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.

The Consortium for Advanced Simulation of Light Water Reactors

The Consortium for Advanced Simulation of Light Water Reactors (CASL), the US Department of Energy's (US DOE) Energy Innovation Hub specific to Nuclear Energy, has been formed, headquartered at the Oak Ridge National Laboratory (ORNL) with Dr. Douglas B. Kothe as Director of CASL.

The basic mission of the CASL is to create a virtual reactor (VR) to computationally model and predictively simulate the operation of light water reactors, with a view to (i) decreasing overall capital and operating costs associated with LWRs (ii) decreasing spent nuclear fuel volume generated by LWRs (iii) improving nuclear safety performance, especially by developing computational tools which better predict ageing, degradation and failure of LWR materials and components. The objective is both to impact the sustainability program for current generation light water reactors, as well as to impact the design of future generation nuclear reactors.

The Core partners in CASL are Oak Ridge National Lab (ORNL), Electric Power Research Institute (EPRI), Idaho National Lab (INL), Los Alamos National Lab (LANL), Massachusetts Institute of Technology (MIT), North Carolina State University (NCSU), Sandia National Labs, Tennessee Valley Authority (TVA), University of Michigan, and Westinghouse Electric Company.

The operational structure and mission statement of CASL explicitly incorporates the vision US Secretary of Energy Dr. Chu has articulated, for example, of 'Bell Labs-like institutions which are mission-driven but solve fundamental problems as well'. See here.

In CASL Director Dr. Kothe's words, (CASL):

• Focuses on a single topic, with work spanning the gamut, from basic research through engineering development to partnering with industry in commercialization
• (creates) Large, highly integrated and collaborative creative teams working to solve priority
technology challenges
• Embraces both the goals of understanding and use, without erecting barriers between
basic and applied research (emphasis added).

Link

To develop the VR, CASL has been organized into five technical focus areas (FAs) to perform the necessary work ranging from basic science, model development, and software engineering, to applications:

Advanced Modeling Applications (AMA) – The primary interface of the CASL VR with the applications related to existing physical reactors, the challenge problems, and full-scale validation. In addition, AMA will provide the necessary direction to the VR development by developing the set of functional requirements, prioritizing the modeling needs, and performing assessments of capability.

Virtual Reactor Integration (VRI) – Develops the CASL VR tools integrating the models, methods, and data developed by other Focus Areas within a software framework. VRI will collaborate with AMA to deliver usable tools for performing the analyses, guided by the functional requirements developed by AMA.

Models and Numerical Methods (MNM) – Advances existing and develops new fundamental modeling capabilities for nuclear analysis and associated integration with solver environments utilizing large-scale parallel systems. The primary mission of MNM is to deliver radiation transport and T-H components that meet the rigorous physical model and numerical algorithm requirements of the VR. MNM will collaborate closely with MPO for sub-grid material and chemistry models and will connect to VRI for integration and development of the CASL VR.

Materials Performance and Optimization (MPO) – Develops improved materials performance models for fuels, cladding, and structural materials to provide better prediction of fuel and material failure. The science work performed by MPO will provide the means to reduce the reliance on empirical correlations and to enable the use of an expanded range of materials and fuel forms.

Validation and Uncertainty Quantification (VUQ) – The quantification of uncertainties and associated validation of the VR models and integrated system are essential to the application of modeling and simulation to reactor applications. Improvements in the determination of operating and safety margins will directly contribute to the ability to uprate reactors and extend their lifetimes. The methods proposed under VUQ will significantly advance the state of the art of nuclear analysis and support the transition from integral experiments to the integration of small-scale separate-effect experiments

The European PERFECT Project shares many of the goals of the CASL, in developing 'virtual reactors', though the PERFECT project aims to develop 2, one each for the reactor pressure vessel and the internal structures. The first will concentrate on modeling irradiation degradation, while the second will concentrate on the corrosion faced by internal structures.

