NRC Seeks Prompt Action on Fukushima Near-Term Task Force Recommendations

The US Nuclear Regulatory Commission today directed its staff to complete several actions over the next 45 days in response to the 12 recommendations of its Near-Term Task Force on the Fukushima Nuclear Accident, which submitted its report on July 12.

The Commission has asked the staff to produce a paper by September 9, outlining which of the recommendations 2-12 (recommendation 1 was that the entire regulatory framework be holistically reviewed) should be implemented immediately, including a public dialogue on the process, the schedule for which will be announced soon. It has also asked the staff to produce, by October 3, another paper prioritizing recommendations 2-12, explaining the rationale, and also engaging both the public and other stakeholders. Over the next 18 months, the staff will consider recommendation 1, to review the entire regulatory framework, balancing risk -informed regulation, and defence-in-depth considerations.

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.

NRC Briefing: Severe Accidents and Level 3 PRA

The staff of the US Nuclear Regulatory Commission today held a briefing for the Commissioners on Severe Accidents, and Options for proceeding with Probabilistic Risk Asessment - Level 3 (PRA Level 3).

Traditionally, PRA/PSA Level 3 has not been a strong regulatory requirement because the results of Level 1 (usually the core damage frequency, CDF) and the results of Level 2 (large early release fraction/frequency, LERF) can be used as surrogate proxies for the types of Risk Metrics that a Level 3 PSA/PRA might generate, which could include the following: the number of early fatalities; the number of early injuries, the number of latent cancer fatalities, or the total population dose at different locations; as well as the individual early and latent fatality risk, and the economic cost of mitigation actions taken following a severe accident. The CDF, for example, can proxy for the latent cancer risk, while the LERF could proxy for the prompt fatality risk.

However, a number of potential benefits are foreseen for a full PRA Level 3 analysis, including feedback into risk-informed regulatory guidance for new reactors and use of risk insights in forthcoming SMR design reviews. In addition, capabilities such as modeling of radionuclide aqueous dispersion modes, and multi-unit risk assessment could also be addressed. The modeling of radionuclide dispersal in the event of a severe accident which is initiated by an external event such as an earthquake, tsunami or hurricane also calls for additional modeling efforts, since meteorological variables such as windspeed & direction, ambient precipitation and humidity may not correspond to what is normally expected for that site at that time of year (for example).

NRC Staff plan to use an existing SPAR (Standardized Plant Assessment Risk) model as the basis for proceeding to PSA Level 3. The SPAR model is essentially a plant-specific PSA/PRA Level 1 designed to incorporate both external and internal initiating events, recent modifications of which include capabilities to yield LERFs. Since external hazards are site-specific, much greater value can be expected to be derived if the SPAR model selected for development to PRA Level 3 is for a NPP site that is either representative of the entire population of NPPs, or, has a larger than average number and type of external hazards.

NRC Releases Near-term Report of Fukushima Task Force

The United States Nuclear Regulatory Commission has released the Near-Term Task Force Review of Insights from the Fukushima Da-ichi Accident: Recommendations for Enhancing Reactor Safety in the 21st Century.

The report has 12 recommendations, which I cite here in full, verbatim.

Clarifying the Regulatory Framework
1. The Task Force recommends establishing a logical, systematic, and coherent regulatory framework for adequate protection that appropriately balances defense-in-depth and risk considerations.
Ensuring Protection
2. The Task Force recommends that the NRC require licensees to reevaluate and upgrade as necessary the design-basis seismic and flooding protection of structures, systems, and components for each operating reactor.
3. The Task Force recommends, as part of the longer term review, that the NRC evaluate potential enhancements to the capability to prevent or mitigate seismically induced fires and floods.
Enhancing Mitigation
4. The Task Force recommends that the NRC strengthen station blackout mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events.
5. The Task Force recommends requiring reliable hardened vent designs in boiling water reactor facilities with Mark I and Mark II containments. (Section 4.2.2)
6. The Task Force recommends, as part of the longer term review, that the NRC identify insights about hydrogen control and mitigation inside containment or in other buildings as additional information is revealed through further study of the Fukushima Dai-ichi accident.
7. The Task Force recommends enhancing spent fuel pool makeup capability and instrumentation for the spent fuel pool.
8. The Task Force recommends strengthening and integrating onsite emergency response capabilities such as emergency operating procedures, severe accident management guidelines, and extensive damage mitigation guidelines .
Strengthening Emergency Preparedness
9. The Task Force recommends that the NRC require that facility emergency plans address prolonged station blackout and multiunit events.
10. The Task Force recommends, as part of the longer term review, that the NRC pursue additional emergency preparedness topics related to multiunit events and prolonged station blackout.
11. The Task Force recommends, as part of the longer term review, that the NRC should pursue emergency preparedness topics related to decisionmaking, radiation monitoring, and public education.
Improving the Efficiency of NRC Programs
12. The Task Force recommends that the NRC strengthen regulatory oversight of licensee safety performance (i.e., the Reactor Oversight Process) by focusing more attention on defense-in-depth requirements consistent with the recommended defense-in-depth framework.

