In This Issue
Fall Issue of The Bridge on Nuclear Energy Revisited
September 15, 2020 Volume 50 Issue 3
The desire to reduce the carbon intensity of human activities and strengthen the resilience of infrastructure key to economic prosperity and geopolitical stability shines a new spotlight on the value and challenges of nuclear energy.

Why the Unique Safety Features of Advanced Reactors Matter

Wednesday, September 16, 2020

Author: José N. Reyes Jr., Finis Southworth, and Brian G. Woods

Over the past two decades, significant efforts have been devoted to creating a new paradigm for the fabrication and deployment of nuclear power plants. These efforts include development of a variety of reactor designs aimed at increasing efficiency, flexibility, and safety.

From the standpoint of safety, an advanced reactor (AR) is defined as a nuclear reactor that, for all design basis accidents, ensures no offsite consequences and does so without requiring operator actions, AC or DC power, or the addition of coolant for an unlimited duration. Among these are small modular reactors (SMRs), factory-fabricated reactors that produce 300 MWe or less, enabling reduced construction times and more competitive overnight capital costs.

New SMR designs come in a variety of configurations, from a single small reactor to a multimodule plant. Some use different coolants, moderators, and fuels than the existing light water reactor (LWR) fleet. One of the most appealing features of the new designs is their unique safety features that enable off-grid operation for new applications and greater resilience to environmental impacts of climate change.

Quantifying the Safety of Nuclear Reactors

All nuclear reactor designs must satisfy three fundamental safety functions in the event of a significant abnormal event: stop the fission chain reaction, ensure adequate cooling of the nuclear fuel, and prevent the release of radioactivity into the biosphere. Nuclear reactors are designed with intrinsic safety features and engineered systems that are deterministically proven to achieve these three safety functions for specific scenarios.

Because the number of possible scenarios and failure modes is very large and continuously evolving, the need for a more general quantification of nuclear ­safety was identified shortly after deployment of the first US commercial nuclear power plants in the 1960s. In 1975 the US Nuclear Regulatory Commission (NRC 1975) published the first probabilistic risk assessment (PRA) for nuclear power plants in its pioneering WASH-1400 report. PRA methods have ­greatly improved since then, but the fundamental PRA approach to quantifying nuclear plant safety remains the same. It is applied to all nuclear power plants in the United States and helps safety analysts identify potential areas for improvement in the plant design.

The PRA approach provides a quantitative measure of the risk of unwanted consequences. The magnitude of the calculated risk can then be interpreted as a measure of nuclear plant safety. In simplest terms, risk is the frequency of an event times the consequences per event:

Risk = (Event Frequency) × (Consequences / Event)

For US nuclear power plants, the core damage frequency (CDF) is the figure of merit for the first level of PRA. The NRC’s qualitative safety goals for all new reactors require that the CDF not exceed 10−4 events per reactor-year. That is, for new designs, a core damage event must not occur with a frequency of more than once every 10,000 reactor-years.

The safety features of advanced reactors are designed to reduce both the frequency of core damage events and their consequences beyond those of the existing nuclear fleet. The DOE-sponsored international effort for ­Generation IV reactors took this concept further by considering whether prevention of core damage required offsite power, onsite emergency AC power, or even DC power (DOE 2002). At the time, only one concept required no power to prevent core damage and that was the very high temperature gas-cooled reactor.

In this paper we focus on AR technologies of relatively higher maturity such as LWR-based SMRs, high-temperature gas-cooled reactors (HTGRs), and sodium-cooled fast reactors (SFRs).

Early Implementation of New Safety Criteria and Designs

After the Fukushima core damage of March 2011, the American Society of Mechanical Engineers (ASME 2012) called for reactors to meet new safety criteria that would ensure no social impact and obviate the need for significant land withdrawal due to an accident. The report has not gained traction in the US regulatory community. However, a NuScale SMR  design sub­mitted for certification largely follows the ASME safety strategy, HTGR designers are endeavoring to ensure “no social impact,” and the Next Generation Nuclear Plant (NGNP) adopted the ASME design strategy in 2008, with the steam cycle HTGR. The NGNP must show that, both under design basis and beyond design basis events, no radionuclide releases offsite will exceed 10 CFR 20 limits.[1]

The PRISM (Power Reactor Innovative Small ­Module) design by General Electric is a small SFR with inherent and passive safety aspects. The small size, passive decay heat removal, and inherent safety benefits of metallic fuel make SFRs very forgiving under severe transients. A few startup companies are pursuing SFRs both for their safety benefits and because of the ability to recycle the fuel and close the fuel cycle.