American Nuclear Society Annual Meeting - ANS 2010

The American Nuclear Society (ANS) will hold its 2010 Annual Meeting in San Diego later this month (June 13-17 2010). It will likely be the world's largest conference of nuclear science and technology professionals, and its packed program is breathtaking in the scope, breadth and depth of coverage it provides of the hottest current topics in nuclear science, technology and policy.

I will simply indicate a few of what I consider very interesting sessions and add a comment or two by way of context. As can be expected, most of these are in areas of my research or consulting interest.

First, the Conference will include an Embedded Topical Meeting on the Safety and Management of Nuclear Hydrogen Production, Control and Management - the second such (the first having been held at ANS 2007). Among other interesting papers in this session is one on Probabilistic Safety Analysis of a hydrogen production plant using the Sulphur-Iodine process, with process heat derived from a High Temperature Test Reactor by a Korean group. This directly relates to topics I have discussed in my earlier papers: Safety Issues in Nuclear Hydrogen Production with with the Very High Temperature Reactor and Nuclear Hydrogen Production: Scoping the Safety Issues.

Secondly, the Conference will include a Session on Key Licensing and Regulatory Issues for Small and Medium Reactors, followed by a panel discussion with panelists from INL and the US NRC - I have discussed this topic earlier in other blog posts, and its importance can scarcely be over-emphasized. A group from GE will be discussing the licensing strategy for the PRISM (Power Reactor Innovative Small Module) liquid sodium-cooled reactor, while a group from KAERI (Korean Atomic Energy Research Institute) will discuss the SMART (System-integrated Modular Advanced Reactor) - a water-cooled reactor with integral steam generators that is designed for power (about 100 MWe per module), seawater desalination, and process heat applications. A separate session on Safety Analysis and Licensing of non-LWR Reactor Concepts, should similarly be of strong interest - discussing gas-cooled and liquid-sodium cooled reactors from both an experimental and simulation perspective.

A related session will cover the Thermal Hydraulics of the VHTR (gas-cooled variant), relevant in the context of the licensing of the Next Generation Nuclear Plant. This session will cover ongoing experimental and computational/simulation of VHTR thermalhydraulics at the Oregon State University and INL - particularly on Loss of Flow and Pressurized Conduction Cooldown events in High temperature Reactors. The important issue of scaling - the ability to draw numerical comparisons and conclusions that are valid for real reactors from experiments and simulations done on smaller systems - will be the topic of a paper from Oregon State that should be of particular interest.

The issue of Scaling Methods will also be the topic of a special Tutorial Session, to be conducted by Dr. Pradip Saha of GE and Prof. Jose Reyes of Oregon State - that will discuss issues of scaling particularly with reference to LWRs - methods of dimensional analysis, method of similitude and normalization of governing equations will be discussed.

The topic of Nuclear Fuel and Structural Materials for Next Generation Nuclear Reactors will be the focus of another Embedded Topical Meeting, a topic I have worked on and discussed in several earlier papers and presentations (and blog posts: here, here and here).

I need hardly add that the Conference promises to be extremely interesting indeed!

Probabilistic Safety Analysis and Management Conference, PSAM-10

The 10th International Probabilistic Safety Analysis and Management Conference (PSAM10), organized by the International Association for Probabilistic Safety Assessment and Management (IAPSAM), begins in Seattle next month (June 7-11, 2010). The conference will deal with probabilistic safety analysis and risk assessment in a number of industrial settings, including aviation, maritime and space, as well as civil engineering applications such as water treatment facilities - but will have a particular focus on the nuclear industry. The conference is sponsored in part by Scandpower Risk Management, a major nuclear risk consultancy and division of the Lloyd's Register group.

