As mentioned before, designers need more formal methods to choose the most appropriate technology for a specific task. Bad decisions at the design stage are often the source of later operational problems that are very hard to fix.

One practical way to examine the viability of technology for future implementation is to consider the interaction features associated with that technology. Another way is to consider the characteristics of the work domain and the context of use. This would include the practical, economic, organisational, operational, regulatory and contextual considerations.

In examining the key entities, principles and relationships between technologies and the contexts where technologies are used, five key dimensions or contexts are identified that could serve as a framework for evaluation and deployment of HSI technologies for new designs. Figure 7.1 identifies these dimensions and specific HSI topics and elements associated with each. The five dimensions or contexts shown (human factors, technology, operations, organisation and regulations) are defined in the following sections.

V endor info Market readiness

efficiency and satisfaction with which specific users can perform specific tasks with specific systems in a given context.

Human performance can be affected positively or negatively by technology and this is determined by a number of factors: the work and environmental context within which a tool is used, the experience and skill with which a person can use the tool, the design of the tool and how usable it is in the given task context. Optimal human performance could therefore be expressed in terms of criteria such as comfort, efficiency, situation awareness, low workload, low error probability and so on. It also means that human performance can often be improved by providing some form of task support to help in reducing operator workload, reducing visual and cognitive complexity of the HSI and making information more accessible.

NIST is a one-third scale (height) facility located at OSU in Corvallis, Oregon, USA, that replicates the entire NuScale Power Module and reactor building pool.

Electrically heated, it brings the system up to operating temperature and pressure. Stability testing ensures that throughout the expected operating conditions, natural circulation is stable. Furthermore, tests validate computer models including thermal efficiency, performance and safety calculations (Reyes, 2010; Houser et al., 2013). The stainless steel facility operates at full system pressure and temperature and simulates: reactor vessel, rod bundle, core shroud with riser, pressurizer, sump recirculation valves, helical coil steam generator, feedwater pump, containment vessel, and containment cooling pool. NIST scaling ratios are summarized in Table 8.5. A photo and schematic of the NIST containment, pool and pressure vessel is shown in Figure 8.3 (Houser et al., 2013).

ADS nozzles

Pump suction nozzles —

Steam line nozzles

SG helical coils

8.3.1 SMART integral test loop (SMART-ITL) facility

SMART-ITL located at KAERI, Republic of Korea, is a four-loop full-height test facility operating at prototypic conditions (pressure, temperature), and with the area ratio of 1:49 (Park, 2011; NEI, 2013). Its maximum power is 2 MW, or about 30% of

Reactor

pressure

vessel

Reactor

building

pool

Figure 8.3 NIST containment, pool and pressure vessel (from Houser et al., 2013 presentation) the full power scaled down by area ratio. The facility consists of a primary system (simulating integral reactor vessel) with four steam generators, a four-loop secondary system that incorporates four trains of the passive residual heat removal system, each with a heat exchanger, an emergency cool-down tank and a makeup tank, together with valves and connecting pipes. SMART-ITL also incorporates several auxiliary systems. Further information is available in NEI (2013).

The nuclear sector commonly clusters nuclear power plant (NPP) life-cycle costs as

capital cost, operation and maintenance, fuel and decommissioning.

• Total capital investment cost (or capital cost): an all-inclusive plant capital cost, or lump-sum up-front cost. This cost is the base construction cost plus contingency, escalation, interest during construction (IDC), owner’s cost (including utility’s start-up cost), commissioning (non-utility start-up cost), and initial fuel core costs for a reactor (EMWG, 2007).

• Operation and maintenance (O&M): costs inclusive of, but not limited to: (i) actions focused on scheduling, procedures, and work/systems control and optimization; (ii) performance of routine, preventive, predictive, scheduled and unscheduled actions aimed at preventing equipment failure or decline with the goal of increasing efficiency, reliability, and safety (Sullivan et al., 2010).

• Fuel cost, the sum of the costs for the fissile/fertile materials (natural uranium, low-

Handbook of Small Modular Nuclear Reactors. http://dx. doi. Org/10.1533/9780857098535.3.239

Copyright © 2015 Elsevier Ltd. All rights reserved.

enrichment uranium, highly enriched uranium, mixed oxide fuel, uranium-thorium, etc.) and the enrichment process of the fuel in fissile materials, plus other materials used in the fuel assemblies (zirconium, graphite, etc.), services required to produce the needed materials (mining, milling, conversion, enrichment, fabrication), fuel fabrication, shipment and handling, costs of spent-fuel disposal or reprocessing and waste (including low-level, high-level and transuranic waste) disposal.

• Decommissioning: costs for the administrative and technical actions taken to allow the removal of some or all of the regulatory controls from a facility. The actions will ensure the long-term protection of the public and the environment, and typically include reducing the levels of residual radionuclides in the materials and on the site of the facility, to allow the materials’ safe recycling, reuse, or disposal as ‘exempt waste’ or as ‘radioactive waste’ and to allow the release of the site for unrestricted use or other use (IAEA, 2007a).

