The Central Argentina de Elementos Modulares (CAREM) design has been under development for a number of years and was first introduced at an IAEA conference in 1984. The Comision National de Energfa Atomica (CNEA) drew on its experience in the design of research reactors to develop CAREM with the primary goals of enhanced safety and reduced costs. It is a simplified integral design with natural circulation of the primary coolant. The prototype design has a capacity of 27 MWe and commercial-sized units are expected to have capacities of 100-300 MWe with forced circulation being employed for units greater than 150 MWe. The design uses self-pressurization, i. e. it does not use sprays or heaters to maintain normal system pressure. The core is composed of 61 hexagonal fuel assemblies with 108 fuel pins per assembly and 25 control rods, which are positioned using internal hydraulic control rod drive mechanisms. The primary coolant flows over the shell-side of 12 compact helical coil steam generators. On the secondary side, the feed water and output steam are collected into a common set of annular pipes that surrounds the reactor pressure vessel.

The CAREM design uses passive safety systems and anticipates a post-accident grace period of 36 hours without operator action or power. Several test facilities have been built to test novel components such as the hydraulic control rod drive mechanisms and validate the safety analysis methods. The CNEA received the support

of the Argentinian government in 2009 to construct the prototype plant adjacent to the existing Atucha nuclear site. Plant construction began in 2014. Key parameters and a representative graphic for the CAREM design are given in Figure 2.1. [1]

For the reader interested in small modular reactors (SMR), this Handbook provides a thorough and authoritative introduction to today’s hottest new development in nuclear plant design and deployment. Building on the success of the large nuclear plants, SMRs offer the potential to expand the use of clean, reliable nuclear energy to a broad range of customers and energy applications.

The early commercial nuclear power reactors designed and built from the 1950s to the mid-1970s were low-power plants (up to a few hundred megawatts) and were built to demonstrate the commercial viability of nuclear energy. These plants were comparable to their fossil-fuelled counterparts, both in output and construction time (a few years). They were moderately successful; however, their unit capital costs ($/kW) were substantially higher than for comparable fossil plants. As the nuclear plant cost (numerator) kept increasing to improve performance and safety, it became necessary to also increase the output power (denominator); thus the plant size increased rapidly from a few hundreds of megawatts to nearly 2000 MWe today. Such a drastic increase had several effects: only a few manufacturers, either large conglomerates or state-owned enterprises, remained in operation worldwide; plant costs became stratospheric, creeping into tens of billions of dollars; and the time from contract signing to initiating power production exceeded a decade.

Thus, looking to the past for help in finding the answers to a troubled present, the SMRs have become the nuclear power version of Back to the Future. Started in the 1990s, new SMR designs emerged worldwide and have gained increasing momentum in the new millennium. The new small plants have several traits in common with earlier designs, such as: size (from tens to a few hundreds of megawatts), relative simplicity, and a reasonably short construction time. Also, SMRs can cover a wide range of applications and deployment times. Those proposed for power producing applications in the short term are designs of the light water reactor (LWR) type, while SMRs best suited for other applications such as fuel breeding and waste burning employ different coolants and are deployable over the long term.

The developers of SMRs, even the near-term LWRs, are for the most part quite different from the large LWR manufacturers. They include smaller sized manufacturers as well as new enterprises. For example, the two leading SMR vendors in the United States are currently Generation mPower, a subsidiary of Babcock and Wilcox, which is an established reactor manufacturer but no longer active in the large LWRs market, and NuScale Power, an entirely new enterprise. Both of these vendors were recently selected by the US Department of Energy to receive major federal funding to facilitate licensing of their SMR designs. Holtec International, which is an established company but novel to the reactor design arena, is also working to bring its SMR product to the market.

This Handbook is composed of 20 chapters structured into four parts, each chapter being authored by a known expert in the field.