The Plenary Speaker in the nuclear track will be Dr. George Apostolakis, the MIT Professor and nuclear PSA expert who joined the US Nuclear Regulatory Commission as a Commissioner last month. Dr. Apostolakis has done pioneering work on licensing issues and probabilistic safety analysis of gas cooled and fast reactors that is of particular relevance to the US Next Generation Nuclear Plant project. His group has also contributed a paper at PSAM10 on how the computational burden in estimating failure probabilities in a passive thermal-hydraulic system may be reduced using artificial neural networks (ANNs) and Quadratic Response Surface Models (that I find to be of particular interest, given my own past background in using similar techniques).

The Apostolakis group also has another contributed paper on a new class of importance measures for PSAs which they call the limit exceedance factor (LEF)- defined as the factor by which the failure probability of a given component in a nuclear plant must be multiplied so that it results in an end-state probability (such as the core damage frequency CDF) exceeding a specified limit, for example, 1E-6. This is shown to be particularly relevant in the technology neutral framework (TNF) for assessing reactors that the NRC has developed - where, rather than specific design basis events (DBEs) being considered, a set of licensing basis events (LBEs) is considered instead, whose frequency and dose must satisfy certain limits. This paper is particularly of interest, since it applies the methodology to sodium-cooled reactors, which are of interest both in the SMR and Gen-IV contexts.

There are several other contributed papers from the US NRC, of which a paper on the Standardized Plant Assessment Risk (SPAR) model, developed for the NRC by the Idaho National Laboratory (INL) detailing its application to the AP1000 Reactor, and planned extensions to the ABWR, ESBWR, US-EPR and US-ABWR reactor designs is of particular interest to me, and there are also papers from INL on other aspects of SPAR development.

Dr. Philippe Hessel of the Canadian Nuclear Safety Commission (CNSC) will present a paper on the methodology used by the CNSC staff to carry out safety assessments of reactor licensing submissions which contain both probabilistic and deterministic arguments.

A paper on preliminary design-phase Probabilistic Risk Assessment of The NuScale Reactor, a modular, scalable 45 MWe Light Water Reactor (SMR) - is also of great interest, given the current excitement in small and modular reactors. Of the many other interesting papers, there are also papers on risk analysis of a Mars base and another on risk analysis for a crewed Mars mission - from a group based at NASA Moffett Field.

In the session on Ageing Management of Nuclear Power Plants - a paper on a new class of PRA risk measures that are able to (i) overcome the limitation imposed by the current inability to use dynamic failure rate data on component failure rates, and (ii) the limitation arising from not including passive components in the PRA - by a group from the Pacific Northwest National Laboratory - seemed very interesting, because these risk measures are claimed to enable better plant ageing management, and also help prioritize directions in materials degradation research.

In addition to all these, the conference will also cover a multitude of risk analysis areas such as those in seismic or hurricane hazards, fire hazard, the hazard from lightning events (especially critical for electrical power distribution grids); as well as other energy sectors such as risk assessment for geological sequestration (both of spent nuclear fuel and carbon dioxide) as well as for the use of hydrogen as a fuel in transportation applications, and miscellaneous nuclear and non-nuclear applications in medicine.

What is remarkable about the meeting is that it brings together practitioners of Probabilistic Safety Analysis and Risk Management from a variety of disciplines, while retaining a strong emphasis on nuclear-related PSA applications, with the potential for the different application domains of PSA to cross-fertilize, as well as being an opportunity for the practitioners from each discipline to learn from each other.

2nd Canada-China Joint Workshop on Supercritical Water-cooled Reactors (CCSC-2010)

The 2nd Canada-China Joint Workshop on Supercritical Water-cooled Reactors was held in Toronto earlier this week. (The 1st workshop had been held in Shanghai, China in April 2008.) The Supercritical Water-cooled Reactor (SCWR) is a Generation IV water-cooled reactor concept that holds the most promise for higher efficiency, on account of its higher operating temperature range, the hoped-for single phase (supercritical) operation (i.e., not having to deal with two-phase flow), the thermophysical properties (especially thermal conductivity and specific heat) of supercritical water, and the resulting saving in balance of plant pumps and compressors and secondary loop tubing and systems. What adds to the attractiveness of the concept is the possibility of realizing it within the Pressure Tube (PT) reactor design envelope, and moreover, the possibility of advanced fuel cycles involving thorium fuel within the concept.

