary success — The probability that an adversary will successfully complete a pathway and generate a consequence.

• Consequences — The effects resulting from the successful completion of the adversary’s intended action described by a pathway, including the effects of mitigation measures.

• Physical protection resources — The staffing, capabilities, and costs required to provide PP, such as background screening, detection, interruption, and neutralization, and the sensitivity of these resources to changes in the threat sophistication and capability.

Measures can be estimated with qualitative and quantitative approaches, which can include documented engineering judgment and formal expert elicitation. Measures can also be estimated using probabilistic methods (such as Markov chains and event trees) and two-sided simulation methods (such as war-gaming techniques). GenlV International Forum (2011b) reviews a number of system analysis techniques relevant for PR&PP studies.

SMRs are defined broadly to include a range of technologies, particularly a range of diverse fuels and coolants. US SMR reactor technology currently chosen for NRC licensing and near-term commercial deployment (2022-2025) is based on LWR technology. Most of these SMRs will be integral pressurized water reactors (iPWRs) as discussed in Part II of this Handbook. LWRs have a well-established framework of regulatory requirements, a technical basis for these requirements, and supporting regulatory guidance that provides acceptable approaches for meeting NRC requirements. The NRC uses a standard review plan (SRP), NUREG-0800, to review licensing applications for these reactor designs. NUREG-0800 was revised in January 2014 to provide general review guidance for SMRs. The NRC will require design-specific SRPs for the licensing of SMRs. These SRPs are under development by the NRC and SMR vendors. The SRP for the B&W mPower design is expected to be published in mid-2014 to support the submittal of the mPower design certification application scheduled in 2015. Additionally, the NRC has a well-established set of validated analytical codes and methods and a well-established infrastructure for conducting safety research needed to support its independent safety review of an LWR plant design and the technical adequacy of a licensing application. It should be emphasized that the near-term SMRs, particularly the iPWRs, will be subject to the same licensing process and the same safety requirements and standards as new large LWRs resulting in no diminution of safety.

New SMRs can be licensed under either of two existing regulatory approaches. The first approach is the traditional ‘two-step’ process described in Title 10, Part 50, ‘Domestic Licensing of Production and Utilization Facilities,’ of the Code of Federal Regulations (10 CFR Part 50), which requires first a construction permit (CP) and then a separate operating license (OL). The second approach is the new ‘one — step’ licensing process described in 10 CFR Part 52, ‘Licenses, Certifications, and Approvals for Nuclear Power Plants,’ which incorporates a combined construction and operating license (COL). It should be noted that the both the SMR vendor B&W and the proposed license applicant TVA have chosen the two-step Part 50 process for licensing of the mPower SMR at the Clinch River, Tennessee site. This two — step process was chosen by TVA and approved by the NRC for this first-of-a-kind (FOAK) design to reduce the risk of delay in obtaining a final design certification under the Part 52 process. It is assumed that all other U. S. applicants for FOAK SMRs and nth of a kind SMRs will adhere to the Part 52 NRC licensing process.

Key to the licensing of any new reactor design, including SMRs, is the approach that the regulatory authority will utilize in assessing the safety basis for the reactor and associated safety systems. Regardless of the licensing process chosen, the licensing authority will need to consider the relative merits of a deterministic versus a risk-informed performance-based approach to assess and approve the safety case for SMRs.