NuScale Power provided the PRA results for its 160 MWt SMR to the NRC as part of its design certification application,[2] showing that the CDF for all internal events was determined to be several orders of magnitude smaller than the NRC safety goal. While this significant reduction in CDF represents a major advancement in safety, the greater contribution relative to public perception of risk is the reduction in consequences: Even if a one-in-a-billion-year event were to occur, the dose at the site boundary would not likely exceed regulatory limits (NRC 2018; NuScale Power 2015). A site boundary emergency planning zone is much smaller than the 10-mile radius currently required for large 3000 MWt reactors. Furthermore, these rare events would evolve slowly such that the early release fraction, another NRC measure of safety, would essentially be zero.

The modular high-temperature gas-cooled reactor (MHTGR) design shows similar results. The MHTGR will use robust TRISO (tristructural isotropic) particle fuel in a low-power-density reactor core with a strong negative temperature coefficient and a solid high heat capacity moderator to ensure passive shutdown, passive heat removal, and low fission product release (INL 2011b). The PRA indicates significant margin to the NRC safety goals and no evacuation required beyond the site boundary as doses are less than prescribed by the EPA Protective Action Guides[3] (Inamati et al. 1987).

Advanced Reactor Safety Features

The numerous AR designs under development each have unique features that enhance safety through a few shared characteristics.

With SMRs, the amount of radioactive material and the corresponding heat generation rates range from 1/15 to 1/3 those of typical large reactors. The ratio is even smaller for microreactors (<5 MWe). This means that the source term available for release is inherently much smaller in each core. The smaller heat generation rates mean that free convection heat transfer and conduction are sufficient to remove heat without pumps needing external power.

Another characteristic of many advanced reactors is the compatibility of the coolant, moderator, and fuel, translating into much less severe off-normal events. In some AR designs, the safety of the reactor is not dependent on the coolant at all, since the methods of decay heat removal rely on phenomena such as conduction through solid material and thermal radiation. Some advanced reactors incorporate additional barriers to fission product release, including shield buildings or special fuel coatings (e.g., TRISO-coated particle fuel).

Given these characteristics, some advanced ­reactors do not require safety-related offsite or onsite power to keep their reactors cooled following an upset (e.g., ­station blackout, loss of coolant). Some might be housed in an underground structure or pool to enhance seismic resilience. The following sections briefly describe some specific AR safety features.

Accident-Tolerant Fuels

The severe reactor accidents at the Fukushima power station led many to question whether a better fuel system, with the same operational performance, could be used in light water reactors to enhance accident tolerance.

Research is underway to define fuels that would reduce chemical reactivity with steam, improve fuel thermal and cladding properties, and enhance fission product retention. Areas of focus are coatings on cladding, different cladding materials, additives to change fuel properties, and different fuel forms (Bragg-Sitton et al. 2014; DOE 2015; Zinkle et al. 2014). Lead test rods for some of these concepts have been loaded into existing US reactors for testing (Reed 2019; WNN 2020).

For sodium-cooled fast reactors, the two major fuel forms are mixed (uranium-plutonium) oxide fuels and metallic fuel. While the former has been studied extensively and remains the fuel choice internationally, in the United States metallic fuel was developed because it has relatively high heat metal densities and thermal conductivity, improved compatibility among the fuel system components, intrinsic passive safety characteristics, simpler fabrication processes, and less stringent quality control requirements than the oxide system. Metallic fuel is also of greater interest to SMR ­developers (­Carmack et al. 2009; Ogata 2012).

TRISO fuel is used in high-temperature gas- and salt-cooled reactors in either cylindrical compacts or pebbles, each containing thousands of particles, and has been under study internationally for more than 50 years (IAEA 1997; Petti et al. 2012). It is fabricated with exceptionally high quality (defects are on the order of 1/100,000 particles) and is quite robust under irradiation and high-temperature accident testing. This has enabled the development of a “functional containment” safety strategy that uses this fuel as a non­structural ­barrier to radionuclide release. A topical report on TRISO fuel performance is under review by the Nuclear Regulatory Commission (EPRI 2019).