However, a number of challenges also exist, which must be resolved through R&D, before the concept can become a realistic design. Within the Generation IV International Forum, Canada leads R&D work on the SCWR concept, and the purpose of the workshop this week was for Canadian and Chinese researchers to share the results of their respective R&D projects on materials, thermalhydraulics, water chemistry, and fuel cycle issues, in addition to more explicit considerations involving safety and licensing related foresight.

Over the three days of the workshop, there were two broad parallel tracks - sessions devoted to (i) materials issues and chemistry; and (ii) sessions devoted to thermalhydraulics, with an interspersed session each on reactor physics, licensing and safety, and nuclear hydrogen production with SCWR heat. Much of the work presented at the conference comprised sharply focused investigations along pre-established R&D priorities that had been scoped out in the basic SCWR R&D plan - both experimental and simulational investigations were presented. 

A significant departure from standard PHWR (CANDU) design that is being considered in the PT-SCWR (CANDU-SCWR) concept involves vertical pressure tubes (as opposed to the horizontal pressure tubes that are standard in PHWRs). Thus, two papers comparing supercritical and subcritical heat transfer correlations and characteristics in vertical pressure tubes, one each from Canada and China, were of particular interest.

Since supercritical water presents significant operating challenges, experimental work often uses surrogate fluids such as supercritical carbon dioxide. An entire session on the thermalhydraulics track was therefore devoted to surrogate fluids. Use of surrogate fluids then necessitates an understanding of two kinds of scaling issues - between experimental loop and a real reactor; and between surrogate fluid and real supercritical water (the 'working fluid').

Two very interesting papers discussed these issues. One paper, from Canada, discussed the supercritical thermalhydraulic loop currently being constructed at the University of Ottawa, while the other, from China, discussed fluid-to-fluid scaling issues. In developing fluid-to-fluid scaling, similarity relationships are often employed, for example, by using dimensionless variables like the ratio of actual pressure to critical pressure - which directly scales with the ratio of temperature to critical temperature for the two different fluids - in the same way. Although the relevant ranges of temperature and pressure at which the behavior develops can be different - the dimensionless ratio behaves in the same way - thus the behavior of the fluid with more easily reachable temperature and pressures (the modelling fluid or surrogate fluid) can be used to perform detailed experimental studies, while the behavior of the fluid with the more stressful operating conditions (the working fluid) can be inferred from the similarity scaling relationship. (Such invariant scaling relationships occur quite widely elsewhere in physics also, with quantities like the magnetization or the superfluid density, for example, in spin glasses or superconductors.) More details are available here [1].

Prof. David Novog's group from McMaster University, and Prof. Guy Marleau's group from Ecole Polytechnique (Montreal) presented papers on safety issues for the Supercritical Water-cooled Reactor.

Overall, the conference covered significant ground in its three days and also included one side trip to NRCan's Material Technology Laboratory (MTL) at Ottawa and another to AECL's Chalk River Laboratories (CRL).

References

1. Groeneveld, D.C., Tavoularis, S., et al Nucl. Eng. Technology vol. 40 no. 2, 107-116, 2007.

Two Energy Materials Conferences in Karlsruhe

Karlsruhe, the Southwest German town, home to the Forschungszentrum Karlsruhe [The Karsruhe Research Center - a major German center for nuclear research) and the Karlsruhe Institut fur Technologie, will host two separate Conferences on Materials for Energy Applications this year - in July and October respectively.

