Sunday, November 21, 2010

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.

Thursday, June 3, 2010

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!

Tuesday, May 18, 2010

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.

Thursday, April 29, 2010

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.

Wednesday, March 17, 2010

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.

Saturday, March 13, 2010

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.

Saturday, March 6, 2010

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.