To determine the outcomes of the system’s response to a threat, analysts compare pathways and assess the system to integrate findings and interpret results.

9.3.5.1 Pathway comparison

Analysts perform a pathway analysis by considering multiple pathway segments. In general, measures are estimated for the individual segments of a pathway and must then be aggregated to yield a net measure for the pathway. Although measures for different pathways may be aggregated, it is generally more valuable to use the measures to identify the most vulnerable pathways. The objective of the system evaluation is then the identification of the most vulnerable pathways and the measures associated with them.

9.3.5.2 System assessment and presentation of results

The final steps in PR&PP evaluations are to integrate the findings of the analysis and interpret the results in order to arrive at an assessment of the SMR. Results include best estimates for descriptors that characterize the results, distributions reflecting the uncertainty associated with those estimates, and appropriate displays to communicate uncertainties.

As indicated earlier, new SMRs in the US can be licensed by the NRC under either of two existing regulatory approaches that require a CP, an OL, or a combined CP/ OL (COL). Licensing options will be briefly outlined below to the extent one or more of the options might be similar to licensing processes in other countries. Important to note in any licensing option is whether (a) the design submitted is preliminary or final, (b) the design submitted has been previously approved or certified by a competent and capable regulatory authority, and (c) the chosen site has been previously characterized and approved (in the US through an Early Site Permit (ESP) process), or is seeking approval. Fundamentally, any chosen licensing strategy will depend on the status of the safety review of the reactor design, and the status of the review of the safety and environmental characteristics of the site.

The NRC has two unique licensing processes. First, the NRC licensing process has a requirement that reactor designs that will be certified through 10 CFR Part 52, Subpart B, must go through a rulemaking process to provide an opportunity for public involvement. No other country requires a rulemaking process in order for a reactor design to be certified by a cognizant regulatory authority. A design certification rule (DCR) is intended to freeze the design in a more final and formal manner. Any changes to the design must be done by amendment to the DCR which requires a formal and lengthy rulemaking process. Second, another licensing option in the US is for a utility or other end-user to ‘bank’ a potential SMR site through the ESP process under 10 CFR Part 52. If an ESP is granted by the NRC, then the applicant can file a subsequent COL application that would reference both the ESP and the DCR. An ESP has a limited time for use (within 10-20 years), but it can be transferred to another potential applicant that might want to use the site for a nuclear power plant.

• Option 1: Preliminary design provided and site needs approval. This option begins with the submittal of an application for a CP that contains site safety information, a complete environmental report (ER), preliminary design information, and a preliminary plan for operational programs. Once the CP is issued, an OL application is submitted that provides final design information and operational programs (with implementation schedules). Upon review and approval of the OL and completion of construction, the NRC grants authorization to load fuel.

• Option 2: Final design provided and site needs approval. This option starts with the submittal of an application for a CP that contains site safety information, a complete ER, final design information, and a preliminary plan for operational programs. Once the site is approved and the CP is issued, an OL application is submitted with a description of the operational programs (with implementation schedules) and confirmation that safety configurations and systems are as described in the final design application. Once construction is complete and the OL is issued, the NRC grants authorization to load fuel.

• Option 3: Design approved/certified and site needs approval. This option is the same as option 2 except that the COL application references a certified design, i. e., a DCR. The COL further contains site safety information, a complete ER, final design information with inspections, tests, analyses, and acceptance criteria (ITAAC), and a description of the operational programs (with implementation schedules). Once the site is approved and the COL is issued and ITAAC are met, the NRC grants authorization to load fuel.

• Option 4: design approved/certified and site approved. This option is the same as option 3 in that a COL application is submitted that references the DCR, but it also references a previously approved site. Once the COL is issued and ITAAC are met, the NRC would grant authorization to load fuel.

The NRC and the US nuclear industry see significant advantages in the Part 52 process and it is the preferred, if not the only, licensing process for all applications that are not FOAK designs. Advantages to this process and key considerations for international licensing are:

• standardized design that will remain final and stable for most applications;

• early identification and resolution of all licensing issues;

• few licensing exemptions required;

• public and transparent licensing process;

• predictable and efficient licensing process that reduces financial risk.

