Advanced fuel cycles options for LWRs and IMF benchmark definition

International Nuclear Information System (INIS)

Breza, J.; Darilek, P.; Necas, V.

2008-01-01

In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

Thorium fuel cycle management

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Breza, J.; Necas, V.

2010-01-01

In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

Self-Sustaining Thorium Boiling Water Reactors

International Nuclear Information System (INIS)

Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra; Seifried, Jeffrey E.; Zhang, Guanheng; Varela, Christopher R.; Fratoni, Massimiliano; Vijic, Jasmina J.; Downar, Thomas; Hall, Andrew; Ward, Andrew; Jarrett, Michael; Wysocki, Aaron; Xu, Yunlin; Kazimi, Mujid; Shirvan, Koroush; Mieloszyk, Alexander; Todosow, Michael; Brown, Nicolas; Cheng, Lap

2015-01-01

The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

Self-Sustaining Thorium Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Greenspan, Ehud [Univ. of California, Berkeley, CA (United States); Gorman, Phillip M. [Univ. of California, Berkeley, CA (United States); Bogetic, Sandra [Univ. of California, Berkeley, CA (United States); Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States); Zhang, Guanheng [Univ. of California, Berkeley, CA (United States); Varela, Christopher R. [Univ. of California, Berkeley, CA (United States); Fratoni, Massimiliano [Univ. of California, Berkeley, CA (United States); Vijic, Jasmina J. [Univ. of California, Berkeley, CA (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Hall, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Ward, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Jarrett, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Wysocki, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Xu, Yunlin [Univ. of Michigan, Ann Arbor, MI (United States); Kazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Shirvan, Koroush [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mieloszyk, Alexander [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todosow, Michael [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, Nicolas [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap [Brookhaven National Lab. (BNL), Upton, NY (United States)

2015-03-15

The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

Potential of axial fuel management strategies in thorium-fuelled CANDU's

International Nuclear Information System (INIS)

Milgram, M.S.

1978-06-01

Three axial fuel management strategies are compared for use in a CANDU-PHW reactor operating on a self-sufficient, equilibrium thorium cycle. Two of these strategies are familiar ones for uranium reactors, and the third seeks to take advantage of the nuclear characteristics of the Th 232 → U 233 transmutation chain to improve the economics of the fuel cycle by periodically removing the fuel from the reactor. This results in an approximately 50% increase in burnup and an approximately 15% decrease in heavy element fuel inventory at a channel power of 6 MW, relative to the other strategies. (author)

Thorium-Based Fuel Cycles in the Modular High Temperature Reactor

Institute of Scientific and Technical Information of China (English)

CHANG Hong; YANG Yongwei; JING Xingqing; XU Yunlin

2006-01-01

Large stockpiles of civil-grade as well as weapons-grade plutonium have been accumulated in the world from nuclear power or other programs of different countries. One alternative for the management of the plutonium is to incinerate it in the high temperature reactor (HTR). The thorium-based fuel cycle was studied in the modular HTR to reduce weapons-grade plutonium stockpiles, while producing no additional plutonium or other transuranic elements. Three thorium-uranium fuel cycles were also investigated. The thorium absorption cross sections of the resolved and unresolved resonances were generated using the ZUT-DGL code based on existing resonance data. The equilibrium core of the modular HTR was calculated and analyzed by means of the code VSOP'94. The results show that the modular HTR can incinerate most of the initially loaded plutonium amounting to about 95.3% net 239Pu for weapons-grade plutonium and can effectively utilize the uranium and thorium in the thorium-uranium fuel cycles.

Homogeneous Thorium Fuel Cycles in Candu Reactors

Energy Technology Data Exchange (ETDEWEB)

Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

2009-06-15

The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

Efficiency Of Transuranium Nuclides Transmutation

International Nuclear Information System (INIS)

Kazansky, Yu.A.; Klinov, D.A.; Semenov, E.V.

2002-01-01

One of the ways to create a wasteless nuclear power is based on transmutation of spent fuel nuclides. In particular, it is considered that the radioactivity of the nuclear power wastes should be the same (or smaller), than radioactivity of the uranium and the thorium extracted from entrails of the Earth. The problem of fission fragments transmutation efficiency was considered in article, where, in particular, the concepts of transmutation factor and the ''generalised'' index of biological hazard of the radioactive nuclides were entered. The transmutation efficiency has appeared to be a function of time and, naturally, dependent on nuclear power activity scenario, from neutron flux, absorption cross-sections of the nuclides under transmutation and on the rate of their formation in reactors. In the present paper the efficiency of the transmutation of transuranium nuclides is considered

On the use of spinel-based nuclear fuels for the transmutation of actinides

International Nuclear Information System (INIS)

Konings, R.J.M.; Bakker, K.; Boshoven, J.G.; Hein, H.; Huntelaar, M.E.; Zhang, H.; Meeldijk, J.D.; Woensdregt, C.F.

1997-01-01

The properties of spinel-based nuclear fuels for the transmutation of actinides are investigated. The results of laboratory experiments, thermodynamic calculations and irradiations in the High Flux Reactor (HFR) at Petten are presented, and allow us to evaluate the potential of spinel as an inert matrix for fuels and targets for transmutation. (author)

The thorium fuel cycle

International Nuclear Information System (INIS)

Merz, E.R.

1977-01-01

The utilization of the thorium fuel cycle has long since been considered attractive due to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Although the uranium ore as well as the separative work requirements are usually lower for any thorium-based fuel cycle in comparison to present uranium-plutonium fuel cycles of thermal water reactors, interest by nuclear industry has hitherto been marginal. Fast increasing uranium prices, public reluctance against widespread Pu-recycling and expected retardations for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, it could be learned in the meantime that problems associated with reprocessing and waste handling, but particularly with a remote refabrication of 233 U are certainly not appreciably more difficult than for Pu-recycling. This may not only be due to psychological constraints but be based upon technological as well as economical facts, which have been mostly neglected up till now. In order to diversify from uranium as a nuclear energy source it seems to be worthwhile to greatly intensify efforts in the future for closing the Th/ 233 U fuel cycle. HTGR's are particularly promising for economic application. However, further R and D activites should not be solely focussed on this reactor type alone. Light and heavy-water moderated reactors, as well as even fast breeders later on, may just as well take advantage of a demonstrated thorium fuel cycle. A summary is presented of the state-of-the-art of Th/ 233 U-recycling technology and the efforts still necessary to demonstrate this technology all the way through to its industrial application

Studies on the feasibility of the LWRs waste-thorium in-core fuel cycle in the Gas Turbine-Modular Helium Reactor

International Nuclear Information System (INIS)

Talamo, Alberto

2006-01-01

The capability to operate on LWRs waste constitutes one of the major benefits of the Gas Turbine-Modular Helium Reactor; in this paper, it has been evaluated the possibility to incinerate the LWRs waste and to simultaneously breed fissile 233 U by fertile thorium. Since a mixture of pure 239 Pu-thorium has shown a quite poor neutron economy, the LWRs waste-thorium fuel performance has been also tested when plutonium and thorium are allocated in different TRISO particles. More precisely, when fissile and fertile actinides share the same TRISO kernel, the resonance at 0.29eV of the fission and capture microscopic cross sections of 239 Pu diminishes also the absorption rate of fertile 232 Th and thus it degrades the breeding process. Consequently, in the present studies, two different types of fuel have been utilized: the Driver Fuel, made of LWRs waste, and the Transmutation Fuel, made of fertile thorium. Since, in the thermal neutron energy range, the microscopic capture cross section of 232 Th is about 80-100 times smaller than the fission one of 239 Pu, setting thorium in particles with a large kernel and LWRs waste in particles with a small one makes the volume integrated reaction rates better equilibrated. At the light of the above consideration, which drives to load as much thorium as possible, for the Transmutation Fuel they have been selected the JAERI TRISO particles packed 40%; whereas, for the Driver Fuel they have been tested different packing fractions and kernel radii. Since no configuration allowed the reactor to work, the above procedure has been repeated when fertile particles are packed 20%; the latter choice permits over one year of operation, but the build up of 233 U represents only a small fraction of the depleted 239 Pu. Finally, the previous configuration has been also investigated when the fertile and fissile fuels share the same kernel or when the fertile fuel axially alternates with the fissile one. (author)

Review of thorium fuel reprocessing experience

International Nuclear Information System (INIS)

Brooksbank, R.E.; McDuffee, W.T.; Rainey, R.H.

1978-01-01

The review reveals that experience in the reprocessing of irradiated thorium materials is limited. Plants that have processed thorium-based fuels were not optimized for the operations. Previous demonstrations of several viable flowsheets provide a sound technological base for the development of optimum reprocessing methods and facilities. In addition to the resource benefit by using thorium, recent nonproliferation thrusts have rejuvenated an interest in thorium reprocessing. Extensive radiation is generated as the result of 232 U-contamination produced in the 233 U, resulting in the remote operation and fabrication operations and increased fuel cycle costs. Development of the denatured thorium flowsheet, which is currently of interest because of nonproliferation concerns, represents a difficult technological challenge

Moving towards sustainable thorium fuel cycles

International Nuclear Information System (INIS)

Hyland, B.; Hamilton, H.

2011-01-01

The CANDU reactor has an unsurpassed degree of fuel-cycle flexibility as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle design. These features facilitate the introduction and full exploitation of thorium fuel cycles in CANDU reactors in an evolutionary fashion. Thoria (ThO 2 ) based fuel offers both fuel performance and safety advantages over urania (UO 2 ) based fuel, due its higher thermal conductivity which results in lower fuel-operating temperatures at similar linear element powers. Thoria fuel has demonstrated lower fission gas release than UO 2 under similar operating powers during test irradiations. In addition, thoria has a higher melting point than urania and is far less reactive in hypothetical accident scenarios owing to the fact that it has only one oxidation state. This paper examines one possible strategy for the introduction of thorium fuel cycles into CANDU reactors. In the short term, the initial fissile material would be provided in a heterogeneous bundle of low-enriched uranium and thorium. The medium term scenario uses homogeneous Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. In the long term, the full energy potential from thorium would be realized through the recycle of the U-233 in the used fuel. With U-233 recycle in CANDU reactors, plutonium would then only be required to top up the fissile content to achieve the desired burnup. (author)

Safety and Regulatory Issues of the Thorium Fuel Cycle

Energy Technology Data Exchange (ETDEWEB)

Ade, Brian [ORNL; Worrall, Andrew [ORNL; Powers, Jeffrey [ORNL; Bowman, Steve [ORNL; Flanagan, George [ORNL; Gehin, Jess [ORNL

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2), add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium-Uranium-Plutonium Fuel Cycle

Science.gov (United States)

Marshalkin, V. Ye.; Povyshev, V. M.

2017-12-01

It is shown for a closed thorium-uranium-plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain 54 kg of fission products, 0.8 kg of thorium, 0.10 kg of uranium isotopes, 0.005 kg of plutonium isotopes, 0.002 kg of neptunium, and "trace" amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

Energy Technology Data Exchange (ETDEWEB)

Rahman, Fariz Abdul; Lee, John C. [University of Michigan, Ann Arbor, MI (United States); Franceschini, Fausto; Wenner, Michael [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

2012-07-01

As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning the legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and

Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw

2005-01-01

In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: 235 U, which represents the 20% of the fresh uranium, 233 U, which is produced by the transmutation of fertile 232 Th, and 239 Pu, which is produced by the transmutation of fertile 238 U. In order to compensate the depletion of 235 U with the breeding of 233 U and 239 Pu, the quantity of fertile nuclides must be much larger than that one of 235 U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of 235 U. At the same time, the amount of 235 U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k eff and mass evolution, reaction rates, neutron flux and spectrum at the

Method for investigating the applicability of thorium-based fuels in existing BWRs

International Nuclear Information System (INIS)

Bjoerk, Klara Insulander; Fhager, Valentin; Demaziere, Christophe

2009-01-01

In this paper, a methodology for investigating the applicability of thorium based fuels in existing Boiling Water Reactors is described. The same tools and methods as those used in this study are also used for design of new fuel batches for existing reactors. As an illustration of the proposed methodology and for the purpose of comparing different fissile components for thorium-based fuel, three different thorium-based fuel designs were developed, and a low enriched uranium fuel design was created in parallel as a reference. The different fuel assemblies, which are based on the mechanical fuel design GE14, were analyzed with the Studsvik Scandpower CASMO-4E code. The linear reactivity model was used to decide a suitable initial fissile content for meeting the burnup goal of 55 MWd/kgHM and the designs were optimized for attaining an even power distribution and for keeping boiling transition factors low. Such optimizations facilitate core designs that would meet prevalent demands on Linear Heat Generation Rate (LHGR) and Critical Power Ratio (CPR). The hence created fuel designs were then analyzed with regards to thermal performance, reactivity coefficients, delayed neutron fractions and control rod worths. An equilibrium core loading pattern was also developed with each one of the thus created fuel designs and analyzed with the three dimensional reactor analysis code SIMULATE-3, also developed by Studsvik Scandpower. The results of the two simulations were compared and found to be coherent, and no major obstacles were met using the given tools and methods. The analysis of the performances of the fuel assemblies and the corresponding loaded cores is reported in a companion paper (K. Insulander Bjoerk, V. Fhager, and C. Demaziere, 'Comparison of Thorium-Based Fuels with Different Fissile Components in Existing BWRs'). (author)

Systematic study on Thorium fuel

International Nuclear Information System (INIS)

Shibata, Toshikazu; Kimura, Itsuro; Iwata, Shiro; Furuya, Hirotaka; Suzuki, Susumu.

1988-01-01

Introduced is the activities of the Joint Research Project Team on Thorium Fuel organized by mainly university researchers in Japan and supported by the Ministry of Education, Science and Culture for seven years since 1980. Four major groups were organized; (1) nuclear data, reactor physics and design, (2) nuclear fuel, (3) down stream and (4) biological effects of thorium. The first group covered measurements and analysis on nuclear data of thorium related nuclides, experiment and analysis on nuclear characteristics of thorium containing cores, basic engineering on a thorium molten salt reactor, and designs of several types of reactors. Fabrication and irradiation tests of thorium oxide fuel, and basic studies on new type thorium fuels (e.g. carbide and nitride) were studied by the second group. The third group covered the use of solutions in reprocessing of spent fuel, behavior of fission products, immobilization of high level radioactive waste, and continuous reprocessing for a molten salt reactor. The fourth group performed the trace study for patients who had been intravascularly injected with thorotrast for diagnosis of war injuries during the Second World War. (author)

Evaluation of thorium based nuclear fuel. Actinide waste

International Nuclear Information System (INIS)

Wichers, V.A.

1995-06-01

Use of thorium based fuel has recently been proposed as a possible way to reduce the amount of actinide waste from nuclear power. To examine this possibility, burnup calculations were done of five once-through Thorium Heavy Water Reactor (THWR) systems, and three THWR systems with uranium recycle. The natural uranium once-through system was adopted as reference. The studied THWR fuel systems differed in the choice of fissile makeup fuel and exit burnup. The HWR was chosen because of its good neutron economy. Actinide waste production (in mass per GW e a) and radiotoxicity (in ALI per GW e a) for storage times up to 10 6 a were calculated for each system. The study shows that the THWR system with uranium recycle and High Enriched Uranium (U-235) makeup fuel performed best, producing both the lowest amount of plutonium and actinide waste with the lowest radiotoxicity. Relative to the natural uranium in HWR once-through system, radiotoxicity is reduced by a factor varying between 2 and 50 for the full range of storage times up to 10 6 a. (orig.)

Transmutation Capability of a Once-Through Molten-Salt and Other Transmuting Reactors

International Nuclear Information System (INIS)

Greenspan, E.; Lowenthal, M.; Barnes, D.; Kawasaki, D.; Kimball, D.; Matsumoto, H.; Sagara, H.; Vietez, E.R.

2002-01-01

A preliminary assessment is done of the transmutation characteristics of three reactor technologies: a multi-batch liquid metal (LM) cooled transmuter, a once-through molten-salt (MS) transmuter and a pebble bed (PB) transmuter. It was found that for the same fractional transmutation and same k eff drop with burnup (Δk effBU ), lead-bismuth offers smaller peak-to-average core power density, and it requires a smaller pumping power but a larger and heavier core than a sodium cooled transmuter. 99 Tc cannot effectively serve as a burnable absorber to reduce Δk effBU of LM transmuters. However, addition of thorium can greatly flatten k eff and almost double the fractional transmutation of the LWR spent fuel from ∼20% to ∼40%. If the 'once-through' MS transmuter is operated with continuous complete removal of fission products, it can achieve ∼85% fractional transmutation provided that the equilibrium concentration of actinides in the MS can reach 4 mole %. If the fission products are not actively removed, the fractional transmutation is reduced to ∼75%. The fractional transmutation of a PB transmuter can exceed 40%. More thorough analysis is required to better quantify the transmutation capability of the different transmuter technologies. (authors)

Fabrication routes for Thorium and Uranium233 based AHWR fuel

International Nuclear Information System (INIS)

Danny, K.M.; Saraswat, Anupam; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun

2011-01-01

India's economic growth is on a fast growth track. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear Energy is best suited to meet this demand without causing undue environmental impact. Considering the large thorium reserves in India, the future nuclear power program will be based on Thorium- Uranium 233 fuel cycle. The major characteristic of thorium as the fuel of future comes from its superior fuel utilization. 233 U produced in a reactor is always contaminated with 232 U. This 232 U undergoes a decay to produce 228 Th and it is followed by decay chain including 212 Bi and 208 Tl. Both 212 Bi and 208 Tl are hard gamma emitters ranging from 0.6 MeV-1.6 MeV and 2.6 MeV respectively, which necessitates its handling in hot cell. The average concentration of 232 U is expected to exceed 1000 ppm after a burn-up of 24,000 MWD/t. Work related to developing the fuel fabrication technology including automation and remotization needed for 233 U based fuels is in progress. Various process for fuel fabrication have been developed i.e. Coated Agglomerate Pelletisation (CAP), impregnation technique (Pellet/Gel), Sol Gel Micro-sphere Pelletisation (SGMP) apart from Powder to Pellet (POP) route. This paper describes each process with respect to its advantages, disadvantages and its amenability to automation and remotisation. (author)

Studies on the safety and transmutation behaviour of innovative fuels for light water reactors

International Nuclear Information System (INIS)

Schitthelm, Oliver

2012-01-01

Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its 238 U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

Candu reactors with thorium fuel cycles

International Nuclear Information System (INIS)

Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.

2006-01-01

Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 550 0 C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

Radkowsky Thorium Fuel Project

International Nuclear Information System (INIS)

Todosow, Michael

2006-01-01

In the early/mid 1990's Prof. Alvin Radkowsky, former chief scientist of the U.S. Naval Reactors program, proposed an alternate fuel concept employing thorium-based fuel for use in existing/next generation pressurized water reactors (PWRs). The concept was based on the use of a 'seed-blanket-unit' (SBU) that was a one-for-one replacement for a standard PWR assembly with a uranium-based central 'driver' zone, surrounded by a 'blanket' zone containing uranium and thorium. Therefore, the SBU could be retrofit without significant modifications into existing/next generation PWRs. The objective was to improve the proliferation and waste characteristics of the current once-through fuel cycle. The objective of a series of projects funded by the Initiatives for Proliferation Prevention program of the U.S. Department of Energy (DOE-IPP) - BNL-T2-0074,a,b-RU 'Radkowsky Thorium Fuel (RTF) Concept' - was to explore the characteristics and potential of this concept. The work was performed under several BNL CRADAs (BNL-C-96-02 and BNL-C-98-15) with the Radkowsky Thorium Power Corp./Thorium Power Inc. and utilized the technical and experimental capabilities in the Former Soviet Union (FSU) to explore the potential of this concept for implementation in Russian pressurized water reactors (VVERs), and where possible, also generate data that could be used for design and licensing of the concept for Western PWRs. The Project in Russia was managed by the Russian Research Center-?'Kurchatov Institute' (RRC-KI), and included several institutes (e.g., PJSC 'Electrostal', NPO 'LUCH' (Podolsk), RIINM (Bochvar Institute), GAN RF (Gosatomnadzor), Kalininskaja NPP (VVER-1000)), and consisted of the following phases: Phase-1 ($550K/$275K to Russia): The objective was to perform an initial review of all aspects of the concept (design, performance, safety, implementation issues, cost, etc.) to confirm feasibility/viability and identify any 'show-stoppers'; Phase-2 ($600K/$300K to Russia

Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se; Gudowski, Waclaw [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)

2005-11-15

In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: {sup 235}U, which represents the 20% of the fresh uranium, {sup 233}U, which is produced by the transmutation of fertile {sup 232}Th, and {sup 239}Pu, which is produced by the transmutation of fertile {sup 238}U. In order to compensate the depletion of {sup 235}U with the breeding of {sup 233}U and {sup 239}Pu, the quantity of fertile nuclides must be much larger than that one of {sup 235}U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of {sup 235}U. At the same time, the amount of {sup 235}U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k {sub eff} and mass

Thorium based fuel options for the generation of electricity: Developments in the 1990s

International Nuclear Information System (INIS)

2000-05-01

The IAEA has maintained an interest in the thorium fuel cycle and its worldwide utilization within its framework of activities. Periodic reviews have assessed the current status of this fuel cycle, worldwide applications, economic benefits, and perceived advantages with respect to other nuclear fuel cycles. Since 1994, the IAEA convened a number of technical meetings on the thorium fuel cycle and related issues. Between 1995 and 1997 individual contributions on the thorium fuel cycle were elicited from experts from France, Germany, India, Japan, the Russian Federation and the USA. These contributions included evaluations of the status of the thorium fuel cycle worldwide; the new incentives to use thorium due to large stockpiles of plutonium produced in nuclear reactors; new reactor concepts utilizing thorium; strategies for thorium use; and an evaluation of toxicity of the thorium fuel cycle waste compared to that from other fuel cycles. The results of this updated evaluation are summarized in this publication

Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

Science.gov (United States)

Gorman, Phillip Michael

The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was

The thorium fuel cycle

International Nuclear Information System (INIS)

Merz, E.R.

1977-01-01

The utilization of the thorium fuel cycle has long since been considered attractive owing to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Rapidly increasing uranium prices, public reluctance for widespread Pu recycling and expected delays for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, problems associated with reprocessing and waste handling, particularly with re-fabrication by remote handling of 233 U, are certainly not appreciably more difficult than for Pu recycling. To divert from uranium as a nuclear energy source it seems worth while intensifying future efforts for closing the Th/ 233 U fuel cycle. HTGRs are particularly promising for economic application. However, further research and development activities should not concentrate on this reactor type alone. Light- and heavy-water-moderated reactors, and even future fast breeders, may just as well take advantage of a demonstrated thorium fuel cycle. (author)

Research and development of thorium fuel cycle

International Nuclear Information System (INIS)

Oishi, Jun.

1994-01-01

Nuclear properties of thorium are summarized and present status of research and development of the use of thorium as nuclear fuel is reviewed. Thorium may be used for nuclear fuel in forms of metal, oxide, carbide and nitride independently, alloy with uranium or plutonium or mixture of the compound. Their use in reactors is described. The reprocessing of the spent oxide fuel in thorium fuel cycle is called the thorex process and similar to the purex process. A concept of a molten salt fuel reactor and chemical processing of the molten salt fuel are explained. The required future research on thorium fuel cycle is commented briefly. (T.H.)

A review on the heterogeneous thorium fuel concept for PWR applications

International Nuclear Information System (INIS)

Joo, H. K.; Noh, J. M.; Yoo, J. W.; Kim, K. H.

2001-08-01

Seed-blanket unit (SBU) and whole assembly seed and blanket (WASB) are being investigated for the PWR application as well as homogeneous thorium fuel under the US NERI program. For the verification of HELIOS capability for thorium analysis, the characteristics of heterogeneous thorium fuels was evaluated by HELIOS color-set calculation and compared with the calculation results of the US NERI. The infinite multiplication factors from HELIOS calculation are in good agreement with CASMO-4 except for SBU which uses metallic fuel for seed material. The maximum relative difference in power distribution is occurred in WASB case, and is about 5% compared to MCNP. The isotopic concentrations for Am-241, Am-243, and Cm-244 of HELIOS agree well with CASMO-4's, but show a significant discrepancy from MOCUP mainly caused by the old data of cross section and decay constants in ORIGEN. The nonproliferation characteristic of thorium-based fuel such as critical mass, spontaneous fission rate, decay heat generation rate are superior to the conventional uranium fuel. Even though the diversion of U-233 produced in blanket is a technically difficult, the enrichment of uranium isotopes including U-233 is slightly over the limit for safeguard aspects. The urnaium contents in thorium fuel is need to be adjusted in order to meet the safeguard limit. A preliminary assessment of fuel economics was performed based on the uranium utilization and SWU utilization. The natural uranium utilization factors of heterogeneous thorium-based fuel increased by 10δ18%, but the SWU utilization factor decreased by 6-δ11% compared to uranium fuel. The cost of uranium purchase of 50USI/KgU and SWU cost of 110USI/SWU-Kg, recommended by OECD/NEA, gives a comparable economics of thorium-based fuel to uraium fuel. The detailed fuel cycle analysis will take account of the other factors like the variation of uranium purchase cost and SWU cost, fabrication cost of thorium fuel, thorium purchase cost, the capcity

A review on the heterogeneous thorium fuel concept for PWR applications

Energy Technology Data Exchange (ETDEWEB)

Joo, H. K.; Noh, J. M.; Yoo, J. W.; Kim, K. H

2001-08-01

Seed-blanket unit (SBU) and whole assembly seed and blanket (WASB) are being investigated for the PWR application as well as homogeneous thorium fuel under the US NERI program. For the verification of HELIOS capability for thorium analysis, the characteristics of heterogeneous thorium fuels was evaluated by HELIOS color-set calculation and compared with the calculation results of the US NERI. The infinite multiplication factors from HELIOS calculation are in good agreement with CASMO-4 except for SBU which uses metallic fuel for seed material. The maximum relative difference in power distribution is occurred in WASB case, and is about 5% compared to MCNP. The isotopic concentrations for Am-241, Am-243, and Cm-244 of HELIOS agree well with CASMO-4's, but show a significant discrepancy from MOCUP mainly caused by the old data of cross section and decay constants in ORIGEN. The nonproliferation characteristic of thorium-based fuel such as critical mass, spontaneous fission rate, decay heat generation rate are superior to the conventional uranium fuel. Even though the diversion of U-233 produced in blanket is a technically difficult, the enrichment of uranium isotopes including U-233 is slightly over the limit for safeguard aspects. The urnaium contents in thorium fuel is need to be adjusted in order to meet the safeguard limit. A preliminary assessment of fuel economics was performed based on the uranium utilization and SWU utilization. The natural uranium utilization factors of heterogeneous thorium-based fuel increased by 10{delta}18%, but the SWU utilization factor decreased by 6-{delta}11% compared to uranium fuel. The cost of uranium purchase of 50USI/KgU and SWU cost of 110USI/SWU-Kg, recommended by OECD/NEA, gives a comparable economics of thorium-based fuel to uraium fuel. The detailed fuel cycle analysis will take account of the other factors like the variation of uranium purchase cost and SWU cost, fabrication cost of thorium fuel, thorium purchase cost

Conceptual design study of small long-life PWR based on thorium cycle fuel

International Nuclear Information System (INIS)

Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

2014-01-01

A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of 233 U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation

Impact of Transmutation Scenarios on Fuel Transportation

International Nuclear Information System (INIS)

Saturnin, A.; Duret, B.; Allou, A.; Jasserand, F.; Fillastre, E.; Giffard, F.X.; Chabert, C.; Caron-Charles, M.; Garzenne, C.; Laugier, F.

2015-01-01

Minor actinides transmutation scenarios have been studied in the frame of the French Sustainable Radioactive Waste Management Act of 28 June 2006. Transmutation scenarios supposed the introduction of a sodium-cooled fast reactor fleet using homogeneous or heterogeneous recycling modes for the minor actinides. Americium, neptunium and curium (MA) or americium alone (Am) can be transmuted together in a homogeneous way embedded in FR-MOX fuel or incorporated in MA or Am-Bearing radial Blankets (MABB or AmBB). MA transmutation in Accelerator Driven System has also been studied while plutonium is being recycled in SFR. Assessments and comparisons of these advanced cycles have been performed considering technical and economic criteria. Transportation needs for fresh and used transmutation fuels is one of these criteria. Transmutation fuels have specific characteristics in terms of thermal load and neutron emissions. Thermal, radiation and criticality constraints have been taken into account in this study to suggest cask concepts for routine conditions of transport, to estimate the number of assemblies to be transported in a cask and the number of annual transports. Comparison with the no transmutation option, i.e. management of uranium and plutonium in SFRs, is also presented. Regarding these matters, no high difficulties appear for assemblies with limited content of Am (homogeneous or heterogeneous recycling modes). When fuels contain curium, technical transport uncertainties increase because of the important heat release requiring dividing fresh fuels and technological innovations development (MABB and ADS). (authors)

Multiple tier fuel cycle studies for waste transmutation

International Nuclear Information System (INIS)

Hill, R.N.; Taiwo, T.A.; Stillman, J.A.; Graziano, D.J.; Bennett, D.R.; Trellue, H.; Todosow, M.; Halsey, W.G.; Baxter, A.

2002-01-01

As part of the U.S. Department of Energy Advanced Accelerator Applications Program, a systems study was conducted to evaluate the transmutation performance of advanced fuel cycle strategies. Three primary fuel cycle strategies were evaluated: dual-tier systems with plutonium separation, dual-tier systems without plutonium separation, and single-tier systems without plutonium separation. For each case, the system mass flow and TRU consumption were evaluated in detail. Furthermore, the loss of materials in fuel processing was tracked including the generation of new waste streams. Based on these results, the system performance was evaluated with respect to several key transmutation parameters including TRU inventory reduction, radiotoxicity, and support ratio. The importance of clean fuel processing (∼0.1% losses) and inclusion of a final tier fast spectrum system are demonstrated. With these two features, all scenarios capably reduce the TRU and plutonium waste content, significantly reducing the radiotoxicity; however, a significant infrastructure (at least 1/10 the total nuclear capacity) is required for the dedicated transmutation system

The economics of thorium fuel cycles

International Nuclear Information System (INIS)

James, R.A.

1978-01-01

The individual cost components and the total fuel cycle costs for natural uranium and thorium fuel cycles are discussed. The thorium cycles are initiated by using either enriched uranium or plutonium. Subsequent thorium cycles utilize recycled uranium-233 and, where necessary, either uranium-235 or plutonium as topping. A calculation is performed to establish the economic conditions under which thorium cycles are economically attractive. (auth)

An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

Science.gov (United States)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutron multiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

International Nuclear Information System (INIS)

Kim, Taek Kyum; Downar, Thomas J.

