USDOE Issues Funding Opportunity Announcement for SMRs

The US Department of Energy (USDOE) issued a draft Funding Opportunity Announcement (FOA) this week, intended to support activities related to the design and licensing of Small Modular Reactors (SMRs), defined as reactors with an electric output of 300MWe or less; which can be manufactured remotely, transported to point of construction, with on-site assembly largely limited to system integration of components for operation.

Importantly, the USDOE is interested in designs with passive safety (e.g., against consequences of a nuclear accident) as well as inherent safety (e.g., against natural catastrophes such as earthquakes, windstorms or floods), in addition to designs with long inter-refueling periods, low capital cost outlays, low maintenance and operating costs, and high proliferation resistance. The stated intention is to support up to 2 reactor designs through the USNRC design and licensing process, with the ability to be deployed ‘expeditiously’ being an important merit criterion. 2022, a decade from now, is the target year for commercial operation.

Proponents may choose to pursue licensing from the USNRC under either 10 CFR 50 or 10 CFR 52. Stakeholders are encouraged to form consortia, and proponents are encouraged to form design-centered working groups (DCWGs) across the supply and value chains e.g., SMR manufacturers, power utilities, local bodies; and the activities funded by USDOE are required to draw at least 50% of their total resources required, from internal sources. The total amount of funding available from USDOE is estimated to be $452 M, subject to Congressional appropriations. The current draft FOA will be issued in final form after feedback from, and consultation with, stakeholders.

Gaps in Current Probabilistic Risk Analysis (PRA) Methodology

Probabilistic Safety Analysis (PSA) of nuclear reactors (in the IAEA's usage), or Probabilistic Risk Analysis (PRA) in the USNRC's usage, is a technology that is being continually refined, both in response to those of its existing inadequacies that are already known to the original analysts and reviewers, and also in response to events that specifically underline one or more of such gaps. The Fukushima nuclear disaster, for example, increased the perceived urgency of addressing major gaps in nuclear reactor safety analyses and PSA/PRA techniques.

The USNRC was recently briefed on Severe Accidents and Options for Proceeding with Level 3 Probabilistic Risk Analysis (PRA Level 3). Meeting Agenda, Slides presented by Biff Bradley from the Nuclear Energy Institute (NEI) , Stewart Lewis of the Electric Power Research Institute (EPRI), Karl N. Fleming (of KNF Consulting), and NRC Staff. Meeting Transcript.

The two main gaps as seen by the US NRC Staff include:

Modeling of Consequential Linked Events
Current PSA techniques have not focused on risk implications of event sequences where a consequent initiating event occurs while a plant is responding to the first. PSA/PRA methodologies traditionally have not considered the risk implications of initiating events leading to accidents at multiple units at the same site - such as the near-simultaneous swamping by the tsunami of diesel generating systems supplying emergency power to several different nuclear reactors, each of which then suffered core damage as a consequence.

Aqueous Dispersion of Radionuclides
The risk implications of a containment breach have traditionally been considered in PRA Level 3, but the focus has been on atmospheric dispersion. Fukushima showed that the possibility of aqueous dispersion of radionuclides, must also be studied and modeled, both from spent reactor fuel pools and from the reactor core itself. The water in the sprays used to cool the spent fuel pools and the core, as emergency measures in severe accident mitigation, resulted in both internal and external floods, and the radiological consequences of radionuclide dispersal through such events deserve to be better understood.

IAEA Course on Natural Circulation Phenomena - Harbin, China

The IAEA is conducting a course on Natural Circulation Phenomena and Passive Safety Systems in Advanced Water-Cooled Reactors at the College of Nuclear Science and Technology of the Harbin Engineering University at Harbin in Northern China, July 11-15, 2011.

Several recent advanced reactor designs (both large reactors like the ESBWR and the AP-1000, and integral pressurized water reactor [iPWR] designs in the SMR category, like mPower, NuScale and the newly announced Westinghouse SMR) propose natural circulation systems for decay heat removal. Other evolutionary designs also propose natural circulation convective systems for heat transfer in regular operation. The site of the first AP-1000 units to be constructed anywhere, the Sanmen Nuclear Power Plant (where two AP-1000 units are currently under construction) with planned in-service dates in 2013-14, is in China. China is also planning 1400 MWe and 1700 MWe variant designs based on the AP-1000, with the CAP-1400 said to be in an advanced design stage. Presumably, these will also utilize passive safety systems based on natural circulation, and this strong interest in natural circulation cooling in China is one of the main reasons that the location of this IAEA course is in Harbin.

The agenda for the course comprises both introductory and advanced lectures, by distinguished researchers in the field of natural circulation cooling, including Drs. Jose Reyes, Dilip Saha, Nusret Aksan, and F. D'Auria, among others. Particularly of interest is the lecture by Dr. Reyes on Thursday 14 July on Flow Stagnation in Single and Two-Phase Natural Circulation Loops [literature citation], which will discuss mechanisms which can interrupt natural circulation - for example, in a PWR, loss of heat sink could result in reverse heat transfer in the steam generator, interrupting single phase natural circulation. This and other mechanisms that interrupt both single- and two-phase natural circulation were studied by Dr. Reyes' group at Oregon State University in special thermalhydraulic loops constructed for the purpose. Scaling relationships are critical in understanding the applicability of results obtained from such loops to real reactor systems, and Dr. Reyes also presents a lecture on Integral System Experiment Scaling Methodology, while Dr. Dilip Saha presents a related lecture on Experimental Validation and Database of Simple Loop Facilities. Developing reliability models of passive safety systems utilizing natural circulation is critical to safety analysis of such reactors, and Prof. F. D'Auria will present a lecture on Reliability of Thermalhydraulic Passive Safety Systems. Generally speaking, studies of natural circulation phenomena are complicated by the fact that the driving force is usually quite weak, as compared, for example, to turbulence or friction that may also be present in the flow.