• Part One (Fundamentals of small modular nuclear reactors) provides a comprehensive introduction to SMR technologies, existing commercial designs, and fundamental design strategies. The three authors contributing to this section have been eminent proponents of SMRs since the 1990s and have led the development of integral pressurized water reactor (iPWR) designs, which is the prevailing design strategy for SMRs and the focus of this Handbook. Part One is articulated over three chapters: 1. Small modular reactors (SMRs) for producing nuclear energy: an introduction 2. Small modular reactors (SMRs) for producing nuclear energy: international developments and 3. Integral pressurized water reactors (iPWRs) for producing nuclear energy: a new paradigm.

• Part Two (Small modular nuclear reactor technologies) reviews the key technologies which are fundamental to the iPWR design, focusing on what is new and different, while also providing insight on potential opportunities and challenges. Six chapters individually address the following technologies: reactor core and fuel; key reactor system components; monitoring and control; instrumentation and control technologies for small modular reactors; human-system interfaces; safety; and proliferation resistance and physical protection. The six authors of this part are internationally recognized authorities in their field and are not associated with any of the current iPWR designs.

• Part Three (Implementation and applications) addresses four key areas critical to successful deployment of SMRs: economics and financing; licensing and manufacturing methods; hybrid energy systems using SMRs. As was the case in Part Two, the four authors of Part Three are recognized authorities in their field.

• Part Four (International R&D and deployment) provides an overview of the worldwide deployment of SMRs. The first six chapters focus on countries that are most active in the development and deployment of SMRs: the United States, the Republic of Korea, Argentina, the Russian Federation, China, and Japan. The authors are directly involved in their country’s activities. Finally, the last chapter addresses how SMR development and deployment can represent a key contribution to the growth of developing countries. It is a reminder that SMRs promise to be not only a better, more economic machine, but also promote improved living conditions and quality of life.

It is hoped that this Handbook will be useful to those with a general interest in SMRs, as well as to those looking for more specific information. It is further hoped that this Handbook will serve as a guide, through its copious references, to further learning.

Mario D. Carelli and Daniel T. Ingersoll

In 2010, Holtec International introduced their entry into the SMR competition: the 140 MWe Small Modular Underground Reactor (HI-SMUR). Although not a traditional reactor vendor, Holtec assembled a diverse team led by the newly created subsidiary, SMR LLC, to develop the HI-SMUR design. In late 2011, several design changes were made: the external steam generator system was changed substantially from a two-stage horizontal arrangement to a vertical arrangement, the reactor vessel was shortened by approximately 40%, and the power was increased to 160 MWe.

Figure 2.12 SMR-160 (United States) — SMR, LLC (HOLTEC International) © HOLTEC International SMR, LLC.

Still identified as HI-SMUR, it is more frequently being referred to as SMR-160 in recent press releases.

The SMR-160 is a compact loop configuration since the steam generator, and hence a portion of the primary coolant, is external to the reactor vessel. As an additional modification to a classic loop-type design, the SMR-160 pressurizer is located at the top of the steam generator vessel. The reactor pressure vessel has a significantly higher height-to-diameter aspect ratio compared to other SMRs to enhance the natural circulation of the primary coolant. The core uses full-height traditional 17 X 17 pin array PWR fuel assemblies and the core is expected to be refueled as a single cartridge that will be discharged to a below-grade spent fuel pool for cooling before being moved to a dry storage facility. A SMR-160 plant is currently planned to include a single reactor unit. Like most contemporary SMR designs, the reactor vessel and spent fuel pool will be entirely below grade level. Key parameters and a representative graphic for the SMR-160 design are given in Figure 2.12. [10]

The principal mission adopted for commercial SMRs has been the generation of electricity. All reactor coolant types address this mission. For those plants designed