2002-01-01

Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight-pitch light water reactor (LWR) lattice is reported. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight-pitch LWRs have considerable advantages in terms of conversion ratio, reactivity control, nonproliferation characteristics, and a reduced production of long-lived radiotoxic wastes. Because of the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11 to 17% higher than that of 238 U fertile fuels. However, unlike 238 U fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing 238 U with 232 Th in advanced high-conversion intermediate spectrum LWRs, such as the reduced-moderator reactor or the supercritical reactor

Comparison of thorium-based fuels with different fissile components in existing BWRs

International Nuclear Information System (INIS)

Bjoerk, Klara Insulander; Fhager, Valentin; Demaziere, Christophe

2009-01-01

Three different types of thorium based BWR fuel have been developed, in each of which thorium was combined with a different fissile component, the three components being reactor grade plutonium, uranium enriched to 20% in uranium 235 and pure uranium 233. A BWR nuclear bundle design, based on the geometrical fuel assembly design GE14, was developed for each of these fissile components. The properties and performance of the corresponding fuel assemblies were investigated via full core calculations carried out for an existing BWR and compared with the ones of an ordinary Low Enriched Uranium (LEU) fuel, which was developed for reference. The fuel assemblies and cores were designed to meet existing fuel design criteria, and were then analyzed with regards to reactivity coefficients, delayed neutron fractions, control rod worths and shutdown margins. The results show that all three alternatives seem to be feasible, although some difficulties remain with complying with the thermal limits, and with the moderator temperature and coolant void coefficients of the U-233 containing fuel being positive under some circumstances. (author)

Practical introduction of thorium fuel cycles

International Nuclear Information System (INIS)

Kasten, P.R.

1982-01-01

The pracitcal introduction of throrium fuel cycles implies that thorium fuel cycles compete economically with uranium fuel cycles in economic nuclear power plants. In this study the reactor types under consideration are light water reactors (LWRs), heavy water reactors (HWRs), high-temperature gas-cooled reactors (HTGRs), and fast breeder reactors (FBRs). On the basis that once-through fuel cycles will be used almost exclusively for the next 20 or 25 years, introduction of economic thorium fuel cycles appears best accomplished by commercial introduction of HTGRs. As the price of natural uranium increases, along with commercialization of fuel recycle, there will be increasing incentive to utilize thorium fuel cycles in heavy water reactors and light water reactors as well as in HTGRs. After FBRs and fuel recycle are commercialized, use of thorium fuel cycles in the blanket of FBRs appears advantageous when fast breeder reactors and thermal reactors operate in a symbiosis mode (i.e., where 233 U bred in the blanket of a fast breeder reactor is utilized as fissile fuel in thermal converter reactors)

Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

International Nuclear Information System (INIS)

Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T.; Takaki, N.; Yamaguchi, A.; Watanabe, H.; Unesaki, H.

2012-01-01

Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

Energy Technology Data Exchange (ETDEWEB)

Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

2013-07-01

Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

International Nuclear Information System (INIS)

Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

2013-01-01

Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

Effects of actinide compositional variability in the US spent fuel inventory on partitioning-transmutation systems

International Nuclear Information System (INIS)

Ludwig, S.B.; Michaels, G.E.; Hanson, B.D.

1992-01-01

Partitioning and transmutation (P-T) is an advanced waste management concept by which certain undesirable nuclides in spent fuel are first isolated (partitioned) and later destroyed (transmuted) in a nuclear reactor or other transmutation device. There are wide variabilities in the nuclide composition of spent fuel. This implies that there will also be wide variabilities in the transmutation device feed. As a waste management system, P-T must be able to accept (all) spent fuel. Variability of nuclide composition (i.e., the feed material for transmutation devices) may be important because virtually all transmutation systems propose to configure transuranic (TRU) nuclides recovered from discharged lightwater reactor (LWR) spent fuel in critical or near-critical cores. To date, all transmutation system core analyses assume invariant nuclide concentrations for startup and recycle cores. Using the US Department of Energy's (DOE's) Characteristics Data Base (CDB) and the ORIGEN2 computer code, the current and projected spent fuel discharges until the year 2016 have been categorized according to combinations of fuel burnup, initial enrichment, fuel age (cooling time) and reactor type (boiling-water or pressurized-water reactors). The variability of the infinite multiplication factor (k ∞ ) is calculated for both fast (ALMR) and thermal (accelerator-based) transmuter systems

Multiple Tier Fuel Cycle Studies for Waste Transmutation

International Nuclear Information System (INIS)

Hill, R.N.; Taiwo, T.A.; Stillman, J.A.; Graziano, D.J.; Bennett, D.R.; Trellue, H.; Todosow, M.; Halsey, W.G.; Baxter, A.

2002-01-01

As part of the U.S. Department of Energy Advanced Accelerator Applications Program, a systems study was conducted to evaluate the transmutation performance of advanced fuel cycle strategies. Three primary fuel cycle strategies were evaluated: dual-tier systems with plutonium separation, dual-tier systems without plutonium separation, and single-tier systems without plutonium separation. For each case, the system mass flow and TRU consumption were evaluated in detail. Furthermore, the loss of materials in fuel processing was tracked including the generation of new waste streams. Based on these results, the system performance was evaluated with respect to several key transmutation parameters including TRU inventory reduction, radiotoxicity, and support ratio. The importance of clean fuel processing (∼0.1% losses) and inclusion of a final tier fast spectrum system are demonstrated. With these two features, all scenarios capably reduce the TRU and plutonium waste content, significantly reducing the radiotoxicity; however, a significant infrastructure (at least 1/10 the total nuclear capacity) is required for the dedicated transmutation system. (authors)

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Jin Ha; Kim, Myung Hyun

2016-01-01

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Status and development of the thorium fuel cycle

International Nuclear Information System (INIS)

Yi Weijing; Wei Renjie

2003-01-01

A perspective view of the thorium fuel cycle is provided in this paper. The advantages and disadvantages of the thorium fuel cycle are given and the development of thorium fuel cycle in several types of reactors is introduced. The main difficulties in developing the thorium fuel cycle lie in the reprocessing and disposal of the waste and its economy, and the ways tried by foreign countries to solve the problems are presented in the paper

Towards proliferation-resistant thorium fuels

International Nuclear Information System (INIS)

Alhaj, M. Yousif; Mohamed, Nader M.A.; Badawi, Alya; Abou-Gabal, Hanaa H.

2017-01-01

Thorium-plutonium mixture is proposed as alternative nuclear reactor fuel to incinerate the increasing stockpile plutonium. However, this fuel will produce an amount of uranium with about 90% 233U at applicable discharge burnups (60GWD/MTU). This research focuses on proposing an optimum non proliferative thorium fuel, by adding a small amount of 238U to reduce the attractiveness of the resultant uranium. Three types of additive which contain 238U were used: 4.98% enriched, natural and depleted uranium. We found that introducing uranium to the fresh thorium-plutonium fuel reduces its performance even if the uranium was enriched up to 5%. While uranium admixtures reduce the quality of the reprocessed uranium, it also increases the quality of the plutonium. However, this increase is very low compared to the reduced quality of uranium. We also found that using uranium as admixture for thorium-plutonium mixed fuel increases the critical mass of the extracted uranium by a factor of two when using only 1% admixture of uranium. The higher the percentage of uranium admixture the higher the critical mass of the reprocessed one.

Feasibility assessment of the once-through thorium fuel cycle for the PTVM LWR concept

International Nuclear Information System (INIS)

Rachamin, R.; Fridman, E.; Galperin, A.

2015-01-01

Highlights: • The PTVM LWR is an innovation reactor concept operating in a “breed & burn” mode. • An advanced once-through thorium fuel cycle for the PTVM LWR concept is proposed. • The PTVM LWR concept makes use of a seed-blanket geometry. • A novel fuel management scheme based on two separate fuel flow routes is analyzed. • The analysis indicates a potential for utilizing the fuel in an efficient manner. - Abstract: This paper investigates the feasibility of a once-through thorium fuel cycle for the novel reactor-design concept named the pressure tube light water reactor with variable moderator control (PTVM LWR). The PTVM LWR operates in a “breed & burn” mode, which makes it an attractive system for utilizing thorium fuel in a once-through mode. The “breed & burn” mode can emphasize the in situ generation as well as incineration of 233 U, which are the basic foundations of the once-through thorium fuel cycle. The PTVM LWR concept makes use of a seed–blanket geometry, whereby the core is divided into separated regions of thorium-based fuel channel assemblies (blanket) and low-enriched uranium (LEU) based fuel channel assemblies (seed). A novel fuel in-core management scheme based on two separate fuel flow routes (i.e., seed route and blanket route) is proposed and analyzed. Neutronic performance analysis indicates that the proposed novel fuel in-core management scheme has the potential to utilize both LEU- and thorium-based fuel in an efficient manner. The once-through thorium cycle, presented and discussed in this paper, provide interesting research leads and can serve as a bridge between current LEU-based fuel cycles and a thorium fuel cycle based on recycling of 233 U

Thorium-Based Fuels Preliminary Lattice Cell Studies for Candu Reactors

International Nuclear Information System (INIS)

Margeanu, C.A.; Rizoiu, A.C.

2009-01-01

The choice of nuclear power as a major contributor to the future global energy needs must take into account acceptable risks of nuclear weapon proliferation, in addition to economic competitiveness, acceptable safety standards, and acceptable waste disposal options. Candu reactors offer a proven technology, safe and reliable reactor technology, with an interesting evolutionary potential for proliferation resistance, their versatility for various fuel cycles creating premises for a better utilization of global fuel resources. Candu reactors impressive degree of fuel cycle flexibility is a consequence of its channel design, excellent neutron economy, on-power refueling, and simple fuel bundle. These features facilitate the introduction and exploitation of various fuel cycles in Candu reactors in an evolutionary fashion. The main reasons for our interest in Thorium-based fuel cycles have been, globally, to extend the energy obtainable from natural Uranium and, locally, to provide a greater degree of energy self-reliance. Applying the once through Thorium (OTT) cycle in existing and advanced Candu reactors might be seen as an evaluative concept for the sustainable development both from the economic and waste management points of view. Two Candu fuel bundles project will be used for the proposed analysis, namely the Candu standard fuel bundle with 37 fuel elements and the CANFLEX fuel bundle with 43 fuel elements. Using the Canadian proposed scheme - loading mixed ThO 2 -SEU CANFLEX bundles in Candu 6 reactors - simulated at lattice cell level led to promising conclusions on operation at higher fuel burnups, reduction of the fissile content to the end of the cycle, minor actinide content reduction in the spent fuel, reduction of the spent fuel radiotoxicity, presence of radionuclides emitting strong gamma radiation for proliferation resistance benefit. The calculations were performed using the lattice codes WIMS and Dragon (together with the corresponding nuclear data

Potential of thorium-based fuel cycle for PWR core to reduce plutonium and long-term toxicity

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

The cross section libraries and calculation methods of the participants were inter-compared through the first stage benchmark calculation. The multiplication factor of unit cell benchmark are in good agreement, but there is significant discrepancies of 2.3 to 3.5 %k at BOC and at EOC between the calculated infinite multiplication factors of each participants for the assembly benchmark. Our results with HELIOS show a reasonable agreement with the others except the MTC value at EOC. To verify the potential of the thorium-based fuel to consume the plutonium and to reduce the radioactivity from the spent fuel, the conceptual core with ThO{sub 2}-PuO{sub 2} or MOX fuel were constructed. The composition and quantity of plutonium isotopes and the radioactivity level of spent fuel for conceptual cores were analyzed, and the neutronic characteristics of conceptual cores were also calculated. The nuclear characteristics for ThO{sub 2}-PuO{sub 2} thorium fueled core was similar to MOX fueled core, mainly due to the same seed fuel material, plutonium. For the capability of plutonium consumption, ThO{sub 2}-PuO{sub 2} thorium fuel can consume plutonium 2.1-2.4 times MOX fuel. The fraction of fissile plutonium in the spent ThO{sub 2}-PuO{sub 2} thorium fuel is more favorable in view of plutonium consumption and non-proliferation than MOX fuel. The radioactivity of spent ThO{sub 2}-PuO{sub 2} thorium and MOX fuel batches were calculated. Since plutonium isotopes are dominant for the long-term radioactivity, ThO{sub 2}-PuO{sub 2} thorium has almost the same level of radioactivity as in MOX fuel for a long-term perspective. (author). 22 figs., 11 tabs.

Minimization of the fission product waste by using thorium based fuel instead of uranium dioxide

Energy Technology Data Exchange (ETDEWEB)

Galahom, A. Abdelghafar, E-mail: Agalahom@yahoo.com

2017-04-01

This research discusses the neutronic characteristics of VVER-1200 assembly fueled with five different fuel types based on thorium. These types of fuel based on mixing thorium as a fertile material with different fissile materials. The neutronic characteristics of these fuels are investigated by comparing their neutronic characteristics with the conventional uranium dioxide fuel using the MCNPX code. The objective of this study is to reduce the production of long-lived actinides, get rid of plutonium component and to improve the fuel cycle economy while maintaining acceptable values of the neutronic safety parameters such as moderator temperature coefficient, Doppler coefficient and effective delayed neutrons (β). The thorium based fuel has a more negative Doppler coefficient than uranium dioxide fuel. The moderator temperature coefficient (MTC) has been calculated for the different proposed fuels. Also, the fissile inventory ratio has been calculated at different burnup step. The use of Th-232 as a fertile material instead of U-238 in a nuclear fuel is the most promising fuel in VVER-1200 as it is the ideal solution to avoid the production of more plutonium components and long-lived minor actinides. The reactor grade plutonium accumulated in light water reactor with burnup can be recycled by mixing it with Th-232 to fuel the VVER-1200 assembly. The concentrations of Xe-135 and Sm-151 have been investigated, due to their high thermal neutron absorption cross section.

Transmutation studies in France, R and D programme on fuels and targets

International Nuclear Information System (INIS)

Boidron, M.; Chauvin, N.; Garnier, J.C.; PIllon, S.; Vambenepe, G.

2001-01-01

For the management of high level and long-lived radioactive waste, a large and continuous research and development effort is carried out in France, to provide a wide range of scientific and technical alternatives along three lines, partitioning and transmutation, disposal in deep geological formations and long term interim surface or subsurface storage. For the line one, and in close link with the partitioning studies, research is carried out to evaluate the transmutation potential of long-lived waste in appropriate reactors configurations (scenarios) relying on current technologies as well as innovative reactors. Performed to evaluate the theoretical feasibility of the Pu consumption and waste transmutation from the point of view of the reactor cores physics to reach the equilibrium of the material fluxes (i.e. consumption = production) and of the isotopic compositions of the fuels, these studies insure the 'scientific' part of the transmutation feasibility. For the technological part of the feasibility of waste transmutation in reactors, a large programme on fuel development is underway. This includes solutions based on the advanced concepts for plutonium fuels in PWR and the development of specific fuels and targets for transmutation in fast reactors in the critical or sub-critical state. For the waste transmutation in fast reactors, an important programme has been launched to develop specific fuels and targets with experiments at various stages of preparation in different experimental reactors including Phenix. Composite fuels as well as particle fuels are considered. This programme is presented and recent results concerning the preparation of the experiments, the characterisation of the compounds properties, the thermal and mechanical modelling and the behaviour of U free fuels are given. (author)

Thorium in nuclear fuel

International Nuclear Information System (INIS)

Stankevicius, Alejandro

2012-01-01

We revise the advantages and possible problems on the use of thorium as a nuclear fuel instead of uranium. The following aspects are considered: 1) In the world there are three times more thorium than uranium 2) In spite that thorium in his natural form it is not a fisil, under neutron irradiation, is possible to transform it to uranium 233, a fisil of a high quality. 3) His ceramic oxides properties are superior to uranium or plutonium oxides. 4) During the irradiation the U 233 due to n,2n reaction produce small quantities of U 232 and his decay daughters' bismuth 212 and thallium 208 witch are strong gamma source. In turn thorium 228 and uranium 232 became, in time anti-proliferate due to there radiation intensity. 5) As it is described in here and experiments done in several countries reactors PHWR can be adapted to the use of thorium as a fuel element 6) As a problem we should mentioned that the different steps in the process must be done under strong radiation shielding and using only automatized equipment s (author)

Concept of a subcritical transmutation system with fast neutron spectrum and liquid fuel

International Nuclear Information System (INIS)

Tittelbach, S.

2002-11-01

The annual amount of nearly 9500 t of spent fuel from worldwide industrial nuclear energy utilization has to be disposed as high level waste. The retention of nuclear waste from the biosphere has to be assured until the radiological risk decreases to tolerable levels. The long-term radiological risk of spent fuel is dominated by actinide elements, i.e. plutonium, americium and curium. It is intended to reduce this amount of high level waste by Partitioning and Transmutation, so that the radiotoxicity of the disposed waste falls short of the reference value of fresh fuel decaying naturally after about thousand years. For this time period the retention of high level waste can be assured by technical means. The scope of this work is the design of a subcritical fast transmutation system with liquid metal cooling and liquid metal fuel. The lead bismuth eutectic has been choosen as the liquid metal coolant and fuel carrier. To dissolve at least 3 at% of transuran elements, a minimum fuel temperature of 600 C is required. The calculations were carried out with a fuel composition, which results from two plutonium recycling steps in a thorium fuel cycle. Two homogeneous and two heterogeneous blankets have been designed and evaluated leading to one preferred heterogeneous blanket design, which has been investigated in more detail. This blanket design merges the positive properties of a solid fuel system (better control of fuel and reactivity because of smaller and closed fuel volumina) and a liquid fuel system (continous charge and discharge or extraction of fission products). The blanket design is based on the core design of fast breeder liquid metal reactors. It consists of hexagonal fuel elements housing up to six annular shaped fuel cylinders. The hexagonal shape of the fuel elements leads to three fuel zones positioned concentrically around the central spallation target. There is a strong heterogeneous distribution of power and heat flux in this blanket design. Besides

Transmutation of minor actinides in a Candu thorium borner

International Nuclear Information System (INIS)

Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Yildiz, K.; Sahin, N.; Altinok, T.; Alkan, M.

2007-01-01

The paper investigates the prospects of exploitation of rich world thorium reserves in CANDU reactors. Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium can be used as a booster fissile fuel material in form of mixed ThO 2 /PuO 2 fuel in a CANDU fuel bundle in order to assure reactor criticality. Two different fuel compositions have been selected for investigations: 1) 96% thoria (ThO 2 ) + 4% PuO 2 and 2) 91% ThO 2 + 5% UO 2 + 4 PuO 2 . The latter is used for the purpose of denaturing the new 2 33U fuel with 2 38U. The behavior of the criticality k ∞ and the burn-up values of the reactor have been pursued by full power operation for > ∼ 8 years. The reactor starts with k ∞ = ∼ 1.39 and the criticality drops down asymptotically to values k ∞ > 1.06, still tolerable and usable in a CANDU reactor. Reactor criticality k ∞ remains nearly constant between the 4th year and 7th year of plant operation and then a slight increase is observed thereafter, along with a continuous depletion of thorium fuel. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn up can be achieved with the same fuel (> 160 000 MW.D/MT). The reactor criticality would be sufficient until a great fraction of the thorium fuel is burnt up, provided that the fuel rods could be fabricated to withstand such high burn up levels. Fuel fabrication costs and nuclear waste mass for final disposal per unit energy could be reduced drastically. There is a great quantity of weapon grade plutonium accumulated in nuclear stockpiles. In the second phase of investigations, weapon grade plutonium is used as a booster fissile fuel material in form of mixed ThO 2 /PuO 2 fuel in a CANDU fuel bundle in order to assure the initial criticality at startup. Two different fuel compositions have been used: 1) 97% thoria (ThO 2 ) + 3% PuO 2 and 2) 92% ThO 2 + 5% UO 2 + 3% PuO 2 . The

Thorium fuel cycle analysis

Energy Technology Data Exchange (ETDEWEB)

Yamaji, K [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

1980-07-01

Systems analysis of the thorium cycle, a nuclear fuel cycle accomplished by using thorium, is reported in this paper. Following a brief review on the history of the thorium cycle development, analysis is made on the three functions of the thorium cycle; (1) auxiliary system of U-Pu cycle to save uranium consumption, (2) thermal breeder system to exert full capacity of the thorium resource, (3) symbiotic system to utilize special features of /sup 233/U and neutron sources. The effects of the thorium loading in LWR (Light Water Reactor), HWR (Heavy Water Reactor) and HTGR (High Temperature Gas-cooled Reactor) are considered for the function of auxiliary system of U-Pu cycle. Analysis is made to find how much uranium is saved by /sup 233/U recycling and how the decrease in Pu production influences the introduction of FBR (Fast Breeder Reactor). Study on thermal breeder system is carried out in the case of MSBR (Molten Salt Breeder Reactor). Under a certain amount of fissile material supply, the potential system expansion rate of MSBR, which is determined by fissile material balance, is superior to that of FBR because of the smaller specific fissile inventory of MSBR. For symbiotic system, three cases are treated; i) nuclear heat supply system using HTGR, ii) denatured fuel supply system for nonproliferation purpose, and iii) hybrid system utilizing neutron sources other than fission reactor.

Oxide fuels and targets for transmutation

International Nuclear Information System (INIS)

Sudreau, F.; Bonnerot, J.M.; Warin, D.; Gaillard-Groleas, G.; Ferroud-Plattet, M.P.

2007-01-01

Full text of publication follows. Direction 1 of the French Act dated 30 December 1991 on the management of high-level, long-lived radioactive waste involves exploring solutions designed to separate long-lived radionuclides from the spent fuel and to transmute them under neutron flux into shorter half-lives or stable elements. In the French research programme conducted by CEA, these radionuclides are mainly minor actinides (americium, neptunium and curium) and fission products (particularly caesium, iodine and technetium). Within this context, this paper aims at illustrating the vast programme that CEA has performed in order to demonstrate the scientific and technical feasibility of minor actinide transmutation. An important part of the research was carried out in collaboration with French research (CNRS) and industrial (EDF, AREVA) organisations, and also in the framework of international co-operation programmes with the European Institute for Transuranium Elements in Karlsruhe (ITU), the US Department of Energy (DOE), the Japanese Atomic Energy Research Institute (now JAEA) and Central Research Institute of Electric Power Industry (CRIEPI) and the Russian Ministry for Atomic Energy (ROSATOM). Such research made it possible to evaluate the capacity of MOX fuels to be used as a support for minor actinide transmutation (homogeneous method). Simulations of pressurised water reactor (PWR) fuels have revealed the limits of this transmutation method, which are mainly related to the pressurization of the fuel rods and the formation of high active californium. On the contrary, for sodium-cooled fast reactor fuels possibly designed with large expansion plenums a first experimental demonstration of the transmutation of americium and neptunium has been successful in the Phenix reactor. Various studies designed to demonstrate the theoretical and experimental feasibility of transmutation using an inert support (heterogeneous method) have been carried out in HFR (EFTTRA

Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-31

In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

The thorium fuel cycle in water-moderated reactor systems

International Nuclear Information System (INIS)

Critoph, E.

1977-01-01

Current interest in the thorium cycle, as an alternative to the uranium cycle, for water-moderated reactors is based on two attractive aspects of its use - the extension of uranium resources, and the related lower sensitivity of energy costs to uranium price. While most of the scientific basis required is already available, some engineering demonstrations are needed to provide better economic data for rational decisions. Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. There appear to be no major feasibility problems associated with the use of thorium, although development is required in the areas of fuel testing and fuel management. The use of thorium cycles implies recycling the fuel, and the major uncertainties are in the associated costs. Experience in the design and operation of fuel reprocessing and active-fabrication facilities is required to estimate costs to the accuracy needed for adequately defining the range of conditions economically favourable to thorium cycles. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An ''inventory'' of uranium of between 1 and 2Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium), is some two decades

Evaluation of plutonium, uranium, and thorium use in power reactor fuel cycles

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.

1977-01-01

% 233 U(''denatured'' fuel). Thorium based fuels appear to be able to utilize metal fuels advantageously, but additional irradiation performance testing is needed

Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

International Nuclear Information System (INIS)

Todosow, M.; Raitses, G.; Galperin, A.

2009-01-01

Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

Energy Technology Data Exchange (ETDEWEB)

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Drying characteristics of thorium fuel corrosion products

Energy Technology Data Exchange (ETDEWEB)

Smith, R.-E. E-mail: rzl@inel.gov

2004-07-01

The open literature and accessible US Department of Energy-sponsored reports were reviewed for the dehydration and rehydration characteristics of potential corrosion products from thorium metal and thorium oxide nuclear fuels. Mixed oxides were not specifically examined unless data were given for performance of mixed thorium-uranium fuels. Thorium metal generally corrodes to thorium oxide. Physisorbed water is readily removed by heating to approximately 200 deg. C. Complete removal of chemisorbed water requires heating above 1000 deg. C. Thorium oxide adsorbs water well in excess of the amount needed to cover the oxide surface by chemisorption. The adsorption of water appears to be a surface phenomenon; it does not lead to bulk conversion of the solid oxide to the hydroxide. Adsorptive capacity depends on both the specific surface area and the porosity of the thorium oxide. Heat treatment by calcination or sintering reduces the adsorption capacity substantially from the thorium oxide produced by metal corrosion.

The Physics of transmutation systems : system capabilities and performances

International Nuclear Information System (INIS)

Finck, P. J.

2002-01-01

This document is complementary to a document produced by Prof. Salvatores on ''The Physics of Transmutation in Critical or Subcritical Reactors and the Impact on the Fuel Cycle''. In that document, Salvatores describes the fundamental of transmutation, through basic physics properties and general parametric studies. In the present document we try to go one step further towards practical implementation (while recognizing that the practical issues such as technology development and demonstration, and economics, can only be mentioned in a very superficial manner). Section 1 briefly overviews the possible objectives of transmutation systems, and links these different objectives to possible technological paths. It also describes the overall constraints which have to be considered when developing and implementing transmutation systems. In section 2 we briefly overview the technological constraints which need to be accounted for when designing transmutation systems. In section 3 we attempt to provide a simplified classification of transmutation systems in order to clarify later comparisons. It compares heterogeneous and homogeneous recycle strategies, and single and multi-tier systems. Section 4 presents case analyses for assessing the transmutation performance of various individual systems, starting with LWR's ((1) generic results; (2) multirecycle of plutonium; (3) an alternative: transmutation based on a Thorium fuel cycle), followed by Gas-Cooled Reactors (with an emphasis on the ''deep burn'' approach), and followed by Fast Reactors and Accelerator Driven systems ((1) generic results; (2) homogeneous recycle of transuranics; (3) practical limit between Fast Reactors and Accelerator Driven Systems) Section 5 summarizes recent results on integrated system performances. It focuses first on interface effects between the two elements of a dual tier system, and then summarizes the major lessons learned from recent global physics studies

Light water reactors with a denatured thorium fuel cycle

International Nuclear Information System (INIS)

1978-05-01

Discussed in this paper is the performance of denatured thorium fuel cycles in PWR plants of conventional design, such as those currently in operation or under construction. Although some improvement in U 3 O 8 utilization is anticipated in PWRs optimized explicitly for the denatured thorium fuel cycle, this paper is limited to a discussion of the performance of denatured thorium fuels in conventional PWRs and consequently the data presented is representative of the use of thorium fuel in existing PWRs or those presently under construction. In subsequent sections of this paper, the design of the PWR, its performance on the denatured thorium fuel cycle, safety, accident and environmental considerations, and technological status and R and D requirements are discussed

Technical meeting on 'Review of solid and mobile fuels for partitioning and transmutation systems'. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

The topics covered during the Meeting were divided into two Sessions. Session 1 - Qualification of Solid and Mobile Fuels delt with: Neutronic, fuel and material properties of a molten salt transmuter; and Preliminary analysis of transmutation fuels for KALIMER. Session 2 - Reactor Physics and Safety Characteristics of Transmutation Systems based on Solid and Mobile Fuel Types included the following: Activity in NEA for P and T area; IAEA activities in the area of partitioning and transmutation; The R and D activity in Brazil: A conceptual fast energy amplifier ADS cooled by helium double stata Th/U fuel cycle; Closed fuel cycle and contemporary tendencies of the nuclear facilities development; Current Russian activities in P and T area; Pyrochemical reprocessing and nuclear spent fuel disposal project; Fuel selection criteria specific for double stratum minor actinide burners.

Transmutation Fuel Campaign Description and Status

International Nuclear Information System (INIS)

Jon Carmack; Kemal O. Pasamehmetoglu

2008-01-01

This report contains a technical summary package in response to a Level 2 milestone in the transmutation fuel campaign (TFC) management work-package calling for input to the Secretarial decision. At present, the form of the Secretarial decision package is not fully defined, and it is not clear exactly what will be required from the TFC as a final input. However, it is anticipated that a series of technical and programmatic documents will need to be provided in support of a wider encompassing document on GNEP technology development activities. The TFC technical leadership team provides this report as initial input to the secretarial decision package which is being developed by the Technical Integration Office (TIO) in support of Secretarial decision. This report contains a summary of the TFC execution plan with a work breakdown structure, high level schedule, major milestones, and summary description of critical activities in support of campaign objectives. Supporting documents referenced in this report but provided under separate cover include: (1) An updated review of the state-of-the art for transmutation fuel development activities considering national as well as international fuel research and development testing activities. (2) A definition of the Technology Readiness Level (TRL) used to systematically define and execute the transmutation fuel development activities

Analysis on specific nuclear data for reactors physics computations applied to CANDU reactors using thorium-based fuels

International Nuclear Information System (INIS)

Visan, Iuliana E.

2010-01-01

The purpose of this work is to analyze the evaluated nuclear data from ENDF libraries IAEA69 (69 energy groups library) and IAEA172 (172 energy groups library), respectively, in WIMS library format and to represent neutron fission yield, absorption and fission cross-section dependence for 233 Uranium, 232 Thorium isotopes and some actinides of interest on the incident energy. Our interest for these two isotopes is mainly based on the importance of 233 Uranium as 'fissile nucleus' in Thorium-Uranium fuel cycle. Nowadays, nuclear data evaluation for the actinides generated in Thorium-Uranium fuel cycle is seen as a world-wide priority. The fissile nucleus, 233 Uranium 'plays' the same function in Thorium-Uranium fuel cycle as the 235 Uranium in 'the classic' Uranium-Plutonium fuel cycle. As opposed to natural Uranium which contains 0.7 % of the fissile isotope 235 Uranium, natural Thorium doesn't contain fissile isotopes, being composed entirely by the fertile isotope 232 Thorium. Graphical evolutions of interest parameters versus the incident energy are presented. Our interest was also to observe the behavior of these nuclear data for fast, resonance and thermal energy groups, respectively. The ENDF nuclear data libraries are constantly up-dated, so that we can observe an improvement of the IAEA172 library, which disposes of evaluated nuclear data at higher energies (about 20 MeV), as opposed to IAEA69 library (which includes evaluated nuclear data below 10 MeV). Based on our graphical representation, a good agreement between the considered libraries has been observed, sustaining nuclear data validity. (authors)

An assessment of once-through homogeneous thorium fuel economics for light water reactors

International Nuclear Information System (INIS)

Joo, Hyung Kook; Noh, Jae Man; Yoo, Jae Woon

2001-01-01

The fuel economics of an once-through homogeneous thorium fuel concept for PWR was assessed by doing a detailed core analysis. In addition to this, the fuel economics assessment was also performed for two other ways enhancing the economic potential of thorium fuel; thorium utilization in the mixed core with uranium fuel assembly and Duplex thorium fuel concepts. As a results of fuel economics assessment, the thorium fuel cycle does not show any economic incentives in preference to uranium fuel cycle under the 18-months fuel cycle for PWR. However, the utilization of thorium is the mixed core with uranium fuel assembly and Duplex thorium fuel cycle and show superior fuel economics to uranium fuel under the longer fuel cycle scheme. The economic potential of once-through thorium fuel cycle is expected to be increased further by utilizing the Duplex thorium fuel in the mixed core with uranium fuel assembly

Some thorium fuel cycle strategies

International Nuclear Information System (INIS)

Duret, M.F.; Hatton, H.

