Symposia

Welcoming Remarks

In his welcoming remarks, MIT Professor Mujid Kazimi, Director of MIT’s Center for Advanced Nuclear Energy Systems (CANES), noted that MIT and INL have had a strong program of collaboration on advanced reactors since 1998, when a workshop was held to explore directions for advanced reactor technology. Five large projects were selected at that time, including a high temperature reactor, a fast reactor and a new fuel material, appropriate topics for the many options under discussion even today.

The sponsor of the workshop, CANES was formally created in September 2000. The first CANES Symposium was held in April 2001 and focused on examining the role of nuclear energy in a sustainable environment. Since then, external events helped make nuclear energy more desirable, and continued improvements in the existing LWRs was key in the birth of a second era for nuclear energy.

Issues for the nuclear industry have changed over the past several years. In 2000 questions included (1) Will new nuclear plants be economic? (2) Can license extension of existing plants be stable? (3) How can nuclear energy higher education survive with few students? (4) Will nuclear energy be recognized as clean energy? And (5) When will the US see the next nuclear plant order?

By 2006 questions of primary important are (1) Which nuclear plant will be most economic? (2) Can licensing of new plants be streamlined? (3) How will nuclear energy education cope with the high load of students? (4) Which environmental group will be the last to switch on nuclear energy? And (5) When will the US introduce reprocessing? Today’s workshop is an important step in the creation of an ACE to be supported by INL within CANES, as the INL strengthens its relation with MIT and all other national universities.

Kazimi then asked the audience to be interactive with the speakers to define the high priority R&D efforts.

In his welcoming remarks, Andrew Klein, Director of Education, Training and Research Partnerships at INL, noted six critical attributes for INL:

• Nuclear Programs: A robust portfolio of nuclear science and technology programs
• Synergistic Programs: A synergistic portfolio of national and homeland security, energy and environmental programs
• Science Base: A robust science base to attract the best staff and create a culture of scientific inquiry
• Revitalize Education: A central role in revitalizing nuclear science and engineering education and training
• Partner/Collaborate: Extensive national and international collaborations
• Modern Infrastructure: Forefront research facilities, support infrastructure and management systems

He said INL has a central role in revitalizing nuclear science and engineering education and training, which will entail creating a hub-node network with five national and three Idaho universities, building a Center for Advanced Energy Studies (CAES) by 2008. This will enable increased faculty and student participation in research activities of INL, university research and education programs aligned with the mission of the INL as well as work with industry to improve training. Klein observed that universities are important partners in the development and success of INL.

Session 1: Priorities and Goals for Advanced Nuclear Systems

Speakers: David Hill, Ernest Moniz

Moderator: Neil Todreas

David Hill, Deputy Laboratory Direction, Science & Technology at INL, spoke on priorities for Next Generation Nuclear Plants (NGNP). This program assumes we will need a lot of nuclear energy in a system that deals effectively with spent fuel and proliferation issues. Priorities for NGNP are:

• Hydrogen: Thermochemical and high temperature electrolysis
• Fuels: Fabrication, irradiation, and safety testing
• Materials: Graphite and high temperature alloy testing & qualification
• Design and safety methods: Development of a phenomena identification and ranking table (PIRT), and validation experiments.

He also discussed R&D priorities for the Global Nuclear Energy Program (GNEP), proposed by the President, yet to be approved by Congress, with regard to LWR separations, a fast test reactor (with sodium technology) and transuranic fuel fabrication and recycle with metal or oxide fuels. The R&D priorities noted were LWR spent fuel separations technology and future fast burner reactor (ABTR) technology (cost reduction, in-service inspection, simulation and safety). TRU fuel development will involve fuel fabrication techniques and TRU sample preparation. A major concern is design of small reactors for countries that do not have large power grids.