The July Conference (EnMat 2010) will mainly deal with materials for non-nuclear energy applications - Hydrogen Storage, Fuel Cells, Thermoelectrics, and related topics (though there will also be a plenary talk on Fusion Materials - this is especially interesting since Fusion does represent, well, a fusion of hydrogen and nuclear technologies). Extremely interestingly, a Fusion plant can be conceived as a complete hydrogen economy - it uses two isotopes of hydrogen - deuterium and tritium as fuel, generates (or breeds) tritium as a byproduct, and the resulting fusion heat can be used to split water either thermo-chemically or electrochemically to yield molecular hydrogen - which can be used in fuel cells to generate electricity, or burnt in internal combustion engines directly. [I discussed this fascinating possibility in my presentation Nuclear Hydrogen Production: Re-examining the Fusion Option and the accompanying paper at the Canadian Hydrogen Association Meeting in 2007.] Fusion does indeed look even more interesting when viewed from the Hydrogen Economy prism.

The October Conference (NuMat 2010) will deal mainly with Materials for Nuclear Applications - fuel materials as well as structural materials for nuclear plants. NuMat 2010 will be a combined venue for several conferences on related topics which have previously been occurring separately, and there will be 6 major themes at NuMat 2010:
* Thermodynamics and Thermophysics of Nuclear Fuels
* Materials Models and Simulations for Nuclear Fuels
* Radiation Stability of Complex Microstructures
* Molten Salts for Nuclear Applications
* Structural and Functional Materials for Fission Reactors
* Structural Materials Modelling and Simulation

Small and Modular Reactors

Small and modular nuclear reactors (those with a thermal power output below 200 MWTh) have become of strong interest, both in Canada and worldwide for a number of reasons. In Canada, the interest arises from the following sources:
(i) The need to replace the NRU (National Research Universal) reactor with another multipurpose research reactor, as recommended by the NRCan Expert Review Panel on Medical Isotope Production last year [Recommendation I, p. xi Executive Summary; also on p. 73 of the main body of the report]
(ii) The interest expressed by energy providers (as well as industrial users such as in mining and tar sands extraction) in off-grid electric and/or thermal process power in modular and scalable units - partly from remote siting considerations and also from emissions reduction considerations
(iii) University nuclear reactors for training and research
(iv) Reactors for dedicated medical radioisotope production.

Separately of the off-grid power reactor interest from resource extractive industries, there is also interest in small reactors as a possible solution for developing countries and first-time nuclear countries who have small or under-developed electric grids. They are also an attractive option for small gas- or coal- fired generating units as a direct replacement, where grid and transmission infrastructure already exist, as in rural areas of developed nations.

As well, the lower up-front capital cost of the smaller power reactors is a motivating consideration for the increased interest, as is the potential for upward scalability in total power output by addition of more units in a more graded manner. Given the lower radionuclide inventory as well as some passive safety features in some of the small reactor designs, they become of additional interest from both the safety and the proliferation-resistance standpoints.

Although some reactors have a thermal output as low as 20 KWTh, the 200 MWTh threshold is chosen to define the upper limit of 'small reactors' from the point of view of accumulation of the radionuclide inventory - which is much smaller below a threshold of 200 MWTh. As well, given that some reactors may have passive safety features, the balance of engineered safety requirements that are imposed could be different for smaller reactors than for large reactors. Consequently, it is possible to justify what has come to be called a 'graded approach' in the safety assessment of small reactors - a smaller reactor will have safety requirements commensurate to the relative risk, compared to a larger reactor, and not necessarily identical ones. This graded approach could reflect itself, for example, in the containment structure requirement, or in extent of the exclusion zone, where the regulatory requirement may not necessarily be identical to that for large power reactors.

The Canadian Nuclear Safety Commission (CNSC) is currently in the process of developing regulatory & licensing guides and related requirements for small nuclear reactors based on these considerations, and will be holding appropriate stakeholder consultations, information sessions and technical workshops during the course of this year to disseminate information and solicit feedback before finalizing the requirements.