These licensing advantages can benefit the international application, licensing and deployment of US-certified or other regulatory-approved SMR designs as will be discussed later in this chapter.

There are some options, however, for wider levels of international cooperation using a globalised modular approach to the build and development of small reactors. A single product that has access to a global market will need to offer routes for international offset and localisation of content and workscope. Depending on the approach to modularity throughout the plant, i. e. both the nuclear island components and the balance of plant, it could be possible to support a localisation scope objective where the mechanical and process interface can be defined at a module boundary. In this way the modularisation within a whole power plant can open up opportunities for localisation. Equally the fixed interfaces across these module boundaries will limit options for change. History has shown that the level of control across modular interface boundaries will need to be managed under robust change control to ensure that interface compatibility problems do not arise and cause delays on site. The small reactor that is moving towards a commodity product is going to need to balance the business case requirements of localisation against the value of standardisation.

Hybrid systems analyses are conducted in two stages: simplified steady-state analysis to determine technical feasibility of a candidate system architecture, followed by detailed dynamic analysis of promising integrated system configurations for system optimization, detailed performance analysis, and control system design.

Top-level system analyses are performed using steady-state modeling and analysis tools. These analyses help establish system and subsystem boundary conditions, such as operational temperatures, pressures, and flow rates necessary to maintain overall energy balance, providing the essential information to determine if a proposed configuration is technically feasible. Introduction of preliminary economic analysis tools then allows a system designer to determine the potential economic value (return on investment) associated with the proposed integrated system configuration.

Dynamic system modeling integrates more detailed component and subsystem models, using validated component models and modeling tools where possible. For advanced subsystems, such as advanced SMR concepts employing non-water coolants or novel heat exchanger designs, validated component models may not be available. In these instances the model developer should use reasonable assumptions for subsystem design parameters and performance behavior using validated modeling tools. Separate effects tests and subsystem testing may be used to later validate those assumptions.

Dynamic simulation of hybrid system performance is where some of the key questions regarding the reliability, resiliency, and efficiency of hybrid system operation can begin to be answered. From the perspective of the reactor subsystem, these questions might include the following:

• To what extent can the power of a single reactor module be varied to accommodate load variations?

• What benefits, if any, can be realized by incorporating multiple small reactor modules in the integrated system?

• What factors limit the rate and magnitude of increasing/decreasing reactor power, including the frequency or magnitude of the proposed fluctuations in system load, frequency, and voltage?

• If rapid response time is required by the balance-of-plant, what mechanisms can be used to buffer the reactor(s) from rapid transients, and how can these be implemented with high reliability?

• What steps are necessary to develop and demonstrate system-level control that will minimize individual reactor cycling?

• How will the overall system safety analysis be performed to reflect the relationship between reactor safety characteristics and the overall system safety?

• How can electricity demand changes be managed? A decrease in electricity demand would trigger a switch from electricity generation to thermal power storage or direct use as thermal energy in the integrated industrial process. What type of interface (valving, controls, operational strategies) might be required to divert reactor thermal power to the appropriate subsystem, and what is the characteristic response associated with those components?

A well-defined dynamic simulation will provide an excellent platform for control system development. The hybrid system control must first establish control hierarchy, including priorities for electricity production (i. e. first meeting the electricity demand before considering other output streams) or allocation of the thermal energy based on the current market price for the output commodity. Second, the control system requires specific state estimators (temperature, pressure, etc.) to provide input to the control algorithms, recognizing that optimal placement of instrumentation necessary to provide these data is critical to reliable control system performance. A validated simulation of the integrated system provides an excellent virtual test bed for sensitivity studies associated with the control system design and optimization of the control system architecture.

Results of the dynamic simulation will identify areas of significant uncertainty or significant sensitivity that will impact the prioritization of subsystem and integrated system testing. While modeling and simulation can provide a vast amount of understanding on the potential performance of an integrated system, there is great value in translating simulations to hardware demonstrations (particularly non-nuclear, electrically heated demonstrations) prior to building an integrated nuclear hybrid system prototype.

