Global Nuclear Energy Partnership Research Articles

Whatever Happened to the Australian Role in the Global Nuclear Energy Partnership?

In 2003, Iraq was invaded, ostensibly to remove a nuclear threat, by a coalition led by George W. Bush. At the same time select allies were invitited to participate in Bush's Global Nuclear Energy Partnership which aimed to limit the spread of nuclear enrichment and reprocessing. This came as climate change gave emphasis to the development of nuclear energy, especially in Asia. With an abundant supply of uranium and strict nuclear safeguards, Australia was well placed to provide a site for the full suite of nuclear services. The recent AUKUS nuclear submarine agreement has underscored its failure to do so. This article makes the case for the adoption of nuclear power as a necessary step in the development of advanced manufacturing; the provision of a domestic capability to fuel nuclear attack submarines and other naval craft; and as a contribution to global nuclear non‐proliferation.

Technology readiness assessment of partitioning and transmutation in Japan and issues toward closed fuel cycle

This paper treats a technology readiness assessment of partitioning and transmutation in Japan and issues toward closed fuel cycles. The generic technology readiness level in this study is based on the definition in the Global Nuclear Energy Partnership: TRL 3 shows the status that critical function are proved and elemental technologies are identified; TRL 4 represents the level that the related technologies are validated at bench-scale in laboratory environment; and TRL 5 indicates the completion of the development related to the subsystem and elemental technologies. The reviewed technological area includes the partitioning and transmutation technologies for minor actinide cycle: fast breeder reactor, and accelerator driven system for minor actinide transmutation; partitioning processes; and minor actinide bearing fuels. The assessments reveal that the TRLs stay around the final step of concept development (TRL 3) and the first half step of elemental technology development (TRL 4) because each system requires more development of its elemental technologies. The fast breeder reactor is assessed to be at TRL 4 because critical experiments with americium and neptunium would be additionally required, and the technology for the accelerator driven system reaches TRL 3 due to several researches. Consequently, a common key issue is how nuclear calculation methodology will be validated for the MA-bearing-fuelled core; however, critical experiments with several kilogrammes of americium or more are difficult in the existing experimental facilities. On the other hand, engineering scale tests of the MA partitioning processes using actual spent fuel, and engineering scale of fabrication and irradiation tests with the separated materials are required to achieve the second half step of elemental technology development (TRL 5); however, they could be a massive investment. Therefore, laboratory-scale tests using actual material are proposed. The tests simulate a nuclear fuel cycle: partitioning actual spent fuels; and fabricating MA bearing fuels from the extracted materials; and then irradiating them. It should be that the tests advance technological readiness from TRL 4 to TRL 4+, which is a reasonable and feasible pathway.

Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio

If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology’s “Future of the Nuclear Fuel Cycle” proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas.The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al.No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time-dependent measures of performance including uranium usage, TRU inventory, and radiotoxicity to evaluate the complex impacts of transition from the current uranium-fueled LWR system, and other more realistic impacts that may not be intuited from the time-independent steady-state conditions of the end-state fuel cycle. These analyses were performed using the Verifiable Fuel Cycle Simulation Model VISION.Compared with static calculations, dynamic results paint a different picture of option space and the urgency of starting a FR fleet. For example, in a static analysis, there is a sharp increase in uranium utilization as CR exceeds 1.0 (burner versus breeder). However, in dynamic analyses that examine uranium use over the next 1 to 2 centuries, behavior as CR crosses the 1.0 threshold is smooth, and other parameters such as the time required outside of reactors to recycle fuel become important.Overall, we find that there is no unambiguously superior value of TRU CR; preferences depend on the relative importance of different fuel cycle system objectives.