Postscript
US Secretary of Energy Steven Chu outlined the interest in small and modular reactors in his Wall Street Journal op-ed on March 23, 2010, a summary of his Congressional testimony of 3-3-2010.

In his 2011 budget request, President Obama requested $39 million for a new program specifically for small modular reactors. Although the Department of Energy has supported advanced reactor technologies for years, this is the first time funding has been requested to help get SMR designs licensed for widespread commercial use.

Right now we are exploring a partnership with industry to obtain design certification from the Nuclear Regulatory Commission for one or two designs. These SMRs are based on proven light-water reactor technologies and could be deployed in about 10 years.

Expanding on the likely advantages of small modular reactors, he said:

Small modular reactors would be less than one-third the size of current plants. They have compact designs and could be made in factories and transported to sites by truck or rail. SMRs would be ready to "plug and play" upon arrival.

If commercially successful, SMRs would significantly expand the options for nuclear power and its applications. Their small size makes them suitable to small electric grids so they are a good option for locations that cannot accommodate large-scale plants. The modular construction process would make them more affordable by reducing capital costs and construction times.

Their size would also increase flexibility for utilities since they could add units as demand changes, or use them for on-site replacement of aging fossil fuel plants. Some of the designs for SMRs use little or no water for cooling, which would reduce their environmental impact. Finally, some advanced concepts could potentially burn used fuel or nuclear waste, eliminating the plutonium that critics say could be used for nuclear weapons.
[...]
To achieve this potential, we are bringing together some of our nation's brightest minds to work under one roof in a new research center called the Nuclear Energy Modeling and Simulation Hub.

Update The Consortium for Advanced Simulation of Light Water Reactors (CASL), the Energy Innovation Hub specific to Nuclear Energy, has been formed, with Dr. Douglas B. Kothe as Director. The Core partners are Oak Ridge National Lab (ORNL), Electric Power Research Institute (EPRI), Idaho National Lab (INL), Los Alamos National Lab (LANL), Massachusetts Institute of Technology (MIT), North Carolina State University (NCSU), Sandia National Labs, Tennessee Valley Authority (TVA), University of Michigan, and Westinghouse Electric Company.

The operational structure and mission statement of CASL explicitly incorporates the vision Prof. Chu has articulated, for example, of 'Bell Labs-like institutions which are mission-driven but solve fundamental problems as well'. See here.

In CASL Director Dr. Kothe's words, (CASL):

• Focuses on a single topic, with work spanning the gamut, from basic research through engineering development to partnering with industry in commercialization
• (creates) Large, highly integrated and collaborative creative teams working to solve priority
technology challenges
• Embraces both the goals of understanding and use, without erecting barriers between
basic and applied research (emphasis added).

A second Energy Innovation Hub also announced is the Joint Center for Artificial Photosynthesis (JCAP), a partnership between Caltech and Lawrence Berkeley Laboratory:

JCAP research will be directed at the discovery of the functional components necessary to assemble a complete artificial photosynthetic system: light absorbers, catalysts, molecular linkers, and separation membranes. The Hub will then integrate those components into an operational solar fuel system and develop scale-up strategies to move from the laboratory toward commercial viability. The ultimate objective is to drive the field of solar fuels from fundamental research, where it has resided for decades, into applied research and technology development, thereby setting the stage for the creation of a direct solar fuels industry.