The SMART core is designed to produce a thermal energy of 330 MW with 57 fuel assemblies of a 17 X 17 array and shown in Figure 15.6. The SMART core design providing an inherent safety is characterized by:

Figure 15.6 SMART core loading pattern and control rod arrangement.

• longer cycle operation with a two-batch reload scheme;

• low core power density;

• adequate thermal margin of more than 15%;

• inherently free from Xenon oscillation instability;

• minimum rod motion for the load follows with coolant temperature control.

SMART fuel management is designed to achieve a maximum cycle length between refueling. A simple two-batch refueling scheme without reprocessing returns a cycle of 990 effective full power days (EFPD) for a 36 month operation.

As noted throughout this discussion, current large PWRs require a safety-related AC power bus backed by an emergency diesel generator (EDG) to provide assurance that active safety-related equipment will function when required. There are usually two EDGs per large LWR onsite. The EDGs are governed by plant technical specifications and must be tested at least once a month. Every six months, each EDG must be shown to come up to speed and voltage, and begin the loading sequence within 10 seconds of receiving a start signal (NRC, 2012b). Although necessary to prove the safety basis of the plant, this kind of testing can be problematical for the EDGs. In the event that one EDG is declared inoperable, the remaining EDG must be tested within 24 hours. As a result, it is not unusual for EDGs at large LWRs to be tested quite frequently, leading to higher maintenance requirements.

By removing the need for safety-related diesel generators in the various iPWR designs, the planned diesel generators become ancillary power supplies that enhance iPWR defense-in-depth by maintaining the availability of normal operating equipment that is not safety-related. Testing of these ancillary diesels can be less frequent and fast load testing can be relaxed such that the ancillary diesels have an opportunity to warm up for a few minutes before loading. Therefore, less maintenance can be anticipated for the iPWR ancillary diesel generators.

Current large PWRs maintain EDG-backed safety-related AC buses, nominally at 480 Volts and 4160 V AC to operate active safety equipment. In addition, current LWRs require safety-related power to operate instrumentation during and following an accident. This is provided by battery-backed 120 V DC and low voltage AC through an inverter. The iPWR designs will not require high-voltage safety-related buses, because there is no active safety-related equipment component with this need. Power to reposition any iPWR safety-related valves will be required initially. Only battery-backed 120 V DC and low voltage AC through an inverter will be required to be safety-related, which will meet the needs for repositioning valves and powering instrumentation. So, the electrical distribution systems of current large LWRs and iPWR designs will be similar, but the approach and amount of testing and maintenance will differ.

There are several reasons why the treatment of HSIs in the nuclear industry is different from other industries. For example, there are probably more regulations, guidelines and standards for the consideration of human factors than in most other industries. SMRs in general and non-light-water reactor (non-LWR) designs in particular face even more challenges. For example, due to the emphasis on new technologies, higher levels of automation, new function allocations and the quest for minimal staffing and lower O&M costs, designers need to cope with a large number of rules, regulations, standards and guidelines. These include, for example, codes of federal regulations like 10 CFR 50.54, the Standard Review Plan NUREG-0800, the requirements for the review of an organisation’s HFE program (NUREG-0711), guidelines for HSI design, such as NUREG-0700 (‘Human-System Interface Design Review Guidelines’), ISO 11064 (‘Ergonomic Design of Control Centres’) and IEC 60964 (‘Nuclear Power Plants — Control Room Design’), and requirements for the integration of HFE in other engineering processes as described in IEEE 1023 (‘IEEE Recommended Practice for the Application of Human Factors Engineering to Systems,

Equipment, and Facilities of Nuclear Power Generating Stations and Other Nuclear Facilities’). There are also many regulations dealing with occupational health and safety, building regulations, and several more.

All of these regulatory and best practice expectations lead to probably the biggest challenges that SMR designers face: the integration of human factors in the systems engineering process throughout the project life cycle. Because a large part of advanced reactor design would be FOAK engineering, human factors engineers need to cope with many organisational, technical, regulatory and methodological questions that are new to the nuclear industry.