MIT Professor Ernest Moniz, Co-Director of the MIT Laboratory for Energy and the Environment and Co-Chair of the MIT Energy Research Council, said it is very important to move forward with a nuclear power agenda. Global electricity demand growth and potentially stringent constraints on greenhouse gas emissions have reawakened interest in nuclear power. But nuclear power is going to have to do a lot better in the future, he said. For nuclear power, the TW (terawatt, or million megawatts) scale means tripling current global nuclear capacity by mid-century. Economics, safety, waste, and nonproliferation goals must be met if nuclear power is to reach the TW scale by that time. If it is realized, there would inevitably be a spread of nuclear power to new regions, including some with low security standards. Proliferation risks lie primarily with enrichment and reprocessing plants. Technical and institutional non-proliferation measures need to be in place before such expansion occurs.

Moniz put forward the following priorities for advancing nuclear power:

• Build new plants to deliver on schedule at low cost.
• Prepare the field for the growth of nuclear power internationally through the “fuel leasing” concept: major issues are security of supply, political asymmetries and incentives.
• Establish process and program plan for moving spent nuclear fuel (SNF) as soon as possible from nuclear sites to one or more federal locations for interim storage and security.
• Establish a substantial R&D program that will develop and evaluate options for nuclear fuel cycle development up to mid-21^st century and beyond for both open and closed fuel cycles, emphasizing the importance of establishing strong modeling and simulation.

Session 2: Role of Fuel Cycle Simulation

Speakers: Pavel Hejzlar, Michael Golay, and Andrew Kadak

Moderator: Neil Todreas

Dr. Pavel Hejzlar, Associate Director of CANES, spoke on the importance of fuel cycle simulation, using insights from initial MIT simulation results over the past three years. He emphasized the need for fuel cycle simulations, since goals are changing faster than technology development time. Further, advanced fuel cycle simulation tools will be necessary to provide guidance to policy makers for setting the goals, to DOE on decisions which reactor/fuel cycle options to choose for development and where to focus R&D, and to labs and industries. Simulation will be needed to indicate the most economical solutions (economics will be a key part of the model) and to quantify proliferation risks of various options.

Reduction of fuel cycle cost is an important priority. To that end, a recent MIT study considered two options:

• For the base case, separation and recycling costs will be paid from electricity generated from the recycled fuel, by ABR or the “Combined Non-Fertile and UO₂ Assembly” (CONFU) design developed by MIT. This is an energy view.
• Recycling can be treated as a form of waste management, and its cost should be assumed by electricity generated from the original fuel (LWRs). This is a waste view and reduces significantly incremental cost since the fee generates interest.

It turns out that given a realistic 2.0% nuclear growth per year, after the Yucca mountain repository is built, the fee of 1 mill per kWhr (or a slightly higher fee) can serve for waste recycling as well.

Fuel cycle simulations can provide significant insights into system behavior and can shape the R&D direction, reactor design and policy decisions. But Hejzlar noted that the goal of rapid reduction of transuranic waste (TRU) inventory in LWRs may be counterproductive to a large deployment of fast reactors. A key challenge for the fast reactor designers will be the significant reduction of capital cost if these systems are to make a substantial impact on TRU management. Further, simulations must incorporate other aspects, in particular proliferation resistance indices

Professor Michael Golay (Department of Nuclear Science and Engineering, MIT) spoke on elements of proliferation resistance. The overall goal of proliferation resistance is prevention of materials useful in nuclear weapons from being extracted from facilities for use in civilian nuclear power applications, he said. The major state-based proliferation threats are:

• Diversion of weapons material from facilities which under the United Nations Non-Proliferation Treaty (NPT) are forbidden for such use (i.e., from “declared”facilities)
• Misuse of declared facilities for clandestine production of weapons materials
• Abandoning the NPT and converting previously declared facilities to the production of weapons material.

Essential elements of proliferation resistance assessment are as follows:

• Definition of a safeguards context (sometimes requiring detailed expertise, and vulnerable to inconsistent treatment)
• Subjective expert data elicitation because objective data needed for an integrated assessment may be unavailable and the absence of which must be compensated for via subjective expertise
• The effects of proliferation resistance attributes upon the probabilities of proliferation-related events

The most recent proliferation assessment effort, begun in 2002, is the PR & PP Expert Group Study for GEN IV Nuclear Systems, involving an international experts group sponsored by DOE & NNSA. Its goal is to develop a comprehensive quantitative methodology. To date it has developed six proliferation resistance measures along with a general framework. However, proliferation resistance assessment has been slow in its progress to becoming a routine element of nuclear system performance and is of little importance in causing nuclear system improvements for the following reasons:

• It is weak, due to inconsistent development funding
• There is an absence of social demand for proliferation resistance
• There is no emergent consensus concerning how to perform proliferation resistance assessment.