CNSC Presentations at NRC-RIC 2010

Two senior officials of the Canadian Nuclear Safety Commission (CNSC) made presentations at the US Nuclear Regulatory Commission Regulatory Information Conference 2010 (NRC-RIC 2010) last week. The President of the CNSC, Dr. Michael Binder, spoke at NRC-RIC-2010 on A Canadian Regulator's Perspective on International Cooperation. He noted that there were now 48 CANDU-type power reactors in 7 different countries (plus 3 reactors under construction - 1 in Argentina and 2 in India). Emphasizing that national regulators have responsibilities toward customer countries, which he considered an international extension of Canada's safety mandate, he outlined the three phases of Canada's current engagement with the regulatory mechanism in a customer country: (i) With the national regulating agency in the buyer country (ii) On-site, at the end-use location of Canadian-origin technology (e.g. at the site of a CANDU reactor). (iii) Training of regulators as well as interactions at the university level. As the Nuclear Renaissance unfolds, he also indicated that the pattern of Canadian inernational regulatory engagement might move beyond bilateral engagements and evolve to encompass more multilateral mechanisms such as the Multinational Design Evaluation Program (MDEP), with greater harmonization of codes and standards and perhaps including a code of conduct for vendors of nuclear technology.

The Vice President of the CNSC's Technical Services Branch, Terry Jamieson took the theme of International Cooperation in Nuclear Regulation forward, speaking on the MDEP's Role in Converging Codes and Standards. He outlined the efforts of the Codes and Standards Working Group (CSWG) of the MDEP, and indicated that the present focus of the group was on the pressure boundary components. Although different countries had their own codes and standards regarding pressure boundary components, the American Society of Mechanical Engineers (ASME) codes were used as a basis for comparison, focusing first on Class I Pressure Vessels. The objective was to eventually evolve a harmonized set of standards (since full convergence was not found feasible). Next steps will focus on codes for Class I piping, pumps and valves, and later on codes for components beyond those at the pressure boundary.

Nuclear Regulatory Commission Regulatory Information Conference NRC-RIC-2010

The US Nuclear Regulatory Commission (NRC) will be holding its annual Regulatory Information Conference (NRC-RIC) from March 9 to March 11, 2010. The conference will bring together a variety of stakeholders in the nuclear sector with regulators and technical specialists, both from the NRC and from US national laboratories. While most attendees will be from within the US, there will also be a large number of attendees from other countries, including Canada, who will share their own experiences and provide their own insights into nuclear regulatory affairs.

Apart from plenary sessions addressed by NRC Chairman and Commissioners, there will also be technical sessions on a number of cutting-edge issues at the interface of regulation and technology. These include a session on Materials Degradation at the Containment and Reactor Coolant System Pressure Boundary [Audio], which will discuss probabilistic analysis tools for carrying out the assessment of materials degradation at the pressure boundary, incorporating insights from investigations of the Pressurized Thermal Shock phenomenon. This is expected to contribute, for example, to better understanding of the probability of leak before rupture of piping systems. There are also at least two technical sessions on international issues: one on International Coordination between countries pursuing New Nuclear Power, another on International Cooperation on New Reactors [including the activities of the Multinational Design Evaluation Program (MDEP)]. Another session is devoted to discussing regulatory applications of International Experience in Operating Nuclear Reactors.

A technical session on Regulatory and Policy Issues for Small Modular Reactors should prove particularly interesting - since there is now great interest in the possibility of constructing small and modular reactors (including for isotope production; research; and local, off-grid power or heat applications). [Audio of event].

A separate session discussing the interest in Small and Modular Reactors will also be held [Audio].There will also be a session devoted to new developments in Probabilistic Risk Analysis (PRA) for nuclear power plants, including a talk on peer review of the US NRC's SPAR (Standardized Plant Analysis Risk) model. [Audio]

There is also a session on technical, policy and R&D issues related to the licensing of the Next Generation Nuclear Plant (NGNP), a gas-cooled reactor currently under development. [Session; Audio]. This includes a talk on the US NRC's efforts to develop an evaluation model(EM) for the NGNP. Other sessions of interest include a poster session on Central and Eastern US Seismic Source Characterization (SSC) model development, which may have implications for characterizing seismic sources in Canada as well.

Overall, the conferences promises to be quite interesting indeed.

Update: The US NRC published a Commission Paper (SECY-10-0034) on Potential Policy, Licensing, and Key Technical Issues for Small Modular Nuclear Reactor Designs on 3-28-2010.