Table 1.2 Reactor characteristics by coolant Coolant PWR1 Light water BWR2 Light water HTGR Helium3 Helium4 SFR5 Sodium Lead6 LFR Lead-bismuth7 Power (MWt/MWe) 530/180 750/250 250/100 625/283 840/311 700/300 280/101.5 Power density (kWt/liter core) 69 39.5 3.2 6.8 215 116 160 Specific power (kWt/kg HM) 26.8 11.6 89.7 ~120 83.6 14.5 30.8 Fuel geometry Rods Rods Pebbles Prismatic graphite blocks Rods Rods Rods Fuel material/cladding UO2/Zr-4 UO2/Zr uo2/triso UCO/TRISO (U+Pu)/SS (U+Pu)N/SS UO2b/8 Primary system temperature inlet/outlet (°C) 295/319 190/285 250/750 325/750 360C/499C 420/540 340/490 Primary operating pressure (MPa) 14.2 6.9 7.0 6.0 0.1 0.1 0.1 Secondary operating pressure (MPa) 5.7 NA 13.3 16.7 14.7 18 6.7 Plant thermal efficiency (%) 34 33.3 42 45 37 43a 36.3 1 Pers. Comm, D. Langley (mPower) to N. Todreas (MIT), Jan 2013. 2 VK-300 — Gabaraev et al. (2004); Kuznetsov et al. (2001). 3 HTR-PM — Zhang et al. (2009); Zhang (2012). 4 SC-HTGR — AREVA (2012). 5 PRISM — Triplett et al. (2012). 6 BREST — Smirnov (2012); Glazov et al. (2007)a. 7 SVBR-100 — Toshinsky and Petrochenko (2012); MOX and N fuel options proposed13. 8 Likely EP823 or EP450. NA — Not applicable since the BWR only has a primary system. Numerical values of characteristics are rounded.

Table 1.3 Reactor coolant properties of significance1 Coolant Water2 Helium3 Sodium4 Lead4 Lead-bismuth4 (0.445Pb-0.555Bi) PWR BWR Atomic weight 18 4 23 207 208 Phase change at 1 atm Melting point (°C) 0 NA 98 327 124 Boiling point (°C) 100 -267 892 1737 1670 Density, p (kg/m3) 704.9 754.7 3.54 880 10536 10180 Specific heat, cp (J/kg K) 5739 5235 5191 1272 147 146 Heat capacity, pcp (MJ/m3 K) 4.05 3.95 0.018 1.07 1.55 1.49 Heat transfer capability 0.543 0.585 0.31 66 15 15 Thermal conductivity, к (W/m K) 3.80 1.90 0.65 18.1 2.81 2.75 Heat transfer coefficient (X10-4) h (W/m2 K) Dynamic viscosity (X104), р (kg/ms) 0.846 0.945 4.0 2.6 20 15 Kinematic Viscosity (X107), v = р/р (m2/s) 1.20 1.26 1.13 2.95 1.91 1.47 Thermal expansion coefficient 326 250 — 29 11 13 (X105), a (1/°C) Prandtl number, Pr 0.89 0.85 0.66 0.005 0.020 0.015 1 Typical reactor values. 2 Property values at PWR average and BWR inlet conditions from Todreas and Kazimi (2012). 3 Property values at 537 0C and 6 MPa from Petersen (1970). 4 Property values at 450 0C from Hejzlar et al. (2009).

to be deployable to remote locations, whether placed terrestrially or dispatched as barge-mounted reactors, the added co-generation capabilities for desalinization and district heating exist. Of the water-cooled SMRs the Russian PWR and BWR systems have been designed for these additional missions. Additionally, propulsion as accomplished by Russian ice-breaker vessels using the KLT-40S reactor and its planned replacement, the RITM-200 reactor, is a further reactor mission.

The helium gas-cooled reactor can operate at high enough outlet coolant temperature, 750 °C in initial designs, to provide a process heat capability. This process heat can be used directly for various industrial processes such as shale oil recovery and the production of hydrogen by relatively high-temperature thermochemical cycles. Hydrogen production from water by electrolysis can be accomplished at the lower outlet temperature of the sodium — and lead-cooled reactors, on the order of 500­550 °C, but these SMRs have not embraced this mission due to current shrinking US interest.