1979-02-01

The report deals with the problem of introducing an advanced nuclear fuel cycle based on thorium in Canada. It is pointed out that timing and introduction rate are important considerations, certain choices of these variables leading to undesirable business fluctuations in some of the industries involved in the production of nuclear energy. (author)

An evaluation of once-through homogeneous thorium fuel cycle for light water reactors

International Nuclear Information System (INIS)

Joo, H. K.; Noh, J. M.; Yoo, J. W.

2002-01-01

The other ways enhancing the economic potential of thorium fuel has been assessed ; the utilization of lower enriched uranium in thorium-uranium fuel, duplex thorium fuel concept, thorium utilization in the mixed core with uranium fuel assembly and thorium blanket utilization in the uranium core. The fuel economics of the proposed ways of thorium fuel increased compared to the previous homogeneous thorium fuel cycle. Compared to uranium fuel cycle, however, they do not show any economic incentives. From the view of proliferation resistance potential, thorium fuel option has the advantage to reduce the inventory of plutonium production. Any of proposed thorium options are less economical than uranium fuel option, the thorium fuel option has the potential to be utilized in the future for the sake of the effective consumption of excessive plutonium and the preparation against the using up of uranium resource

Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

Energy Technology Data Exchange (ETDEWEB)

Trianti, N.; Su' ud, Z.; Riyana, E. S. [Nuclear Physics and Biophysics Research Division Department of Physics - Institut Teknologi Bandung (ITB) Jalan Ganeca 10 Bandung 40132 (Indonesia)

2012-06-06

A preliminary design study for the utilization of thorium added with {sup 231}Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of {sup 233}U to {sup 231}Pa in burn-up process. Optimizations of the content of {sup 231}Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 {approx} 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

Thorium fuels for heavy water reactors. Romanian experience

International Nuclear Information System (INIS)

Glodeanu, F.; Mirion, I.; Mehedinteanu, S.; Balan, V.

1984-01-01

The renewed interest in thorium fuel cycle due to the increased demand for fissile materials has resulted in speeding up the related research and development activities. For heavy water reactors the thorium cycles, especially SSET, are very promising and many efforts are made to demonstrate their feasibility. In our country, at INPR, the research and development activity has been initiated in the following areas: the conceptual design of thorium bearing fuel elements; fuel modelling; nuclear grade thorium dioxide powder technology; mixed oxide fuel technology. In the design area, the key factors in performance limitation, especially at extended burnup have been accounted and different remedies proposed. An irradiation programme has been settled and will start this year. The modelling activities are focused on mixed oxide behaviour and material data measurements are in progress. In the nuclear grade thorium powder technology area, a good piece of work has been done to develop an integrated technology for monasite processing (thorium being a by-product in lanthanides extraction). As regards the mixed oxide fuel technology, efforts have been made to obtain (ThU)O 2 pellets with good homogeneity and high density at different compositions. Besides the mixing powders route, other non-conventional technologies for refabrication like: microspheres, pellet impregnation and clay extrusion are studied. Experimental fuel rods for irradiation testing have been manufactured. (author)

Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

Energy Technology Data Exchange (ETDEWEB)

Calabrese, R., E-mail: rolando.calabrese@enea.i [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Vettraino, F.; Artioli, C. [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Sobolev, V. [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Thetford, R. [Serco Technical and Assurance Services, 150 Harwell Business Centre, Didcot OX11 0QB (United Kingdom)

2010-06-15

Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R and D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic-ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic-metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg{sub HM} discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW{sub (th)} loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life

A Simplified Supercritical Fast Reactor with Thorium Fuel

Directory of Open Access Journals (Sweden)

Peng Zhang

2014-01-01

Full Text Available Super-Critical water-cooled Fast Reactor (SCFR is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

The sphinx project: experimental verification of design inputs for a transmuter with liquid fuel based on molten fluorides

International Nuclear Information System (INIS)

Hron, M.; Uhlir, J.; Vanicek, J.

2002-01-01

The current proposals for high-active long-lived (more then 10 4 years) waste from spent nuclear fuel disposal calls forth an increasing societal mistrust towards nuclear power. These problems are highly topical in the Czech Republic, a country which is operating nuclear power and accumulating spent fuel from PWRs and is further located on an inland and heavily populous Central European region. The proposed project, known under the acronym SPHINX (SPent Hot fuel Incineration by Neutron flux) deals with a solution to some of the principle problems through a very promising means of radioactive waste treatment. In particular, high-level wastes from spent nuclear fuel could be treated using this method, which is based on the transmutation of radionuclides through the use of a nuclear reactor with liquid fuel based on molten fluorides (Molten Salt Transmutation Reactor - MSTR) which might be a subcritical system driven by a suitable neutron source. Its superiority also lies in the fact that it makes possible to utilize actinides contained, by others, in spent nuclear fuel and so to reach a positive energy effect. After the first three-year stage of Research and Development which has been focused mostly on computer analyses of neutronics and corresponding physical characteristics, the next three-year stage of this programme will be devoted to experimental verification of inputs for the design of a demonstration transmuter using molten fluoride fuel. The Research and Development part of the SPHINX project in the area of fuel cycle of the MSTR is focused in the first place on the development of suitable technology for the preparation of an introductory liquid fluoride fuel for MSTR and subsequently on the development of suitable fluoride pyrometallurgical technology for the separation of the transmuted elements from the non-transmuted ones. The idea of the introductory fuel preparation is based on the reprocessing of PWR spent fuel using the Fluoride Volatility Method

INR Recent Contributions to Thorium-Based Fuel Using in CANDU Reactors

International Nuclear Information System (INIS)

Prodea, I.; Mărgeanu, C. A.; Rizoiu, A.; Olteanu, G.

2014-01-01

The paper summarizes INR Pitesti contributions and latest developments to the Thorium-based fuel (TF) using in present CANDU nuclear reactors. Earlier studies performed in INR Pitesti revealed the CANDU design potential to use Recovered Uranium (RU) and Slightly Enriched Uranium (SEU) as alternative fuels in PHWRs. In this paper, we performed both lattice and CANDU core calculations using TF, revealing the main neutron physics parameters of interest: k-infinity, coolant void reactivity (CVR), channel and bundle power distributions over a CANDU 6 reactor core similar to that of Cernavoda, Unit 1. We modelled the so called Once Through Thorium (OTT) fuel cycle, using the 3D finite-differences DIREN code, developed in INR. The INR flexible SEU-43 bundle design was the candidate for TF carrying. Preliminary analysis regarding TF burning in CANDU reactors has been performed using the finite differences 3D code DIREN. TFs showed safety features improvement regarding lower CVRs in the case of fresh fuel use. Improvements added to the INR ELESIMTORIU- 1 computer code give the possibility to fairly simulate irradiation experiments in INR TRIGA research reactor. Efforts are still needed in order to get better accuracy and agreement of simulations to the experimental results. (author)

Transmutation of minor actinide using BWR fueled mixed oxide

International Nuclear Information System (INIS)

Susilo, Jati

2000-01-01

Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

International Nuclear Information System (INIS)

Jeong, Chang Joon; Park, Chang Je

2006-02-01

The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO 2 UO 2 and ThO 2 UO 2 -DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future

Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Park, Chang Je

2006-02-15

The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO{sub 2}UO{sub 2} and ThO{sub 2}UO{sub 2}-DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future.

Fuels and targets for the transmutation of high activity long lived radioactive wastes

International Nuclear Information System (INIS)

Pillon, S.; Warin, D.

2010-01-01

The authors present and comment the different strategies which can be adopted to transmute minor actinides (concerned reactors, in fast breeder reactors, in accelerator driven systems or ADS), and the chemical composition of transmutation fuels (actinide compounds, inert matrices, fuels and targets). They describe the behaviour of refractory ceramic fuels during their service life under irradiation with their different damage origins (neutrons, fission by-products, alpha particles), the fabrication of transmutation fuels and targets through different processes (metallurgical, co-precipitate, sol-gel, wax, infiltration of radioactive materials, VIPAC/SPHEREPAC) and the reprocessing or recycling of these transmutation fuels and targets

Fuel design for the U.S. accelerator driven transmutation system

International Nuclear Information System (INIS)

Meyer, M. K.; Hayes, S. L.; Crawford, D. C.; Pahl, R. G.; Tsai, H.

2002-01-01

The U.S. concept for actinide transmutation is currently envisioned as a system to destroy plutonium as well as minor actinides in a single or two tier system. In order to maximize the actinide destruction rate, an inert matrix fuel is used. The effectiveness of transmutation in reducing the actinide inventory is linked to the development of a robust fuel system, capable of achieving very high burnup. Very little fuel performance data has been generated to date on inert matrix systems, and there are several issues specific to the behavior of higher actinides that do not allow extension of the existing uranium-plutonium fuel database to these new fuels. These issues include helium production, fuel-cladding-chemical-interaction, and americium migration. In the early 1990's, two U-Pu-Zr metal alloy fuel elements containing 1.2 wt.% Am and 1.3 wt.% Np were fabricated and irradiated to approximately 6 at.% burnup in the Experimental Breeder Reactor-II. Postirradiation examination results were not published; however the recent interest in fuel for actinide transmutation has prompted a reexamination of this data. The results of the postirradiation examination of this experiment, including gas sampling, metallography, and gamma scanning are discussed. Available data on inert matrix fuels and other fuels incorporating actinides are used to assess the implications of minor-actinide specific issues on transmuter fuel. Considerations for the design of nitride and oxide fuels, metallic fuels, and metal-matrix dispersion fuels are discussed

Fuel Design for the U.S. Accelerator Driven Transmutation System

International Nuclear Information System (INIS)

Meyer, M.K.; Hayes, S.L.; Crawford, D.C.; Pahl, R.G.; Tsai, H.

2002-01-01

The U.S. concept for actinide transmutation is currently envisioned as a system to destroy plutonium as well as minor actinides in a single or two tier system. In order to maximize the actinide destruction rate, an inert matrix fuel is used. The effectiveness of transmutation in reducing the actinide inventory is linked to the development of a robust fuel system, capable of achieving very high burnup. Very little fuel performance data has been generated to date on inert matrix systems, and there are several issues specific to the behavior of higher actinides that do not allow extension of the existing uranium-plutonium fuel database to these new fuels. These issues include helium production, fuel-cladding-chemical-interaction, and americium migration. In the early 1990's, two U-Pu-Zr metal alloy fuel elements containing 1.2 wt.% Am and 1.3 wt.% Np were fabricated and irradiated to approximately 6 at.% burnup in the Experimental Breeder Reactor-II. Postirradiation examination results were not published; however the recent interest in fuel for actinide transmutation has prompted a reexamination of this data. The results of the postirradiation examination of this experiment, including gas sampling, metallography, and gamma scanning are discussed. Available data on inert matrix fuels and other fuels incorporating actinides are used to assess the implications of minor-actinide specific issues on transmuter fuel. Considerations for the design of nitride and oxide fuels, metallic fuels, and metal-matrix dispersion fuels are discussed. (authors)

Transmutation of transuranium elements in a gas-cooled accelerator-driven system

International Nuclear Information System (INIS)

Biss, Klaus Hendrik

2014-01-01

The peaceful usage of nuclear energy by light and boiling water reactors is connected with a buildup of long-lived high-level radioactive waste. Compared to the direct disposal, partitioning and transmutation (P and T) is considered as an effective way to reduce this waste in its quantity by converting it into short-lived radio nuclides. By that the long term radiotoxicity is reduced compared to direct disposal. Subcritical systems, which are powered by spallation processes for free neutron production to maintain the nuclear chain reaction, allow a target-oriented transmutation. As a subcritical system a gas-cooled accelerator driven system (ADS) for transmutation of transuranic elements has been modeled in this thesis to evaluate the reduction of the radio toxicity by P and T. The simulation of neutron-physical processes is based on the Monte Carlo computer program MCNPX. The development of an equilibrium core made it possible to study the transmutation and operating behavior for several fuel variations in a magnesium oxide matrix and develop a simplified burnup method. Americium as part of the fuel has a stabilizing effect on the neutron multiplication due to its conversion into plutonium during the operation. Thorium was investigated as an alternative matrix for the fuel in order to replicate the stabilizing effect of americium by the conversion of thorium in 233 U. By that a consistent operating cycle in the later P and T-process is ensured. Calculation of the nuclide composition at the end of a P and T-process leads to an expansion of the mathematical description of the mass reduction (transmutation efficiency) by the material located in the reactor. The achieved transmutation efficiency with the investigated ADS is 98.8 %. The transmutation time was examined with different operating strategies regarding the number, size and thermal power of use of transmutation facilities to determine the effort for the P and T-process depending on efficiency. It turns out

The evolutionary adoption of thorium beginning with its application in niche LWR fuels

International Nuclear Information System (INIS)

Drera, Saleem

2015-01-01

Since the inception of nuclear energy, the use of thorium as a nuclear fuel has been envisioned. Thorium boasts benefits, however, drawbacks which are both economic and technical including its the lack of a naturally occurring fissile isotope implies that its utility is inherently more difficult. The implementation of thorium as a nuclear fuel requires that it must provide sound technical advantages in combination with attractive economics as compared to standard uranium fuel. Revolutionary thorium concepts such as molten salt reactors and accelerator driven systems may provide theoretical merit, however, their exotic nature and associated technical challenges label them as long-term solutions at best. A near-to-medium term solution for thorium must be based on an evolutionary approach utilizing light/heavy water reactor platforms. While thorium does not provide a near-to-medium term complete replacement of uranium, it does provide substantial benefit within niche applications. To license and bring to market these niche fuels, Thor Energy and an international consortium of entities (including: Fortum, KAERI, Westinghouse, NNL, ITU, IFE, and a few other minor entities) have initiated a fuel development and irradiation test program to characterize the performance of these thoria-containing fuels. (author)

Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

International Nuclear Information System (INIS)

Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

1999-01-01

High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

Economic analysis of thorium-uranium fuel cycle introduced into PWRs

International Nuclear Information System (INIS)

Fan Li; Sun Qian

2014-01-01

Using PWR of Daya Bay Unit l as the reference reactor, a validated computer code was used to calculate the fuel cycle costs for uranium fuel cycle and thorium-uranium fuel cycle over the following 20 0perational years respectively. The calculation results show that the thorium-uranium fuel cycle is economically competitive with the uranium fuel cycle when reprocessing mode is adopted. For thorium-uranium fuel cycle, if the price of natural uranium is higher than 120 $ /pound U_3O_8, the fuel cycle cost of the direct disposal mode is greater than that of the reprocessing mode. Therefore, when the uranium price may maintain a high level long-termly, adopting reprocessing mode will benefit the economic advantage for the thorium-uranium fuel cycle introduced into PWRs. (authors)

Effects of actinide compositional variability in the U.S. spent fuel inventory on partitioning-transmutation systems

International Nuclear Information System (INIS)

Ludwig, S.B.; Michaels, G.E.; Hanson, B.D.

1993-01-01

The partitioning and transmutation concept (P-T) has as a mission the reduction by many orders of magnitude of certain undesirable nuclides in the waste streams. Given that only a very small fiction of spent fuel can be rejected by a P-T enterprise, a P-T system must therefore be capable of accommodating a wide range of spent fuel characteristics. Variability of nuclide composition (i.e. the feed material for transmutation devices) may be important because virtually all transmutation systems propose to configure TRU nuclides recovered from discharged LWR fuel in critical or near-critical cores. To date, all transmutation system core analyses assume nonvariable nuclide concentrations for startup and recycle cores. Using the Department of Energy (DOES) Characteristic Data Base (CDB) and the ORIGEN2 computer code, the current and projected spent fuel discharges until the year 2016 have been categorized according to combinations of fuel burnup, initial enrichment, fuel age (cooling time) and reactor type (boiling-water or pressurized-water reactor). In addition to quantifying the variability of nuclide composition in current and projected LWR fuel discharge, the variability of the infinite multiplication factor (K ∞ ) is calculated for both fast (ALMR) and thermal (accelerator-based) transmuter systems. It is shown that actinide compositional variations are potentially significant and warrant further investigation. (authors)

Advanced Reactor Technology Options for Utilization and Transmutation of Actinides in Spent Nuclear Fuel

International Nuclear Information System (INIS)

2009-09-01

Renewed interest in the potential of nuclear energy to contribute to a sustainable worldwide energy mix is strengthening the IAEA's statutory role in fostering the peaceful uses of nuclear energy, in particular the need for effective exchanges of information and collaborative research and technology development among Member States on advanced nuclear power technologies (Articles III-A.1 and III-A.3). The major challenges facing the long term development of nuclear energy as a part of the world's energy mix are improvement of the economic competitiveness, meeting increasingly stringent safety requirements, adhering to the criteria of sustainable development, and public acceptability. The concern linked to the long life of many of the radioisotopes generated from fission has led to increased R and D efforts to develop a technology aimed at reducing the amount of long lived radioactive waste through transmutation in fission reactors or accelerator driven hybrids. In recent years, in various countries and at an international level, more and more studies have been carried out on advanced and innovative waste management strategies (i.e. actinide separation and elimination). Within the framework of the Project on Technology Advances in Fast Reactors and Accelerator Driven Systems (http://www.iaea.org/inisnkm/nkm/aws/fnss/index.html), the IAEA initiated a number of activities on utilization of plutonium and transmutation of long lived radioactive waste, accelerator driven systems, thorium fuel options, innovative nuclear reactors and fuel cycles, non-conventional nuclear energy systems, and fusion/fission hybrids. These activities are implemented under the guidance and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Fast Reactors (TWG-FR). This publication compiles the analyses and findings of the Coordinated Research Project (CRP) on Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste (2002

Studies on the safety and transmutation behaviour of innovative fuels for light water reactors; Untersuchungen zum Sicherheits- und Transmutationsverhalten innovativer Brennstoffe fuer Leichtwasserreaktoren

Energy Technology Data Exchange (ETDEWEB)

Schitthelm, Oliver

2012-07-01

Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its {sup 238}U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

Comparison of the Environment, Health, And Safety Characteristics of Advanced Thorium- Uranium and Uranium-Plutonium Fuel Cycles

Science.gov (United States)

Ault, Timothy M.

The environment, health, and safety properties of thorium-uranium-based (''thorium'') fuel cycles are estimated and compared to those of analogous uranium-plutonium-based (''uranium'') fuel cycle options. A structured assessment methodology for assessing and comparing fuel cycle is refined and applied to several reference fuel cycle options. Resource recovery as a measure of environmental sustainability for thorium is explored in depth in terms of resource availability, chemical processing requirements, and radiological impacts. A review of available experience and recent practices indicates that near-term thorium recovery will occur as a by-product of mining for other commodities, particularly titanium. The characterization of actively-mined global titanium, uranium, rare earth element, and iron deposits reveals that by-product thorium recovery would be sufficient to satisfy even the most intensive nuclear demand for thorium at least six times over. Chemical flowsheet analysis indicates that the consumption of strong acids and bases associated with thorium resource recovery is 3-4 times larger than for uranium recovery, with the comparison of other chemical types being less distinct. Radiologically, thorium recovery imparts about one order of magnitude larger of a collective occupational dose than uranium recovery. Moving to the entire fuel cycle, four fuel cycle options are compared: a limited-recycle (''modified-open'') uranium fuel cycle, a modified-open thorium fuel cycle, a full-recycle (''closed'') uranium fuel cycle, and a closed thorium fuel cycle. A combination of existing data and calculations using SCALE are used to develop material balances for the four fuel cycle options. The fuel cycle options are compared on the bases of resource sustainability, waste management (both low- and high-level waste, including used nuclear fuel), and occupational radiological impacts. At steady-state, occupational doses somewhat favor the closed thorium option while low

Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

International Nuclear Information System (INIS)

Washington, J.; King, J.; Shayer, Z.

2017-01-01

Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO_2, Pu_3Si_2, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO_2) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO_2), Pu_0_._3_1ZrH_1_._6Th_1_._0_8, and PuZrO_2MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B_4C, CdO, Dy_2O_3, Er_2O_3, Eu_2O_3, Gd_2O_3, HfO_2, In_2O_3, Lu_2O_3, Sm_2O_3, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO_2MgO (8 wt% Pu) target fuel with a coating of Lu_2O_3 resulted in the highest rate of plutonium transmutation with the greatest reduction in curium

Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Washington, J., E-mail: jwashing@gmail.com [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Shayer, Z., E-mail: zshayer@mines.edu [Department of Physics, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States)

2017-03-15

Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO{sub 2}, Pu{sub 3}Si{sub 2}, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO{sub 2}) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO{sub 2}), Pu{sub 0.31}ZrH{sub 1.6}Th{sub 1.08}, and PuZrO{sub 2}MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B{sub 4}C, CdO, Dy{sub 2}O{sub 3}, Er{sub 2}O{sub 3}, Eu{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, HfO{sub 2}, In{sub 2}O{sub 3}, Lu{sub 2}O{sub 3}, Sm{sub 2}O{sub 3}, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO{sub 2}MgO (8 wt% Pu) target

Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

Energy Technology Data Exchange (ETDEWEB)

Han, Kyu Hyun [Korea Institute of Nuclear Safety, 19, Guseong-dong, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

2008-07-01

Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

Spent Nuclear Fuel Option Study on Hybrid Reactor for Waste Transmutation

International Nuclear Information System (INIS)

Hong, Seong Hee; Kim, Myung Hyun

2016-01-01

DUPIC nuclear fuel can be used in hybrid reactor by compensation of subcritical level through (U-10Zr) fuel. Energy production performance of Hyb-WT with DUPIC is grateful because it has high EM factor and performs waste transmutation at the same time. However, waste transmutation performance should be improved by different fissile fuel instead of (U-10Zr) fuel. SNF (Spent Nuclear Fuel) disposal is one of the problems in the nuclear industry. FFHR (Fusion-Fission Hybrid Reactor) is one of the most attractive option on reuse of SNF as a waste transmutation system. Because subcritical system like FFHR has some advantages compared to critical system. Subcritical systems have higher safety potential than critical system. Also, there is suppressed excess reactivity at BOC (Beginning of Cycle) in critical system, on the other hand there is no suppressed reactivity in subcritical system. Our research team could have designed FFHR for waste transmutation; Hyb-WT. Various researches have been conducted on fuel and coolant option for optimization of transmutation performance. However, Hyb-WT has technical disadvantage. It is required fusion power (Pfus) which is the key design parameter in FFHR is increased for compensation of decreasing subcritical level. As a result, structure material integrity is damaged under high irradiation condition by increasing Pfus. Also, deep burn of reprocessed SNF is limited by weakened integrity of structure material. Therefore, in this research, SNF option study will be conducted on DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactor) fuel, TRU fuel and DUPIC + TRU mixed fuel for optimization of Hyb-WT performance. Goal of this research is design check for low required fusion power and high waste transmutation. In this paper, neutronic analysis is conducted on Hyb-WT with DUPIC nuclear fuel. When DUPIC nuclear fuel is loaded in fast neutron system, supplement fissile materials need to be loaded together for compensation of low criticality

Spent Nuclear Fuel Option Study on Hybrid Reactor for Waste Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

DUPIC nuclear fuel can be used in hybrid reactor by compensation of subcritical level through (U-10Zr) fuel. Energy production performance of Hyb-WT with DUPIC is grateful because it has high EM factor and performs waste transmutation at the same time. However, waste transmutation performance should be improved by different fissile fuel instead of (U-10Zr) fuel. SNF (Spent Nuclear Fuel) disposal is one of the problems in the nuclear industry. FFHR (Fusion-Fission Hybrid Reactor) is one of the most attractive option on reuse of SNF as a waste transmutation system. Because subcritical system like FFHR has some advantages compared to critical system. Subcritical systems have higher safety potential than critical system. Also, there is suppressed excess reactivity at BOC (Beginning of Cycle) in critical system, on the other hand there is no suppressed reactivity in subcritical system. Our research team could have designed FFHR for waste transmutation; Hyb-WT. Various researches have been conducted on fuel and coolant option for optimization of transmutation performance. However, Hyb-WT has technical disadvantage. It is required fusion power (Pfus) which is the key design parameter in FFHR is increased for compensation of decreasing subcritical level. As a result, structure material integrity is damaged under high irradiation condition by increasing Pfus. Also, deep burn of reprocessed SNF is limited by weakened integrity of structure material. Therefore, in this research, SNF option study will be conducted on DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactor) fuel, TRU fuel and DUPIC + TRU mixed fuel for optimization of Hyb-WT performance. Goal of this research is design check for low required fusion power and high waste transmutation. In this paper, neutronic analysis is conducted on Hyb-WT with DUPIC nuclear fuel. When DUPIC nuclear fuel is loaded in fast neutron system, supplement fissile materials need to be loaded together for compensation of low criticality

JAERI R & D on accelerator-based transmutation under OMEGA program

Energy Technology Data Exchange (ETDEWEB)

Takizuka, T.; Nishida, T.; Mizumoto, M. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1995-10-01

The overview of the Japanese long-term research and development program on nuclide partitioning and transmutation, called {open_quotes}OMEGA,{close_quotes} is presented. Under this national program, major R&D activities are being carried out at JAERI, PNC, and CRIEPI. Accelerator-based transmutation study at JAERI is focused on a dedicated transmutor with a subcritical actinide-fueled subcritical core coupled with a spallation target driven by a high intensity proton accelerator. Two types of system concept, solid system and molten-salt system, are discussed. The solid system consists of sodium-cooled tungsten target and metallic actinide fuel. The molten-salt system is fueled with molten actinide chloride that acts also as a target material. The proposed plant transmutes about 250 kg of minor actinide per year, and generates enough electricity to power its own accelerator. JAERI is proposing the development of an intense proton linear accelerator ETA with 1.5 GeV-10 mA beam for engineering tests of accelerator-based transmutation. Recent achievements in the accelerator development are described.

The use of thorium as an alternative nuclear fuel

International Nuclear Information System (INIS)

Wilson, D.J.

1982-04-01

The use of thorium as an alternative or supplementary nuclear fuel is examined and compared with uranium. A description of various reactor types and their suitability to thorium fuel, and a description of various aspects of the fuel cycle from mining to waste disposal, are included. Comments are made on the safety and economics of each aspect of the fuel cycle and the extension of the lifetime of nuclear fuel

Growth scenarios with thorium fuel cycles in pressurised heavy water reactors

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1991-01-01

Since India has generous deposits of thorium, the availability of thorium will not be a limiting factor in any growth scenario. It is fairly well accepted that the best system for utilisation of thorium is the heavy water reactor. The growth scenarios possible using thorium in HWRs are considered. The base has been taken as 50,000 tons of natural uranium and practically unlimited thorium. The reference reactor has been assumed to be the PHWR, and all other growth scenarios are compared with the growth scenario provided by the once-through natural cycle in the PHWR. Two reactor types have been considered: the heavy water moderated, heavy water cooled, pressure tube reactor, known as the PHWR; and the heavy water moderated and cooled pressure vessel kind, similar to the ATUCHA reactor in Argentina. For each reactor, a number of different fuel cycles have been studied. All these cycles have been based on thorium. These are: the self-sustaining equilibrium thorium cycle (SSET); the high conversion ratio high burnup cycle; and the once through thorium cycle (OTT). The cycle have been initiated in two ways: one is by starting the cycle with natural uranium, reprocessing the spent fuel to obtain plutonium, and use that plutonium to initiate the thorium cycle; the other is to enrich the uranium to about 2-3% U-235 (the so-called Low Enriched Uranium or LEU), and use the LEU to initiate the thorium cycle. Both cases have been studied, and growth scenarios have been projected for every one of the possible combinations. (author). 1 tab

Fuels and materials for transmutation. A status report

International Nuclear Information System (INIS)

2005-01-01

The safe and efficient management of spent fuel from the operation of commercial nuclear power plants is an important issue. Worldwide, more than 250 000 tons of spent fuel from reactors currently operating will require disposal. These numbers account for only high-level radioactive waste generated by present-day power reactors. Nearly all issues related to risks to future generations arising from the long-term disposal of such spent nuclear fuel is attributable to only about 1% of its content. This 1% is made up primarily of plutonium, neptunium, americium and curium (called transuranic elements) and the long-lived isotopes of iodine and technetium. When transuranics are removed from discharged fuel destined for disposal, the toxic nature of the spent fuel drops below that of natural uranium ore (that which was originally mined for the nuclear fuel) within a period of several hundred to a thousand years. This significantly reduces the burden on geological repositories and the problem of addressing the remaining long-term residues can thus be done in controlled environments having timescales of centuries rather than millennia stretching beyond 10 000 years. Transmutation is one of the means being explored to address the disposal of transuranic elements. To achieve this, advanced reactor systems, appropriate fuels, separation techniques and associated fuel cycle strategies are required. This report describes the current status of fuel and material technologies for transmutation and suggests technical R and D issues that need to be resolved. It will be of particular interest to nuclear fuel and material scientists involved in the field of partitioning and transmutation (P and T), and in advanced fuel cycles in general. (author)

Neutronic and Isotopic Simulation of a Thorium-TRU's fuel Closed Cycle in a Lead Cooled ADS

International Nuclear Information System (INIS)

Garcia-Sanz, J. M.; Embid, M.; Fernandez, R.; Gonzalez, E. M.; Perez-Parra, A.

2000-01-01

The FACET group at CIEMAT is studying the properties and potentialities of several lead-cooled ADS designs for actinide and fission product transmutation. The main characteristics of these systems are the use of lead as primary coolant and moderator and fuels made by transuranics inside a thorium oxide matrix. The strategy assumed in this simulation implies that every discharge of the ADS will be reprocessed and would produce four waste streams: fission and activation products, remaining ''232 Th, produced ''233 U and remaining TRU's. The ''233 U is separated for other purposes; the remaining TRU are recovered altogether and mixed with the adequate amount of ''232 Th and fresh TRUs coming from LWR spent fuel. The simulations performed in this study have been focused primarily in the evolution of the fuel isotopic composition during and after each ADS burn-up cycle. (Author) 10 refs

Reprocessing in the thorium fuel cycle

International Nuclear Information System (INIS)

Merz, E.

1984-01-01

An overview of the authors personal view is presented on open questions in regard to still required research and development work for the thorium fuel cycle before its application in a technical-industrial scale may be tackled. For a better understanding, all stations of the back-end of the thorium fuel cycle are briefly illustrated and their special features discussed. They include storage and transportation measures, all steps of reprocessing, as well as the entire radioactive waste treatment. Knowledge gaps are, as far as they are obvious, identified and proposals put forward for additional worthwile investigations. (orig.) [de

Feasibility and desirability of employing the thorium fuel cycle for power generation - 254

International Nuclear Information System (INIS)

Sehgal, B.R.

2010-01-01

Thorium fuel cycle for nuclear power generation has been considered since the very start of the nuclear power era. In spite of a very large amount of research, experimentation, pilot scale and prototypic scale installations, the thorium fuel was not adopted for large scale power generation [1,2]. This paper reviews the developments over the years on the front and the back-end of the thorium fuel cycle and describes the pros and cons of employing the thorium fuel cycle for large generation of nuclear power. It examines the feasibility and desirability of employing the thorium fuel cycle in concert with the uranium fuel cycle for power generation. (authors)

Radiotoxicity study of a boiling water reactor core design based on a thorium-uranium fuel concept

International Nuclear Information System (INIS)

Nunez C, A.; Espinosa P, G.