Feasibility Study on Pyroprocessing of Spent Nuclear Fuels for Nuclear Power Plants

For the nuclear non-proliferation strategy as well as the energy production in the spent nuclear fuels, it is proposed to justify the analysis of the Pu/U non-separation characteristics in pyroprocessing. The model of the pyroprocessing development in Korea for nuclear proliferation is constructed successfully. Pyroprocessing is a promising option following the Global Nuclear Energy Partnership as well as the Non-nuclear Proliferation Treaty. In spite of the skeptical review about the Global Nuclear Energy Partnership, it is reasonable for Korea to develop pyroprocessing technology, due to the request for high-level waste repository and the strong necessity of the stable nuclear fuel supply. The possibility of license should be considered, which is the earlier stage comparing to economics, safety, environmental aspect, plant technology, and social acceptance. The volume reduction of spent nuclear fuels is another issue due to the time limitation of the site storage by the year 2016. Recycling of a sodium fast reactor is supposed to develop with pyroprocessing by the year 2040.

Enhancing TRU burning and Am transmutation in Advanced Recycling Reactor

This paper presents about conceptual designs of Advanced Recycling Reactor (ARR) focusing on enhancement in transuranics (TRU) burning and americium (Am) transmutation. The design has been conducted in the context of the Global Nuclear Energy Partnership (GNEP) seeking to close nuclear fuel cycle in ways that reduce proliferation risks, reduce the nuclear waste in the US and further improve global energy security. This study strives to enhance the TRU burning and the Am transmutation, assuming the development of related technologies in this study, while the ARR based on mature technologies was designed in the previous study. It has followed that the provided TRU burning core is designed to burn TRU at 28 kg/TW thh, by adding moderator pins of B 4C (Enriched B-11) and the Am transmutation core will be able to transmute Am at 34 kg/TW thh, by locating Am blanket of AmN around the TRU burning core. It indicates that these concepts improve TRU burning by 40–50% than the previous core and can transmute Am effectively, keeping the void reactivity acceptable.

A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10–15 years and would take 3–5 years to construct.

Nuclear mechanism for TRU burning oxide fueled core in Advanced Recycling Reactor

This paper presents an approximation approach to predict the core characteristics based on parametric survey and an analysis of nuclear mechanism in a conceptual nuclear design for enhanced transuranics (TRU) burning mixed oxide fueled and sodium cooled fast reactor which can be realized in the near future. The design study of Advanced Recycling Reactor was conducted in the context of the program for the industry in Global Nuclear Energy Partnership initiatives, including a core in the first plant for demonstration and cores with enhanced TRU burning capability for the future plants. Both concepts for the first plant; low core height and large volume fraction of structure are deployed, seeking small TRU conversion ratio (CR) * and small void reactivity which are crucial in the design, but different approaches. In this paper, the TRU CR and the sodium void reactivity have been approximated with a single equation in these concepts, based on the theoretical formula related to the chain reaction in the reactor and the calculation results. Shortening the core height and increasing the structure volume fraction will enhance TRU enrichment through increased neutron leakage and capture, which will reduce a ratio of U-238 to sum of Pu-239 and Pu-241 so that TRU CR decreases from 0.78 to 0.57. A small ratio of sodium loss to plutonium fissile will be effective also in the reduction of positive reactivity effect by spectral hardening. On the other hand, when this ratio and geometrical buckling of flux reduce, negative contribution by the neutron leakage becomes small. Theses relations related to the positive void reactivity have been formularized by the approximation with few parameters within several percent respectively as well as the TRU CR, indicating that one of dominating parameters is the ratio of sodium loss to plutonium fissile in the void reactivity at large fast reactors. * = (1 − net loss of TRU/loss of heavy metal).