The US Advanced Liquid-Metal Reactor (ALMR) program in the 1980s resulted in the design of the 160 MWe Power Reactor Inherently Safe Module (PRISM) sodium — cooled reactor. The PRISM design was one of the first advanced reactor designs to employ significant use of passive safety features and was designed as a power module to be used in multiple three-unit packs to form a large electrical-capacity power plant. The PRISM design was originally intended to be a breeder reactor for improved uranium resource management, but more recently has become focused on recouping the unused energy content in discharged LWR fuel and also to consume the very long-lived higher actinide elements that dominate the long-term hazard in a geologic repository.

General Electric, now teamed with Hitachi, has resumed development of the 311 MWe PRISM design, although the name has been changed to Power Reactor Innovative and Small Module. The design has similarities to LWR-based integral

Figure 2.22 PRISM (United States) — General Electric-Hitachi (GEH) © GE Hitachi Nuclear Energy.

designs except that the internal steam generators are replaced with intermediate heat exchanges that transfer heat to two secondary sodium loops. External secondary heat exchangers are coupled to a supercritical Rankine power-conversion unit. The design uses U-Pu-Zr metal fuel and can accommodate actinide waste products from LWR spent fuel. Four electromagnetic pumps circulate the sodium coolant in the pool-type primary system, which operates at nearly atmospheric pressure.

The current deployment strategy is to couple two PRISM modules into a single ‘power block’ with a shared turbine-generator. One or more power blocks would be co-located with a small electro-refining fuel recycle facility. Although there was significant regulator review of the PRISM design earlier, there are no immediate plans to license PRISM for commercial power production. Key parameters and a representative graphic for the PRISM design are given in Figure 2.22. [21]

The ACP-100 is an integral pressurized water reactor being developed by the Nuclear Power Institute of China (NPIC) for the China National Nuclear Corporation (CNNC). The design draws heavily on technology developed by CNNC for the larger LWR plants, including the CNP-600 and the ACP-600/1000. The forced circulation of the primary coolant is driven by externally mounted reactor coolant pumps. Also external to the reactor pressure vessel are the control rod drive mechanisms and the pressurizer. The core comprises 57 partial-height CF2-type fuel assemblies. Reactivity is controlled using control rods and boron shim dispersed in the primary coolant. The reactor vessel and other primary system components are contained in a traditional large-volume containment structure that is 29 m diameter and 45 m tall. Safety-grade batteries provide backup power for up to 72 hours in the case of a station blackout and sufficient water is provided in the spent fuel pool to allow 7 days’ grace period before fuel is uncovered.

The 100 MWe design is intended to provide electricity and also process steam for water desalination, district heating and industrial applications. It is primarily intended for inland locations within China as a replacement for or alternative to

Reactor vessel

Figure 2.2 ACP-100 (China) — China National Nuclear Corporation (CNNC) © F. Zhong.

coal-fired plants. Approval for the design project was given in 2010 and the budget for the construction of two ACP-100 modules was approved in 2011. In July 2012, Hunan Hengyang city government signed an agreement of cooperation with CNNC for the potential deployment of ACP-100 units in Putian, Fujian. Approval of the preliminary safety analysis report is expected in 2014. Key parameters and a representative graphic for the ACP-100 design are given in Figure 2.2. [2]

N. Todreas

Massachusetts Institute of Technology, Cambridge, MA, USA

1.1 Introduction

Just what are small modular reactors (SMRs)? This question is first answered simply along with a brief history of the evolution of this class of reactors. Subsequent sections detail the incentives and challenges to achieving successful commercial deployments, the different types of SMRs based on coolants employed, and, finally, the current status and future trends in the worldwide effort to develop and deploy this reactor type.

1.1.1 Defining SMRs

‘Small’ refers to the reactor power rating. While no definitive range exists, a power rating from approximately 10 to 300 MWe has generally been adopted. The minimum rating assures that the reactor delivers power suitable for the practical industrial application of interest. The maximum rating constrains these designs to power levels at which the expected advantages of serial production and incremental deployment as well as the match to electric grid siting opportunities and constraints can be realized.