2007-01-01

Full text: The innovative design of a Boiling Water Reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233 U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR. A comparison of the toxicity of the spent fuel showed that toxicity is lower in the thorium cycle than other commercial fuels as UO 2 and MOX (uranium and plutonium) in case of the one-through cycle for LWR. (Author)

Environmental impact of radioactive releases from recycle of thorium-based fuel using current containment technology

International Nuclear Information System (INIS)

Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Morse, L.E.; Meyer, H.R.; Till, J.E.; Yalcintas, M.G.

1980-01-01

The analysis of thorium mining and milling suggests that the resulting doses should be similar to those from uranium operations. An absolute comparison cannot be made at this time, however, due to differences in some assumptions utilized by the various investigators and the lack in some cases of site-specific meteorology and population data at thorium resource sites in the western United States. A distinct difference resulting from the short half-life of 220 Rn (T/sub 1/2/ = 55.6 s) in the thorium decay chain compared to that for 222 Rn (T/sub 1/2/ = 3.82 d) in uranium decay was noted for emissions following mill shutdown. This effect is to make potential releases following thorium mill shutdown of lesser consequence than in the uranium case. Thorium tailings activity would also decrease relatively rapidly due to the comparatively short half-life (T/sub 1/2 = 5.75 y) of 228 Ra. Doses due to airborne releases from thorium-uranium carbide fuel refabrication are significantly less than that due to fuel reprocessing. Tritium is the principal contributor to reprocessing plant doses while carbon-14, 131 Cs, and 232 U account for most of the remaining dose. A tenfold increase in reprocessing plant CF for tritium reduces both individual and population doses by about 60%. For refabrication operations, a near linear dependence upon dose with 232 U content of the fuel was calculated between concentrations of 10 ppM and 5000 ppM. Comparison of (Th, U)C and (U, Pu)C showed little difference in dose commitment, but the presence of 232 U in the (Th, U) fuel causes a notable increase in the refabrication plant dose over that previously calculated for (U, Pu) type fuels

Transmutation technologies to solve the problem of long-term spent nuclear fuel storage

International Nuclear Information System (INIS)

Hosnedl, P.; Valenta, V.; Blahut, O.

2000-01-01

The paper gives a brief description of the transmutation process for actinides and long-lived fission products which are present in spent nuclear fuel. Transmutation technologies can solve the problem of long-term spent nuclear fuel storage and reduce the requirements for storage time and conditions. The basic data and requirements for the detailed design of the transmutor are summarized, and the views upon how to address the fuel purification and dry reprocessing issues are discussed. The results of activities of SKODA JS are highlighted; these include, for instance, the fluoride salt-resistant material MONICR, test loops, and electrowinners. The preliminary design of the transmutor is also outlined. Brief information regarding activities in the field of transmutation technologies in the Czech Republic and worldwide is also presented. The research and design activities to be developed for the whole design of the demonstration and basic units are summarized. It is emphasized that SKODA JS can join in international cooperation without constraints. The Attachment presents a simple assessment of how the radioactivity balance can be reduced, based on the actinide and long-lived fission product transmutation half-lives, is presented in the Attachment. (author)

Definition of Technology Readiness Levels for Transmutation Fuel Development

International Nuclear Information System (INIS)

Jon Carmack; Kemal O. Pasamehmetoglu

2008-01-01

To quantitatively assess the maturity of a given technology, the Technology Readiness Level (TRL) process is used. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Transmutation fuel development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the transmutation fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Transuranic Fuel Development Campaign

Multi-faceted evaluation for nuclear fuel cycles with transmutation

International Nuclear Information System (INIS)

Nishihara, Kenji

2015-03-01

Environment impact, economy and proliferation resistance were estimated for nuclear fuel cycles involving transmutation by fast reactor and accelerator-driven system in equilibrium state. As a result, the transmutation scenario using only fast reactor was superior to the scenarios combined with accelerator-driven system in all estimation, but the differences were insignificant. (author)

Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

International Nuclear Information System (INIS)

Edwards, G.W.R.; Priest, N.D.; Richardson, R.B.

2013-01-01

The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Edwards, G.W.R.; Priest, N.D.; Richardson, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

Recycling and transmutation of spent fuel as a sustainable option for the nuclear energy development

International Nuclear Information System (INIS)

Maiorino, Jose R.; Moreira, Joao M.L.

2013-01-01

The objective of this paper is to discuss the option of recycling and transmutation of radioactive waste against Once-through Fuel Cycle (OTC) based on uranium feed under the perspective of sustainability. We use a qualitative analysis to compare OTC with closed fuel cycles based on studies already performed such as the Red Impact Project and the comparative study on accelerator driven systems and fast reactors for advanced fuel cycles performed by the Nuclear Energy Agency. The results show that recycling and transmutation fuel cycles are more attractive than the OTC from the point of view of sustainability. The main conclusion is that the decision about the construction of a deep geological repository for spent fuel disposal must be reevaluated. (author)

Radiological implications of plutonium recycle and the use of thorium fuels in power reactor operations

Energy Technology Data Exchange (ETDEWEB)

Macdonald, H. F.

1976-01-15

As economically attractive sources of natural uranium are gradually depleted attention will turn to recycling plutonium or to the use of thorium fuels. The radiological implications of these fuel cycles in terms of fuel handling and radioactive waste disposal are investigated in relation to a conventional /sup 235/U enriched oxide fuel. It is suggested that a comparative study of this nature may be an important aspect of the overall optimization of future fuel cycle strategies. It is shown that the use of thorium based fuels has distinct advantages in terms of neutron dose rates from irradiated fuels and long term proportional to decay heating commitment compared with conventional uranium/plutonium fuels. However, this introduces a ..gamma.. dose rate problem in the fabrication and handling of unirradiated /sup 233/U fuels. For both plutonium and thorium fuels these radiological problems increase during storage of the fuel prior to reactor irradiation. The novel health physics problems which arise in the handling and processing of thorium fuels are reviewed in an appendix.

Integral benchmarks with reference to thorium fuel cycle

International Nuclear Information System (INIS)

Ganesan, S.

2003-01-01

This is a power point presentation about the Indian participation in the CRP 'Evaluated Data for the Thorium-Uranium fuel cycle'. The plans and scope of the Indian participation are to provide selected integral experimental benchmarks for nuclear data validation, including Indian Thorium burn up benchmarks, post-irradiation examination studies, comparison of basic evaluated data files and analysis of selected benchmarks for Th-U fuel cycle

Transmutation of DUPIC spent fuel in the hyper system

International Nuclear Information System (INIS)

Kim, Y.H.; Song, T.Y.

2005-01-01

In this paper, the transmutation of TRUs of the DUPIC (Direct Use of Spent PWR Fuel in CANDU) spent fuel has been studied with the HYPER system, which is an LBE-cooled ADS. The DUPIC concept is a synergistic combination of PWRs and CANDUs, in which PWR spent fuels are directly re-utilized in CANDU reactors after a very simple re-fabrication process. In the DUPIC-HYPER fuel cycle, TRUs are recovered by using a pyro-technology and they are incinerated in a metallic fuel form of U-TRU-Zr. The objective of this study is to investigate the TRU transmutation potential of the HYPER core for the DUPIC-HYPER fuel cycle. All the previously-developed HYPER core design concepts were retained except that fuel is composed of TRU from the DUPIC spent fuel. In order to reduce the burnup reactivity swing, a B 4 C burnable absorber is used. The HYPER core characteristics have been analyzed with the REBUS-3/DIF3D code system. (authors)

Effects of an LMR-based partitioning-transmutation system on US nuclear fuel cycle health risk

International Nuclear Information System (INIS)

Michaels, G.E.; Reich, W.J.

1992-01-01

Health risks for the current US nuclear fuel cycle and for an illustrative partitioning and transmutation (P-T) fuel cycle based on Liquid Metal Reactor (LMR) technology are calculated and compared. Health risks are calculated for all non-reactor fuel cycle steps, including reprocessing, transportation, and high-level waste (HLW) disposal. Uranium mining and milling health risks have been updated to include recent occupational injury and death statistics, and the radiological health risk to the general public posed by the uranium mining overburden. In addition, the radiological health risks for transportation have been updated to include latent cancer fatalities associated with both normal transport and accidents. Given the assumptions of the study, it is shown that the deployment of an LMR-based P-T system is expected to reduce overall nuclear fuel cycle health risk

Transmutation of radioactive waste: Effect on the nuclear fuel cycle

International Nuclear Information System (INIS)

Rasmussen, N.C.; Pigford, T.H.

1997-01-01

A committee of the National Research Council reviewed three concepts for transmuting radionuclides recovered from the chemical reprocessing of commercial light-water-reactor (LWR) fuel: LWR transmutation reactors fueled with recycled actinides, advanced liquid-metal reactors (ALMRs), and accelerator-driven subcritical reactors for transmutation of waste (ATW). The concepts were evaluated in terms of: (1) the extent to which waste disposal would benefit from transmutation, (2) time required to reduce the total inventory of radionuclides in the waste and fuel cycle, (3) the complexity of the overall transmutation system, (4) the extent of new development required, and (5) institutional and economic problems of operating such systems. Transmutation could affect geologic disposal of waste by reducing the inventory of transuranics (TRUs), fission products, and other radionuclides in the waste. Reducing the inventory of transuranics does not necessarily affect radiation doses to people who use contaminated ground water if the dissolution rate of transuranics in waste is controlled by elemental solubilities. However, reducing inventories of Am and Pu would decrease potential hazards from human intrusion. The likelihood for underground nuclear criticality would also be reduced. The long-lived fission products Tc-99, I-129, Cs-135 and others typically contribute most to the long-term radiation doses to future populations who use contaminated water from the repository. Their transmutation requires thermal or epithermal neutrons, readily available in LWR and ATW transmutors. ALMR and LWR transmutors would require several hundred years to reduce the total transuranic inventory by even a factor of 10 at constant electric power, and thousands of years for a hundred-fold reduction. For the same electrical power, the ATW could reduce total transuranic inventory about tenfold more rapidly, because of its very high thermal-neutron flux. However, extremely low process losses would be

A preliminary assessment of thorium as a fuel for thermal reactors

International Nuclear Information System (INIS)

Duret, M.F.; Halsall, M.J.

1965-08-01

This report presents a preliminary survey of the contributing factors to total fuel cost, for a variety of potentially economic fuel cycles. U 238 and Th 232 are considered as fertile materials with enrichments of U 235 , U 233 and plutonium. The total fuel cost comprises: (a) a fuel supply cost which is derived from a manufacturing cost and the fuel burnup, (b) as spent fuel credit, and (c) inventory charges. The interrelation of these three factors and their dependence on reactor flux, fuel burnup, the heat rating of the fuel and the interest rate are considered for each combination of fertile and fissile material. Results show inventory charges, spent fuel credit (hence processing costs) and fabrication costs have an important effect on the fuelling cost. In addition, using highly rated thorium-based fuels will affect reactor capital costs. Further, more detailed studies of reactor design, fuel design and fuel management schemes are now required to verify the attraction of using thorium as a reactor fuel. (author)

The physics design of accelerator-driven transmutation systems

International Nuclear Information System (INIS)

Venneri, F.

1995-01-01

Nuclear systems under study in the Los Alamos Accelerator-Driven Transmutation Technology program (ADTT) will allow the destruction of nuclear spent fuel and weapons-return plutonium, as well as the production of nuclear energy from the thorium cycle, without a long-lived radioactive waste stream. The subcritical systems proposed represent a radical departure from traditional nuclear concepts (reactors), yet the actual implementation of ADTT systems is based on modest extrapolations of existing technology. These systems strive to keep the best that the nuclear technology has developed over the years, within a sensible conservative design envelope and eventually manage to offer a safer, less expensive and more environmentally sound approach to nuclear power

The physics design of accelerator-driven transmutation systems

Energy Technology Data Exchange (ETDEWEB)

Venneri, F. [Los Alamos National Laboratory, NM (United States)

1995-10-01

Nuclear systems under study in the Los Alamos Accelerator-Driven Transmutation Technology program (ADTT) will allow the destruction of nuclear spent fuel and weapons-return plutonium, as well as the production of nuclear energy from the thorium cycle, without a long-lived radioactive waste stream. The subcritical systems proposed represent a radical departure from traditional nuclear concepts (reactors), yet the actual implementation of ADTT systems is based on modest extrapolations of existing technology. These systems strive to keep the best that the nuclear technology has developed over the years, within a sensible conservative design envelope and eventually manage to offer a safe, less expensive and more environmentally sound approach to nuclear power.

Uranium production in thorium/denatured uranium fueled PWRs

International Nuclear Information System (INIS)

Arthur, W.B.

1977-01-01

Uranium-232 buildup in a thorium/denatured uranium fueled pressurized water reactor, PWR(Th), was studied using a modified version of the spectrum-dependent zero dimensional depletion code, LEOPARD. The generic Combustion Engineering System 80 reactor design was selected as the reactor model for the calculations. Reactors fueled with either enriched natural uranium and self-generated recycled uranium or uranium from a thorium breeder and self-generated recycled uranium were considered. For enriched natural uranium, concentrations of 232 U varied from about 135 ppM ( 232 U/U weight basis) in the zeroth generation to about 260 ppM ( 232 U/U weight basis) at the end of the fifth generation. For the case in which thorium breeder fuel (with its relatively high 232 U concentration) was used as reactor makeup fuel, concentrations of 232 U varied from 441 ppM ( 232 U/U weight basis) at discharge from the first generation to about 512 ppM ( 232 U/U weight basis) at the end of the fifth generation. Concentrations in freshly fabricated fuel for this later case were 20 to 35% higher than the discharge concentration. These concentrations are low when compared to those of other thorium fueled reactor types (HTGR and MSBR) because of the relatively high 238 U concentration added to the fuel as a denaturant. Excellent agreement was found between calculated and existing experimental values. Nevertheless, caution is urged in the use of these values because experimental results are very limited, and the relevant nuclear data, especially for 231 Pa and 232 U, are not of high quality

Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

Science.gov (United States)

Uyttenhove, W.; Sobolev, V.; Maschek, W.

2011-09-01

A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2-x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW -1 h -1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

International Nuclear Information System (INIS)

Uyttenhove, W.; Sobolev, V.; Maschek, W.

2011-01-01

A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2-x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW -1 h -1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

Study of deep subcritical electronuclear systems and feasibility of their application for energy production and radioactive waste transmutation

International Nuclear Information System (INIS)

Adam, J.; Baldin, A.; Vladimirova, N.

2010-01-01

Physical substantiation for investigation of new schemes of electronuclear power production and transmutation of long-lived radioactive wastes based on nuclear relativistic technologies is presented. 'E and T RAW' ('Energy and Transmutation of Radioactive Wastes') is aimed at complex study of interaction of relativistic beams of the Nuclotron-M with energies up to 10 GeV in quasi-infinite targets. Feasibility of application of natural/depleted uranium or thorium without the use of uranium-235, as well as utilization of spent fuel elements of atomic power plants is demonstrated based on analysis of results of known experiments, numerical, and theoretical works. The 'E and T RAW' project will provide fundamentally new data and numerical methods necessary for design of demonstration experimental-industrial setups based on the proposed scheme

Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

Directory of Open Access Journals (Sweden)

Gholamzadeh Zohreh

2014-12-01

Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

Introduction of Thorium in the Nuclear Fuel Cycle. Short- to long-term considerations

International Nuclear Information System (INIS)

Allibert, M.; Merle-Lucotte, E.; Ghetta, V.; Ault, T.; Krahn, S.; Wymer, R.; Croff, A.; Baron, P.; Chauvin, N.; Eschbach, R.; Rimpault, G.; Serp, J.; Bergeron, A.; Bromley, B.; Floyd, M.; Hamilton, H.; Hyland, B.; Wojtaszek, D.; McDonald, M.; Collins, E.; Cornet, S.; Michel-Sendis, F.; ); Feinberg, O.; Ignatiev, V.; Hesketh, K.; Kelly, J.F.; Porsch, D.; Vidal, J.; Taiwo, T.; Uhlir, J.; Van Den Durpel, L.; Van Den Eynde, G.; Vitanza, C.; Butler, Gregg; Cornet, Stephanie; Dujardin, Thierry; Greneche, Dominique; Nordborg, Claes; Rimpault, Gerald; Van Den Durpel, Luc; Michel-Sendis, Franco

2015-01-01

Since the beginning of the nuclear era, significant scientific attention has been given to thorium's potential as a nuclear fuel. Although the thorium fuel cycle has never been fully developed, the opportunities and challenges that might arise from the use of thorium in the nuclear fuel cycle are still being studied in many countries and in the context of diverse international programmes around the world. This report provides a scientific assessment of thorium's potential role in nuclear energy both in the short to longer term, addressing diverse options, potential drivers and current impediments to be considered if thorium fuel cycles are to be pursued. (authors)

Analysis of advanced european nuclear fuel cycle scenarios including transmutation and economical estimates

International Nuclear Information System (INIS)

Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

2013-01-01

In this work the transition from the existing Light Water Reactors (LWR) to the advanced reactors is analyzed, including Generation III+ reactors in a European framework. Four European fuel cycle scenarios involving transmutation options have been addressed. The first scenario (i.e., reference) is the current fleet using LWR technology and open fuel cycle. The second scenario assumes a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel. The third scenario is a modification of the second one introducing Minor Actinide (MA) transmutation in a fraction of the FR fleet. Finally, in the fourth scenario, the LWR fleet is replaced using FR with MOX fuel as well as Accelerator Driven Systems (ADS) for MA transmutation. All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for a period of 200 years looking for equilibrium mass flows. The simulations were made using the TR-EVOL code, a tool for fuel cycle studies developed by CIEMAT. The results reveal that all scenarios are feasible according to nuclear resources demand (U and Pu). Concerning to no transmutation cases, the second scenario reduces considerably the Pu inventory in repositories compared to the reference scenario, although the MA inventory increases. The transmutation scenarios show that elimination of the LWR MA legacy requires on one hand a maximum of 33% fraction (i.e., a peak value of 26 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation). On the other hand a maximum number of ADS plants accounting for 5% of electricity generation are predicted in the fourth scenario (i.e., 35 ADS units). Regarding the economic analysis, the estimations show an increase of LCOE (Levelized cost of electricity) - averaged over the whole period - with respect to the reference scenario of 21% and 29% for FR and FR with transmutation scenarios respectively, and 34% for the fourth scenario. (authors)

Actinide partitioning-transmutation program final report. III. Transmutation studies

International Nuclear Information System (INIS)

Wachter, J.W.; Croff, A.G.

1980-07-01

Transmutation of the long-lived nuclides contained in fuel cycle wastes has been suggested as a means of reducing the long-term toxicity of the wastes. A comprehensive program to evaluate the feasibility and incentives for recovering the actinides from wastes (partitioning) and transmuting them to short-lived or stable nuclides has been in progress for 3 years under the direction of Oak Ridge National Laboratory (ORNL). This report constitutes the final assessment of transmutation in support of this program. Included are (1) a summary of recent transmutation literature, (2) a generic evaluation of actinide transmutation in thermal, fast, and other transmutation devices, (3) a preliminary evaluation of 99 Tc and 129 I transmutation, and (4) a characterization of a pressurized-water-reactor fuel cycle with and without provisions for actinide recovery and transmutation for use in other parts of the ORNL program. The principal conclusion of the report is that actinide transmutation is feasible in both thermal and fast reactors, subject to demonstrating satisfactory fuel performance, with relatively little impact on the reactor. It would also appear that additional transmutation studies are unwarranted until a firm decision to proceed with actinide transmutation has been made by the responsible authorities

Critical review of analytical techniques for safeguarding the thorium-uranium fuel cycle

International Nuclear Information System (INIS)

Hakkila, E.A.

1978-10-01

Conventional analytical methods applicable to the determination of thorium, uranium, and plutonium in feed, product, and waste streams from reprocessing thorium-based nuclear reactor fuels are reviewed. Separations methods of interest for these analyses are discussed. Recommendations concerning the applicability of various techniques to reprocessing samples are included. 15 tables, 218 references

A review of the potential for actinide redistribution in CANDU thorium cycle fuels

International Nuclear Information System (INIS)

Cameron, D.J.

1978-02-01

Actinide redistribution resulting from large radial temperature gradients is an accepted feature of the technology of fast reactor (U,Pu)O 2 fuels. A thorium cycle in CANDU reactors would require the use of oxide fuels with two or more components, raising the question of actinide redistribution in these fuels. The mechanisms proposed to explain redistribution in (U,Pu)O 2 fuels are reviewed, and their relevance to fuels based on ThO 2 is discussed. The fuel primarily considered is (Th,U)O 2 but some reference is made to (Th,Pu)O 2 . At this early stage of thorium fuel cycle technology, it is not possible to consider quantitatively the question of redistribution in specific fuels. However, some guidelines can be presented to indicate to fuel engineers conditions which might result in significant redistribution. It is concluded that redistribution is probably of little concern in high density, CANDU, thorium cycle fuel whose centre temperature is limited to 2350 K. If this centre temperature is exceeded, or if the fuel contains substantial inter-connected porosity, the potential for redistribution is significant and warrants more serious study. (author)

Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

Directory of Open Access Journals (Sweden)

Igor Shamanin

2015-01-01

Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel

International Nuclear Information System (INIS)

Francois, J.L.; Nunez C, A.

2003-01-01

In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

Comparison for thorium fuel cycle facilities of two different capacities for implementation of safeguards

International Nuclear Information System (INIS)

Gangotra, Suresh; Grover, R.B.; Ramakumar, K.L.

2013-01-01

Highlights: • Facilities for implementation of safeguards for thorium fuel cycle have been compared. • Two concepts have been compared. • In one concept, the facilities are designed in hub and spoke concept. • In second concept the facilities are designed as self-contained concept. • The comparison is done on a number of factors, which affect safeguardability and proliferation resistance. -- Abstract: Thorium based nuclear fuel cycle has many attractive features, its inherent proliferation resistance being one of them. This is due to the presence of high energy gamma emitting daughter products of U 232 associated with U 233 . This high energy gamma radiation also poses challenges in nuclear material accounting. A typical thorium fuel cycle facility has a number of plants including a fuel fabrication plant for initial and equilibrium core, a reprocessed U 233 fuel fabrication plant, a reprocessing plant, a fuel assembly/disassembly plant and associated waste handling and management plants. A thorium fuel cycle facility can be set up to serve reactors at a site. Alternatively, one can follow a hub and spoke approach with a large thorium fuel cycle facility acting as a hub, catering to the requirements of reactors at several sites as spokes. These two concepts have their respective merits and shortcomings in terms of engineering and economics. The present paper is aimed at comparing the merits and challenges for implementation of safeguards on the two concepts viz. a large fuel cycle hub catering to reactors at several sites versus a small fuel cycle facility dedicated to reactors at a single site

Comparison for thorium fuel cycle facilities of two different capacities for implementation of safeguards

Energy Technology Data Exchange (ETDEWEB)

Gangotra, Suresh, E-mail: sgangotra@yahoo.co.in; Grover, R.B.; Ramakumar, K.L.

2013-09-15

Highlights: • Facilities for implementation of safeguards for thorium fuel cycle have been compared. • Two concepts have been compared. • In one concept, the facilities are designed in hub and spoke concept. • In second concept the facilities are designed as self-contained concept. • The comparison is done on a number of factors, which affect safeguardability and proliferation resistance. -- Abstract: Thorium based nuclear fuel cycle has many attractive features, its inherent proliferation resistance being one of them. This is due to the presence of high energy gamma emitting daughter products of U{sup 232} associated with U{sup 233}. This high energy gamma radiation also poses challenges in nuclear material accounting. A typical thorium fuel cycle facility has a number of plants including a fuel fabrication plant for initial and equilibrium core, a reprocessed U{sup 233} fuel fabrication plant, a reprocessing plant, a fuel assembly/disassembly plant and associated waste handling and management plants. A thorium fuel cycle facility can be set up to serve reactors at a site. Alternatively, one can follow a hub and spoke approach with a large thorium fuel cycle facility acting as a hub, catering to the requirements of reactors at several sites as spokes. These two concepts have their respective merits and shortcomings in terms of engineering and economics. The present paper is aimed at comparing the merits and challenges for implementation of safeguards on the two concepts viz. a large fuel cycle hub catering to reactors at several sites versus a small fuel cycle facility dedicated to reactors at a single site.

Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

International Nuclear Information System (INIS)

Brown, N.R.; Powers, J.J.; Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G.; Todosow, M.; Worrall, A.; Gehin, J.C.; Kim, T.K.; Taiwo, T.A.

2015-01-01

Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems

Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

Energy Technology Data Exchange (ETDEWEB)

Brown, N.R., E-mail: nbrown@bnl.gov [Brookhaven National Laboratory, Upton, NY (United States); Powers, J.J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G. [Argonne National Laboratory, Argonne, IL (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States); Worrall, A.; Gehin, J.C. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Kim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States)

2015-08-15

Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10{sup 5} eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.

Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

International Nuclear Information System (INIS)

Powers, J.J.; Worrall, A.; Gehin, J.C.; Harrison, T.J.; Sunny, E.E.

2013-01-01

Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

Performance of Energy Multiplier Module (EM2) with long-burn thorium fuel cycle

International Nuclear Information System (INIS)

Choi, Hangbok; Schleicher, Robert; Gupta, Puja

2015-01-01

Energy Multiplier Module (EM 2 ) is a helium-cooled fast reactor being developed by General Atomics for the 21 st century grid. It is designed as a modular plant with a net electric output of 265 MWe with an evaporative heat sink and 240 MWe with an air-cooled heat sink. EM 2 core performance is examined for the baseline loading of low-enriched uranium (LEU) as fissile material with depleted uranium (DU) as fertile material and compared to the alternate LEU with thorium loading. The latter has two options: a heterogeneous loading of thorium fuel in the place of DU that produces a longer fuel cycle, and homogeneously mixed thorium-uranium fuel loading. Compared to the baseline LEU/DU core, the cycle length of both thorium options is reduced due to higher neutron absorptions by thorium. However, for both, heterogeneous and homogenous thorium loading options, the fuel cycle length is over 24 years without refueling or reshuffling of fuel assemblies. The physics properties of the EM 2 thorium core are close to those of the baseline core which constitute low excess reactivity, negative fuel temperature coefficient, and very small void reactivity. However, unlike the case of baseline EM 2 , the homogeneous thorium fuel loading provides additional advantage in reducing the power peaking of the core, which in turn reduces the cladding material neutron damage rate by 23%. It is interpreted that the relatively slow 233 U buildup as compared to 239 Pu for baseline core retards reactivity increase without the need for a complicated fuel loading pattern of the heterogeneous fuel loading, while maintaining the peak power density low. Therefore both the heterogeneous and homogeneous thorium loading options will be feasible in the EM 2

Safety analysis of thorium-based fuels in the General Electric Standard BWR

International Nuclear Information System (INIS)

Colby, M.J.; Townsend, D.B.; Kunz, C.L.

1980-06-01

A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

A competitive thorium fuel cycle for pressurized water reactors of current technology

International Nuclear Information System (INIS)

Galperin, A.; Radkowsky, A.; Todosow, M.

2002-01-01

Two important issues may influence the development and public acceptance of the nuclear power worldwide: a reduction of proliferation potential and spent fuel disposal requirements of the nuclear fuel cycle. Both problems may be addressed effectively by replacement of uranium by thorium fertile part of the fuel. A practical and competitive fuel design to satisfy the described design objectives and constraints may be achieved by seed-blanket core, proposed by A. Radkowsky and implemented in Shippingport reactors. The main idea is to separate spatially the uranium part of the core (seed) from the thorium part of the core (blanket), and thus allow two separate fuel management routes for uranium and thorium parts of the fuel. The uranium part (seed) is optimized to supply neutrons to the subcritical thorium blanket. The blanket is designed to generate and bum insitu 233 U. (author)

Evaluation of subcritical hybrid systems loaded with reprocessed fuel

International Nuclear Information System (INIS)

Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L.

2015-01-01

Highlights: • Accelerator driven systems (ADS) and fusion–fission systems are investigated for transmutation and fuel regeneration. • The calculations were performed using Monteburns code. • The results indicate the most suitable system for achieve transmutation. - Abstract: Two subcritical hybrid systems containing spent fuel reprocessed by Ganex technique and spiked with thorium were submitted to neutron irradiation of two different sources: ADS (Accelerator-driven subcritical) and Fusion. The aim is to investigate the nuclear fuel evolution using reprocessed fuel and the neutronic parameters under neutron irradiation. The source multiplication factor and fuel depletion for both systems were analysed during 10 years. The simulations were performed using MONTEBURNS code (MCNP/ORIGEN). The results indicate the main differences when irradiating the fuel with different neutron sources as well as the most suitable system for achieving transmutation

Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economical estimates

International Nuclear Information System (INIS)

Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

2013-01-01

Four European fuel cycle scenarios involving transmutation options have been addressed from a point of view of resources utilization and economics. Scenarios include the current fleet using Light Water Reactor (LWR) technology and open fuel cycle (as a reference scenario), a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel and two fuel cycles with Minor Actinide (MA) transmutation in a fraction of the FR fleet or in dedicated Accelerator Driven Systems (ADS).Results reveal that all scenarios are feasible according to nuclear resources demand. Regarding the economic analysis, the estimations show an increase of LCOE - averaged over the whole period - with respect to the reference scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios respectively.

Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economical estimates

Energy Technology Data Exchange (ETDEWEB)

Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

2013-07-01

Four European fuel cycle scenarios involving transmutation options have been addressed from a point of view of resources utilization and economics. Scenarios include the current fleet using Light Water Reactor (LWR) technology and open fuel cycle (as a reference scenario), a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel and two fuel cycles with Minor Actinide (MA) transmutation in a fraction of the FR fleet or in dedicated Accelerator Driven Systems (ADS).Results reveal that all scenarios are feasible according to nuclear resources demand. Regarding the economic analysis, the estimations show an increase of LCOE - averaged over the whole period - with respect to the reference scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios respectively.

Transmutation and accelerator driven systems

International Nuclear Information System (INIS)

Shapira, J.P.

2001-01-01

Full text: Today, countries who are presently involved in nuclear energy are facing many challenges to maintain this option open for the next few decades. Among them, management of nuclear wastes produced in nuclear reactors and in fuel cycle operations has become a very strong environmental issue among the public. In most countries with sizeable commercial nuclear programs, deep geological disposal of ultimate highly active and long-lived nuclear wastes is considered as the reference long-term management scheme. But, many questions arise on the possibility to demonstrate that such wastes can be dealt in such a way as to protect the future generations and the environment. The characteristics of nuclear wastes, the various back end policies concerning spent fuels and the nuclear wastes long-term management options will be first described. Then recent proposals, based on transmutation, especially those using accelerator driven systems (ADS) and/or thorium will be presented. Finally, the possibility for the nuclear physics community to play a part in alleviating the nuclear wastes burden will be pointed out. (author)

Electrochemical reduction of CerMet fuels for transmutation using surrogate CeO2-Mo pellets

Science.gov (United States)

Claux, B.; Souček, P.; Malmbeck, R.; Rodrigues, A.; Glatz, J.-P.

2017-08-01

One of the concepts chosen for the transmutation of minor actinides in Accelerator Driven Systems or fast reactors proposes the use of fuels and targets containing minor actinides oxides embedded in an inert matrix either composed of molybdenum metal (CerMet fuel) or of ceramic magnesium oxide (CerCer fuel). Since the sufficient transmutation cannot be achieved in a single step, it requires multi-recycling of the fuel including recovery of the not transmuted minor actinides. In the present work, a pyrochemical process for treatment of Mo metal inert matrix based CerMet fuels is studied, particularly the electroreduction in molten chloride salt as a head-end step required prior the main separation process. At the initial stage, different inactive pellets simulating the fuel containing CeO2 as minor actinide surrogates were examined. The main studied parameters of the process efficiency were the porosity and composition of the pellets and the process parameters as current density and passed charge. The results indicated the feasibility of the process, gave insight into its limiting parameters and defined the parameters for the future experiment on minor actinide containing material.