Professor Andrew Kadak of MIT’s Department of Nuclear Science and Engineering spoke on nuclear waste aspects in fuel cycle simulation.

He posed the following set of questions/challenges:

• What is the waste form to be generated by GNEP? Do we assume all long-lived isotopes are “consumed”?
• What is the environment (now versus 10,000 years to time-to-peak dose of 700,000 years)?
• What is the effect of water or water vapor on the waste form?
• What are the chemical transport processes possible with the waste forms (e.g., radionuclides, molecules, colloids)?
• What are the key dose contributing isotopes?

There is a need for fundamental understanding to develop simulations of behavior, but such simulations depend on the knowledge of local environments. For completeness, fuel cycle simulations must define and understand waste forms for SNF disposal in a repository. But, Kadak said, we cannot assume that volume reduction is a benefit: rather, we must understand what is being disposed in terms of chemical constituency, heat load, and ability to be degraded and transported. Further, we need to capture the range of environmental conditions present for the time scale of concern.

Session 3: Materials Modeling and Simulation

Speakers: Sidney Yip, Paul Meakin, Tomas Diaz de la Rubia, and Jim Roberto

Moderator: Prof. Jacopo Buongiorno

Professor Sidney Yip of MIT’s Department of Nuclear Science and Engineering provided a broad overview, noting the intersection of materials modeling for advanced systems with simulation in the radiation environment. Computational science has become an important capability by solving complex science with computing capability. In this regard, he cited two recent reports, (1) Report to the President: Computational Science – Ensuring America’s Competitiveness, completed in June 2005 by the President’s Information Technology Advisory Committee, and (2) Simulation-Based Engineering Science - Revolutionizing Engineering Science through Simulation, February 2005, Report of the National Science Foundation’s Blue Ribbon Panel on Simulation-Based Engineering Science. A major problem being addressed is radiation damage, a project of MIT Professor Ron Ballinger, on the nano-scale, micro-scale, and macro-scale levels. Five MIT professors are working on each scale of this project.

The MIT-CANES-ACE Initiative (March 2006), intends to integrate experimental techniques with modeling and simulation to understand materials performance in aggressive environments by leveraging the extensive multiscale materials modeling and simulation interests and capabilities across MIT in partnership with INL and other labs, and in collaboration with a number of universities. In this connection, Prof. Yip also noted an MIT course, Introduction to Modeling and Simulation, given in Spring 2006. A key concept is modeling as the physicalization of a concept, while simulation is its computational realization. The course has fourteen lecturers from 7 different MIT departments.

Paul Meakin, Director of INL’s Center for Advanced Modeling and Simulation (CAMS) spoke on R&D priorities for advanced reactors. He described the mission of CAMS, which is to ensure that the Idaho National Laboratory has the computing resources (human, hardware, software, communications and collaborations) needed to support its goal of becoming a preeminent national nuclear energy laboratory with synergistic world-class multi-program capabilities.  When mature, CAMS programs will span the entire spectrum of INL science and engineering. Initially, however, constrained by resources, CAMS is focusing on a few themes of importance to NE and of secondary benefit to other programs while building infrastructure for the entire Laboratory, addressing its high-performance computing needs. The plan is to take a balanced approach to upgrade capabilities in human resources, computer hardware, software/code, visualization, infrastructure, communication, and collaboration. Unique materials challenges are exacerbated by nuclear fission and fusion energy system requirements. CAMS is planned as a virtual nuclear reactor center to develop high-resolution computer models for advanced nuclear reactors in a national lab/university/industry partnership.