‘Modular’ refers to the unit assembly of the nuclear steam supply system (NSSS) which, when coupled to a power conversion system or process heat supply system, delivers the desired energy product. The unit assembly can be assembled from one or several submodules. The desired power plant can then be created from one or several modules as necessary to deliver the desired power rating. Importantly the deployment of modules can also be sequenced over time both to match regional load growth and to levelize the timing of capital spending over a prescribed time horizon. Construction of the plant by assembly of factory-built elements or modules is the technique of modular construction. Although it is an integral part of the construction strategy envisioned for all SMRs, this technique is not uniquely applied to SMRs. Rather, it is now being employed for relevant construction elements of nuclear power plants of all power ratings, although the modules for large plants are considerably different in size, not typically amenable to rapid assembly as is being proposed for SMRs.

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Copyright © 2015 Elsevier Ltd. All rights reserved.

‘Reactor’ is a term more broadly applied to vessels in which all manner of chemical processes are conducted. However, in our case, reactor refers to a system in which a controlled nuclear fission process is conducted.

In 2011, Westinghouse introduced a new SMR design, the W-SMR. It is rated at 800 MWt with an expected electrical capacity of greater than 225 MWe. The W-SMR design draws heavily from Westinghouse’s AP1000®[1] design experience, although it uses an integral primary system configuration rather than the traditional loop-type configuration used for the AP1000® plant. Except for eight external, horizontally mounted primary coolant pumps, all primary system components are contained within the 3.5 m diameter by 24.7 m tall reactor pressure vessel, including the control rod drive mechanisms, pressurizer, and steam generator. The reactor core is composed of 89 fuel assemblies with 17 X 17 fuel pins and an active core height of 2.4 m. In addition to the 37 control rods, soluble boron in the primary coolant is used for reactivity shim. The internal control rod drive mechanisms are based on the AP1000® plant magnetic jack units. The steam generator is a recirculating once-through, straight-tube bundle that surrounds the central hot leg riser. The low-quality steam produced in the steam generator is passed to a steam dryer drum mounted outside the containment vessel.

The reactor pressure vessel is contained within a 9.8 m diameter by 27.1 m tall steel containment vessel. The passive safety systems and residual heat removal systems are modeled closely after the AP1000® plant systems and are expected to provide a 7-day post-accident grace period without operator action. A W-SMR plant consists of a single reactor, although multiple plants can be constructed at the same site.

Ameren-Missouri has announced the intent to build up to six W-SMR plants at the existing Callaway Nuclear Generation Station site in Missouri. Key parameters and a representative graphic for the W-SMR design are given in Figure 2.13. [11]

Control rod drive mechanisms

Reactor coolant
pumps

Reactor core

Figure 2.13 W-SMR (United States) — Westinghouse Electric Company (used with permission of Westinghouse Electric Company LLC).

Certainly this criterion is best met by reactor concepts using conventional components and systems operating at coolant temperatures and pressures within the envelope of significant operating experience. Water as a coolant for SMRs has been selected explicitly because of the satisfaction of these conditions. Experience with water reactors using the essential design features selected for water-cooled SMRs goes back to the beginnings of the nuclear electricity generation and propulsion age. The major caveat regarding the achievable reliability of water-cooled SMRs relates to those having selected the integral configuration, the placement of all NSSS components and piping within a single pressure vessel. While the Otto Hahn merchant vessel successfully used this reactor configuration and operated commercially for nine years, the potential reduction in operational reliability of this configuration due to its limited accessibility for primary system component monitoring, maintenance and repair can be confidently assessed only through many more years of operating reactor experience.