Thorium fuel performance assessment in HTRs

Energy Technology Data Exchange (ETDEWEB)

Allelein, H.-J. [Forschungszentrum Jülich, D-52425 Jülich (Germany); RWTH Aachen, D-52072 Aachen (Germany); Kania, M.J.; Nabielek, H. [Forschungszentrum Jülich, D-52425 Jülich (Germany); Verfondern, K., E-mail: k.verfondern@fz-juelich.de [Forschungszentrum Jülich, D-52425 Jülich (Germany)

2014-05-01

Thorium as a nuclear fuel is receiving renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTR development employed thorium together with high-enriched uranium. After 1980, most HTR fuel systems switched to low-enriched uranium. After completing fuel development for AVR and THTR with BISO coated particles, the German program expanded efforts on a new program utilizing thorium and high-enriched uranium TRISO coated particles for advanced HTR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of LTI inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with HTI-BISO coatings. The improved performance of the HEU (Th,U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 °C in normal operations and 1600 °C in accidents, with burnups up to 13% FIMA and fast fluences to 8 × 10{sup 25} m{sup −2} (E > 16 fJ), the results exceed the design limits on manufacturing and operational requirements for the German HTR Modul concept, which were: <6.5 × 10{sup −5} for manufacturing; <2 × 10{sup −4} for normal operating conditions; and <5 × 10{sup −4} for accident conditions. These

Transmutation of nuclear waste. Status report RAS programme 1994: Recycling and transmutation of actinides and fission products

Energy Technology Data Exchange (ETDEWEB)

Cordfunke, E H.P.; Gruppelaar, H; Franken, W M.P.

1995-07-01

This report describes the status and progress of the Dutch RAS programme on `Recycling and Transmutation of Actinides and Fission Products` over the year 1994, which is the first year of the second 4-year programme. This programme is outlined and a short progress report is given over 1994, including a listing of 23 reports and publications over the year 1994. Highlights of 1994 were: The completion of long-lived fission-product transmutation studies, the initiation of small-scale demonstration experiments in the HFR on Tc and I, the issue of reports on the potential of the ALMR (Advanced Liquid Metal Reactor) for transmutation adn the participation and international cooperation on irradiation experiments with actinides in inert matrices. The remaining chapters contain more extended contributions on recent developments and selected topics, under the headings: Benefits and risks of partitioning and transmutation, Perspective of chemical partitioning, Inert matrices, Evolutionary options (MOX), Perspective of heavy water reactors, Perspective of fast burners, Perspective of accelerator-based systems, Thorium cycle, Fission-product transmutation, End scenarios, and Executive summary and recommendations. (orig.).

Transmutation of nuclear waste. Status report RAS programme 1994: Recycling and transmutation of actinides and fission products

International Nuclear Information System (INIS)

Cordfunke, E.H.P.; Gruppelaar, H.; Franken, W.M.P.

1995-07-01

This report describes the status and progress of the Dutch RAS programme on 'Recycling and Transmutation of Actinides and Fission Products' over the year 1994, which is the first year of the second 4-year programme. This programme is outlined and a short progress report is given over 1994, including a listing of 23 reports and publications over the year 1994. Highlights of 1994 were: The completion of long-lived fission-product transmutation studies, the initiation of small-scale demonstration experiments in the HFR on Tc and I, the issue of reports on the potential of the ALMR (Advanced Liquid Metal Reactor) for transmutation adn the participation and international cooperation on irradiation experiments with actinides in inert matrices. The remaining chapters contain more extended contributions on recent developments and selected topics, under the headings: Benefits and risks of partitioning and transmutation, Perspective of chemical partitioning, Inert matrices, Evolutionary options (MOX), Perspective of heavy water reactors, Perspective of fast burners, Perspective of accelerator-based systems, Thorium cycle, Fission-product transmutation, End scenarios, and Executive summary and recommendations. (orig.)

TRU transmutation using ThO2-UO2 and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

International Nuclear Information System (INIS)

Bae, Gonghoon; Hong, Sergi

2012-01-01

The objective of this work is to design new LWR fuel assemblies which are able to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO 2 -UO 2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO 2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO 2 pins. Additionally, the use of thorium contributes to the extension of the fuel cycle length. All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) code. The results show that the new concept of fuel assembly has the TRU destruction rates of ∼40% and ∼25% per 1200 EFPD (Effective Full Power Day) over the TRU FCM pins and the overall fuel assembly, respectively, without degradation of FTC and MTC

Transmutation Technology Development

Energy Technology Data Exchange (ETDEWEB)

Song, T. Y.; Park, W. S.; Kim, Y. H. (and others)

2007-06-15

The spent fuel coming from the PWR is one of the most difficult problems to be solved for the continuous use of nuclear power. It takes a few million years to be safe under the ground. Therefore, it is not easy to take care of the spent fuel for such a long time. Transmutation technology is the key technology which can solve the spent fuel problem basically. Transmutation is to transmute long-lived radioactive nuclides in the spent fuel into short-lived or stable nuclide through nuclear reactions. The long-lived radioactive nuclides can be TRU and fission products such as Tc-99 and I-129. Although the transmutation technology does not make the underground disposal totally unnecessary, the period to take care of the spent fuel can be reduced to the order of a few hundred years. In addition to the environmental benefit, transmutation can be considered to recycle the energy in the spent fuel since the transmutation is performed through nuclear fission reaction of the TRU in the spent fuel. Therefore, transmutation technology is worth being developed in economical aspect. The results of this work can be a basis for the next stage research. The objective of the third stage research was to complete the core conceptual design and verification of the key technologies. The final results will contribute to the establishment of Korean back end fuel cycle policy by providing technical guidelines.

Assessment of the thorium fuel cycle in power reactors

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.; Allen, E.J.

1977-01-01

A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

International Nuclear Information System (INIS)

Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

2010-01-01

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

Transmutation of long-lived nuclides in the fuel cycle of Brest-type reactors

International Nuclear Information System (INIS)

Lopatkin, A.V.; Orlov, V.V.; Filin, A.I.

2001-01-01

Transmutation of long-lived nuclides produced as a result of nuclear generation, should be set up proceeding from the principle of reasonable sufficiency, expressed as radiation equivalence between the radwaste sent to disposal and source natural uranium. In this case, introduction of fast reactors of new generation (such as BREST or other reactors based on similar philosophy) will resolve transmutation problems even with the thermal-to-fast reactor capacity ratio of 2:1. The authors of the 'Strategy of nuclear power development in Russia' foresee, and substantiate their prediction, that fast reactors of the new generation will account for no less than 2/3 of nuclear capacity in future large-scale nuclear power sector. Fast reactors will be the basis of a transmutation fuel cycle, which will remove the need of creating additional transmutation facilities. (author)

Transmutation Fuels Campaign FY-09 Accomplishments Report

Energy Technology Data Exchange (ETDEWEB)

Lori Braase

2009-09-01

This report summarizes the fiscal year 2009 (FY-08) accomplishments for the Transmutation Fuels Campaign (TFC). The emphasis is on the accomplishments and relevance of the work. Detailed description of the methods used to achieve the highlighted results and the associated support tasks are not included in this report.

Comparison of the radiological impacts of thorium and uranium nuclear fuel cycles

International Nuclear Information System (INIS)

Meyer, H.R.; Witherspoon, J.P.; McBride, J.P.; Frederick, E.J.

1982-03-01

This report compares the radiological impacts of a fuel cycle in which only uranium is recycled, as presented in the Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors (GESMO), with those of the light-water breeder reactor (LWBR) thorium/uranium fuel cycle in the Final Environmental Statement, Light Water Breeder Reactor Program. The significant offsite radiological impacts from routine operation of the fuel cycles result from the mining and milling of thorium and uranium ores, reprocessing spent fuel, and reactor operations. The major difference between the impacts from the two fuel cycles is the larger dose commitments associated with current uranium mining and milling operations as compared to thorium mining and milling. Estimated dose commitments from the reprocessing of either fuel type are small and show only moderate variations for specific doses. No significant differences in environmental radiological impact are anticipated for reactors using either of the fuel cycles. Radiological impacts associated with routine releases from the operation of either the thorium or uranium fuel cycles can be held to acceptably low levels by existing regulations

Nuclear data covariances and sensitivity analysis, validation of a methodology based on the perturbation theory; application to an innovative concept: the molten thorium salt fueled reactor

International Nuclear Information System (INIS)

Bidaud, A.

2005-10-01

Neutron transport simulation of nuclear reactors is based on the knowledge of the neutron-nucleus interaction (cross-sections, fission neutron yields and spectra...) for the dozens of nuclei present in the core over a very large energy range (fractions of eV to several MeV). To obtain the goal of the sustainable development of nuclear power, future reactors must have new and more strict constraints to their design: optimization of ore materials will necessitate breeding (generation of fissile material from fertile material), and waste management will require transmutation. Innovative reactors that could achieve such objectives - generation IV or ADS (accelerator driven system) - are loaded with new fuels (thorium, heavy actinides) and function with neutron spectra for which nuclear data do not benefit from 50 years of industrial experience, and thus present particular challenges. After validation on an experimental reactor using an international benchmark, we take classical reactor physics tools along with available nuclear data uncertainties to calculate the sensitivities and uncertainties of the criticality and temperature coefficient of a thorium molten salt reactor. In addition, a study based on the important reaction rates for the calculation of cycle's equilibrium allows us to estimate the efficiency of different reprocessing strategies and the contribution of these reaction rates on the uncertainty of the breeding and then on the uncertainty of the size of the reprocessing plant. Finally, we use this work to propose an improvement of the high priority experimental request list. (author)

Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

International Nuclear Information System (INIS)

Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

1978-10-01

This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured 235 U) startup fuels, first-recycle and equilibrium (denatured 233 U) thorium--uranium LWR fuels, and to recover the plutonium generated in the 238 U denaturant as well. 12 figures, 3 tables

Some recent contributions of basic nuclear science to nuclear waste transmutation

International Nuclear Information System (INIS)

Schapira, J.P.

2001-01-01

Nuclear waste transmutation aims at alleviating some long-term risks associated with actinides and with some long-lived fission products. Proposals of using accelerator driven system (ADS) to efficiently burn actinides in uranium free fuels have revitalized some basic researches in the field of nuclear and reactor physics. This is the case for high intensity accelerator in the ADS context and for the neutron source which relies to a large extent on basic nuclear physics related to spallation. There is also an experimental program called MUSE at Cadarache to study the sub-critical reactor physics with regard to its neutronics. A second area where basic research is involved is the measurement of new or more reliable neutron cross sections specific to transmutation and also to the thorium fuel cycle considered as a long-term option for ''clean'' energy production with reduced actinide production. This second area will possibly be covered by a new facility called n-TOF developed at CERN. (author)

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

Indian Academy of Sciences (India)

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

The thorium fuel cycle in water-moderated reactor systems

International Nuclear Information System (INIS)

Critoph, E.

1977-05-01

Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An 'inventory' of uranium of between 1 and 2 Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium) is some two decades

Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

International Nuclear Information System (INIS)

Zhou, Shengcheng; Wu, Hongchun; Zheng, Youqi

2018-01-01

Highlights: •ADS reloading scheme is optimized to raise discharge burnup and lower reactivity loss. •ADS is flexible to be combined with various pyro-chemical reprocessing technologies. •ADS is flexible to transmute MAs from different spent PWR fuels. -- Abstract: A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

FCRD Transmutation Fuels Handbook 2015

Energy Technology Data Exchange (ETDEWEB)

Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. It is, therefore, important to understand the properties of U-Pu-Zr alloys, both with and without minor actinide additions. In addition to requiring extensive safety precautions, alloys containing U and Pu are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phase-transformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, and that general acceptance of results sometimes indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, it attempts to provide information about how well the property is known and how much variation exists between measurements. Although the handbook includes some references to publications about modeling

Uranium-thorium fuel cycle in a very high temperature hybrid system

International Nuclear Information System (INIS)

Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.

2011-01-01

Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)

Progress on the Application of Metallic Fuels for Actinide Transmutation

International Nuclear Information System (INIS)

Kennedy, J. Rory; Fielding, Randall; Janney, Dawn; Mariani, Robert; Teague, Melissa; Egeland, Gerald

2015-01-01

Full text of publication follows: Idaho National Laboratory (INL) is developing actinide bearing alloy metallic fuels intended for effecting the transmutation of long-lived isotopes in fast reactor application as part of a partitioning and transmutation strategy. This presentation will report on progress in three areas of this effort: demonstration of the fabrication of fuels under remote (hot cell) conditions directly coupled to the product from the Pyro-processing of spent fuel as part of the Joint Fuel Cycle Studies (JFCS) collaboration with the Korean Atomic Energy Research Institute (KAERI); the chemical sequestration of lanthanide fission products to mitigate fuel-cladding-chemical-interaction (FCCI); and transmission electron microscopy (TEM) and atom probe tomography (APT) studies on the as-cast microstructure of the metallic fuel alloy. For the JFCS efforts, we report on the implementation of the Glove-box Advanced Casting System (GACS) as a prototype casting furnace for eventual installation into the INL Hot Fuel Examination Facility (HFEF) where the recycled fuel will be cast. Results from optimising process parameters with respect to fuel characteristics, americium volatility, materials interaction, and lanthanide fission product carry over distribution will be discussed. With respect to the lanthanide carry over from the Pyro-processing product, encouraging studies on concepts to chemically sequester the FCCI promoting lanthanides within the fuel matrix thus inhibiting migration and interaction with the cladding will be presented. Finally, in relation to advanced modelling and simulation efforts, detailed investigations and interpretation on the nano-scale as cast microstructure of possible recycle fuel composition containing U, Pu, Am, Np as well as carry-over lanthanide species will be discussed. These studies are important for establishing the initial conditions from which advanced physics based fuel performance codes will run. (authors)

Optimization of plutonium and minor actinide transmutation in an AP1000 fuel assembly via a genetic search algorithm

Energy Technology Data Exchange (ETDEWEB)

Washington, J., E-mail: jwashing@gmail.com; King, J., E-mail: kingjc@mines.edu

2017-01-15

Highlights: • We model a modified AP1000 fuel assembly in SCALE6.1. • We couple the NEWT module of SCALE to the MOGA module of DAKOTA. • Transmutation is optimized based on choice of coating and fuel. • Greatest transmutation achieved with PuZrO{sub 2}MgO fuel pins coated with Lu{sub 2}O{sub 3}. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, which contains approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are the preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. Previous simulation work demonstrated the potential to transmute transuranic elements in a modified light water reactor fuel pin. This study optimizes a quarter-assembly containing target fuels coated with spectral shift absorbers for the transmutation of plutonium and minor actinides in light water reactors. The spectral shift absorber coating on the target fuel pin tunes the neutron energy spectrum experienced by the target fuel. A coupled model developed using the NEWT module from SCALE 6.1 and a genetic algorithm module from the DAKOTA optimization toolbox provided performance data for the burnup of the target fuel pins in the present study. The optimization with the coupled NEWT/DAKOTA model proceeded in three stages. The first stage optimized a single-target fuel pin per quarter-assembly adjacent to the central instrumentation channel. The second stage evaluated a variety of quarter-assemblies with multiple target fuel pins from the first stage and the third stage re-optimized the pins in the optimal second stage quarter-assembly. An 8 wt% PuZrO{sub 2}MgO inert matrix fuel pin with a 1.44 mm radius and a 0.06 mm Lu{sub 2}O{sub 3} coating in a five target fuel pin per quarter-assembly configuration represents the optimal combination for the

Neutronic behavior of thorium fuel cycles in a very high temperature hybrid system

International Nuclear Information System (INIS)

Rodriguez Garcia, Lorena; Milian Perez, Daniel; Garcia Hernandez, Carlos; Milian Lorenzo, Daniel; Velasco, Abanades

2013-01-01

Nuclear energy needs to guarantee four important issues to be successful as a sustainable energy source: nuclear safety, economic competitiveness, proliferation resistance and a minimal production of radioactive waste. Pebble bed reactors (PBR), which are very high temperature systems together with fuel cycles based in Thorium, they could offer the opportunity to meet the sustainability demands. Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. This paper shows the main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a variety of fuel cycles with Thorium (Th+U 233 , Th+Pu 239 and Th+U). The parameters related to the neutronic behavior like deep burn, nuclear fuel breeding, Minor Actinide stockpile, power density profiles and other are used to compare the fuel cycles using the well-known MCNPX computational code. (author)

Neutronic behavior of thorium fuel cycles in a very high temperature hybrid system

Energy Technology Data Exchange (ETDEWEB)

Rodriguez Garcia, Lorena; Milian Perez, Daniel; Garcia Hernandez, Carlos; Milian Lorenzo, Daniel, E-mail: dperez@instec.cu, E-mail: cgh@instec.cu, E-mail: dmilian@instec.cu [Higher Institute of Technologies and Applied Sciences, Havana (Cuba); Velasco, Abanades, E-mail: abanades@etsii.upm.es [Department of Simulation of Thermo Energy Systems, Polytechnic University of Madrid (Spain)

2013-07-01

Nuclear energy needs to guarantee four important issues to be successful as a sustainable energy source: nuclear safety, economic competitiveness, proliferation resistance and a minimal production of radioactive waste. Pebble bed reactors (PBR), which are very high temperature systems together with fuel cycles based in Thorium, they could offer the opportunity to meet the sustainability demands. Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. This paper shows the main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a variety of fuel cycles with Thorium (Th+U{sup 233}, Th+Pu{sup 239} and Th+U). The parameters related to the neutronic behavior like deep burn, nuclear fuel breeding, Minor Actinide stockpile, power density profiles and other are used to compare the fuel cycles using the well-known MCNPX computational code. (author)

Performance of a transmutation advanced device for sustainable energy application

International Nuclear Information System (INIS)

Garcia, C.; Rosales, J.; Garcia, L.; Perez-Navarro, A.; Escriva, A.; Abanades, A.

2009-01-01

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

Performance of a transmutation advanced device for sustainable energy application

Energy Technology Data Exchange (ETDEWEB)

Garcia, C.; Rosales, J.; Garcia, L. [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Perez-Navarro, A.; Escriva, A. [Universidad Politecnica de Valencia, Valencia (Spain). Inst. de Ingenieria Energetica; Abanades, A. [Universidad Politecnica de Madrid (Spain). Grupo de Modelizacion de Sistemas Termoenergeticos

2009-07-01

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economic estimates

International Nuclear Information System (INIS)

Rodríguez, Iván Merino; Álvarez-Velarde, Francisco; Martín-Fuertes, Francisco

2014-01-01

Highlights: • Four fuel cycle scenarios have been analyzed in resources and economic terms. • Scenarios involve Once-Through, Pu burning, and MA transmutation strategies. • No restrictions were found in terms of uranium and plutonium availability. • The best case cost and the impact of their uncertainties to the LCOE were analyzed. - Abstract: Four European fuel cycle scenarios involving transmutation options (in coherence with PATEROS and CP-ESFR EU projects) have been addressed from a point of view of resources utilization and economic estimates. Scenarios include: (i) the current fleet using Light Water Reactor (LWR) technology and open fuel cycle, (ii) full replacement of the initial fleet with Fast Reactors (FR) burning U–Pu MOX fuel, (iii) closed fuel cycle with Minor Actinide (MA) transmutation in a fraction of the FR fleet, and (iv) closed fuel cycle with MA transmutation in dedicated Accelerator Driven Systems (ADS). All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for 200 years, looking for long term equilibrium mass flow achievement. The simulations were made using the TR E VOL code, capable to assess the management of the nuclear mass streams in the scenario as well as economics for the estimation of the levelized cost of electricity (LCOE) and other costs. Results reveal that all scenarios are feasible according to nuclear resources demand (natural and depleted U, and Pu). Additionally, we have found as expected that the FR scenario reduces considerably the Pu inventory in repositories compared to the reference scenario. The elimination of the LWR MA legacy requires a maximum of 55% fraction (i.e., a peak value of 44 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation) or an average of 28 units of ADS plants (i.e., a peak value of 51 ADS units). Regarding the economic analysis, the main usefulness of the provided economic results is for relative comparison of

Flowchart evaluations of irradiated fuel treatment process of low burnup thorium

International Nuclear Information System (INIS)

Linardi, M.

1987-01-01

A literature survey has been carried out, on some versions of the acid-thorex process. Flowsheets of the different parts of the process were evaluated with mixer-settlers experiments. A low burnup thorium fuel (mass ratio Th/U∼100/1), proposed for Brazilian fast breeder reactor initial program, was considered. The behaviour of some fission products was studied by irradiated tracers techniques. Modifications in some of the process parameters were necessary to achieve low losses of 233 U and 232 U and 232 Th. A modified acid-thorex process flowsheet, evaluated in a complete operational cycle, for the treatment of low burnup thorium fuels, is presented. High decontamination factors of thorium in uranium, with reasonable decontamination of uranium in thorium, were achieved. (author) [pt

Advances in Metallic Fuels for High Burnup and Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

2016-10-01

Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

International Nuclear Information System (INIS)

Trellue, Holly R.

2004-01-01

The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

DE-NE0000735 - FINAL REPORT ON THORIUM FUEL CYCLE NEUP PROJECT

Energy Technology Data Exchange (ETDEWEB)

Krahn, Steven [Vanderbilt Univ., Nashville, TN (United States); Ault, Timothy [Vanderbilt Univ., Nashville, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-30

The report is broken into six chapters, including this executive summary chapter. Following an introduction, this report discusses each of the project’s three major components (Fuel Cycle Data Package (FCDP) Development, Thorium Fuel Cycle Literature Analysis and Database Development, and the Thorium Fuel Cycle Technical Track and Proceedings). A final chapter is devoted to summarization. Various outcomes, publications, etc. originating from this project can be found in the Appendices at the end of the document.

Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

Science.gov (United States)

Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

2018-03-01

The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

Design and safety studies on the European Facility for Industrial Transmutation (EFIT) with CERMET fuel

International Nuclear Information System (INIS)

Chen, X.N.; Rineiski, A.; Liu, P.; Matzerath Boccaccini, C.; Flad, M.; Gabrielli, F.; Maschek, W.; Morita, K.

2008-01-01

European R and D for ADS design and fuel development is driven in the 6 th FP of the EU by the EUROTRANS Programme [1]. In EUROTRANS two ADS design routes are followed, the XT-ADS and the EFIT. The XT-ADS is designed to provide the experimental demonstration of transmutation in an Accelerator Driven System. The EFIT development, the European Facility for Industrial Transmutation, aims at a generic conceptual design of a full transmuter. A key issue of the R and D work is the choice of an adequate fuel to be used in an Accelerator Driven Transmuter (ADT) like EFIT. Various fuel forms have been assessed. CERCER and CERMET fuels, specifically with the matrices MgO and Mo, have finally been selected and are now under closer investigation. Within EUROTRANS, a special domain named 'AFTRA', is responsible to more deeply assess the behavior of these dedicated fuels and to provide the fuel data base for the core design of the EFIT. The EFIT concept has to be optimized towards: a good transmutation efficiency, high burnup, low reactivity swing, low power peaking, adequate subcriticality, reasonable beam requirements and a high safety level. The final recommendation on fuels by AFTRA gave a ranking of these fuels based on the mentioned criteria. The composite CERMET fuel (Pu 0.5 ,Am 0.5 )O 2-x - Mo (with the isotope 92 Mo comprising 93% of the molybdenum) has been recommended as the primary candidate for the EFIT. This CERMET fuel fulfils adopted criteria for fabrication and reprocessing, and provides excellent safety margins. Disadvantages include the cost for enrichment of 92 Mo and a lower specific transmutation rate of minor actinides, because of the higher neutron absorption cross-section of the matrix. The composite CERCER fuel (Pu 0.4 ,Am 0.6 )O 2-x - MgO has therefore been recommended as a backup solution as it might offer a higher consumption rate of minor actinides, and can be manufactured for a lower unit cost. This paper is in fact a sequel to our last paper [2

Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

Energy Technology Data Exchange (ETDEWEB)

Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

2017-07-01

The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

Development of CERMET fuels for minor actinides transmutation

International Nuclear Information System (INIS)

Haas, D.; Fernandez, A.; Naestren, C.; Staicu, D.; Somers, J.; Maschek, W.; Chen, X.

2006-01-01

The sub-critical Accelerator Driven System (ADS) is now being considered as a potential means to burn long-lived transuranium nuclides. The preferred fuel for such a fast neutron reactor is uranium-free, highly enriched with plutonium and minor actinides. Requirements for ADS transmutation fuels are linked with the core design and safety parameters, the fuel properties and the ease of reprocessing. This study concerns the properties of metals as matrices, with the particular case of Mo. To improve the neutronic characteristics, enriched molybdenum (Mo-92) is required. To overcome the high enrichment cost, it is proposed to recover the matrix by pellet dissolution, and to recycle it for further use. Irradiation programmes are also planned to examine the in-reactor properties of the material. Based on the current status of the research, the results are promising, but irradiation results are still missing. (authors)

Study Of Thorium As A Nuclear Fuel.

Directory of Open Access Journals (Sweden)

Prakash Humane

2017-10-01

Full Text Available Conventional fuel sources for power generation are to be replacing by nuclear power sources like nuclear fuel Uranium. But Uranium-235 is the only fissile fuel which is in 0.72 found in nature as an isotope of Uranium-238. U-238 is abundant in nature which is not fissile while U-239 by alpha decay naturally converted to Uranium- 235. For accompanying this nuclear fuel there is another nuclear fuel Thorium is present in nature is abundant can be used as nuclear fuel and is as much as safe and portable like U-235.

Prospective thorium fuels for future nuclear energy generation

International Nuclear Information System (INIS)

Lainetti, Paulo E.O.

2017-01-01

In the beginning of the Nuclear Era, many countries were interested on thorium, particularly during the 1950 1970 periods. Nevertheless, since its discovery almost two centuries ago, the use of thorium has been restricted to gas mantles employed in gas lighting. The future world energy needs will increase and, even if we assumed a conservative contribution of nuclear generation, it will be occur a significant increasing in the uranium prices, taking into account that uranium, as used in the present thermal reactors, is a finite resource. Nowadays approximately the worldwide yearly requirement of uranium for about 435 nuclear reactors in operation is 65,000 metric t. Therefore, alternative solutions for future must be developed. Thorium is nearly three times more abundant than uranium in The Earth's crust. Despite thorium is not a fissile material, 232 Th can be converted to 233 U (fissile) more efficiently than 238 U to 239 Pu. Besides this, thorium is an environment alternative energy source and also inherently resistant to proliferation.. Many countries had initiated research on thorium in the past, Nevertheless, the interest evanesced due new uranium resources discoveries and availability of enriched uranium at low prices from obsolete weapons. Some papers evaluate the thorium resources in Brazil over 1.200.000 metric t. Then, the thorium alternative must be seriously considered in Brazil for strategic reasons. A brief history of thorium and its utilization are presented, besides a very short discussion about prospective thorium nuclear fuels for the next generation of nuclear reactors. (author)

Prospective thorium fuels for future nuclear energy generation

Energy Technology Data Exchange (ETDEWEB)

Lainetti, Paulo E.O., E-mail: lainetti@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-07-01

In the beginning of the Nuclear Era, many countries were interested on thorium, particularly during the 1950 1970 periods. Nevertheless, since its discovery almost two centuries ago, the use of thorium has been restricted to gas mantles employed in gas lighting. The future world energy needs will increase and, even if we assumed a conservative contribution of nuclear generation, it will be occur a significant increasing in the uranium prices, taking into account that uranium, as used in the present thermal reactors, is a finite resource. Nowadays approximately the worldwide yearly requirement of uranium for about 435 nuclear reactors in operation is 65,000 metric t. Therefore, alternative solutions for future must be developed. Thorium is nearly three times more abundant than uranium in The Earth's crust. Despite thorium is not a fissile material, {sup 232}Th can be converted to {sup 233}U (fissile) more efficiently than {sup 238}U to {sup 239}Pu. Besides this, thorium is an environment alternative energy source and also inherently resistant to proliferation.. Many countries had initiated research on thorium in the past, Nevertheless, the interest evanesced due new uranium resources discoveries and availability of enriched uranium at low prices from obsolete weapons. Some papers evaluate the thorium resources in Brazil over 1.200.000 metric t. Then, the thorium alternative must be seriously considered in Brazil for strategic reasons. A brief history of thorium and its utilization are presented, besides a very short discussion about prospective thorium nuclear fuels for the next generation of nuclear reactors. (author)

Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

International Nuclear Information System (INIS)

Dudnikov, A. A.; Alekseev, P. N.; Subbotin, S. A.

2007-01-01

Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

Assessment of the insertion of reprocessed fuel spiked with thorium in a PWR core

Energy Technology Data Exchange (ETDEWEB)

Castro, Victor F.; Monteiro, Fabiana B.A.; Pereira, Claubia, E-mail: victorfc@fis.grad.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Reprocessed fuel by UREX+ technique and spiked with thorium was inserted in a PWR core and neutronic parameters have been analyzed. Based on the Final Safety Analysis Report (FSAR) of the Angra-2 reactor, the core was modeled and simulated with SCALE6.0 package. The neutronic data evaluation was carried out by the analysis of the effective and infinite multiplication factors, and the fuel evolution during the burnup. The conversion ratio (CR) was also evaluated. The results show that, when inserting reprocessed fuel spiked with thorium, the insertion of burnable poison rods is not necessary, due to the amount of absorber isotopes present in the fuel. Besides, the conversion ratio obtained was greater than the presented by standard UO{sub 2} fuel, indicating the possibility of extending the burnup. (author)

1. round table - Spent fuels composition. Back-end of the fuel cycle and reprocessing, plutonium and other nuclear materials management. 2. round table - Separation-transmutation. 3. round table - Scenarios for a long term inventory of nuclear materials and wastes

International Nuclear Information System (INIS)

2005-01-01

The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. The debate comprises 4 public hearings (September 2005: Bar-le-Duc, Saint-Dizier, Pont-du-Gard, Cherbourg), 12 round-tables (October and November 2005: Paris, Joinville, Caen, Nancy, Marseille), a synthesis meeting (December 2005, Dunkerque) and a closing meeting (January 2006, Lyon). This document is the synthesis of the round table debates which took place at Paris on the reprocessing of spent fuels. Three aspects are discussed: the risks linked with the recovery of valorizable materials, the economical viability of the separation/transmutation option, and the future of the nuclear option in the French energy policy. Six presentations (transparencies) are attached with these proceedings which treat of: the reprocessing/recycling to the test, perspectives of future wastes, present day wastes/valorizable materials and future scenarios, critical analysis scenarios, why reprocessing spent fuels?, processing of spent fuels and recycling, separation and transmutation of long-lived radioactive wastes, thorium-uranium cycle. (J.S.)

Equilibrium transuranic management scheme for diverse fuel cycle analysis

International Nuclear Information System (INIS)

Haas, Jason; Lee, John C.