These costs contribute in different ways to the economics of a NPP. In general it is possible to compare them using one of the most important indicator for policy makers. This indicator, usually called levelized unit electricity cost (LUEC) or levelized cost of electricity (LCOE), represents a unit generation cost of electricity, accounting for all the NPP life-cycle costs and is expressed in terms of energy currency, typically [$/KW h]. For both large reactors (LRs) and SMRs the capital cost is the main component (50-75%) of the LCOE, followed by O&M and fuel, as shown in Table 10.1. From this consideration arises the opportunity to analyse in detail the nature of the capital cost item. In accordance with the glossary provided by EMWG (2007), Table 2 lists and clusters all the main accounts included in the capital cost.

If the construction time increases, almost all the cost items, apart from the equipment, are affected by such increase. In particular, the cost items affected by a time schedule increase are as follows:

• Labour cost: on the construction site of a reactor plant, thousands of people are employed.

• Rent fees for building infrastructures (e. g. special cranes).

• Escalation: the amount of all the cost items tends to increase because of a generalized inflation mechanism; the inflation rate may specifically relate to the price dynamics of the main inputs, such as structural materials, energy, etc.

• Interest during construction: financial costs related to the capital remuneration increase with the investment duration.

In addition to the cost increase, each day of construction schedule delay represents a loss in terms of missed electricity generation and potential revenues.

Once construction and commissioning are completed, the NPP enters the operation mode. In this phase almost all the costs are fixed (Parsons and Du, 2009). A large part of the operation and fuel costs is independent of the electricity generated (fixed costs). Even if the plant has a low capacity factor, the labour cost, which is the main component of O&M costs, does not change, and neither does most of the maintenance cost.

As a consequence of the essentially fixed nuclear generation costs, the NPP manager’s interest is generally to run the plant at its target (i. e. nominal) capacity. For this reason nuclear power is most suited for base load production.

Nuclear regulators and the IAEA are participating in the Multinational Design Evaluation Program (MDEP). Thirteen nations and the IAEA participate in MDEP which is facilitated by OECD/NEA. MDEP is a program where regulatory organizations jointly cooperate in sharing information about the review of specific new reactor designs. These new designs require detailed evaluation of their safety as well as development of ITAAC for their manufacturing and construction. Since new reactor designs, including SMRs, will be exported and a portion of their systems and components may be manufactured outside the country of origin, it is important that all regulatory authorities interact to share relevant information, experience and expertise. MDEP also supports activities to facilitate the international convergence of codes, standards, and safety goals.

Reliability impacts the economic performance of a plant by increasing the capacity factor and avoiding large costs that operators face for sudden loss of energy supply (both electricity and heat). Furthermore, customers have come to expect on-demand availability of high-quality energy products. Therefore, energy systems should be judged on their ability to respond to market demands, on a minute, hourly, daily, weekly, and seasonal basis. A hybrid system must be capable of making temporal adjustments on timescales that ensure power quality standards on the grid (i. e., power factor, voltage, frequency, and phase); other hybrid system products can be selected based on the ability of the production subsystem to handle variation in production rates and schedules. One of the key advantages that a hybrid system could offer is the additional possibility of redirecting energy production to more lucrative products given market situation changes in a short time while continuing to meet grid demand. This flexibility provides a clear benefit in terms of economic performance and capability to cope with fluctuations in demand.

The DOE-NE created its NEUP in 2009 to consolidate university support under one initiative and better integrate university research within DOE-NE’s technical programs including SMRs. The objective of the program is to engage US colleges and universities to conduct R&D, enhance infrastructure at these institutions, and support student education. The IRP represents the program directed component of the NEUP program designed to address near-term, significant needs of the DOE — NE R&D programs. The IRP awards are typically for three-year periods and more significant funding amounts.

An ART-related R&D award for three years was made in 2011 with work starting in 2012 to develop the path forward to a test reactor and ultimately a commercial FHR. While not explicitly focused on a small modular reactor design, the results from this IRP will provide input to and allow the ART program to leverage these efforts on work conducted in support of the SmAHTR concept, and thus is included in this chapter. The Massachusetts Institute of Technology (MIT) is leading the three-team project and will irradiate materials in liquid salts at prototypic conditions in the MIT reactor and conduct other experiments to validate reactor models and viability [12]. The University of California at Berkeley (UCB) will conduct thermal hydraulic experiments using simulants to predict heat transfer and accident behavior of the fluoride salt, and the University of Wisconsin at Madison will undertake corrosion experiments on candidate materials of construction. MIT and UCB are to develop pre-conceptual designs of a test and commercial power reactor.

The reactor enclosure system is composed of double vessels (reactor vessel and guard vessel) and a thick flat plate of the reactor head. In this design, the reactor vessel size is 12 m in diameter and 16.5 m in height. IHTS main piping is 144 m long per loop system. The reactor system is supported by a skirt-type support structure which joints the reactor head and the reactor vessel by bolts. This will provide access holes for in-service inspection devices to inspect the reactor and guard vessels. The core support structure is a detached skirt type structure which has no welds between the core support structure and reactor vessel bottom head. This is just put on the flange forged with a reactor vessel bottom head to allow a free thermal expansion.