Network Computing Infrastructure to Share Tools and Data in Global Nuclear Energy Partnership

CCSE/JAEA (Center for Computational Science and e-Systems/Japan Atomic Energy Agency) integrated a prototype system of a network computing infrastructure for sharing tools and data to support the U.S. and Japan collaboration in GNEP (Global Nuclear Energy Partnership). We focused on three technical issues to apply our information process infrastructure, which are accessibility, security, and usability. In designing the prototype system, we integrated and improved both network and Web technologies. For the accessibility issue, we adopted SSL-VPN (Security Socket Layer-Virtual Private Network) technology for the access beyond firewalls. For the security issue, we developed an authentication gateway based on the PKI (Public Key Infrastructure) authentication mechanism to strengthen the security. Also, we set fine access control policy to shared tools and data and used shared key based encryption method to protect tools and data against leakage to third parties. For the usability issue, we chose Web browsers as user interface and developed Web application to provide functions to support sharing tools and data. By using WebDAV (Web-based Distributed Authoring and Versioning) function, users can manipulate shared tools and data through the Windows-like folder environment. We implemented the prototype system in Grid infrastructure for atomic energy research: AEGIS (Atomic Energy Grid Infrastructure) developed by CCSE/JAEA. The prototype system was applied for the trial use in the first period of GNEP.

International cooperation in the sphere of peaceful uses of nuclear energy as a resolution of energy supply and nuclear proliferation problems

Comparative analysis of the Russian and the US initiatives. The article gives an outline of such a promising branch of international cooperation as cooperation in the sphere of peaceful uses of nuclear energy, in particularly its multilateral aspects – initiatives of States based on the multilateral principle of uses of nuclear power. The comparative analysis of the two large-scale initiatives in the field ofmultilateral approaches to the nuclear fuel cycle – these are the Russian initiative on the development of the Global infrastructure of nuclear energy and the American Global nuclear energy partnership –made in the article discloses the main principles of work of the abovementioned mechanisms of interaction as well as their advantages and disadvantages. The goal of such an analysis is to figure out which one has a greater potential for international security and future development of the nuclear energy sector.

High‐level radioactive waste management in the USA

High‐level radioactive waste (HLW) disposal policy in the USA since 1987 has focused on a site in volcanic tuffs 305 meters beneath Yucca Mountain, Nevada, with current plans calling for the repository to be opened in 2017 subject to approval by the Obama Administration. Yet the offsite radiation release standards of the US Environmental Protection Agency are still being finalized, and there is significant doubt over whether a repository at Yucca Mountain can be successfully licensed by the Nuclear Regulatory Commission. Moreover, the proposed repository has a capacity cap insufficient to accept the total HLW stream from the civilian and military sectors, and focus on a single site is unnecessarily risky. Thus, the HLW problem in the USA remains far from being solved. After providing an overview of the nuclear waste regime of the USA, this paper reviews the technical, legal, and political status of the Yucca Mountain project. A large range of current issues and social values are discussed in this context, including the recently proposed Global Nuclear Energy Partnership that would revive spent fuel reprocessing and has important implications for national security and nuclear non‐proliferation.

Destruction rate analysis of transuranic targets in sodium-cooled fast reactor (SFR) assemblies using MCNPX and SCALE 6.0

The purpose of this research was to model destruction rates of potential transuranic (TRU) targets made of americium and other inert matrix materials in the core of a sodium-cooled fast reactor (SFR) using depletion analysis in MCNPX and SCALE 6.0. The research is concentrated on the destruction of long-lived transuranics in an effort to reduce the overall volume and activity of spent nuclear fuel as proposed by the Global Nuclear Energy Partnership (GNEP) Technology Development Plan ( GNEP/TIO, 2007). The model uses a once-through approach in a single assembly of a sodium-cooled fast reactor based off of the Clinch River reactor design. Analysis was performed on both unmoderated and moderated targets in the fast reactor assembly. Optimization of the transmutation scheme could result in a 24-percent reduction in mass of long-lived transuranic actinides. Further iterations and optimization of the target material and geometry could greatly increase the actinide destruction rate. The overall reduction of TRU in spent nuclear fuel wastes will increase the usable life of the geological repository for high-level waste.

The Fast Reactor and Its Fuel Cycle Developments in Japan: Can Japan Unlock its Development Path?