Tomas Diaz de la Rubia (Associate Director for Chemistry and Material Science, Lawrence Livermore National Laboratory) spoke on the multiscale modeling of materials for national security applications. LLNL, which has one of the world’s most powerful computers, is facing the evolving nature of national security imperatives for the future, as both energy security and health security are becoming increasingly important. LLNL’s Stockpile Stewardship Program (SSP) has driven significant scientific and technical achievements, said de la Rubia, as well as presenting major technical challenges such as:

• National Ignition Facility (NIF): to achieve fusion ignition in the laboratory
• Advanced Simulation Capability (ASC): to enhance simulation capability more than a million-fold
• Materials: to understand their properties (and aging) from the atomistic to engineering scale
• Advanced radiography: to image dynamic behavior at engineering scale

The SSP uses predictive models to mitigate risk associated with national policy decisions and to eliminate technological surprises that could be encountered as systems age. Successful implementation requires experiments on the scale of simulations. Developing a suite of scalable materials science codes, LLNL wants to partner and focus its tool set on reactor fuel issues. Ab initio methods have been used to calculate defect and bulk physical properties of actinides. The CALPHAD thermodynamic database has been extended through ab initio calculations to assess the phase stability of plutonium alloys. CALPHAD combined with DICTRA and phase field modeling can predict phase segregation.

A movie was presented showing simulations of defect interactions uncovering the origins of strength and ductility of metals. Microstructural information is used to determine macroscopic mechanical behavior of irradiated materials. Materials modeling and simulation will reduce design uncertainties leading to accelerated insertion of new materials at reduced cost. This will provide a sound scientific basis for the design and development of new nuclear fuels, cladding and structures and will narrow the parameter space that must be experimentally investigated.

Jim Roberto, Deputy Director for Science and Technology at the Oak Ridge National Laboratory (DOE) spoke on materials modeling and simulation for advanced nuclear energy systems at ORNL. ORNL is DOE’s largest multipurpose science and energy laboratory. It is the nation’s largest unclassified scientific computing facility, and a $300M modernization effort is underway.

ORNL’s capabilities in nuclear energy R&D include reactor materials, fuels, chemical separations, theory and modeling, and a number of related facilities for the High Flux Irradiation Reactor (HFIR): Materials irradiation including fast neutron capability; REDC: R&D-scale separations and actinide fuels research; hot cells: irradiated materials and fuels characterization; materials synthesis and characterization facilities; and leadership computing.

Nuclear materials modeling and simulation research at ORNL provides

• A prominent role in materials modeling and simulation for previous and ongoing radiation effects programs including Liquid Metal Fast Breeder Reactors (LMFBR), basic energy sciences (underlying mechanisms), NRC (close association with codes and standards), fusion (primary steward of fundamental radiation effects research in recent years), and Generation IV (leading cross-cutting theory and modeling task)
• Modeling activities tightly linked with strong experimental programs to maximize relevance and impact of both
• Current research efforts cut across division and program lines

Recent activities at ORNL include development of new He-Fe potential based on ab initio calculations of He defect properties in Fe, and Fe-He cross-potential fit for three commonly used iron matrices. ORNL research directions include electronic structure calculations to obtain intrinsic and defect properties in iron and its alloys; large-scale atomistic models to determine material parameters and processes that interact to control migration and non-equilibrium, accumulation of defects, transmutant gases, and solute elements in complex alloys; integrated multiscale (atomistic, mesoscopic, and continuum) microstructural models for predicting deformation and fracture behavior (e.g. plastic instability) in complex alloys with realistic loading conditions; and petascale computational challenges.

Session 4: Fuel and Materials for LWRs

Speakers: R. Yang, M. Kazimi, and J. Tulenko

Moderator: J. Buongiorno

Dr. Rosa Yang, Director of Materials and Chemistry in the Nuclear Power Division, Electric Power Research Institute (EPRI), spoke on LWR fuel reliability status. She emphasized the need for the nuclear industry to refocus its effort to meet the zero-defect goal by 2010. Water chemistry is a major part of EPRI work, particularly with regard to improving plant performance for PWR and BWR reactors as well as providing new licensing criteria for high burnup fuel.