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.

International Topical Meeting on Probabilistic Safety Analysis - PSA 2011

The International Topical Meeting on Probabilistic Safety Analysis (PSA 2011), sponsored by the American Nuclear Society and Sandia National Laboratories, along with a variety of commercial sponsors, will be underway next week in Wilmington, North Carolina from March 13 to March 17, 2011. Dr. George Apostolakis of the US NRC is the Honorary Chair of the Organizing Committee, while the Technical Committee has four Co-Chairs, one each from the US, Europe, Japan and Korea.

A truly large number of technical sessions are planned, and include several sessions on PSA of New Reactors from internal initiating events (including a very interesting paper on incorporating PSA principles into fusion reactor design, and papers on both gas-cooled and sodium-cooled fast reactors). Also PSA of a variety of hazards including fire, seismic, and flood; as well as PSA of non-reactor nuclear applications. There are sessions on incorporating digital information & control (I & C) systems into nuclear plant PSA; sessions on dynamic PSA (incorporating the dynamic, i.e., changing aspects of a system in to the probabilistic safety assessment [including a very interesting paper using genetic algorithms to explore the space within the failure domain where at least one safety limit is violated].

Several sessions explore Ageing in PSAs - one very interesting paper interpolates state transition probabilities in a Markov Model for estimating reliability of passive components such as metal pipes using physics-based models of weld degradation, instead of in-service failure data for the entire piping component. The paper finds that incorporating such time-inhomogeneous and stochastic transition rates into the Markov Model causes it to become non-Markov.

Interesting panel discussions are planned on: Alternative Risk Metrics, which will consider, among other things, how the promised lower risk numerics for new reactors will be maintained over their reactor life; and how risk profiles will be affected by multiple units in a suite of SMRs (small modular reactors); PRA Standards Development (which will examine, among other things, how the regulatory endorsement of PSAs as a risk management tool impacts the development of risk informed applications).

The conference brings together practitioners of PSA from a variety of disciplines and countries, and promises to be very interesting indeed.

5th International Symposium on Supercritical Water-cooled Reactors ISSCWR-5 Vancouver

The 5th International Symposium on Supercritical Water-cooled Reactors (ISSCWR-5) begins on March 14 2011 in Vancouver. The conference gets underway with five plenary addresses by national and international program managers of respective SCWR/HPLWR programs on the morning of the first day, Monday, and then branches off into three parallel technical sessions in the afternoon: on SCWR Core Design; on Materials Issues and on General Thermalhydraulics and Safety, chaired by international authorities in these respective fields. The session on General Thermalhydraulics and Safety will be co-chaired by Sama Bilbao y Leon of Virginia Commonwealth University and Jovica Riznik of the Canadian Nuclear Safety Commission.

This pattern of technical sessions continues also on Tuesday; an important facet of the Tuesday morning sessions will be regulatory considerations: a talk by Alexandre Viktorov of the Canadian Nuclear Safety Commission will be on Regulatory Expectations for Advanced Reactors, while Ima Ituen and David Novog of McMaster University will present on Assessing the Applicability of Canadian Regulations to the SCWR.

On Wednesday morning, there are sessions on Safety Issues and non-Aqueous Fluid Heat Transfer, the latter referring especially to experiments on supercritical carbon dioxide, where considerations on fluid-to-fluid scaling are important in interpreting the results and applying them to the real working fluid, supercritical water. Of the many interesting papers, one which describes a supercritical loop for in-pile testing of materials seemed especially interesting.

On all three days, the pattern of three parallel technical sessions is maintained, testifying to the high level and quality of national and international participation in the conference, and the interesting work on the SCWR that continues apace through the Gen-IV International Forum (GIF). Canada, as the host country [and also the country that formally leads R&D on the SCWR under the GIF] has the highest number of papers - both established groups and newer ones, and both senior researchers and students are presenting papers. Importantly, the Canadian participation shows significant engagement with the SCWR concept, across all major stakeholders: by academic groups, by regulatory authorities, as well as by R&D Labs and industrial firms.

The conference closes on Thursday with a tour of TRIUMF, Canada's national laboratory for nuclear and particle physics, located on the campus of the University of British Columbia. The scenic locale of the conference in Vancouver, and the very interesting papers to be presented, and discussions to be had, plus the social and cultural programs and the tour of TRIUMF promise to make this a most memorable conference in the biannual ISSCWR series.