This paper reviews the history, status, and probable future of fast reactor and associated fuel cycle development in Japan. The fast breeder reactor and its closed fuel cycle have been the cornerstone of Japan's nuclear-energy development program since the 1950s. For economic, technological, and political reasons, Japan's development and implementation of these technologies is significantly delayed. The budget for fast breeder reactor development has steadily declined since the mid-1990s, and its commercialization target has slipped from the 1980s to the 2050s. An accident at the Monju prototype reactor contributed to delays and triggered a fundamental shift from R&D and early commercialization to an emphasis on advanced fuel cycles. Nevertheless, Japan is still committed to fast-reactor development. This paper examines the motivation for its continued commitment to a fast reactor program and concludes that several non-technological factors, such as bureaucratic inertia, commitments to local communities, and an absence of R&D oversight, have contributed to this entrenched position. Japan is currently reorganizing its R&D programs with the goal of operating a demonstration breeder reactor by approximately 2025. This effort is in response to the government sponsored “Nuclear Power Nation Plan” and the Bush Administration's Global Nuclear Energy Partnership. Breeder R&D programs face significant obstacles such as plutonium-stockpile management, spent fuel management, fuel cycle technologies, and arrangements for cost and risk sharing between government, industry and local governments. As a result, it is unlikely that fast breeder reactor (FBR) and fuel cycle development programs will move forward as planned.

Metallic fast reactor fuel fabrication for the global nuclear energy partnership

Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.

Technology Gap Analysis on Sodium-Cooled Reactor Fuel-Handling System Supporting Advanced Burner Reactor Development

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner resactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology.The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phénix, Superphénix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components—such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations—have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

THE GLOBAL NUCLEAR ENERGY PARTNERSHIP

Since the dawn of the atomic age, the United States has sought to encourage the use of nuclear energy while minimizing the proliferation risks associated with it. The latest U.S. initiative that sets out to accomplish this is the Global Nuclear Energy Partnership (GNEP), which, in its current form, potentially includes the spread of sensitive nuclear technologies around the globe. This article examines the concerns surrounding the proliferation of these technologies and surveys their history both domestically and internationally. In identifying these concerns, the author argues that GNEP needs to be considered in the context of the Atoms for Peace program; that it erodes the successful thirty-year U.S. position against reprocessing; and that it allows for the spread of technologies that are not proliferation-resistant.

Thermal Analysis of a Fuel-Handling System for Sodium-Cooled Reactor with Minor Actinide–Bearing Metal Fuel

The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide–bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the sodium coolant. The cladding temperature requirement is maintained below the creep temperature limit to avoid any damage before core installation. The thermal analysis shows that a helium gas–filled cask can accommodate ABR-1000 fresh minor actinide–bearing fuel with 700-W decay heat. The above analysis results revealed the overall requirement for minor actinide–bearing metal fuel handling. The information is thought to be helpful in the design of the ABR-1000 and future sodium-cooled-reactor fuel-handling system.

Technology Gap Analysis on Sodium-Heated Steam Generators Supporting Advanced Burner Reactor Development

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The Advanced Burner Reactor (ABR) is a fast reactor concept that supports the GNEP fuel cycle system. Since the Integral Fast Reactor (IFR) and Advanced Liquid Metal Reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP ABR program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology.A sodium-heated steam generator is one of the key components in the fast reactor system since it provides interface between sodium and water. In this gap analysis, information of fabrication and operation experiences in reactor plant steam generators and prototype steam generators was carefully reviewed, for example the Enrico Fermi Atomic Power Plant, the Prototype Fast Reactor (PFR), and Phénix steam generators; the Babcock & Wilcox helical coil tube, 70 MW; the Westinghouse double-wall tube, 70 MW; the Clinch River Breeder Reactor (CRBR) full-scale evaporator; the Superphénix prototype helical coil tube, 45 MW; the SNR-300 prototype straight tube, 50 MW; the SNR-300 prototype helical coil tube, 50 MW; and the Monju prototype helical coil tube, 50 MW. The results of this evaluation indicate that straight and helical coil tube steam generators are the best immediate candidate designs for producing reliable steam generators for future sodium fast reactor applications. Though the design comparison suggested that the straight tube type has the advantages of compactness and ease of inspection, prototype tests revealed more technical problems than the helical modules. From the viewpoint of tube material, 2¼Cr steel has been well established, and Incoloy® 800, 9Cr, and 12Cr steels are available as higher-performance materials.