There exists an opportunity for industry and government partnership to develop advanced LWR fuel for existing and new LWRs; to leverage the relationship and infrastructure at INL with regard to hot cells, test reactors, analytical capabilities and staff expertise; and to provide favorable responses from preliminary discussions held with DOE. This fuel may require more than 5% enrichment, three to four fuel designs (one per vendor/reactor design) starting with existing innovative concepts and addressing the need to address proprietary issues. A host plant needs to be identified early on, and 80-20 government funding is critical. Fuels with significantly increased reliability and operating margins will enhance the competitive edge of nuclear power and will be commercially available by 2020, but it is necessary for the US to develop and expand its own infrastructures. The next steps for developing advanced LWR fuel are to quantify the fuel cycle economics benefit and amount of spent fuel reduction and to work with industry leadership to gain government funding.

Professor Mujid Kazimi of MIT and Director of CANES, posed the question: “Advanced LWRs: Can we make them worthwhile?” It is evident that we will be using LWRs for a very long time: at least 50% of all nuclear reactors through the 21^st century. Consequently, there have been a considerable number of recent projects for improved LWRs in terms of increasing power density, enhancing operations, and advanced concepts. In addition, there is a need for small reactors of a 15-20 year single batch running time, which are now being considered by various vendors.

One issue is the limits of LWR power density. These limits include:

• Critical Heat Flux (for PWR reactors) and Critical Power (for BWR reactors)
• Maximum fuel temperature below melting
• Peak cladding temperature during loss of coolant accidents (LOCA)
• Moderator-to-fuel ratio below that at peak reactivity
• Velocity implications for pressure drop, lift-off and vibrations.

Methods to overcome these are being considered at CANES, often with support from DOE, INL, or industry. A recent project examined the use of annular fuel with internal cooling for PWRs and BWRs. Annular fuel promises 50% upgrade in PWR power density. The situation is not clear for BWRs. Nanofluids for Nuclear Applications are being investigated by CANES in separate collaborations with INL, AREVA and NRC. A nanofluid is an ‘engineered’ colloid, i.e. base fluid (water, organic liquid, gas) plus nanoparticles (Al₂O₃, ZrO₂, SiO₂, CuO, Cu, Au, C). These collaborations are exploring their potential use in nuclear applications, such as PWR primary coolant and PWR and BWR safety systems. A simple pool boiling wire experiment has shown that a significant increase (30-80%) in critical heat flux (CHF) can be achieved at modest nanoparticle concentrations.  Measurements of nanofluid CHF in a forced convection loop are also underway at MIT.

In another project involving MIT, Westinghouse, Gamma Engineering, and ORNL, ceramic composite SiC clad is being investigated. SiC has advantages over Zircaloy. These include:

• Constant LOCA and RIA performance during transient
• Very low operating and accident corrosion rates
• Potential for eliminating DNB and Dryout issues
• Reduced cladding neutron absorption cross sections

Westinghouse and Gamma are supporting initial irradiations using the MIT reactor.

Kazimi pointed out that we should look at new types of LWR assemblies to improve the waste situation. There are many ideas for improving performance of LWRs from the economic, safety, and waster generation standpoints.

Professor James Tulenko of the Nuclear/Radiological Department at the University of Florida gave an overview of LWR fuels development. Research in nuclear fuels seeks to improve and extend performance and improve and extend performance and reduce costs.  To accomplish this we must understand all the phenomena that occur in the fuel.  We must also consider fabrication, operation and the back end stages of storage, reprocessing, refabrication, and disposal, Tulenko said. Fuel is the consumable part of the reactor; therefore it is important to modify fuel consumption, allowing the fuel to remain longer in the reactor.  We must study the materials making up the fuel assembly, understanding how they perform under irradiation along with their resistance to temperature, pressure, stress, fretting, neutron flux, hydriding, and corrosion.

LWR fuels must be developed so as to be compatible with water. Incompatibility with water eliminates uranium carbide, uranium nitride, matrix materials such as magnesium oxide, and other water soluble materials. Molten materials are not allowed in fuels at anticipated operating conditions. Annular pellets must ensure that material does not relocate into the annulus.