2008-01-01

A key issue cited in the U.S. Department of Energy's report to Congress (2003) on the research path for the Advanced Fuel Cycle Initiative (AFCI) is an accurate estimation of life cycle costs for the construction, operation, decontamination and decommissioning of all fuel cycle facilities. In this report we discuss the methodology and validation of a fuel cycle model based on equilibrium operation. We apply our model to a diverse set of advanced reactors and fuel types in order to determine the most effective transmuting system while simultaneously minimizing fuel cycle costs. Our analysis shows that a nearly instant equilibrium modeling of fuel cycle scenarios can accurately approximate the detailed complex dynamic models developed by national laboratories. Our analysis also shows that the cost of transmuting Spent Nuclear Fuel (SNF) from a UO 2 fueled Pressurized Water Reactor (PWR) is minimized by utilizing the thorium cycle in sodium cooled fast reactors and is near the cost for long term repository storage of SNF at Yucca Mountain. (authors)

Status of thorium technology

International Nuclear Information System (INIS)

Garg, R.K.; Raghavan, R.V.; Karve, V.M.; Narayandas, G.R.

1977-01-01

Although a number of studies have been conducted in various countries to evolve reactor systems based on thorium fuel cycle, its use, so far, is limited to only a few reactors. However, for countries having large reserves of thorium, its utilization is of great significance for their nuclear power programmes. Reasonably assured world resources of thorium in the lower price range have been estimated at more than 500,000 tons of ThO 2 . While most of these resources are in placer deposits in various parts of the world, some vein deposits and uranium ores are other important sources of thorium. Monazite, the most important mineral of thorium, is found in the beach sand deposits along with other heavy minerals like ilmenite, rutile, zircon, and sillimanite etc. Mining of these deposits is usually carried out by suction dredging and separation of monazite from other minerals is effected by a combination of magnetic, electrostatic and gravity separation techniques. Chemical processing of monazite is carried out either by sulphuric acid or caustic treatment, followed by separation of the rare earths and thorium by partial precipitation or leaching. The thorium concentrate is further processed to obtain mantle grade thorium nitrate by chemical purification steps whereas solvent extraction using TBP is adopted for making nuclear-grade material. The purified thorium nitrate is converted to the oxide usually by precipitation as oxalate followed by calcination. The oxide is reduced directly with calcium or converted to the chloride or fluoride and then reduced by calcium or magnesium to obtain thorium metal. Various fuel designs based on the metal or its alloys, mixed oxides or carbides, and dispersed type fuel elements have been developed and accordingly, different fabrication techniques have been employed. Work on irradiation of thorium containing fuel elements and separation of U 233 is being carried out. This paper reviews the status of thorium technology in the world with

Verification of a Subgroup Generation Method for Thorium Fuel Assemblies

International Nuclear Information System (INIS)

Sim, Ohsung; Kim, Myunghyun

2013-01-01

Resonance parameter consists of subgroup level and weight. The subgroup weight is obtained by solving the ultrafine slowing down equation and fixed source problem. That means this cross section library procedure considers conservation of the shielded cross section for pin-cell in order to obtain subgroup parameters. There are some isotopes to be concerned for research such as actinides and thorium. Minor actinides(MA) are existing with very small amount in a spent fuel, but effect is not negligible in a high burnup fuel assemblies. Some MAs have high fission cross sections under thermal neutron spectrum. Thorium isotopes was not investigated as much as uranium, but it has high potential for future application. In this study, a new cross section library to be replaced with HELIOS library was generated and compared for the assembly calculation, specially for assembly with thorium. An average capture cross section value at a certain fuel pin and multiplication factor of assembly were compared with nTRACER calculation with HELIOS library and Monte Carlo calculation of MCNP with ENDF-B/II. The accuracy of library data generated for thorium isotope in nTRACER calculation was tested for WASB model. There was a great improvement in K-eff and capture cross section for this assembly compared with old library, HELIOS library

Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

International Nuclear Information System (INIS)

Stacey, W.M.

2001-01-01

The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (P fus ∼ 10-100 MW, β N ∼ 2-3, Q p ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

International Nuclear Information System (INIS)

Bi, G.; Liu, C.; Si, S.

2012-01-01

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared

Thorium fuel for light water reactors - reducing proliferation potential of nuclear power fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Galperin, A; Radkowski, A [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

1996-12-01

The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of Th utilization is to design a core and a fuel cycle, which would be proliferation-resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WWER-1000, and designated as VVERT. The following are the main results of the preliminary reference design: * The amount of Pu contained in the RTR spent fuel stockpile is reduced by 80% in comparison with a VVER of a current design. * The isotopic composition of the RTR-Pu greatly increases the probability of pre-initiation and yield degradation of a nuclear explosion. An extremely large Pu-238 content causes correspondingly large heat emission, which would complicate the design of an explosive device based on RTR-Pu. The economic incentive to reprocess and reuse the fissile component of the RTR spent fuel is decreased. The once-through cycle is economically optimal for the RTR core and cycle. To summarize all the items above: the replacement of a standard (U-based) fuel for nuclear reactors of current generation by the RTR fuel will provide an inherent barrier for nuclear weapon proliferation. This inherent barrier, in combination with existing safeguard measures and procedures is adequate to unambiguously disassociate civilian nuclear power from military nuclear power. * The RTR concept is applied to existing power plants to assure its economic feasibility. Reductions in waste disposal requirements, as well as in natural U and fabrication expenses, as compared to a standard WWER fuel, provide approximately 20% reduction in fuel cycle (authors).

Anticipated radiological impacts from the mining and milling of thorium for the nonproliferative fuels

International Nuclear Information System (INIS)

Meyer, H.R.; Till, J.E.

1978-01-01

Recent emphasis on proliferation-resistant fuel cycles utilizing thorium--uranium-233 fuels has necessitated evaluation of the potential radiological impact of mining and milling thorium ore. Therefore, an analysis has been completed of hypothetical mine-mill complexes using population and meteorological data representative of a thorium resource site in the Lemhi Pass area of Idaho/Montana, United States of America. Source terms for the site include thorium-232 decay chain radionuclides suspended as dusts and radon-220 and daughters initially released as gas. Fifty-year dose commitments to maximally exposed individuals of 2.4 mrem to total body, 9.5 mrem to bone, and 35 mrem to lungs are calculated to result from facility operation. Radium-228, thorium-228, thorium-232 and lead-212 (daughter of radon-220) are found to be the principal contributors to dose. General population doses for a 50-mile radius surrounding the facility are estimated to be 0.05 man-rem to total body, 0.1 man-rem to bone, and 0.7 man-rem to lungs. Generally speaking, the results of this study indicate that the radiological aspects of thorium mining and milling should pose no significant problems with regard to implementation of thorium fuel cycles

Study of an optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak

Energy Technology Data Exchange (ETDEWEB)

Hong, Bong Guen, E-mail: bghong@jbnu.ac.kr [Department of Quantum System Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of); Kim, Hoseok [Department of Applied Plasma Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of)

2016-11-15

Highlights: • Optimum configuration of a transmutation reactor based on a low aspect ratio tokamak was found. • Inboard and outboard radial build are determined by plasma physics, engineering and neutronics constraints. • Radial build and equilibrium fuel cycle play a major role in determining the transmutation characteristics. - Abstract: We determine the optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak. For self-consistent determination of the radial build of the reactor components, we couple a tokamak systems analysis with a radiation transport calculation. The inboard radial build of the reactor components is obtained from plasma physics and engineering constraints, while outboard radial builds are mainly determined by constraints on neutron multiplication, the tritium-breeding ratio, and the power density. We show that the breeding blanket model has an effect on the radial build of a transmutation blanket. A burn cycle has to be determined to keep the fast neutron fluence plasma-facing material below its radiation damage limit. We show that the radial build of the transmutation reactor components and the equilibrium fuel cycle play a major role in determining the transmutation characteristics.

Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

International Nuclear Information System (INIS)

Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

1980-09-01

A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

International Nuclear Information System (INIS)

Hong, S. G.; Kim, Y. H.; Venneri, F.

2010-01-01

In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

FCRD Advanced Reactor (Transmutation) Fuels Handbook

Energy Technology Data Exchange (ETDEWEB)

Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, important to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and

High-quality thorium TRISO fuel performance in HTGRs

Energy Technology Data Exchange (ETDEWEB)

Verfondern, Karl [Forschungszentrum Juelich GmbH (Germany); Allelein, Hans-Josef [Forschungszentrum Juelich GmbH (Germany); Technische Hochschule Aachen (Germany); Nabielek, Heinz; Kania, Michael J.

2013-11-01

Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10{sup 25} n/m{sup 2} (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10{sup -5} for manufacturing, 2 x 10{sup -4} for normal operating conditions, and 5 x 10{sup -4

High-quality thorium TRISO fuel performance in HTGRs

International Nuclear Information System (INIS)

Verfondern, Karl; Allelein, Hans-Josef; Nabielek, Heinz; Kania, Michael J.

2013-01-01

Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O 2 TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10 25 n/m 2 (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10 -5 for manufacturing, 2 x 10 -4 for normal operating conditions, and 5 x 10 -4 for accident conditions. These

The uranium and thorium separation in the chemical reprocessing of the irradiated fuel of thorium and uranium mixed oxides

International Nuclear Information System (INIS)

Oliveira, E.F. de.

1984-09-01

A bibliographic research has been carried out for reprocessing techniques of irradiated thorium fuel from nuclear reactors. The Thorex/Hoechst process has been specially considered to establish a method for reprocessing thorium-uranium fuel from PWR. After a series of cold tests performed in laboratory it was possible to set the behavior of several parameters affecting the Thorex/Hoechst process. Some comments and suggestions are presented for modifications in the process flosheet conditions. A discussion is carried out for operational conditions such as the aqueous to organic flow ratio the acidity of strip and scrub solutions in the process steps for thorium and uranium recovery. The operation diagrams have been constructed using equilibrium experimental data which correspond to conditions observed in laboratory. (Author) [pt

Survey of thorium utilization in power reactor systems

International Nuclear Information System (INIS)

Schwartz, M.H.; Schleifer, P.; Dahlberg, R.C.

1976-01-01

It is clear that thorium-fueled thermal power reactor systems based on current technology can play a vital role in serving present and long-term energy needs. Advanced thorium converters and thermal breeders can provide an expanded resource base from which the world's growing energy demands can be met. Utilization of a symbiotic system of fast breeders and thorium-fueled thermal reactors can be particularly effective in providing low cost power while conserving uranium resources. Breeder reactors are characterized by high capital costs and very low fuel costs since they produce more fuel than they consume. This excess fuel can be used to fuel thermal converter reactors whose capital costs are low. This symbiosis is optimized when 233 U is bred in the fast breeders and then used to fuel high-conversion-ratio thermal converter reactors operating on the thorium-uranium fuel cycle. The thorium-cycle HTGR, after undergoing more than fifteen years of development in both the United States and Europe, provides for the optimum utilization of our limited uranium resources. Other thermal reactor systems, previously operating on the uranium cycle, also show potential in their capability to utilize the thorium cycle effectively

Review of Brazilian activities related to the thorium fuel cycle and production of thorium compounds at IPEN-CNEN/SP

International Nuclear Information System (INIS)

Lainetti, Paulo E.O.; Freitas, Antonio A.; Mindrisz, Ana C.

2013-01-01

The Brazilian's interest in the nuclear utilization of thorium has started in the 50's as a consequence of the abundant occurrence of monazite sands. Since the sixties, IPEN-CNEN/SP has performed some developments related to the thorium fuel cycle. The production and purification of thorium compounds was carried out at IPEN for about 18 years and the main product was the thorium nitrate with high purity, having been produced over 170 metric tons of this material in the period, obtained through solvent extraction. The thorium nitrate was supplied to the domestic industry and used for gas portable lamps (Welsbach mantle). Although the thorium compounds produced have not been employed in the nuclear area, several studies were conducted. Therefore, those activities and the accumulated experience are of strategic importance, on one hand due to huge Brazilian thorium reserves, on the other hand by the resurgence of the interest of thorium for the Generation IV Advanced Reactors. This paper presents a review of the Brazilian research and development activities related to thorium technology. (author)

Evaluation of alternative partitioning/transmutation scenarios using transmutation in light-water reactors (LWRs)

International Nuclear Information System (INIS)

Collins, E.D.; Renier, J.P.; Del Cul, B.; Spencer, B.

2005-01-01

Previous Advanced Fuel Cycle Initiative (AFCI) studies were made to assess the effects of the existing accumulation of LWR spent fuel in the United States on the capability to partition/transmute actinides using existing and advanced LWR. The concept of treating the oldest fuel first indicated that significant advantages could be gained in both partitioning, transmutation, and in overall cost reduction. The processing scenarios previously evaluated assumed that (1) 2000 MT/year of spent fuel, irradiated to 45 GWd/MT and decayed for 30 years is processed; (2) recovered plutonium and 90% of the neptunium are transmuted in LWR MOX fuel; and (3) minor actinides, consisting of americium, curium, and 10% of the neptunium are transmuted in burnable poison type targets. Results of the previous study showed that significant benefits could be obtained, including (1) lower costs for partitioning and transmutation and for storage of spent fuel, (2) maintenance of proliferation resistance for the fissile plutonium in spent fuels, and (3) extended lifetime for the repository. The lower costs would be achieved primarily because no capital investment for a special transmuter reactor (fast reactor, accelerator-driven system, etc.) would be required. Instead, only existing and new LWRs would be utilized. Moreover, no new storage capacity could be needed for spent fuels and irradiated targets because the number of spent fuel assemblies would remain the same after the scenario is begun. Even though the total inventory of plutonium would rise during the early cycles, ∼98% of the plutonium would be contained in stored spent fuel and would be protected by high radiation (the Spent Fuel Standard). This is because the spent fuel would be reprocessed and re-irradiated at intervals within which the fission products, 137 Cs and 90 Sr, both with half-lives of ∼ 30 years, exist in significantly high concentrations.The lifetime of the repository would be extended significantly because all of

Program on fuels for transmutation: present status and prospects

Energy Technology Data Exchange (ETDEWEB)

Rouault, J.; Garnier, J.C.; Chauvin, N.; Pillon, S. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Combustibles

2001-07-01

The performance calculations of appropriate fuel cycle facilities and reactor configurations (scenarios) relying on current reactor technologies (Pressurized Water Reactor and Fast neutrons Reactors) or innovative reactors (Accelerator Driven Systems) have proved the scientific feasibility of some P and T strategies. To insure the technological feasibility, a large program on fuels and materials is underway, including advanced concepts for PWRs and the development of specific targets (dispersed fuels) for transmutation in Fast Reactors. Experiments in different reactors including Phenix are being prepared. The program is presented and recent results are given. (author)

Program on fuels for transmutation: present status and prospects

International Nuclear Information System (INIS)

Rouault, J.; Garnier, J.C.; Chauvin, N.; Pillon, S.

2001-01-01

The performance calculations of appropriate fuel cycle facilities and reactor configurations (scenarios) relying on current reactor technologies (Pressurized Water Reactor and Fast neutrons Reactors) or innovative reactors (Accelerator Driven Systems) have proved the scientific feasibility of some P and T strategies. To insure the technological feasibility, a large program on fuels and materials is underway, including advanced concepts for PWRs and the development of specific targets (dispersed fuels) for transmutation in Fast Reactors. Experiments in different reactors including Phenix are being prepared. The program is presented and recent results are given. (author)

The Tasse concept (thorium based accelerator driven system with simplified fuel cycle for long term energy production)

International Nuclear Information System (INIS)

Berthou, V.; Slessarev, I.; Salvatores, M.

2001-01-01

Within the framework of the nuclear waste management studies, the ''one-component''. concept has to be considered as an attractive option in the long-term perspective. This paper proposes a new system called TASSE (''Thorium based Accelerator driven System with Simplified fuel cycle for long term Energy production''.), destined to the current French park renewal. The main idea of the TASSE concept is to simplify both the front and the back end of the fuel cycle, and his major goals are to provide electricity with low waste production, and with an economical competitiveness. (author)

Interaction between thorium and potential clad materials

International Nuclear Information System (INIS)

Kale, G.B.; Gawde, P.S.; Sengupta, Pranesh

2005-01-01

Thorium based fuels are being used for nuclear reactors. The structural stability of fuel-clad assemblies in reactor systems depend upon the nature of interdiffusion reaction between fuel-cladding materials. Interdiffusion reaction thorium and various cladding materials is presented in this paper. (author)

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

International Nuclear Information System (INIS)

Ikeda, Kazumi; Kurata, Masaki; Morita, Yasuji; Tsujimoto, Kazufumi; Minato, Kazuo; Koyama, Shin-ichi

2011-01-01

This paper treats technology readiness level (TRL) assessment of Partitioning and Transmutation (P-T) toward closed fuel cycle in JAPAN. The purpose is providing clarified information related to the current maturity of the partitioning and transmutation technologies by applying the methodology of TRL, parallel to attempting to establish common indications among relating technology area. The methodology should be one of useful communication tools between specialists and management level, and also among countries interested in the P-T technologies. The generic TRL in this study is based on the GNEP (Global Nuclear Energy Partnership)'s definition: TRL 3 shows the status that critical function is proved and elemental technologies are identified, TRL 4 represents that relating technologies are validated at bench scale in laboratory environment, and TRL 5 achieves the completion of development related to the subsystem and elemental technologies. Detailed indications are established through discussion of the relating specialists. Reviewed technological area includes P-T and minor actinide (MA) cycle: Fast Breeder Reactor (FBR) and Accelerator driven system (ADS) for MA transmutation, partitioning processes, and MA-bearing fuels. The assessments reveal that TRL spreads around TRL 3 to TRL 4 because each system requires more the development of elemental technologies. Transmutation core of FBR is assessed to be TRL 4 in that MA bearing integral test is required additionally, and ADS becomes TRL 3 because the elemental technologies were identified and the requirements were specified. Consequently, the common key issue is how the nuclear calculation methodology will be validated for MA-bearing-fuelled core, since several percentages of MA changes the void reactivity and the Doppler Effect significantly, which are inherently important in reactor safety. It should be that critical experiments with several kg of americium or more are difficult in the existing experimental

Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

2014-02-15

Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

Short-term storage considerations for spent plutonium-thorium fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Blomeley, L.; Dugal, C.; Masala, E.; Tran, T., E-mail: laura.blomeley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2015-12-15

To support the development of advanced pressurized heavy water reactor (PHWR) fuel cycles, it is necessary to study short-term storage solutions for spent reactor fuel. In this paper, some representational criticality safety and shielding assessments are presented for a particular PHWR plutonium-thorium based fuel bundle concept in a hypothetical aboveground dry storage module. The criticality assessment found that the important parameters for the storage design are neutron absorber content and fuel composition, particularly in light of the high sensitivity of code results to plutonium. The shielding assessment showed that the shielding as presented in the paper would need to be redesigned to provide greater gamma attenuation. These findings can be used to aid in designing fuel storage facilities. (author)

Road-map design for thorium-uranium breeding recycle in PWR - 031

International Nuclear Information System (INIS)

Shengyi, Si

2010-01-01

The paper was focused on designing a road-map to finally approach sustainable Thorium-Uranium ( 232 Th- 233 U) Breeding Recycle in current PWR, without any other change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. At first, the paper presented some insights to the inherence of Thorium-Uranium fuel conversion or breeding in PWR based on the neutronics theory and revealed the prerequisites for Thorium-Uranium fuel in PWR to achieve sustainable Breeding Recycle; And then, various Thorium-based fuels were designed and examined, and the calculation results further validated the above theoretical deductions; Based on the above theoretical analysis and calculation results, a road-map for sustainable Thorium-Uranium breeding recycle in PWR was outlined finally. (authors)

Feasibility of Thorium Fuel Cycles in a Very High Temperature Pebble-Bed Hybrid System

Directory of Open Access Journals (Sweden)

L.P. Rodriguez

2015-08-01

Full Text Available Nuclear energy presents key challenges to be successful as a sustainable energy source. Currently, the viability of the use thorium-based fuel cycles in an innovative nuclear energy generation system is being investigated in order to solve these key challenges. In this work, the feasibility of three thorium-based fuel cycles (232Th-233U, 232Th-239Pu, and 232Th-U in a hybrid system formed by a Very High Temperature Pebble-Bed Reactor (VHTR and two Pebble-Bed Accelerator Driven Systems (ADSs was evaluated using parameters related to the neutronic behavior such as nuclear fuel breeding, minor actinide stockpile, the energetic contribution of each fissile isotope, and the radiotoxicity of the long lived wastes. These parameters were used to compare the fuel cycles using the well-known MCNPX ver. 2.6e computational code. The results obtained confirm that the 232Th-233U fuel cycle is the best cycle for minimizing the production of plutonium isotopes and minor actinides. Moreover, the inclusion of the second stage in the ADSs demonstrated the possibility of extending the burnup cycle duration and reducing the radiotoxicity of the discharged fuel from the VHTR.

Inert matrix fuels for incineration of plutonium and transmutation of americium

International Nuclear Information System (INIS)

Matzke, Hj.

2000-01-01

In conventional U-based nuclear fuels, both Pu and higher actinides (mainly Am, but also Np and Cm) are formed by neutron capture reactions and α- or β-decay. If a strategy of reprocessing is adopted as in some European nations and in Japan, the separated Pu can be recycled as (U, Pu)O 2 (or mixed-oxide-MOX) fuel. The high-level liquid waste of reprocessing is presently vitrified. However, the alternative of separating the minor actinides from the fission products (partitioning) and subsequent transmutation in existing reactors or in new dedicated actinide burners is widely studied as a possible means to reduce the radiotoxicity of the waste

Microgamma Scan System for analyzing radial isotopic profiles of irradiated transmutation fuels

International Nuclear Information System (INIS)

Hilton, Bruce A.; McGrath, Christopher A.

2008-01-01

The U. S. Global Nuclear Energy Partnership / Advanced Fuel Cycle Initiative (GNEP/AFCI) is developing metallic transmutation alloys as a fuel form to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. A micro-gamma scan system is being developed to analyze the radial distribution of fission products, such as Cs-137, Cs-134, Ru-106, and Zr-95, in irradiated fuel cross-sections. The micro-gamma scan system consists of a precision linear stage with integrated sample holder and a tungsten alloy collimator, which interfaces with the Idaho National Laboratory (INL) Analytical Laboratory Hot Cell (ALHC) Gamma Scan System high purity germanium detector, multichannel analyzer, and removable collimators. A simplified model of the micro-gamma scan system was developed in MCNP (Monte-Carlo N-Particle Transport Code) and used to investigate the system performance and to interpret data from the scoping studies. Preliminary measurements of the micro-gamma scan system are discussed. (authors)

Transmutations of nuclear waste. Progress report RAS programme 1995: Recycling and transmutation of actinides and fission products

International Nuclear Information System (INIS)

Gruppelaar, H.; Cordfunke, E.H.P.; Konings, R.J.M.; Bultman, J.H.; Dodd, D.H.; Franken, W.M.P.; Kloosterman, J.L.; Koning, A.J.; Wichers, V.A.

1996-04-01

This report describes the progress of the Dutch RAS programme on 'Recycling and Transmutation of Actinides and Fission Products' over the year 1995, which is the second year of the 4-year programme 1994-1997. An extensive listing of reports and publications from 1991 to 1995 is given. Highlights in 1995 were: -The completion of the European Strategy Study on Nuclear Waste Transmutation as a result of which the understanding of transmutation of plutonium, minor actinides and long-lived fission products in thermal and fast reactors has been increased significantly. Important ECN contributions were given on Am, 99 Tc and 129 I transmutation options. Follow-up contracts have been obtained for the study of 100% MOX cores and accelerator-based transmutation. - Important progress in the evaluation of CANDU reactors for burning very large amounts of transuranium mixtures in inert matrices. - The first RAS irradiation experiment in the HFR, in which the transmutation of technetium and iodine was examined, has been completed and post-irradiation examination has been started. - A joint proposal of the EFTTRA cooperation for the 4 th Framework Programme of the EU, to demonstrate the feasibility of the transmutation of americium in an inert matrix by an irradiation in the HFR, has been granted. - A bilateral contract with CEA has been signed to participate in the CAPRA programme, and the work in this field has been started. - The thesis work on Actinide Transmutation in Nuclear Reactor Systems was succesfully defended. New PhD studies on Pu burning in HTGR, on nuclear data for accelerator-based systems, and on the SLM-technique for separation of actinides were started. - A review study of the use of the thorium cycle as a means for nuclear waste reduction, has been completed. A follow-up of this work is embedded in an international project for the 4th Framework Programme of the EU. (orig./DG)

Transmutation research and fuel cycle (report on discussion at Research Reactor Institute, Kyoto University)

International Nuclear Information System (INIS)

Yamana, Hajimu

1999-01-01

A symposium was held on a topic of 'Transmutation Research' on Dec. 21 and 22, 1999 at Research Reactor Institute, Kyoto University. This meeting was held as a joint-meeting of KUR's specialist meeting and Tokyo University's activity supported by the Grant-in-Aid for Scientific Research of Ministry of Education, Sport and Culture of Japan. This paper describes the overview of the discussions of this joint-meeting, and interprets their significance. Major themes discussed are, needed discussions on the transmutation research, policy and concepts of the organizations doing transmutation researches, a view from university side, transmutation researches in the oversea countries, opinions from various standpoints of the nuclear fuel cycle, conclusive discussions. 'the meanings of the transmutation research should be discussed together with the geological disposal and fast reactor system', 'transmutation may be a cooperative option for the disposal, thus, they should not be in a independent relation', and Balance evaluation will be needed' are the examples of the conclusive remarks of this meeting. (author)

A novel concept of QUADRISO particles Part III: applications to the plutonium-thorium fuel cycle

International Nuclear Information System (INIS)

Talamo, A.

2009-01-01

In the present study, a plutonium-thorium fuel cycle is investigated including the 233 U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing 233 U. The above configuration has been compared with a configuration where fissile (neptunium-plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the 233 U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.

The Thorium-Cycle: safe, abundant power for the new millennium

Science.gov (United States)

Don, May; George, Kim; Peter, Mcintyre; Charles, Meitzler; Robert, Rogers; Akhdior, Sattarov; Mustafa, Yavuz

2001-10-01

A design has been developed for using accelerator-driven thorium fission to produce electric power. A thorium-cycle reactor works by electro-breeding. A pattern of thorium fuel rods is supported in a vessel containing molten lead. A beam of high-energy (1 GeV) protons is targeted in the center of the vessel, and produces a copious flux of energetic neutrons by spallation. The neutrons transmute the thorium nuclei two steps up the periodic table to U233, which fissions rapidly to produce thermal energy. The lead serves as the spallation target, the moderator, and the heat exchange medium to transfer heat from the core to steam exchangers above the core. The thorium cycle has several important advantages over current uranium-cycle fission technology: it is intrinsically stable it cannot melt down; it eats its own waste; it cannot produce bomb-grade isotopes; and there are sufficient thorium reserves to supply the entire Earth’s energy economy for the next millennium. The concept of a thorium-cycle power reactor was first proposed by Rubbia in 1995. Key problems in the original concept were the proton injector (15 MW beam power), reliability of accelerator systems, and parasitic absorption of neutrons by fission products during the life of the core. We have addressed all three problems in a design for a flux-coupled stack of isochronous cyclotrons, delivering a pattern of 7 independent beams to the core. An interdisciplinary collaboration is being formed to develop the concept to a serious design.

Role of thorium in ensuring long term energy security to India

International Nuclear Information System (INIS)

Malhotra, S.K.

2013-01-01

Role of nuclear power in ensuring energy security to the world is inevitable due to a) dwindling fossil fuel resources and b) need for minimising green house gas emission that poses the risk of global climate change. India, keeping in mind its limited uranium and vast thorium resources, is pursuing a three stage nuclear power programme. The first stage is based on reactors that use uranium as fuel. It comprises of the indigenous Pressurised Heavy Water Reactors using natural uranium as fuel and light water reactors that employ enriched uranium as fuel and are to be set up in technical collaboration with other countries. The second stage is based on fast breeder reactors that employ plutonium derived from reprocessing of spent fuel from the first stage reactors. The third stage envisages reactors which will employ thorium based fuel after its irradiation in the second stage reactors. This programme is sequential in nature and has an ultimate objective of securing long term energy security to India through judicial use of its thorium resources. Thorium based reactors offer advantages in terms of better neutronic characteristics of thorium, it being better fertile host for plutonium disposition and better thermo-mechanical properties and slower fuel deterioration of thorium oxide. It is planned to introduce thorium in the Indian Nuclear Power Programme after sufficient (about 200 GWe) capacity build-up in the second stage. DAE is a global leader in the development of the entire thorium fuel cycle. It has a mature technology for extraction of thorium and preparation of thoria pellets. It has long back carried out irradiation of thoria pellets in its research reactors and also in PHWRs, post irradiation examination and reprocessing of irradiated thoria, fabrication of 233 U based fuel. It has KAMINI - the world's only operating reactor employing 233 U as fuel. An Advanced Heavy Water Reactor (AHWR) has been designed as a technology demonstrator for large scale

Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

Energy Technology Data Exchange (ETDEWEB)

Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

2015-11-04

This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and

Nuclear waste transmutation and related innovative technologies

International Nuclear Information System (INIS)

2002-01-01

The main topics of the summer school meeting were 1. Motivation and programs for waste transmutation: The scientific perspective roadmaps; 2. The physics and scenarios of transmutation: The physics of transmutation and adapted reactor types. Impact on the fuel cycle and possible scenarios; 3. Accelerator driven systems and components: High intensity accelerators. Spallation targets and experiments. The sub critical core safety and simulation physics experiments; 4. Technologies and materials: Specific issues related to transmutation: Dedicated fuels for transmutation. Fuel processing - the role of pyrochemistry. Materials of irradiation. Lead/lead alloys. 5. Nuclear data: The N-TOF facility. Intermediate energy data and experiments. (orig./GL)

Study on reprocessing of uranium-thorium fuel with solvent extraction for HTGR

International Nuclear Information System (INIS)

Jiao Rongzhou; He Peijun; Liu Bingren; Zhu Yongjun

1992-08-01

A single cycle process by solvent extraction with acid feed solution is suggested. The purpose is to reprocess uranium-thorium fuel elements which are of high burn-up and rich of 232 U from HTGR (high temperature gas cooled reactor). The extraction cascade tests have been completed. The recovery of uranium and thorium is greater than 99.6%. By this method, the requirement, under remote control to re-fabricate fuel elements, of decontamination factors for Cs, Sr, Zr-Nb and Ru has been reached

Reprocessed and combined thorium fuel cycles in a PER system with a micro heterogeneous approaches

International Nuclear Information System (INIS)

Monteiro, Fabiana B.A.; Castro, Victor F.; Faria, Rochkhudson B. de; Pereira, Claubia; Fortini, Angela

2015-01-01

A micro heterogeneous approaches were used to study the behavior of reprocessed fuel spiked with thorium in a PWR fuel element considering (TRU-Th) cycle. The goal is to achieve a higher burnup using three different configurations to model the fuel element using SCALE 6.0. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a spent PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment. The spent fuel remained in the cooling pool for five years and then reprocessed using the UREX+ technique. Three configurations of micro heterogeneous approaches were analyzed, and the k inf and plutonium evolution during the burnup were evaluated. The preliminary results show that the behavior of advanced fuel based on transuranic elements spiked with thorium, and micro heterogeneous approach are satisfactory in PWRs, and the configuration that use a combination of Th and TRU (configuration 1) seems to be the most promising once has higher values for k inf during the burnup, compared with other configurations. (author)

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

Science.gov (United States)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Separation and transmutation. A picture of the applications in Sweden; Separation och transmutation. Belysning av tillaempning i Sverige

Energy Technology Data Exchange (ETDEWEB)

Grundfelt, Bertil; Lindgren, Maria [Kemakta Konsult AB, Stockholm (Sweden)

2006-11-15

utilised by the Swedish market. According to the IAEA joint convention on the safety of spent fuel management and on the safety of radioactive waste management it is required that the waste management should be performed such that undue burdens are not laid on future generations. The long time periods for implementing and performing transmutation imply a transfer of responsibility to future generations. Therefore, it can be questioned if transmutation as a strategy for the final management of the radioactive waste in a 'phase-out' scenario fulfills the requirements of the convention. Even if the separation and transmutation could be performed efficiently, waste that needs handling and final disposal will remain. Thus, transmutation is not an alternative to final disposal. Transmutation will require a continued production of nuclear power for in the order of 100 years or more. It could be a realistic alternative in a long-term perspective in which a significant fraction of the energy supply is based on new types of reactors replacing the current generation. Hence, long-term commitments to nuclear power requiring long-term political decisions are required.