Processing Neutron Cross Section Covariances using NJOY-99 and PUFF-IV

With the growing demand for multigroup covariances, the National Nuclear Data Center (NNDC) has been experiencing an upsurge in its covariance data processing activities using the two US codes NJOY-99 (LANL) and PUFF-IV (ORNL). The code NJOY-99 was upgraded by incorporating the new module ERRORJ-2.3, while the NNDC served as the active user and provided feedback. The NNDC has been primarily processing neutron cross section covariances on its 64-bit Linux cluster in support of two DOE programs, the Global Nuclear Energy Partnership (GNEP) and the Nuclear Criticality Safety Program (NCSP). For GNEP, the NNDC used NJOY-99.259 to generate multigroup covariance matrices of {sup 56}Fe, {sup 23}Na, {sup 239}Pu, {sup 235}U and {sup 238}U from the JENDL-3.3 library using the 15-, 33-, and 230-energy group structures. These covariance matrices will be used to test a new collapsing algorithm which will subsequently be employed to calculate uncertainties on integral parameters in different fast neutron-based systems. For NCSP, we used PUFF-IV 1.0.4 to verify the processability of new evaluated covariance data of {sup 55}Mn, {sup 239}Pu, {sup 233}U, {sup 235}U and {sup 238}U generated by a collaboration of ORNL and LANL. For the data end-users at large, the NNDC has made available amore » Web site which provides a static visualization interface for all materials with covariance data in the four major data libraries: ENDF/B-VI.8 (47 materials), ENDF/B-VII.0 (26 materials), JEFF-3.1 (37 materials) and JENDL-3.3 (20 materials)« less

Characterization of minor actinide mixed oxide fuel

The global nuclear energy partnership (GNEP) was created in order for ‘fuel-cycle supplier’ nations to provide assured supplies of nuclear fuel to ‘fuel-cycle customer’ nations. The customer nations would utilize the fuel for electricity generation and subsequently return it to the supplier nation after it is spent. This spent fuel would then be reprocessed by the supplier nation in order to recycle the actinide constituents, mainly uranium and plutonium, in advanced nuclear power reactors, and thus reduce waste volumes [1,2]. The International Atomic Energy Agency would control the nuclear materials. One of the thrust areas for the GNEP program is the development of these actinide bearing fuels for transmutation in a fast reactor.

Potential advantages and disadvantages of sequentially building small nuclear units instead of a large nuclear plant

Abstract Renewal of nuclear power programs in countries with modest electricity consumptions and weak electrical grid interconnections has raised the question of optimal nuclear power plants sizes for such countries. The same question would be also valid for isolated or weakly connected regions within a large country. Building large size nuclear power plant could be prevented by technical or financial limits. Research programs have been initiated in the International Atomic Energy Agency and in the USA (within the framework of the Global Nuclear Energy Partnership (GNEP) program) with the aim to inspect under which circumstances small and medium reactors could be the preferred option compared to large nuclear plants. The economy of scale is a clear advantage of large plants. This paper compares, by using probabilistic methods, the net cash flow of large and medium size plants, taking as example a large nuclear plant (around 1200 MW) and four sequentially built smaller plants (300 MW). Potential advantages and disadvantageous of both options have been considered. Main advantages of the sequential construction of several identical small units could be the reduced investor risk and reduced investment costs due to the learning effect. This analysis is a part of studies for the Croatian power generating system development.