Because of all of these requirements, we can expect as much, if not more, regulatory oversight for advanced NPP design projects in the form of regular, mandatory audits, quality and safety management requirements and intensive verification and validation of, for example, HSI designs, human performance with operating procedures, and so on. Regulators will also expect to see evidence of human factors work in provisions made to protect the public, workers and the environment. This includes, for example, attention to situation awareness, safety culture, human reliability, workload and performance shaping factors. All of this contributes to long lead times, not only for engineering and design, but particularly for the licensing processes.

It should also be emphasised that HFE practitioners in the nuclear industry are not excluded from scrutiny and public opinion. It is thus also important for HFE people to show how they contribute to safety to counter misconceptions about hazards (opinions often based on incomplete or out-dated information, or disinformation by antagonists).

The IRIS consortium designed the SPES3 integral test facility at SIET, Piacenza, Italy. Since fairly complete data are publicly available (Carelli et al., 2009), it will be used to illustrate some of the design choices. The facility is characterized by full height and prototypic fluid properties (pressure, temperature, pressure drops). Volume ratio is 1:100; area ratio is the same, 1:100, to maintain the same residence times.

The facility simulates the primary circuit (integral vessel), secondary loop, and the containment system of the IRIS reactor. Comparison of additional selected parameters between the IRIS and SPES3 is shown in Table 8.4. Table 8.4 also indicates which IRIS systems are being represented. A 3-D view of the SPES3 facility is shown in Figure 8.1, while the SPES3 pressure vessel is shown in Figure 8.2 (Carelli et al., 2009). Further details are available in (Carelli et al., 2009).

S. Boarin1, M. Mancini1, M. Ricotti1, G. Locatelli2 1 Politecnico di Milano, Milan, Italy; 2University of Lincoln, Lincoln, UK

10.1 Introduction

A description of the economic and industrial potential features of small modular reactors (SMRs) was given in 2010 by the US Secretary of Energy (Chu, 2010):

‘[…] Small modular reactors would be less than one-third the size of current plants. They have compact designs and could be made in factories and transported to sites by truck or rail. SMRs would be ready to ‘plug and play’ upon arrival. If commercially successful, SMRs would significantly expand the options for nuclear power and its applications. Their small size makes them suitable to small electric grids so they are a good option for locations that cannot accommodate large-scale plants. The modular construction process would make them more affordable by reducing capital costs and construction times. Their size would also increase flexibility for utilities since they could add units as demand changes, or use them for on-site replacement of aging fossil fuel plants. […] These SMRs are based on proven Light Water Reactor technologies and could be deployed in about 10 years’.

The goal of this chapter is to present the most relevant economic and competitive aspects related to the SMR concept.

The first priority should be the development of international nuclear codes and standards that can be adopted and referenced by sovereign licensing authorities. The NRC database of codes and standards can be used as a starting point for this effort. Also it is crucial for each country with existing nuclear power and every embarking nation with nuclear power aspirations, to participate in the development of consensus nuclear standards that may provide a solid foundation for agreement on an international strategy and framework for licensing.

11.4.1 International regulatory guidance for SMRs

IAEA and member states recognize the need to develop international guidance on how to license new SMR designs more effectively. IAEA recommended in its 6th INPRO Dialogue Forum, ‘Licensing and Safety Issues for Small and Medium — Sized Reactors’ (29 July-2 August 2013), that guidance on the application of a graded approach for the licensing and regulation of SMRs be developed. All sovereign nations embarking on SMR deployment must have a licensing process and capabilities that provide reasonable assurance that the operation of a nuclear power plant in the country will be safe and secure. This process must be open and transparent to all stakeholders. However, licensing capabilities to fully review the safety basis may be limited in many countries. Therefore, IAEA guidance on the appropriate grading of the scope and depth of a licensing safety review, particularly with a previously approved/certified SMR, is a high priority. Licensing authorities should not and need not ‘reinvent the wheel’ in the safety review of a previously approved reactor. Licensing authorities must review the complete license application, but the scope and depth of a reactor safety review may be graded based on considerations such as:

• margins of safety in the design;

• unique safety considerations of the site and region;

• operational interaction of the SMR with other industrial applications at the site;

• risk to public at remote locations;

• interactions with the approving regulatory authority to better understand and assess the licensing basis; and

• interactions with the SMR manufacturer to better understand and assess safety, quality, and ITAAC issues.