Study on MAs transmutation of accelerator-driven system sodium-cooled fast reactor loaded with metallic fuel

International Nuclear Information System (INIS)

Han Song; Yang Yongwei

2007-01-01

Through the analysis of the effect of heavy metal actinides on the effective multiplication constant (k eff ) of the core in accelerator-driven system (ADS) sodium-cooled fast reactor loaded with metallic fuel, we gave the method for determining fuel components. the characteristics of minor actinides (MAs) transmutation was analyzed in detail. 3D burn-up code COUPLE, which couples MCNP4c3 and ORIGEN2, was applied to the neutron simulation and burn up calculation. The results of optimized scheme shows that adjusting the proportion of 239 Pu and maintaining the value during the burn-up cycle is an efficient method of designing k eff and keeping stable during the burn-up cycle. Spallation neutrons lead to the neutron spectrum harder at inner core than that at outer core. It is in favor of improving MA's fission cross sections and the capture-to-fission ratio. The total MAs transmutation support ratio 8.3 achieves excellent transmutation effect. For higher flux at inner core leads to obvious differences on transmutation efficiency,only disposing MAs at inner core is in favor of decreasing the loading mass and improving MAs transmutation effect. (authors)

Fuels and targets for incineration and transmutation of actinides: the ITU programme

International Nuclear Information System (INIS)

Fernandez, A.; Glatz, J.P.; Haas, D.; Konings, R.J.M.; Somers, J.; Toscano, E.; Walker, C.T.; Wegen, D.

2000-01-01

The ITU programme for the development of fuels and targets for transmutation of actinides is presented. The fabrication of various types of oxide fuels/targets by dust-free processes is described. Selected results of post-irradiation examinations of irradiation experiments (SUPERFACT, TRABANT-1, EFTTRA-T4) are presented to demonstrate the irradiation behaviour of these fuels/targets. Finally, the future developments at ITU in this field are described, including the new shielded facility (the MA lab) for fabrication of minor actinide fuels. (authors)

Fuels and targets for incineration and transmutation of actinides: the ITU programme

Energy Technology Data Exchange (ETDEWEB)

Fernandez, A.; Glatz, J.P.; Haas, D.; Konings, R.J.M.; Somers, J.; Toscano, E.; Walker, C.T.; Wegen, D. [Eurpean Commission, Joint Research Centre, Institute for Transuranium Elements, Kurlsruhe (Germany)

2000-07-01

The ITU programme for the development of fuels and targets for transmutation of actinides is presented. The fabrication of various types of oxide fuels/targets by dust-free processes is described. Selected results of post-irradiation examinations of irradiation experiments (SUPERFACT, TRABANT-1, EFTTRA-T4) are presented to demonstrate the irradiation behaviour of these fuels/targets. Finally, the future developments at ITU in this field are described, including the new shielded facility (the MA lab) for fabrication of minor actinide fuels. (authors)

A new concept for accelerator driven transmutation of nuclear wastes

International Nuclear Information System (INIS)

Arthur, E.D.

1991-01-01

A new concept for an accelerator-driven transmutation system is described. The central feature of the concept is generation of intense fluxes of thermal neutrons. In the system all long-lived radionuclides comprising high-level nuclear waste can be transmuted efficiently. Transmutation takes place in a unique, low material inventory environment. Presently two principal areas are being investigated for application of the concept. The first is associated with cleanup of defense high-level waste at DOE sites such as Hanford. The second, longer term area involves production of electric power using a coupled accelerator-multiplying blanket system. This system would utilize natural thorium or uranium and would transmute long-lived components of high-level waste concurrently during operation. 5 refs., 5 figs

Neutronic analysis of the PBMR-400 full core using thorium fuel mixed with plutonium or minor actinides

International Nuclear Information System (INIS)

Acır, Adem; Coşkun, Hasan

2012-01-01

Highlights: ► Neutronic calculations for PBMR 400 were conducted with the computer codes MCNP and MONTEBURNS 2.0. ► The criticality and burnup were investigated for reactor grade plutonium and minor actinides. ► We found that the use of these new fuels in PBMRs would reduce the nuclear waste repository significantly. -- Abstract: Time evolution of criticality and burnup grades of the PBMR were investigated for reactor grade plutonium and minor actinides in the spent fuel of light water reactors (LWRs) mixed with thoria. The calculations were performed by employing the computer codes MCNP and MONTEBURNS 2.0 and using the ENDF/B-V nuclear data library. Firstly, the plutonium–thorium and minor actinides–thorium ratio was determined by using the initial k eff value of the original uranium fuel design. After the selection of the plutonium/minor actinides–thorium mixture ratio, the time-dependent neutronic behavior of the reactor grade plutonium and minor actinides and original fuels in a PBMR-400 reactor was calculated by using the MCNP code. Finally, k eff , burnup and operation time values of the fuels were compared. The core effective multiplication factor (k eff ) for the original fuel which has 9.6 wt.% enriched uranium was computed as 1.2395. Corresponding to this k eff value the reactor grade plutonium/thorium and minor actinide/thorium oxide mixtures were found to be 30%/70% and 50%/50%, respectively. The core lives for the original, the reactor grade plutonium/thorium and the minor actinide/thorium fuels were calculated as ∼3.2, ∼6.5 and ∼5.5 years, whereas, the corresponding burnups came out to be 99,000, ∼190,000 and ∼166,000 MWD/T, respectively, for an end of life k eff set equal to 1.02.

Power level effects on thorium-based fuels in pressure-tube heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

2016-06-15

Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

Power level effects on thorium-based fuels in pressure-tube heavy water reactors

International Nuclear Information System (INIS)

Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

2016-01-01

Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

Development of Metallic Fuels for Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

2015-09-01

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

Nuclear waste transmutation

International Nuclear Information System (INIS)

Salvatores, M.; Girard, C.; Delpech, M.; Slessarev, I.; Tommasi, J.

1994-01-01

Waste management strategies foresee the use of a deep geological repository either for final disposal of irradiated fuel or, after reprocessing and reuse of U and Pu for final disposal of long-lived radio-active materials. In the second case, partitioning and transmutation of these materials can be considered to reduce the impact of radiation on man due to the storage. On the basis of the SPIN programme developed by CEA in this field, the main features of transmutation is presented. The goal to achieve and the criteria to use are quite difficult to establish. The rights para-meters to characterize the risk are the potential radiotoxicity in the the repository and the residual radiotoxicity at the outlet. Transmutation studies in CEA used the potential radiotoxicity which is based on well-known parameters and less precise hazardous factors. The second point to appreciate the trans- mutation interest is to dispose of a criteria for the radio-radiotoxicity reduction. As there is no general agreement, we try to have a toxicity as low as possible within reasonable technical limits. To reduce the long term radio- toxicity, Pu, minor actinides and some long-lived fission products have to be transmuted. To assess the feasibility of such trans-mutation in reactors or advanced systems, one has to consider constraints on neutronic balance, safety, fuel cycle, technology , economy. Taking in account the main conclusions of this analysis, parametric studies of homogeneous and heterogenous transmutation permit a choice of promising solutions. Goals are to use every long-lived element with a minimized production of other long- lived elements in order to obtain an appreciable radiotoxicity reduction. It implies multi recycling of Pu which favours fast neutron reactors and different strategies of multi recycling for Np, Am, Cm. Multi recycling makes the results strongly dependant of losses. Researches to obtain the high partitioning efficiency needed are in progress. Calculations

A Simplified Supercritical Fast Reactor with Thorium Fuel

OpenAIRE

Peng Zhang; Kan Wang; Ganglin Yu

2014-01-01

Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure ...

Role of accelerators in the Czech national transmuter project

International Nuclear Information System (INIS)

Bem, P.; Kugler, A.

1999-01-01

The problem of spent nuclear fuel from the so far operated PWRs has become a crucial issue in the Czech Republic. The first attempts to solve this problem by a final deposit of spent fuel into a suitable geological formation have been shown not to be fully acceptable. Therefore, the revival of nuclear transmutation technology application for nuclear incineration of nuclear waste and spent fuel in particular was welcomed. A realistic national project started to be developed in 1996. The four major nuclear research institutions of the country formed a consortium focused on an adoption of the world-wide experience and a development of a national project of a transmutation technology (experimental transmuter LA-0) or an efficient participation in the international effort in that field. Because the LA-0 transmuter concept of subcritical reactor with liquid fuel based on molten fluorides driven by an external neutron source has been adopted, the R and D effort has been focused on three regions. The first is devoted to the problem of a suitable neutron source, the second to a pre-conceptual design of a blanket for burning of actinides contained in spent fuel from PWRs. The third region is devoted to the utilisation of the experience from a specific field of dry (fluorine) reprocessing of spent fuel and a preparation of liquid fuel in the form of molten fluorides for the transmuter LA-0. (R.P.)

Proposed plan for critical experiments supporting thorium fuel cycle development

International Nuclear Information System (INIS)

Gore, B.F.

1978-09-01

A preliminary plan is proposed for critical experiments to provide data needed for the recycle of thorium based nuclear fuels. The sequence of experimentation starts with well moderated solutions followed by highly concentrated low moderated solutions. It then progresses through lattices moderated by water, by water plus soluble poisons, and by fissile solutions, to solutions poisoned by raschig rings and soluble poisons. Final experiments would treat lattices moderated by poisoned fissile solution, and arrays of stored fissile units

Burn-up calculations for a thorium HTR with one and with two types of fuel particle

Energy Technology Data Exchange (ETDEWEB)

Griggs, C. F.

1975-06-15

Cell burn-up calculations have been made on a thorium pin-cell operating with one or with two types of particle. With one particle, the input thorium and uranium are mixed prior to irradiation and all discharged uranium is recycled. With two particles, the fuel is kept in two streams and only the uranium generated from thorium is recycled. The two models are found to give similar power generations from a given initial U-235 input. The choice between the two types of particle is probably not determined by reactor physics considerations but by the value of the fuel credits and by the cost of fuel fabrication and reprocessing.

A review of reprocessing, partitioning, and transmutation of spent nuclear fuel and the implications for Canada

International Nuclear Information System (INIS)

Jackson, D.P.

2006-01-01

The current status of the reprocessing, partitioning, and transmutation of used nuclear fuel are reviewed in the context of assessing the possible application of these technologies to used CANDU fuel. The status of commercial reprocessing is briefly surveyed and recent progress in world R and D programs on the transmutation of FP's and actinides using Accelerator Driven Systems is summarized. The implications of reprocessing for Canada are explored from the point of view of a long strategy for managing used CANDU fuel in terms of the costs of initiating reprocessing domestically at some time in the future including public and occupational radiation doses, and the wastes generated. (author)

Waste arisings from a high-temperature reactor with a uranium-thorium fuel cycle

International Nuclear Information System (INIS)

1979-09-01

This paper presents an equilibrium-recycle condition flow sheet for a high-temperature gas-cooled reactor (HTR) fuel cycle which uses thorium and high-enriched uranium (93% U-235) as makeup fuel. INFCE Working Group 7 defined percentage losses to various waste streams are used to adjust the heavy-element mass flows per gigawatt-year of electricity generated. Thorium and bred U-233 are recycled following Thorex reprocessing. Fissile U-235 is recycled one time following Purex reprocessing and then is discarded to waste. Plutonium and other transuranics are discarded to waste. Included are estimates of volume, radioactivity, and heavy-element content of wastes arising from HTR fuel element fabrication; HTR operation, maintenance, and decommissioning; and reprocessing spent fuel where the waste is unique to the HTR fuel cycle

Accelerator molten-salt breeding and thorium fuel cycle

International Nuclear Information System (INIS)

Furukawa, Kazuo; Nakahara, Yasuaki; Kato, Yoshio; Ohno, Hideo; Mitachi, Kohshi.

1990-01-01

The recent efforts at the development of fission energy utilization have not been successful in establishing fully rational technology. A new philosophy should be established on the basis of the following three principles: (1) thorium utilization, (2) molten-salt fuel concept, and (3) separation of fissile-breeding and power-generating functions. Such philosophy is called 'Thorium Molten-Salt Nuclear Energy Synergetics [THORIMS-NES]'. The present report first addresses the establishment of 233 U breeding fuel cycle, focusing on major features of the Breeding and Chemical Processing Centers and a small molten-salt power station (called FUJI-II). The development of fissile producing breeders is discussed in relation to accelerator molten-salt breeder (AMSB), impact fusion molten-salt breeder, and inertial-confined fusion hybrid molten-salt breeder. Features of the accelerator molten-salt breeder are described, focusing on technical problems with accelerator breeders (or spallators), design principle of the accelerator molten-salt breeder, selection of molten salt compositions, and nuclear- and reactor-chemical aspects of AMSB. Discussion is also made of further research and development efforts required in the future for AMSB. (N.K.)

French fuel cycle strategy and partitioning and transmutation programme

International Nuclear Information System (INIS)

Pradel, Ph.

2007-01-01

The global energy context pleads in favour of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel sources. How we deal with radioactive waste is crucial in this context. The production of nuclear energy in France has been associated, since its inception, with the optimisation of radioactive waste management, including the partitioning and the recycling of recoverable energetic materials. The public's concern regarding long-term waste management led the French government to prepare and pass the Law of December 1991, requesting in particular the study for fifteen years of solutions to minimising even further the quantity and the hazardousness of final waste, via partitioning and transmutation. At the end of these fifteen years of research, it is considered that partitioning techniques which have been validated on real solutions are at disposal. Indeed, aqueous process for separation of minor actinides from the PUREX raffinate has been brought to a point where there is reasonable assurance that industrial deployment can be successful. A key experiment has been the kilogram-scale successful trials in the CEA-Marcoule Atalante facility in 2005 and this result, together with the results obtained in the frame of the successive European projects, constitutes a considerable step forward. For transmutation, CEA has conducted programmes proving the feasibility of the elimination of minor actinides and fission products: fabrication of specific targets and fuels for transmutation tests in the HFR and Phenix reactors, neutronics and technology studies for critical reactors and ADS developments. Scenario studies have also allowed assessing the feasibility, at the level of cycle and fuel facilities, and the efficiency of transmutation in terms of the quantitative reduction of the final waste inventory depending of

Thorium utilization

Energy Technology Data Exchange (ETDEWEB)

Trauger, D B [Oak Ridge National Lab., TN (USA)

1978-01-01

Some of the factors that provide incentive for the utilization of thorium in specific reactor types are explored and the constraints that stand in the way are pointed out. The properties of thorium and derived fuels are discussed, and test and reactor operating experience is reviewed. In addition, symbiotic systems of breeder and converter reactor are suggested as being particularly attractive systems for energy production. Throughout the discussion, the High-Temperature Gas-Cooled Reactor and Molten Salt Reactor are treated in some detail because they have been developed primarily for use with thorium fuel cycles.

Once-through thorium fuel cycle evaluation for TVA's Browns Ferry-3 Boiling Water Reactor

International Nuclear Information System (INIS)

Hopkins, G.C.

1982-05-01

This report documents benchmark evaluations to test thorium lattice predictive methods and neutron cross sections against available data and summarizes specific evaluations of the once-through thorium cycle when applied to the Browns Ferry-3 BWR. It was concluded that appreciable uncertainties in thorium cycle nuclear data cloud the ability to reliably predict the fuel cycle performance and that power reactor irradiations of ThO 2 rods in BWRs are desirable to resolve uncertainties. Benchmark evaluations indicated that the ENDF/B-IV data used in the evaluations should cause an underprediction of U-233/ThO 2 fuel reactivity, and, therefore, the results of the preliminary evaluations completed under the program should be conservative

Model for the behaviour of thorium and uranium fuels at pelletization

International Nuclear Information System (INIS)

Ferreira Neto, Ricardo Alberto

2000-11-01

In this work, a model for the behaviour of thorium-uranium-mixed oxide microspheres in the pelletizing process is presented. This model was developed in a program whose objective was to demonstrate the viability of producing fissile material through the utilization of thorium in pressurized water reactors. This is important because it allows the saving of the strategic uranium reserves, and makes it possible the nuclear utilization of the large brazilian thorium reserves. The objective was to develop a model for optimizing physical properties of the microspheres, such as density, fracture strength and specific surface, so as to produce fuel pellets with microstructure, density, open porosity and impurity content, in accordance with the fuel specification. And, therefore, to adjust the sol-gel processing parameters in order to obtain these properties, and produce pellets with an optimized microstructure, adequate to a stable behaviour under irradiation. The model made it clear that to achieve this objective, it is necessary to produce microspheres with density and specific surface as small as possible. By changing the sol-gel processing parameters, microspheres with the desired properties were produced, and the model was experimentally verified by manufacturing fuel pellets with optimized microstructures, density, open porosity and impurity content, meeting the specifications for this new nuclear fuel for pressurized water reactors. Furthermore it was possible to obtain mathematical expressions that enables to calculate from the microspheres properties and the utilized compaction pressure, the sinter density that will be obtained in the sintered pellet and the necessary compaction pressure to reach the sintered density specified for the fuel. (author)

Preliminary study on characteristics of equilibrium thorium fuel cycle of BWR

International Nuclear Information System (INIS)

Waris, A.; Kurniadi, R.; Su'ud, Z.; Permana, S.

2007-01-01

One of the main objectives behind the transuranium recycling ideas is not merely to utilize natural resource that is uranium much more efficiently, but to reduce the environmental impact of the radio-toxicity of the nuclear spent fuel. Beside uranium resource, there is thorium which has three times abundance compared to that of uranium which can be utilized as nuclear fuel. On top of that thorium is believed to have less radio-toxicity of spent fuel since its produce smaller amount of higher actinides compared to that of uranium. However, the studies on the thorium utilization in nuclear reactor in particular in light water reactors (LWR) are not performed intensively yet. Therefore, the aim of the present study is to evaluate the characteristics of thorium fuel cycle in LWR, especially boiling water reactor (BWR). To conduct the comprehensive investigations we have employed the equilibrium burnup model (1-3). The equilibrium burnup model is an alternative powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor1). We have employed 1368 nuclides in the equilibrium burnup calculation where 129 of them are heavy metals (HMs). This burnup code then is coupled with SRAC cell calculation code by using PIJ module to compose an equilibrium-cell burnup code. For cell calculation, 26 HMs, 66 fission products (FPs) and one pseudo FP have been utilized. The JENDL 3.2 library has been used in this study. References: 1. A. Waris and H. Sekimoto, 'Characteristics of several equilibrium fuel cycles of PWR', J. Nucl. Sci. Technol., 38, p.517-526, 2001 2. A. Waris, H. Sekimoto, and G. Kastchiev, Influence of Moderator-to-Fuel Volume Ratio on Pu and MA Recycling in Equilibrium Fuel Cycles of

Monte Carlo method in ADS transmutation reactor coolant and the research of optimal placement of the fuel

International Nuclear Information System (INIS)

Niu Yunlong; Wei Qianglin; Liu Yibao; Wang Aixing; Zhang Peng

2014-01-01

This paper calculated the effects of different coolants to neutron energy spectrum in different position of the transmutation reactor by Monte Carlo N-Particle Transport Code (MCNP5). After having chosen the coolant and particular parameters, different nuclides in fuel rods of the transmutation reactor were calculated and compared. According to the actual situation, nuclides of 99 Tc and 241 Am were chosen and compared. Then the nonuniform-arrangement scheme of different spent fuels were proposed. By comparison of the diagram, it is found that it is more effective to promote the neutron utilization in the reactor by the non-uniform arrangement scheme, which is more reasonable than traditional uniform one. Thus, it would be helpful for transmutation technology by the application of the scheme. (authors)

The measurements of critical mass with uranium fuel elements and thorium rods

International Nuclear Information System (INIS)

Yao Zhiquan; Chen Zhicheng; Yao Zewu; Ji Huaxiang; Bao Borong; Zhang Jiahua

1991-01-01

The critical experiments with uranium elements and Thorium rods have been performed in zero power reactor at Shanghai Institute of Nuclear Research. The critical masses have been measured in various U/Th ratios. The fuels are 3% 235 U-enriched uranium. The Thorium rods are made from power of ThF 4 . Ratios of calculated values to experimental values are nearly constant at 0.995

A Fast Numerical Method for the Calculation of the Equilibrium Isotopic Composition of a Transmutation System in an Advanced Fuel Cycle

Directory of Open Access Journals (Sweden)

F. Álvarez-Velarde

2012-01-01

Full Text Available A fast numerical method for the calculation in a zero-dimensional approach of the equilibrium isotopic composition of an iteratively used transmutation system in an advanced fuel cycle, based on the Banach fixed point theorem, is described in this paper. The method divides the fuel cycle in successive stages: fuel fabrication, storage, irradiation inside the transmutation system, cooling, reprocessing, and incorporation of the external material into the new fresh fuel. The change of the fuel isotopic composition, represented by an isotope vector, is described in a matrix formulation. The resulting matrix equations are solved using direct methods with arbitrary precision arithmetic. The method has been successfully applied to a double-strata fuel cycle with light water reactors and accelerator-driven subcritical systems. After comparison to the results of the EVOLCODE 2.0 burn-up code, the observed differences are about a few percents in the mass estimations of the main actinides.

All heavy metals closed-cycle analysis on water-cooled reactors of uranium and thorium fuel cycle systems

International Nuclear Information System (INIS)

Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Takaki, Naoyuki

2009-01-01

Uranium and Thorium fuels as the basis fuel of nuclear energy utilization has been used for several reactor types which produce trans-uranium or trans-thorium as 'by product' nuclear reaction with higher mass number and the remaining uranium and thorium fuels. The utilization of recycled spent fuel as world wide concerns are spent fuel of uranium and plutonium and in some cases using recycled minor actinide (MA). Those fuel schemes are used for improving an optimum nuclear fuel utilization as well to reduce the radioactive waste from spent fuels. A closed-cycle analysis of all heavy metals on water-cooled cases for both uranium and thorium fuel cycles has been investigated to evaluate the criticality condition, breeding performances, uranium or thorium utilization capability and void reactivity condition. Water-cooled reactor is used for the basic design study including light water and heavy water-cooled as an established technology as well as commercialized nuclear technologies. A developed coupling code of equilibrium fuel cycle burnup code and cell calculation of SRAC code are used for optimization analysis with JENDL 3.3 as nuclear data library. An equilibrium burnup calculation is adopted for estimating an equilibrium state condition of nuclide composition and cell calculation is performed for calculating microscopic neutron cross-sections and fluxes in relation to the effect of different fuel compositions, different fuel pin types and moderation ratios. The sensitivity analysis such as criticality, breeding performance, and void reactivity are strongly depends on moderation ratio and each fuel case has its trend as a function of moderation ratio. Heavy water coolant shows better breeding performance compared with light water coolant, however, it obtains less negative or more positive void reactivity. Equilibrium nuclide compositions are also evaluated to show the production of main nuclides and also to analyze the isotopic composition pattern especially

Radioactive Wastes Generated From JAERI Partitioning-Transmutation Fuel Cycle

International Nuclear Information System (INIS)

Nakayama, Shinichi; Morita, Yasuji; Nishihara, Kenji

2003-01-01

Production of lower-level radioactive wastes, as well as the reduction in radioactivity of HLW, is an important performance indicator in assessing the viability of a partitioning-transmutation system. We have begun to identify the chemical compositions and to quantify the amounts of radioactive wastes that may be generated by JAERI's processes. Long-lived radionuclides such as 14 C and 59 Ni and spallation products of Pb-Bi coolants are added to the existing inventory of these nuclides that are generated in the current fuel cycle. Spent salts of KCl-LiCl, which is not generated from the current fuel cycle, will be introduced as a waste. (authors)

Nuclear energy from thorium

International Nuclear Information System (INIS)

Coote, G.E.

1977-06-01

Relevant topics in nuclear and reactor physics are outlined. These include: the thorium decay series; generation of fissile from fertile nuclides, in particular U-233 from Th-232; the princiiples underlying thermal breeder reactors; the production of U-232 in thorium fuel and its important influence on nuclear safeguards and the recycling of U-233. Development work is continuing on several types of reactor which could utilise thorium; each of these is briefly described and its possible role is assessed. Other tipics covered include safety aspects of thorium oxide fuel, reprocessing, fabrication of recycle fuel and the possibility of denaturing U-233 by adding natural uranium. It is concluded that previoue arguments for development of the thorium cycle are still valid but those relating to non-proliferation of weapons may become even more compelling. (auth.)

Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

Science.gov (United States)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, k eff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

Science.gov (United States)

Sobolev, V.; Uyttenhove, W.; Thetford, R.; Maschek, W.

2011-07-01

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2-x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of "best estimates" provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

Thorium fuel-cycle development through plutonium incineration by THORIMS-NES (Thorium Molten-Salt nuclear energy synergetics)

International Nuclear Information System (INIS)

Furukawa, K.; Furuhashi, A.; Chigrinov, S.E.

1996-01-01

Thorium fuel-cycle has benefit on not-only trans-U element reduction but also their incineration. The disadvantage of high gamma activity of fuel, which is useful for improving the resistance to nuclear proliferation and terrorism, can overcome by molten fluorides fuel, and practically by THORIMS-NES, symbiotically coupled with fission Molten-Salt Reactor (FUJI) and fissile-producing Accelerator Molten-Salt Breeder (AMSB). This will have wide excellent advantages in global application, and will be deployed by incinerating Pu and Producing 233 U. Some details of this strategy including time schedule are presented. 14 refs, 2 figs, 4 tabs

Thorium utilization: conversion ratio and fuel needs in thermal reactors

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1975-01-01

As a preparatory study for thorium utilization in thermal reactors a study has been made of the fuel comsumption in existing reactor types. A quantitative description is given of the influence of enrichment, burnup, amount of structural material, choise of coolant and control requirements on the convertion ratio. The enrichment is an important factor and a low fuel comsumption can be achieved by increasing the enrichment

Thorium as an energy source. Opportunities for Norway; Thorium som energikilde - Muligheter for Norge

Energy Technology Data Exchange (ETDEWEB)

2008-02-15

Final Recommendations of the Thorium Report Committee: 1) No technology should be idolized or demonized. All carbon-dioxide (Co2) emission-free energy production technologies should be considered. The potential contribution of nuclear energy to a sustainable energy future should be recognized. 2) An investigation into the resources in the Fen Complex and other sites in Norway should be performed. It is essential to assess whether thorium in Norwegian rocks can be defined as an economical asset for the benefit of future generations. Furthermore, the application of new technologies for the extraction of thorium from the available mineral sources should be studied. 3) Testing of thorium fuel in the Halden Reactor should be encouraged, taking benefit of the well recognized nuclear fuel competence in Halden. 4) Norway should strengthen its participation in international collaborations by joining the EURATOM fission program and the GIF program on Generation IV reactors suitable for the use of thorium. 5) The development of an Accelerator Driven System (ADS) using thorium is not within the capability of Norway working alone. Joining the European effort in this field should be considered. Norwegian research groups should be encouraged to participate in relevant international projects, although these are currently focused on waste management. 6) Norway should bring its competence in waste management up to an international standard and collaboration with Sweden and Finland could be beneficial. 7) Norway should bring its competence with respect to dose assessment related to the thorium cycle up to an international standard. 8) Since the proliferation resistance of uranium-233 depends on the reactor and reprocessing technologies, this aspect will be of key concern should any thorium reactor be built in Norway. 9) Any new nuclear activities in Norway, e.g. thorium fuel cycles, would need strong international pooling of human resources, and in the case of thorium, a strong long

Partitioning and Transmutation - Physics, Technology and Politics

International Nuclear Information System (INIS)

Gudowski, W.

2002-01-01

Nuclear reactions can be effectively used to destroy radio toxic isotopes through transmutation processes transforming those isotopes into less radio toxic or stable ones Spent nuclear fuel, a mixture of many isotopes with some of them being highly radio toxic for many hundred thousands of years, may be effectively transmuted through nuclear reactions with neutrons. In a dedicated, well designed transmutation system one can, in principle, reduce the radiotoxicity of the spent nuclear fuel to a level, which will require isolation from the biosphere for the period of time for which engineered barriers can be constructed and licensed (not more than 1-2 thousands of years). En effective transmutation process can not be achieved without a suitable partitioning. Only partitioning of the spent nuclear fuel into predetermined groups of elements makes possible an effective use of neutrons to transmute long-lived radioactive isotopes into short-lived or stable one. However, most of the chemical separation/partitioning processes are element- not isotope-specific, therefore the transmutation of the elements with an existing isotope composition is a typical alternative for transmutation processes. Isotope-specific separation is possible but still very expensive and technologically not matured

Actinide and Fission Product Partitioning and Transmutation

International Nuclear Information System (INIS)

2015-06-01

The benefits of partitioning and transmutation (P and T) have now been established worldwide and, as a result, many countries are pursuing R and D programmes to advance the technologies associated with P and T. In this context, the OECD Nuclear Energy Agency (NEA) has organised a series of biennial information exchange meetings to provide experts with a forum to present and discuss state-of-the-art developments in the field of partitioning and transmutation since 1990. The OECD Nuclear Energy Agency Information Exchange Meeting on Actinides and Fission Products Partitioning and Transmutation is a forum for experts to present and discuss the state-of-the-art development in the field of P and T. Thirteen meetings have been organised so far and held in Japan, the United States, France, Belgium, Spain, the Republic of Korea and the Czech Republic. This 13. meeting was hosted by Seoul National University (Seoul, Republic of Korea) and was organised in co-operation with the International Atomic Energy Agency (IAEA) and the European Community (EC). The meeting covered strategic and scientific developments in the field of P and T such as: fuel cycle strategies and transition scenarios, the role of P and T in the potential evolution of nuclear energy as part of the future energy mix; radioactive waste management strategies; transmutation fuels and targets; advances in pyro and aqueous separation processes; P and T specific technology requirements (materials, spallation targets, coolants, etc.); transmutation systems: design, performance and safety; impact of P and T on the fuel cycle; fabrication, handling and transportation of transmutation fuels. A total of 103 presentations (39 oral and 64 posters) were discussed among the 110 participants from 19 countries and 2 international organisations. The meeting consisted of one plenary session where national and international programmes were presented followed by 5 technical sessions: - Fuel Cycle Strategies and Transition

Actinide nitride ceramic transmutation fuels for the Futurix-FTA irradiation experiment

International Nuclear Information System (INIS)

Voit, St.; McClellan, K.; Stanek, Ch.; Maloy, St.