To improve thermal conductivity, the University of Florida has looked at adding high conducting fibers to the fuel, coating fuel particles prior to pelletizing with high conducting materials and inert matrix fuel carriers. Improving thermal conductivity will reduce swelling, fission gas release, columnar grain growth, can eliminate pellet cracking, and potentially eliminate LOCA concerns and mitigate other accident conditions.  Improved fuel lifetime involves obtaining better pellet characteristics, better burnable poison materials, better cladding, and better fuel assembly structure performance under irradiation.

Finally, there is a great need for a first principles fuel performance code. All current codes are empirical based on irradiation measurements and are limited to the range of materials tested. Development of a code with detailed materials input from databases of nuclear, materials behavior determined from thermodynamic, atomic-level and electronic-level calculations and simulations would allow material performance to be assessed quickly and with minimum expense, with only selected materials requiring expensive and time consuming irradiation confirmatory tests. Extensive work is currently underway, but much more needs to be done.

Session 5: Fuels and Materials for Advanced Reactors

Speakers: R. Ballinger, D. Crawford, T. Allen, and G. Hayner

Moderator. A. Kadak

Professor Ronald Ballinger of MIT’s Department of Nuclear Science and Engineering spoke on the development of better materials for nuclear system applications. There is a close coupling that can be made between experiments and modeling/simulation that was not possible in past years. Modeling of real systems is in the offing, and we now have the experimental tools that allow very detailed interrogation and that can be used in combination with new modeling and simulation capability. The time is right, said Ballinger, to make the connection between engineering materials and the modeling of these systems.

The key areas of investigation are the material/environment interface, which entails modeling of aggressive chemistry, modeling of material evolution, and modeling the interface (with regard to film formation, gas/water/material interaction, crack tip environment, mechanical evolution, microstructural evolution, radiation effects and multi component chemistry) as well as the experimental levels of testing (sophisticated chemistry and mechanical control) and detailed analysis. Types of systems being studied are current LWR systems (PWR and BWR) and future reactor systems (high temperature gas, liquid metal (sodium, Pb, and Pb-Bi), and supercritical systems (CO₂ and water).

The work is predicated upon a strong interface between university and industry through which it is necessary to blend the physics with engineering to make the research viable. Key issues being considered are temperature with future reactors ‑ specifically high-temperature gas reactors ‑ fuel reliability (a major issue for HTGRs), and radiolysis. General corrosion, localized corrosion, thermal effects and radiation effects are all key phenomena being studied and are all interconnected.

Doug Crawford, Manager of Advanced Fuel Programs in the Nuclear Fuels and Materials Department at INL, spoke on advanced fuel development. Presumed Goals for Advanced Fuel Cycle Development include demonstration of feasibility and engineering-scale success of fuel recycle and demonstration of safe and reliable reactor performance with recycled fuel. The research also emphasizes innovative fuel concepts such as high-density, low-enrichment fuel for research reactors, and gas-cooled fast reactor (GFR) fuel. GFR fuel development is a challenging environment in terms of conceptualization, fabrication development, and irradiation tests.

Crawford also discussed fuel qualification for the VHTR concept of NGNP, being addressed in a collaborative program between INL, ORNL, BWXT, and others. The objectives of the collaboration are to demonstrate a reliable and reproducible TRISO fuel fabrication process in the US and to gather data to support licensing of a VHTR using TRISO fuel. Activities include investigation of fabrication parameters, development of production characterization techniques, preparation of fuel specification and quality assurance criteria for production, irradiation testing and out-of-pile safety testing to demonstrate necessary low-breach rates, and modeling of fuel performance to demonstrate an understanding of fuel performance phenomena and impact on fuel reliability.

Fuel design evolutions and innovations are being considered, Crawford said, for research reactors, gas-cooled fast reactors, and sodium-cooled reactors. Low-conversion objectives for fast reactors motivate the use of high-TRU:U and high-TRPu:Pu fuels, with additional challenges to fuel performance. Previous experience with metal alloy fuel and mixed oxide fuel provides a relatively low-risk path for fast reactor transmutation fuel. Remaining fabrication development work includes process development for retention of volatile americium in recycle and fabrication, the reduction of process loss in molds, dies, and equipment, and the development and demonstration of remote fabrication of recycled fuel.