Sodium-cooled reactors have generally had a mixed, albeit limited, record of operating experience. The US Experimental Breeder Reactor II (EBR-II) and British Dounreay Fast Reactor (DFR) records were exemplary, the Russian BOR-60 and BN-600 and the French Phenix reactor experience was on balance satisfactory, while the Japanese Monju experience has been very troubled, principally due to a sodium leakage event as was the Superphenix experience. Similarly the lead-bismuth — cooled Russian submarine reactors operated reliably but with the need for careful attention to coolant chemistry control and freeze prevention after the major accident in 1968 before adequate understanding existed of the need for rigorous control of coolant oxygen concentration to prevent lead oxide slag formation (Toshinsky and Petrochenko, 2012). Helium-cooled reactors, e. g., the experimental reactors AVR and THTR in Germany and the commercial Fort St. Vrain unit in the US, also have had a mixed operating record.

Hence it can be concluded that, based on operating experience, the water-cooled SMR class has a significant advantage over the other coolant types with regard to its promise of operational reliability. The operational reliability of non-water-cooled reactors will be uncertain until sufficient demonstration plant operational experience is accumulated.

The principal coolant characteristics influencing this operational experience — e. g., coolant toxicity, corrosion effect on bounding surfaces, and coolant freezing and boiling temperatures — are shown in Table 1.4. Coolant toxicity has been expressed in terms of radiological, biological and chemical factors.

Biological consequences arise from decay of 210Bi which yields 210Po. The polonium then chemically combines with lead as PbPo(s). Should water enter the primary system due to a failure of the ingress penetration barrier coincident with a steam generator tube leak, it would react with the PbPo(s) to produce H2Po(g), a volatile alpha-emitting aerosol of biological inhalation concern. The designers of the lead-bismuth-cooled SVBR-100 reactor (see Table 1.2), who are well versed in Russian submarine experience, cite that operating experience has resulted in the development of measures for providing adequate radiation safety. For water-cooled reactors, water chemistry measures typically include introduction of boron and lithium in the form of boric acid and lithium hydroxide for corrosion control, although some SMRs, e. g., the B&W mPower design, have eliminated the use of soluble boron for reactivity control. Neutron activation of 6Li and 10B produces tritium, 3T, albeit in small quantities, which nevertheless is a biological hazard if ingested.

Occupational contact hazards of a chemical nature exist for lead through high levels of exposure due to inhalation and occasionally skin contact. Similarly asphyxiation due to accidental immersion in helium (or in nitrogen typically used to inert BWR containments) is a potential hazard. The more significant, well-recognized chemical oxidation reactions of zirconium cladding and sodium are covered as a safety concern under potential energy release in Section 1.4.1.

Of all the coolants, helium, because it is an inert gas, poses the least corrosion potential, and its activation is minimal as demonstrated by the Fort St. Vrain experience that showed very low activity in the coolant compared to light-water reactors. The aggressive attack of lead and lead-bismuth on metal cladding (e. g., in HT-9 and the Russian equivalents EP 823 and EP 450) has forced the limitation of coolant velocity in lead — and lead-bismuth-cooled core designs to 3 m/s. This in turn has necessitated the provision of a large coolant flow area to bound core coolant temperature rise. Hence lead and lead-bismuth cores have fuel pins spaced with a large pitch/diameter square lattice array. However, recent development (Short and Ballinger, 2012) of a composite material for cladding and structural application may mitigate such limitations.

Finally, the operability of liquid metal coolant systems requires trace heaters around piping and components of sodium, lead and lead-bismuth reactors to prevent coolant freezing when insufficient heat is available from power operation or decay heat. The high freezing temperature of lead, 327 °C, compared to the modest values for sodium, 98 °C, and lead-bismuth, 125 °C, renders lead disadvantageous as a reactor coolant in this regard. However, with these high freezing temperatures both