Once the enrichment and number of assemblies required has been determined as outlined above, the designer needs to evaluate the need for BPs in the fuel assemblies. Since sufficient enrichment (and reactivity) has to be present for the lifetime of the fuel, the excess reactivity has to be controlled, particularly during the first cycle of irradiation. In general for large PWRs and for iPWRs, the core-wide (global) reactivity during the cycle is controlled by means of either boron in the coolant, or by use of control rods. The BP content (type of material, number of rods, content in the rods) is not a design requirement, other than it assists in controlling power peaking within (fuel rod powers) and between the assemblies (assembly powers), and it assists in reducing core wide excess reactivity, thereby reducing the soluble boron concentration and resulting in a negative moderator temperature coefficient (MTC).

Materials are chosen for BPs that initially absorbs neutrons (have a high neutron capture cross-section), but upon capture, they become an isotope with a low-absorption cross-section, i. e., during irradiation, they are ‘burnable’. Examples include boron, gadolinium, erbium and dysprosium; the first two are used today on a routine basis in commercial, large PWRs and these are the most likely candidates for iPWRs also.

The absorbing material can either be mixed in with the fuel itself during the manufacture of the fuel pellets, (known as ‘integral BPs’) or can be loaded as separate components into the fuel assemblies (into the guide tubes) and so can be removed at the end of a cycle of operation (known as ‘discrete BPs’).

Boron (specifically B-10) burns out quickly because of its high absorption cross­section, whereas gadolinium, because of self-shielding effects, tends to burn out more slowly. Generally, this makes boron-based BPs more suitable for short cycles of operation, and gadolinium more suitable for longer cycles, but varying the weight percent of the poison material is used to tailor the rate of burnout required.

A comparison of the burnout rates, and extent of the reactivity hold down can be seen in Figure 4.2 for examples containing gadolinium. It can be seen that the overall reactivity doesn’t quite return to the level of the no burnable poison cases. This is due to residual absorption (albeit relatively small) from the remaining gadolinium in the fuel. For fuels that contain boron as the BP, such as IFBA (which is an ‘integral fuel burnable absorber’ technology where a boron coating is sprayed onto the outside of the fuel pellets), there is no residual absorption in the fuel.

Other performance considerations for the designer may include helium build-up (a result of neutron capture in B-10), or thermal conductivity, both of which can limit the fuel performance. For example, gadolinium has a lower thermal conductivity that uranium, and so the fissile enrichment of the carrier material for the gadolinium poison is deliberately lowered to avoid power peaking concerns.

For those iPWRs that are looking at either very long cycle lengths (of the order of a few years), or those that need a much longer lasting reactivity hold down (for example, those that do not use boron in the coolant), erbia is another contender. Erbia has a relatively low absorption cross-section compared with boron-10 or gadolinium, which means it depletes slowly. In addition, all of the isotopes of erbia have reasonable absorption cross-sections, which means the isotopes produced by capture also have a reactivity hold down effect.

The designer has to consider the magnitude of the reactivity hold down, the duration, and the rate of depletion. Combining the number of BP rods, weight percent of the BP in those rods, as well as the BP material type gives the designer sufficient degrees of freedom to achieve the desire outcome. Examples of indicative gadolinium pin locations are provided in Figure 4.3 by way of illustration. Note that the BPs are loaded near the water holes (instrument and guide thimbles) as additional thermalization of neutrons improves their effectiveness.

In many cases the NSSS pressure signals are generated by the safety system instrumentation. Pressurizer pressure and main steam pressure are two types of signal that come to the NSSS control system through isolators from the safety system. Feedwater pressure, on the other hand, is an example of a non-safety signal that is dedicated to the NSSS function.

In the case of feedwater pressure and other NSSS dedicated sensed parameters, the iPWR designer has the flexibility to use newer designs and technologies as the evolution pedigree for safety system qualification is not required. But because some of the NSSS signals are located in accessible places with typically more space than the safety system devices, options to use the traditional sensors can be considered.

Many of the traditional suppliers of pressure sensors have modernized the signal processing (to digital in most cases) and are modifying their products with state-of — the-art processing methods. They are transitioning toward digital processing whenever possible. Also many I&C designers are using wireless set-up options whenever possible. (A discussion on wireless vs. wired solutions is in Section 6.8.3.) These advances within the traditional methods, give the traditional sensor manufacturers an advantage in the NSSS instrumentation market.

The new technology options for NSSS pressure sensors/transmitters are covered under Section 6.2.2.

Many of the advantages mentioned for outage control centres also apply to the EOF and the TSC. The EOF is normally outside the perimeter of the power plant and it serves as the management and coordination centre for the emergency staff that will operate from there in the event of an emergency at the plant. Advanced HSIs will help to manage information about important plant parameters and radiological conditions in the plant and its immediate surroundings.

The TSC is an on-site facility located close to the control room — according to NUREG-0696 (1981), the TSC must be located within a two-minute walk of the main control room. During upset and emergency conditions it provides technical support to plant management and the reactor operating personnel located in the control room. Advanced HSIs with diagnostic features will be important here too to help TSC personnel analyse the plant conditions before and throughout the course of an accident.