Professor Todd Allen of the Department of Engineering Physics at the University of Wisconsin spoke on R&D for advanced metallic cladding and structural materials.  He described the issues for fuel assembly performance in terms of design functions, providing support and protection for the fuel-pin bundle and other components of the subassembly, providing a controlled path for the primary coolant, and providing a compact structural unit that can be easily moved in and out of the core by a refueling machine. Design function also includes interaction with adjacent subassemblies, retaining ring, and core support plates in a manner that assures safe and predictable reactor geometry.

Also discussed were design issues of swelling, creep, fatigue, and toughness as well as reduced limits for weldments. Allen said survival of cladding must be predictable: issues include:

• Wastage (corrosion by the coolant and fuel cladding chemical interaction)
• Strain (Fission gas pressurization, swelling and associated creep of constrained components, and fuel cladding mechanical interaction)
• Microstructural stability during accident transients

George Hayner, Manager of NGNP Materials Program at INL, spoke on materials R&D priorities for the VHTR. >The Next Generation Nuclear Plant (NGNP) is a very high temperature reactor (VHTR) that uses helium as a coolant and graphite as a neutron moderator. Materials R&D activities in support of the NGNP have been ongoing for about two years in the US, primarily at INL and ORNL. Current materials R&D activities being performed are associated with graphite, high temperature metallic alloys, ASME code issues, and composites. Recently, the Energy Policy Act (EPACT) of 2005 initiated activities directed at the construction of an NGNP prototype at the INL site. As a result of the EPACT, the INL drafted a preliminary NGNP Project Management Plan, but the actual US NGNP design has not been selected at this point.

A major component of the research is heat exchanger design. The IHX or PHX is a critical component due to high temperature exposure, creep and fatigue affects, and complexity of the design. The IHX closest to the reactor would probably have to be fabricated from a high temperature mature metallic alloy. Compact heat exchangers using a printed circuit or plate fin design are currently favored rather than a shell in tube IHX or PHX, although the HTTR (Japan) uses a helical coil tube in shell IHX. Development of the IHX or PHX would require, at a minimum, materials selection and qualification, a possible new ASME Code Case, fabrication development, and mockup testing.

Both GFR and VHTR concepts will require composite materials to achieve design goals, most importantly core internal temperature. Presently, there are only two viable candidate composites for use in nuclear power applications: C/C and SiC/SiC. C/C composite are more mature and have clear advantages in cost, manufacturability and some thermo-mechanical properties (eg thermal conductivity.). SiC/SiC has a clear advantage for irradiation stability, specifically a lower level of swelling and retention of mechanical properties. Offers the possibility of a lifetime component for control rod application to NGNP (C/C would require 2-3 replacements over life.) The application of ceramic composites will require substantial investment in ASTM development, NDE development, and must be handled by prototyping and proof testing, but with substantial additional costs compared to more conventional alloys.

Hayner noted that NRC licensing, in part, has previously relied on the ASME code for adequacy of structural design.  The ASME code is not currently ready for the design of VHTR reactors from the materials perspective.  Code qualification is required to support the VHTR.  Further qualification of Grade 91 steel will be needed to provide additional margin for the VHTR primary vessel.  In addition, further qualification of high temperature alloys at higher use temperatures than fossil power is required to support the VHTR. The development of composites for the GFR and the VHTR is needed, Hayner said, for critical components for designs currently being proposed and will be required to a greater extent in the future of nuclear power for higher temperature applications.

Session 6: Risk-Informing the Design Process

Speakers: G. Apostolakis, K. Fleming, R. Denning, and T. Aldemir

Moderator: A. Kadak

Professor George Apostolakis of MIT’s Department of Nuclear Science and Engineering led off the session with the question, Why Risk-Informed Design? He said the USNRC is preparing a new risk-informed licensing process for future reactors. Important uncertainties are being identified early.  He suggested that the combination of the structuralist (i.e., defense in depth) and the rationalist (i.e., risk-based) safety philosophies could be addressed early in the process, but PSA methodological needs are identified early so that improvements can be made resulting in a more risk-informed design.