Table 1.4 Inherent coolant characteristics affecting operational reliability Water1 Helium Sodium2 Lead2 Lead-bismuth2 Radiological 16O(n, p)16N 16N^16O + 5 to 7 MeVr (T1/2 = 7.1 s) None but erosion created dust liftoff from sudden depressurization can cause mechanical clogging 23Na(n, r)24Na (Tm = 15 h) 1.38, 2.76 MeVr s 23Na(n, 2n)22Na (T1/2 = 2.6 yr) 1.28 MeVr 204Pb(n, r)205Pb (T1/2 = 51.5 days) 1.28 MeVr Same as lead plus 209Bi(n, r)210Bi(e) 210Po 210Po (a, r low prob.) 206Pb (T1/2 = 138 days) 5.3 MeV a; 805 keV r Toxicity Biological 6Li(n, a)3T 10B(n, 2a)3T 10B(n, a)7Li(n, na)3T (T1/2 = 12.3 yrs) None None Trace amounts of Po from 205Pb to 210Po by neutron capture and b- decay PbPo(s)+H2O=PbO+H2Po(g) (volatile alpha-emitting aerosol) Chemical None Asphyxiation hazard None Exposure to high levels of lead through inhalation, ingestion or occasionally skin contact can lead to the medical condition known as lead poisoning Same as for lead Corrosion Prevention of stress corrosion cracking of stainless steel requires significant attention. Also significant corrosion- induced crud formation potential None Sodium is practically non­corrosive with respect to stainless steel. Corrosion is lower than for lead or water Aggressive corrosion by: • direct dissolution by a surface reaction • intergranular attack. Oxide film formation tends to inhibit the corrosion rates. Need to limit velocity to about 3 m/s to avoid cladding corrosion. Same as for lead Melting (freezing)/ Boiling points (°C) 0/100 NA 98/883 327/1737 High freezing temp — need trace heating 125/1670 Lower freezing temperature advantageous vs lead 1Lin (1996). 2Todreas et al. (2008).

lead and lead-bismuth eutectic will solidify in ambient air, providing a means for sealing small leaks in the primary coolant boundary. On the other hand, the high boiling points with the attendant low vapor pressures of these liquid metal coolants allow reactor operation at atmospheric pressure without the source of stored energy associated with a high-pressure coolant. Operation at low pressure allows reduction of the required thickness of the pressure vessel and other primary pressure boundary components. Nevertheless, for the heavy lead coolant the dimensioning of these vessels must be carefully evaluated to satisfy seismic design criteria.

M. D. Carelli

Formerly of Westinghouse Electric Co.,Pittsburgh, PA, USA

3.1 Introduction

Over 60 years ago nuclear power has been ushered into mankind, representing a quantum change which rivals and potentially surpasses those of the combustion engine and electricity. It has advantages and disadvantages, proponents and detractors, just like any other human endeavor. The underlying puzzle is why other innovations have successfully overcome — in a reasonably short time — the inevitable initial distrust, setbacks and potentially suffocating legislation, while nuclear power has not yet flourished.

True, nuclear weapons are not exactly the best introduction to nuclear energy, but that potential connection was overcome in the span of a decade; the advanced world economies accepted nuclear power, development plans were outlined and implemented with a variety of prototypes followed by a variety of progressively improving designs. So, the first couple of decades were not that different for nuclear power than they were for previous endeavors, except that nuclear power progressed at a more sedate pace, as would be expected because of the much higher financial exposure, and of course the drastically different consequences of potential accidents. However, at that time (mid-1970s), instead of becoming a staple of human development, nuclear power became a more and more controversial issue, crowned by the 1979 Three Mile Island accident. Since then it has moved in fits and starts, characterized, on one side, by the need for economical and reliable power and, on the other, by three major accidents, plus a variety of minor or nuisance-type, but highly publicized, ones and a substratum of polarized politics. This resulted in divided public opinion, showing roughly one-third firmly pro, one-third firmly con and the remaining one-third going back and forth depending on the latest happenings.

Consequently, for the past 40 years nuclear power, while it has expanded overall, has been unable to fulfil its potential, nor to maintain its promises; it will remain so for the foreseeable future, unless it goes through a drastic change from its current modus operandi. If properly planned and executed, the catalyst for this change could be the deployment of SMRs, starting immediately with the integral PWR designs (iPWRs), whose technology and characteristics are examined in detail in Part II of this Handbook.

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