Use of Containment as a Passive Heat Exchanger

Figure 1 presents the unique containment design for a NuScale SMR, a natural circulation reactor housed in an underground stainless steel–lined concrete pool in a seismic category 1 building resistant to massive earthquakes (>1.0 g at building frequency), very high winds (~470 km/h, exceeding those typical of category 5 storms), floods, and other natural disasters.

Figure 1 

Coolant leaks from the reactor vessel into the containment vessel cannot cause the core to uncover nor the containment to overpressurize. The large surface area of the containment vessel, relative to the heat generated by the reactor core, will completely remove decay heat by condensation, conduction, and ­natural convection to the pool without operator action, AC/DC power, or the need to add water.

Figure 2 

For modular high-temperature gas-cooled reactors (figure 2), decay heat can be removed from the vessel through radiation heat transfer from the outside reactor vessel wall to panels in the containment walls. Naturally circulating water or air (depending on the specific design) in the panels acts as the ultimate decay heat sink. Since radiation heat transfer is proportional to the surface temperatures to the fourth power, the removal of decay heat through radiation becomes significantly more effective as the temperature of the vessel wall increases. These gas reactors are designed to reach high temperatures to allow the efficient removal of decay heat through radiation. Even if the ultimate heat sink were to fail, the heat would radiate into the ground since the reactor is embedded below grade.

Extended Coping Periods without Power or Operator Action

Advanced reactor designs extend the duration, or coping period, that a nuclear reactor can be cooled without the need for active power systems to replenish coolant. Some designs have a 7-day coping period, others can transition from water cooling to air cooling for an unlimited coping period.

Figure 1 illustrates this safety feature for an extended loss of onsite and offsite power. During normal operation, the safety valves are held shut or open against a motive force (i.e., spring or accumulator) using electrical power. Loss of power results in gravity insertion of the control rods and alignment of safety valves to their safe position, requiring no operator action. The reactor safety valves vent steam into the containment where it condenses on its inside surface. The condensate, driven by its gravity head, returns to the reactor vessel through recirculation valves, thus maintaining the core covered with liquid. Heat conduction through the containment wall and free convection on the outside surface of the containment remove the decay heat. Heat is also transferred to the pool via the steam generator using the decay heat removal system. After 1–3 seconds the core decay power drops to ~10 MW thermal, and after 1 day to ~1.1 MW thermal, typical of many university research reactors. The water level in the pool decreases over time due to boiling but heat transfer from the containment remains effective. After 30 days, the core decay power is only ~0.4 MW thermal such that radiative heat transfer and free convection of air on the outside surface of the containment vessel are sufficient to remove all the decay heat. This unlimited coping period is achieved without operator or computer action, AC or DC power, or the need to add water.

For MHTGRs, the high thermal inertia of the ­graphite core results in very slow transients and long coping times. Upon loss of either helium flow or ­helium inventory in the reactor, heatup of the reactor core under decay heat can take 1 to 2 days depending on the design. Peak fuel temperatures will remain below the level where significant fuel damage occurs. The reactor cavity cooling system passively protects the reactor silo concrete from overheating so that restart is enabled after repairs to the reactor coolant system.

Reactor Safety without Control Rods

Most AR cores implement very strong negative reactivity feedback mechanisms such that overheating the system results in a significant decrease in core ­thermal power, or complete shutdown of the reactor, even without the insertion of control rods. The moderator temperature, fuel temperature, and void reactivity coefficients are part of the inherent physics of the core design. The self-limitation of core power to a fraction of full power conditions means that the passive safety systems normally used to remove decay heat are fully capable of keeping the core cooled without control rod insertion.

Reactor Safety Demonstrations

Many AR safety characteristics have been demonstrated in existing reactors. Inherent and passive safety features were demonstrated in EBR-II, a small SFR with metallic fuel, in the 1980s. For MHTGRs, safety demonstrations of the strong negative reactivity feedback and loss of flow tests were conducted at the AVR pebble bed reactor in Germany, the HTR-10 pebble bed in China, and the high-temperature engineering prismatic test reactor in Japan (Buongiorno et al. 2018).