Apostolakis said more data appropriate for gas reactors are needed. PRA insights have been used in advanced reactors design efforts to

• change the configuration of the design
• add a secondary onsite power source
• add a nitrogen accumulator system

Several designs have satisfied the probabilistic goals but not the deterministic criteria.  Mircoturbines have never been used in a NPP emergency power supply system. As such, they will be thoroughly scrutinized during the licensing process. Again, more data are needed. Adding redundant ECCS loops beyond 2x100% capability does not result in significant improvement. This is due to the insensitivity of the CCF models. No quantitative guidance exists as to how the values of the beta factor change when the design changes. However, deliberation allows the inclusion of best engineering practices as well as comparison with other NERAC goals such as sustainability, economics, reliability, proliferation resistance, and physical protection.

Karl Fleming, Senior Engineer at Technology Insights, discussed the use of Probabilistic Risk Assessment (PRA) to support the design and licensing of high temperature gas-cooled reactors. PRA provides a number of opportunities:

• To incorporate risk significant sequences into design basis envelope.
• To incorporate risk insights into the design
• To fulfill the capability for risk informed; performance based design and regulation

But challenges to its use persist, namely:

• lack of design and operational details for reactors that are still in the pre-conceptual or conceptual design stage
• lack of relevant service experience from which to derive a PRA database
• increased emphasis on passive systems to perform safety functions
• need to address events and sequences within and beyond the design basis
• inapplicability of risk metrics such as core damage frequency to reactors with inherent reactor characteristics that are fundamentally different from those of LWRs
• lack of experience by reviewers and regulators with PRA as it has been applied to non-LWRs.

Lessons have been learned from LWR PRA. Risk for LWRs is dominated by beyond-design-basis core damage events for which containment is not specifically designed. Preoccupation with single failure criterion has shifted attention away from multiple dependent failures. The role of inherent reactor characteristics vs. engineered safety features remains underappreciated. The deterministic approach to safety and licensing prior to the advent of PRA led to poor allocation of resources relative to risk contributors.

The most important justifications for early introduction of PRA are its powerful and unmatched capability to systematically enumerate events and event sequences that need to be considered, increased objectivity in selection of design and licensing basis events by identifying events that are most risk significant, and its capability to reveal sources of uncertainty to be considered within regulatory requirements.

Rich Denning and Professor Tunc Aldemir of the Ohio State University Engineering Program discussed Level 2 PRA as a design tool. A distinction was made between Design PRA and Operating PRA. PRA at the design stage is a statement of the design goal for the plant, while PRA during the period of operation is an assessment of the analyst’s perception of the risk. The extent to which design PRA changes in transition to operation depends on quality assurance, compromise to reduce cost, and the reduction of epistemic uncertainty. Safety goals are interpreted at Level 3 but are implemented in design at Level 2 (which requires some dispersion analyses to translate Level 2 outcomes to consequences). Magnitude and timing of radionuclide release determine the potential for offsite consequences: particularly, early fatalities. Level 2 results impact site suitability, need for containment, need for emergency response, and public perception.

Level 2 methods have not advanced to the same degree as Level 1 methods. In general, static trees assume a fixed order of events, branch probabilities are not obtained by means of a systematic process -non-mechanistic approximations are used to compare loads with fragility curves, and epistemic uncertainties are mixed with aleatory uncertainties. Advanced reactor issues take into account the following: For passive safety systems, epistemic phenomenological uncertainties can be the principal contributors to uncertainty; containment failure probabilities can be very small, increasing the difficulty of quantification.

Existing work indicates that uncertainties in the timing of events can significantly affect their predicted consequences in Level 2 PRAs for current plants.  Uncertainty assessment is expected to play a more important role in the Level 2 PRA of Generation IV reactors. A code-independent computational tool is being developed for the mechanized and computationally efficient generation of dynamic event trees for Level 2 PRA. Future work will address epistemic uncertainty quantification (joint NERI project with Purdue University).