Specific Benefits of New Safety Features

Advanced reactor safety features offer a level of functionality, flexibility, and resilience not previously offered by nuclear power.

Off-Grid Operation

Because some AR designs do not require offsite ­power for safety, they could operate off-grid (NRC 2017) to provide heat and power for a wide range of industrial applications. For example, a “six-pack” NuScale plant could generate 200 metric tons of hydrogen per day using high-temperature steam electrolysis without ­carbon emissions. A single module could generate 60 million gallons of desalinated water per day for ­coastal cities (Ingersoll et al. 2014), and an HTGR, with its high outlet temperature (>750°C), would be well suited to supply process heat for the petrochemical industry as well as hydrogen production. (For a study of relevant markets and the associated economics, see INL 2011a.)

Off-grid operation also aids in adapting to the increased frequency of severe weather events due to climate change. If a severe weather event isolates a nuclear plant from the grid, instead of shutting down as required by existing regulation, an AR plant could remain in operational “island mode,” dispatching power in increments as needed to support recovery of the grid. The ability to provide “first responder power” and black-start capability (the ability to resume power generation after a shutdown without relying on the electric grid) are due to the unique safety features of AR designs.

Climate Change Mitigation

Nuclear power must be a major component of strategies to combat climate change because it offers the greatest potential for reduced carbon emissions in the electricity sector. Both the International Panel on Climate Change (https://www.ipcc.ch/sr15/) and the ­International ­Energy Agency (IEA 2019) propose a significant increase in nuclear power to achieve global carbon emission reduction goals.

Because of their smaller footprint, reduced complexity, enhanced safety, and load following capabilities, small and midsize advanced reactors could play a major role in helping states reach their clean energy goals. Retired coal-fired plants could be repurposed, for example, and the use of existing infrastructure such as water supply, switchyard, and electric transmission lines would be a cost-effective approach to add carbon-free energy to the existing grid.

Highly Reliable Long-Term Power for Mission-Critical Facilities

Advanced SMRs and microreactors can provide highly reliable long-term power to mission-critical facilities such as hospitals, data centers, national laboratories, and military bases. For example, the NuScale 12-­module plant design, with a redundant array of independent reactors and island mode capability, can provide 60 MWe to a dedicated microgrid at 99.98 percent reliability for the 60-year life of the plant. This corresponds to only 4 days with zero output over those 60 years (Doyle et al. 2016).

If a catastrophic event damages both the transmission grid and transportation infrastructure such that neither fuel nor power can be delivered to the site for a prolonged period, multimodule plants operating in island mode have a significant advantage. If the microgrid remains intact or can be restored, a 12-module plant can provide 120 MWe to the microgrid of a mission-critical facility for 12 years without the need for new fuel.

Conclusions

Advanced reactors offer unique safety features that significantly reduce both the frequency and the consequences of core damage events and will enable a new level of functionality, flexibility, and resilience for nuclear power. Some features may prevent exceeding regulatory doses at the site boundary even in a highly unlikely event that exceeds design basis.

The improved features may expand the role of ­nuclear power in climate change mitigation, for example reducing CO2 emissions through the repurposing of coal-fired power plants located near population centers. They also enable a variety of off-grid applications, such as hydrogen production, desalination, first responder power for grid recovery, and power to microgrids for mission-­critical facilities.

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[1]  Federal Code of Regulations, Title 10, Part 20: Standards for Protection against Radiation, online at https://www.nrc.gov/reading-rm/doc-collections/cfr/part020/ .

[2]  NuScale Power design certification application submitted to the US Nuclear Regulatory Commission; latest revisions available at https://www.nrc.gov/reactors/new-reactors/design-cert/ nuscale/documents.html#dcApp.

[3]  https://www.epa.gov/radiation/protective-action-guides- pags#pagmanual

About the Author:José Reyes (NAE) is cofounder and chief technology officer of NuScale Power LLC. Finis Southworth is former chief technology officer of AREVA. Brian Woods holds the Henry W. and Janice J. Schuette Endowed Chair in Nuclear Engineering and Radiation Health Physics at Oregon State University.