2007-01-01

Full text of publication follows. The transmutation of plutonium and other minor actinides is an important component of an advanced nuclear fuel cycle. The Advanced Fuel Cycle Initiative (AFCI) is currently considering mono-nitrides as potential transmutation fuel material on account of the mutual solubility of actinide mono-nitrides as well as their desirable thermal characteristics. The feedstock is most commonly produced by a carbothermic reduction/nitridisation process, as it is for this programme. Fuel pellet fabrication is accomplished via a cold press/sinter approach. In order to allow for easier investigation of the synthesis and fabrication processes, surrogate material studies are used to compliment the actinide activities. Fuel compositions of particular interest denoted as low fertile (i.e. containing uranium) and non-fertile (i.e. not containing uranium) are (PuAmNp) 0.5 U 0.5 N and (PuAm) 0.42 Zr 0.58 N, respectively. The AFCI programme is investigating the validity of these fuel forms via Advanced Test Reactor (ATR) and Phenix irradiations. Here, we report on the recent progress of actinide-nitride transmutation fuel development and production for the Futurix-FTA irradiation experiment. Furthermore, we highlight specific cases where the complimentary approach of surrogate studies and actinide development aid in the understanding complex material issues. In order to allow for easier investigation of the fundamental materials properties, surrogate materials have been used. The amount of surrogate in each compound was determined by comparing both molar concentration and lattice parameter mismatch via Vegard Law. Cerium was chosen to simultaneously substitute for Pu, Am and Np, while depleted U was chosen to substitute for enriched U. Another goal of this work was the optimisation of added graphite during carbothermic reduction in order to minimise the duration of the carbon removal step (i.e. heat treatment under H 2 containing gas). One proposed

A road map for the realization of global-scale thorium breeding fuel cycle by single molten-fluoride flow

International Nuclear Information System (INIS)

Furukawa, K.; Arakawa, K.; Erbay, L. B.

2007-01-01

For global survival in this century, we urgently need to launch a completely new global nuclear fission industry. To get worldwide public acceptance of nuclear energy, improvements are essential not only on safety, radio-waste management and economy but also especially nuclear proliferation resistance and safeguards. However, such global fission industry cannot replace the present fossil fuel industry in the next 50 years, unless the doubling-time of nuclear energy is less than 10 years, preferably 5-7 years. Such a doubling-time cannot be established by any kind of classical 'Fission Breeding Power Station' concept. We need a symbiotic system which couples fission power reactors with a system which can convert fertile thorium to fissile U-233, such as a spallation or D/T fusion (if and when it becomes available). For such a purpose, THORIMS-NES [Thorium Molten-Salt Nuclear Energy Synergetic System] has been proposed, which is composed of simple thermal fission power stations (FUJI) and fissile producing Accelerator Molten-Salt Breeder (AMSB). Its system functions are very ambitious, delicate and complex, but can be realized in the form of simple hardware applying the multifunctional 'single-phase molten-fluoride' circulation system. This system has no difficulties relating with 'radiation-damage', 'heat-removal' and 'chemical processing' owing to the simple 'idealistic ionic liquid' character. FUJI is size-flexible (economical even in smaller sizes), fuel self-sustaining without any continuous chemical processing and without core-graphite replacement, and AMSB is based on a single-fluid molten-salt target/blanket concept, which solves most engineering difficulties such as radiation-damage, heat-removal etc., except high-current proton accelerator development. Several AMSBs are accommodated in the regional centers (several ten sites in the world) with batch chemical processing plants including radio-waste management. The integrated thorium breeding fuel cycle is

Feasibility study on the thorium fueled boiling water breeder reactor

International Nuclear Information System (INIS)

PetrusTakaki, N.

2012-01-01

The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

International Nuclear Information System (INIS)

Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

2006-01-01

A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

Energy Technology Data Exchange (ETDEWEB)

Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

2011-11-30

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

Uranium and thorium cycles for sodium fast reactors: Neutronic aspects and associated wastes

International Nuclear Information System (INIS)

Brizi, J.

2010-10-01

Sodium fast reactors (SFR-Na) with uranium 238/plutonium 239(U/Pu) cycle, its technical feasibility has already proven, allow to overcome the problem of natural uranium resources in achieving the regeneration of the fuel fissile element. In addition, a waste management can be performed to reduce the radiotoxicity of actinides produced by the reactor in transmuting the AM in the core (homogeneous transmutation). Another alternative to minimize waste is to use another couple fertile-fissile: the thorium 232 and the uranium 233 (Th/U). The comparison is performed on neutronic and safety aspects and on waste production, in using an evolutive Monte Carlo. Although one does not disclose real clear advantages concerning the radiotoxicity of wastes for a particular cycle, the Th/U cycle reduces the radiotoxicity during periods when it is the highest. The homogeneous transmutation minimizes significantly for both cycles, radiotoxicity of wastes, with different factors depending on the considered time period. However, it is done to the detriment of an important increase of AM in the core. If we consider the nuclear stop, the inventory of the reactor core becomes a waste. The gain provided by the transmutation, taking into account both the core and accumulated waste radio-toxicities, will be quantified, and shows the transmutation does not provide a significant gain if the burning of main fissile elements is not considered when the nuclear is stopped. (author)

1. round table - Spent fuels composition. Back-end of the fuel cycle and reprocessing, plutonium and other nuclear materials management. 2. round table - Separation-transmutation. 3. round table - Scenarios for a long term inventory of nuclear materials and wastes; 1. table ronde - La composition des combustibles uses. L'aval du combustible et le retraitement, la gestion du plutonium et des autres matieres nucleaires. 2. table ronde - Separation-transmutation. 3. table ronde - Scenarii pour un inventaire des matieres et des dechets nucleaires a LT

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. The debate comprises 4 public hearings (September 2005: Bar-le-Duc, Saint-Dizier, Pont-du-Gard, Cherbourg), 12 round-tables (October and November 2005: Paris, Joinville, Caen, Nancy, Marseille), a synthesis meeting (December 2005, Dunkerque) and a closing meeting (January 2006, Lyon). This document is the synthesis of the round table debates which took place at Paris on the reprocessing of spent fuels. Three aspects are discussed: the risks linked with the recovery of valorizable materials, the economical viability of the separation/transmutation option, and the future of the nuclear option in the French energy policy. Six presentations (transparencies) are attached with these proceedings which treat of: the reprocessing/recycling to the test, perspectives of future wastes, present day wastes/valorizable materials and future scenarios, critical analysis scenarios, why reprocessing spent fuels?, processing of spent fuels and recycling, separation and transmutation of long-lived radioactive wastes, thorium-uranium cycle. (J.S.)

Thorium base fuels reprocessing at the L.P.R. (Radiochemical Processes Laboratory) experimental plant

International Nuclear Information System (INIS)

Almagro, J.C.; Dupetit, G.A.; Deandreis, R.A.

1987-01-01

The availability of the LPR (Radiochemical Processes Laboratory) plant offers the possibility to demonstrate and create the necessary technological basis for thorium fuels reprocessing. To this purpose, the solvents extraction technique is used, employing TBP (at 30%) as solvent. The process is named THOREX, a one-cycle acid, which permits an adequate separation of Th 232 and U 233 components and fission products. For thorium oxide elements dissolution, the 'chopp-leach' process (installed at LPR) is used, employing a NO 3 H 13N, 0.05M FH and 0.1M Al (NO 3 ) 3 , as solvent. To adapt the pilot plant to the flow-sheet requirements proposed, minor modifications must be carried out in the interconnection of the existing decanting mixers. The input of the plant has been calculated by Origin Code modified for irradiations in reactors of the HWR type. (Author)

Analysis of the thorium inclusion in the fuel of a fast reactor cooled by lead

International Nuclear Information System (INIS)

Juarez M, L. C.; Francois L, J. L.

2017-09-01

In the present work, we first verified a model of the European reactor cooled with lead (ELFR). The calculations were made with the code Monte Carlo serpent 2.27 and the library of cross sections Jeff-3.1. For this verification, three neutron parameters were compared: the evolution of the neutron multiplication factor, the Doppler constant and the effect of the vacuum fraction of the refrigerant, obtaining a good approximation with the reference values. Subsequently, the inclusion of thorium as a fertile material within the fuel was analyzed and the same neutron parameters were compared with the original fuel. The evolution of criticality for the case of thorium fuel differs significantly with respect to that of the original fuel (without thorium); this is due mainly to the breeding of the fissile isotope 233 U. Therefore, is possible to have a longer fuel cycle, favoring the availability factor of the plant, without compromising the performance of the reactor since both the Doppler constant and the effect of the vacuum fraction of the refrigerant show a similar tendency to those of the original fuel, being negative in both cases. (Author)

Status of nuclear transmutation study

International Nuclear Information System (INIS)

Takizuka, Takakazu

1999-01-01

JAERI is carrying out R and Ds on partitioning and transmutation under the OMEGA Program. The R and Ds include the design study of accelerator-driven transmutation systems and the development of transmutation experimental facilities. Accelerator-driven systems have received much interests due to their potential role as dedicated transmuters in the nuclear fuel cycle for minimizing long-lived waste. Principles of accelerator-driven system, its history, JAERI proposed system concepts, and the experimental program are overviewed. (author)

Data base for a CANDU-PHW operating on the thorium cycle

International Nuclear Information System (INIS)

1979-07-01

This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on various thorium cycles. In all these cycles thorium is the main fertile component of the fuel and all assume recycling of the uranium component. In the reference cycle, the requirements for externally supplied fissile material are met using U-235, with the feed adjusted to provide a fuel burnup of approximately 30,000 MW.d/t(U). Two versions of the reference cycle are treated. In one, the U-235 is supplied in a highly enriched form (93% U-235 in uranium); in the other, the U-235 is supplied at a lower enrichment, such that the uranium present in the feed fuel is ''denatured''. The effects of varying the fuel burnup and the recycle delay time are discussed for the reference cases. Data are also given for thorium cycles using plutonium instead of U-235 to meet requirements for externally supplied fissile material. The special case of ''self sufficient equilibrium thorium cycles'', which require no external source of fissile material for equilibrium operation, is also treated

The cross sections of reactions resulting in transmutation of long-lived radionuclides of exhausted nuclear fuel exposed to fast neutrons

International Nuclear Information System (INIS)

Konodeev, A.Yu.; Korovin, Yu.A.; Erview, K.

1993-01-01

Research is at present concerned with the possible transmutation of long-lived radionuclides of spent nuclear fuel in the flux of fast neutrons from neutron generators which are distinguished by their energy spectrum and density of the flux generated. For this purpose one must know the cross sections of the nuclear reactions resulting in the transmutation and formation of new long-lived radionuclides due to the irradiation. The transmutation rate of radioisotope irradiated with neutrons have a known energy spectrum is determined by calculating the transmutation cross section which is equal to the sum of the cross sections of neutron reactions causing conversion of a particular isotope into another after the decay of short-lived residual nuclei. The presently available neutron cross section data of long-lived radionuclides, i.e., the products of the fission of nuclear fuel, are insufficient for research on these effects as transmutations occur in a flux of high-energy neutrons. This paper presents the cross sections of reactions accounting for the transmutation of the most important long-lived radionuclides of exhausted nuclear fuel during its irradiation with neutron having energies of up to 100 MeV. The neutron cross sections were calculated for 79 Se, 90 Sr, 93 Zr, 99 Tc, 107 Pd, 126 Sn, 129 I, 135 Cs and 137 Cs with a half-life ≥30 years

EFTTRA, a European collaboration for the development of fuels and targets for the transmutation

International Nuclear Information System (INIS)

Babelot, J.F.; Muehling, G.; Prunier, C.; Rome, M.

1994-12-01

In the frame of the research programmes on the transmutation of long lived nuclides, many experimental or theoretical investigations have to be carried out within European collaborations, owing mainly to the costs of such studies. Therefore, a group named 'Experimental Feasibility of Targets for Transmutation' (EFTTRA), has been formed, with participants from CEA (France). ECN (The Netherlands), EDF (France), KFK (Germany) and ITU (European Commission), to organise joint experiments for the study of materials for the transmutation. So far, it was decided to focus the work on the transmutation of 99 Tc (metal), of 129 I (compound), and of Am (in an inert matrix). Irradiations will take place in parallel in the Phenix fast reactor in France, and in the high flux thermal reactor HFR in the Netherlands. These experiments, together with the related post-irradiation examinations, constitute the first phase of the EFTTRA collaboration. In subsequent phases, EFTTRA will contribute to the development of fuels and targets. (orig.)

Analysis of thorium and uranium fuel cycles in an iso-breeder lead fast reactor using extended-EQL3D procedure

International Nuclear Information System (INIS)

Fiorina, Carlo; Krepel, Jiri; Cammi, Antonio; Franceschini, Fausto; Mikityuk, Konstantin; Ricotti, Marco Enrico

2013-01-01

Highlights: ► Extension of EQL3D procedure to calculate radio-toxicity and decay heat. ► Characterization of uranium- and thorium-fueled LFR from BOL to equilibrium. ► Safety improvements for a LFR in a closed thorium cycle. ► Advantages of thorium-fueled LFR in terms of decay heat and radio-toxicity generation. ► Safety, decay heat and radio-toxicity concerns for a Th–Pu beginning-of-life core. - Abstract: Use of thorium in fast reactors has typically been considered as a secondary option, mainly thanks to a possible self-sustaining thorium cycle already in thermal reactors and due to the limited breeding capabilities compared to U–Pu in the fast neutron energy range. In recent years nuclear waste management has become more important, and the thorium option has been reconsidered for the claimed potential to burn transuranic waste and the lower build-up of hazardous isotopes in a closed cycle. To ascertain these claims and their limitations, the fuel cycle isotopic inventory, and associated waste radio-toxicity and decay heat, should be quantified and compared to the case of the uranium cycle using realistic core configurations, with complete recycle of all the actinides. Since the transition from uranium to thorium fuel cycles will likely involve a transuranic burning phase, this transition and the challenges that the evolving fuel actinide composition presents, for instance on reactor feedback parameters, should also be analyzed. In the present paper, these issues are investigated based on core physics analysis of the Lead-cooled Fast Reactor ELSY, performed with the fast reactor ERANOS code and the EQL3D procedure allowing full-core characterization of the equilibrium cycle and the transition cycles. In order to compute radio-toxicity and decay heat, EQL3D has been extended by developing a new module, which has been assessed against ORIGEN-S and is presented here. The capability of the EQL3D procedure to treat full-core 3D geometries allowed to

Transmutation of actinides in power reactors.

Science.gov (United States)

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

Small molten-salt reactors with a rational thorium fuel-cycle

International Nuclear Information System (INIS)

Furukawa, Kazuo; Mitachi, Kohshi; Kato, Yoshio

1992-01-01

In the fission-energy utilization for solving global social and environmental problems including the 'Greenhouse Effect' in the next century, a new strategy should be introduced considering high safety and economy, simplicity, size-flexibility, anti-nuclear proliferation and terrorism, high temperature heat supply, etc., aiming to establish a rational breeding fuelcycle. Thorium Molten-Salt Nuclear Energy Synergetics based on [I] Th utilization, [II] fluid-fuel concept and [III] separation of fissile breeding and power generation functions would be one of the most promising approach. A design study of a standard Molten-Salt Reactor: FUJI-II (350 MWth, 155-161 MWe) ensuring fuel self-sustaining nature (conversion-ratio ∝ 1.0) in spite of small-size, and pilot-plant miniFUJI-II has been proceeded. (orig.)

Development of nitride fuel and pyrochemical process for transmutation of minor actinides

International Nuclear Information System (INIS)

Arai, Yasuo; Akabori, Mitsuo; Minato, Kazuo; Uno, Masayoshi

2010-01-01

Nitride fuel cycle for transmutation of minor actinides has been investigated under the double-strata fuel cycle concept. Mononitride solid solutions containing minor actinides have been prepared and characterised. Thermo-physical properties, such as thermal expansion, heat capacity and thermal diffusivity, have been measured by use of minor actinide nitride and burn-up simulated nitride samples. Irradiation behaviour of nitride fuel has been examined by irradiation tests. Pyrochemical process for treatment of spent nitride fuel has been investigated mainly by electrochemical measurements and nitride formation behaviour in pyrochemical process has been studied for recycled fuel fabrication. Recent results of experimental study on nitride fuel and pyrochemical process are summarised in the paper. (authors)

Minor actinide transmutation in accelerator driven systems

Energy Technology Data Exchange (ETDEWEB)

Friess, Friederike [IANUS, TU Darmstadt (Germany)

2015-07-01

Transmutation of radioactive waste, the legacy of nuclear energy use, gains rising interest. This includes the development of facilities able to transmute minor actinides (MA) into stable or short-lived isotopes before final disposal. The most common proposal is to use a double-strata approach with accelerator-driven-systems (ADS) for the efficient transmutation of MA and power reactors to dispose plutonium. An ADS consists of a sub-critical core that reaches criticality with neutrons supplied by a spallation target. An MCNP model of the ADS system Multi Purpose Research Reactor for Hightech Applications will be presented. Depletion calculations have been performed for both standard MOX fuel and transmutation fuel with an increased content of minor actinides. The resulting transmutation rates for MAs are compared to published values. Special attention is given to selected fission products such as Tc-99 and I-129, which impact the radiation from the spent fuel significantly.

The advanced fuel cycle initiative: the future path for advanced spent fuel treatment and transmutation research in the United States

International Nuclear Information System (INIS)

Herczeg, J.W.

2003-01-01

The U. S. Department of Energy (DOE) has invested over USD 100 million in transmutation research and development over the past three years. The programme has evolved from an accelerator based transmutation programme to a multi-tier reactor and accelerator based programme. These changes have resulted in a significant re-focus of the research and development programme as well as a name change to reflect the new direction. The Advanced Accelerator Application (AAA) programme is now renamed the Advanced Fuel Cycle Initiative (AFCI). Research completed by the AAA programme in Fiscal Year 2002 points to a multi-phased AFCI Programme consisting of two elements that would be conducted in parallel as part of an integrated research effort: an intermediate-term technology element (AFCI Series One), which emphasises advanced technical enhancements to the current commercial nuclear power infrastructure; and a long term technology element (AFCI Series Two), which will require the introduction of next-generation nuclear energy systems to reduce the toxicity of nuclear waste. (author)

Sensitivity analysis of minor actinides transmutation to physical and technological parameters

International Nuclear Information System (INIS)

Kooyman, T.; Buiron, L.

2015-01-01

Minor actinides transmutation is one of the 3 main axis defined by the 2006 French law for management of nuclear waste, along with long-term storage and use of a deep geological repository. Transmutation options for critical systems can be divided in two different approaches: (a) homogeneous transmutation, in which minor actinides are mixed with the fuel. This exhibits the drawback of 'polluting' the entire fuel cycle with minor actinides and also has an important impact on core reactivity coefficients such as Doppler Effect or sodium void worth for fast reactors when the minor actinides fraction increases above 3 to 5% depending on the core; (b) heterogeneous transmutation, in which minor actinides are inserted into transmutation targets which can be located in the center or in the periphery of the core. This presents the advantage of decoupling the management of the minor actinides from the conventional fuel and not impacting the core reactivity coefficients. In both cases, the design and analyses of potential transmutation systems have been carried out in the frame of Gen IV fast reactor using a 'perturbation' approach in which nominal power reactor parameters are modified to accommodate the loading of minor actinides. However, when designing such a transmutation strategy, parameters from all steps of the fuel cycle must be taken into account, such as spent fuel heat load, gamma or neutron sources or fabrication feasibility. Considering a multi-recycling strategy of minor actinides, an analysis of relevant estimators necessary to fully analyze a transmutation strategy has been performed in this work and a sensitivity analysis of these estimators to a broad choice of reactors and fuel cycle parameters has been carried out. No threshold or percolation effects were observed. Saturation of transmutation rate with regards to several parameters has been observed, namely the minor actinides volume fraction and the irradiation time. Estimators of interest that have been

Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

International Nuclear Information System (INIS)

Permana, Sidik

2009-01-01

Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

Current Status of the Transmutation Reactor Technology and Preliminary Evaluation of Transmutation Performance of the KALIMER Core

Energy Technology Data Exchange (ETDEWEB)

Hong, Ser Gi; Sim, Yoon Sub; Kim, Yeong Il; Kim, Young Gyum; Lee, Byung Woon; Song, Hoon; Lee, Ki Bog; Jang, Jin Wook; Lee, Dong Uk

2005-08-15

Recently the most countries using the nuclear power plants for electricity generation have been faced with the problem of the preparation of the repository for the disposition of the nuclear waste generated from LWR. It was well-known that the issues related with long term risk of the radioactive wastes for the future generations are due only to 1% of the total waste. This small fraction of 1% consists of transuranic (TRU) nuclides such as Pu, Np, Am, Cm and the long lived fission products such as Tc and I. For the transuranic (TRU) nuclides, their half lives range from several years to several hundred thousands years and hence their radioactive toxicity can be lasted over very long time period. This has made the change of the rule of the fast spectrum reactor from the economical use of uranium resource through breeding to the reduction of the nuclear waste through the transmutation. The purpose of this study is to obtain the basic knowledge on the nuclear transmutation technology and to suggest the technical solution ways for the future technology development and enhancement through a survey of the state-of-art of the international research on the nuclear transmutation. The increase of the transmutation rate requires the reduction of the breeding ratio. In fact, the transmutation rate is determined by the breeding ratio. The reduction of the breeding ratio can be achieved by reducing the U-238 content in fuel or increasing the neutron leakage through core boundary or absorbing the neutrons by using some absorbers. However, the reduction of the U-238 content results in the degradation of the fuel Doppler coefficient that is one of the most important safety-related parameters and the reduction of the effective delayed neutron fraction that is related with the controllability of the reactor core. Also, the increase of the transmutation rate can lead to the increase of the coolant void reactivity worth unless some ways to reduce the coolant void reactivity are not

Design considerations and evaluations of an accelerator-driven fluid fuel transmuter

Energy Technology Data Exchange (ETDEWEB)

Lizana, P.; Lypsch, F.; Phlippen, P.W. [Institute for Safety Research and Reactor Technology, Juelich (Germany)

1995-10-01

A fluid fuel transmuter is proposed on the basis of circulating lead forming the fluid carrier material for long-lived actinides. Thermalization of neutrons is achieved by the use of graphite in the blanket leading to low actinide concentrations, typically around 100 g/l. An eigenvalue of 0.95 is aimed at and the extraneous source neutrons are provided by the interaction of 1.6 GeV protons with a central lead target (spallation process). Fuel depletion and neutron transport calculations are discussed with a view to the technical feasibility and possible advantageous design modifications.

Comparison of two thorium fuel cycles for use in light water prebreeder/breeder reactor systems (AWBA Development Program)

International Nuclear Information System (INIS)

Merriman, F.C.; McCoy, D.F.; Boyd, W.A.; Dwyer, J.R.

1983-05-01

Light water prebreeder/breeder conceptual reactor systems have been developed which have the potential to significantly improve the fuel utilization of present generation light water reactors. The purpose of this study is to describe and compare two possible types of thorium fuel cycles for use in these light water prebreeder and breeder concepts. The two types of thorium fuel cycles basically differ in the fuel rod design used in the prebreeder cores and the uranium isotopic concentration of fuel supplied to the breeder cores

Transmutation in ASTRID

International Nuclear Information System (INIS)

Grouiller, Jean-Paul; Buiron, Laurent; Mignot, Gérard; Palhier, Raphael

2013-01-01

Summary and future prospects for incorporating Am in ASTRID: → Potential to demonstrate the minor actinide transmutation on an industrial scale in the CFV V1 core of ASTRID: • Homogeneous concept: 2% of Am in a standard fuel; • Heterogeneous concept: 10% on UO 2 in the radial blanket. • The objective of ensuring a balance in the Am (and total minor actinides) flow in the ASTRID fuel cycle may be obtained without any impact on the design of the core and handling systems for the management of the new and spent fuel subassemblies. • Several experimental phases in ASTRID to implement different transmutation scenarios using homogeneous and heterogeneous concepts. ⇒ the availability of facilities involved in the ASTRID material cycles

Maximization of burning and/or transmutation (B/T) capacity in coupled spectrum reactor (CSR) by fuel and core adjustment

International Nuclear Information System (INIS)

Aziz, F.; Kitamoto, Asashi.

1996-01-01

A conceptual design of burning and/or transmutation (B/T) reactor, based on a modified conventional 1150 MWe-PWR system, consisted of two core regions for thermal and fast neutrons, respectively, was proposed herein for the treatments of minor actinides (MA). In the outer region 237 Np, 241 Am, and 243 Am burned by thermal neutrons, while in the inner region 244 Cm was burned mainly by fast neutrons. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio. The maximization of B/T capacity in CSR were done by, first, increasing the radius of the inner region. Second, reducing the coolant to fuel volume ratio, and third, choosing a suitable B/T fuel type. The result of the calculations showed that the equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute up to 808 kg of MA in a single reactor core effectively and safely. (author)

Utilization of thorium in thermal reactors

International Nuclear Information System (INIS)

Srinivasan, K.R.; Nakra, A.N.

1978-01-01

Large deposits of thorium are found in India. 233 U produced by neutron capture in 232 Th is a more valuable fuel for thermal reactors than the plutonium that results from capture in 238 U. These two facts are the main reasons for the interest in utilizing thorium in power reactors. But natural thorium does not contain any fissile material and its capture cross section is nearly two and a half times that of 238 U. These have made the fuelling cost high. However, in certain conditions and certain types of reactors the costs are comparable with those using uranium fuel. The relative cost effectiveness of different fuels is discussed. Apart from long term interest, the short term interest of using thorium fuel in RAPP type reactors is also briefly described. Finally the reactor physics experiments using thorium fuel and their comparison with calculations are presented. (author)

A general overview of generation IV molten salt reactor (MSR) and the use of thorium as fuel

Energy Technology Data Exchange (ETDEWEB)

Yamaguchi, Carlos H.; Stefani, Giovanni L.; Santos, Thiago A., E-mail: carlos.yamaguchi@usp.br, E-mail: giovanni.stefani@ipen.br, E-mail: thiago.santos@ufabc.edu.br [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Fisica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

2017-07-01

The molten salt reactors (MSRs) make use of fluoride salt as primary cooler, at low pressure. Although considered a generation IV reactor, your concept isn't new, since in the 1960 years the Oak Ridge National Laboratory created a little prototype of 8MWt. Over the 20{sup th} century, other countries, like UK, Japan, Russia, China and France also did research in the area, especially with the use of thorium as fuel. This goes with the fact that Brazil possess the biggest reserve of thorium in the world. In the center of nuclear engineering at IPEN is being created a study group connected to thorium reactors, which purpose is to investigate reactors using thorium to produce {sup 233}U and tailing burn, thus making the MSR using thorium as fuel, an object of study. This present work searches to do a general summary about the researches of MSR's, having as focus the utilization of thorium with the goal being to show it's efficiency and utilization is doable. (author)

Nuclear data covariances and sensitivity analysis, validation of a methodology based on the perturbation theory; application to an innovative concept: the molten thorium salt fueled reactor; Analyses de sensibilite et d'incertitude de donnees nucleaires. Contribution a la validation d'une methodologie utilisant la theorie des perturbations; application a un concept innovant: reacteur a sels fondus thorium a spectre epithermique

Energy Technology Data Exchange (ETDEWEB)

Bidaud, A

2005-10-15

Neutron transport simulation of nuclear reactors is based on the knowledge of the neutron-nucleus interaction (cross-sections, fission neutron yields and spectra...) for the dozens of nuclei present in the core over a very large energy range (fractions of eV to several MeV). To obtain the goal of the sustainable development of nuclear power, future reactors must have new and more strict constraints to their design: optimization of ore materials will necessitate breeding (generation of fissile material from fertile material), and waste management will require transmutation. Innovative reactors that could achieve such objectives - generation IV or ADS (accelerator driven system) - are loaded with new fuels (thorium, heavy actinides) and function with neutron spectra for which nuclear data do not benefit from 50 years of industrial experience, and thus present particular challenges. After validation on an experimental reactor using an international benchmark, we take classical reactor physics tools along with available nuclear data uncertainties to calculate the sensitivities and uncertainties of the criticality and temperature coefficient of a thorium molten salt reactor. In addition, a study based on the important reaction rates for the calculation of cycle's equilibrium allows us to estimate the efficiency of different reprocessing strategies and the contribution of these reaction rates on the uncertainty of the breeding and then on the uncertainty of the size of the reprocessing plant. Finally, we use this work to propose an improvement of the high priority experimental request list. (author)

Thorium: Issues and prospects in Malaysia

Energy Technology Data Exchange (ETDEWEB)

AL-Areqi, Wadeeah M.; Majid, Amran Ab.; Sarmani, Sukiman; Bahri, Che Nor Aniza Che Zainul [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 Bangi, Malaysia. walareqi@yahoo.com (Malaysia)

2015-04-29

In Malaysia, thorium exists in minerals and rare earth elements production residue. The average range of thorium content in Malaysian monazite and xenotime minerals was found about 70,000 and 15,000 ppm respectively. About 2,636 tonnes of Malaysian monazite was produced for a period of 5 years (2006-2010) and based on the above data, it can be estimated that Malaysian monazite contains about 184.5 tonnes of thorium. Although thorium can become a major radiological problem to our environment, but with the significant deposit of thorium in Malaysian monazite, it has a prospect as a future alternative fuel in nuclear technology. This paper will discuss the thorium issues in Malaysia especially its long term radiological risks to public health and environment at storage and disposal stages, the prospect of exploring and producing high purity thorium from our rare earth elements minerals for future thorium based reactor. This paper also highlights the holistic approach in thorium recovery from Malaysian rare earth element production residue to reduce its radioactivity and extraction of thorium and rare earth elements from the minerals with minimum radiological impact to health and environment.

Thorium: Issues and prospects in Malaysia

International Nuclear Information System (INIS)

AL-Areqi, Wadeeah M.; Majid, Amran Ab.; Sarmani, Sukiman; Bahri, Che Nor Aniza Che Zainul

2015-01-01

In Malaysia, thorium exists in minerals and rare earth elements production residue. The average range of thorium content in Malaysian monazite and xenotime minerals was found about 70,000 and 15,000 ppm respectively. About 2,636 tonnes of Malaysian monazite was produced for a period of 5 years (2006-2010) and based on the above data, it can be estimated that Malaysian monazite contains about 184.5 tonnes of thorium. Although thorium can become a major radiological problem to our environment, but with the significant deposit of thorium in Malaysian monazite, it has a prospect as a future alternative fuel in nuclear technology. This paper will discuss the thorium issues in Malaysia especially its long term radiological risks to public health and environment at storage and disposal stages, the prospect of exploring and producing high purity thorium from our rare earth elements minerals for future thorium based reactor. This paper also highlights the holistic approach in thorium recovery from Malaysian rare earth element production residue to reduce its radioactivity and extraction of thorium and rare earth elements from the minerals with minimum radiological impact to health and environment

Thorium: Issues and prospects in Malaysia

Science.gov (United States)

AL-Areqi, Wadeeah M.; Majid, Amran Ab.; Sarmani, Sukiman; Bahri, Che Nor Aniza Che Zainul

2015-04-01

In Malaysia, thorium exists in minerals and rare earth elements production residue. The average range of thorium content in Malaysian monazite and xenotime minerals was found about 70,000 and 15,000 ppm respectively. About 2,636 tonnes of Malaysian monazite was produced for a period of 5 years (2006-2010) and based on the above data, it can be estimated that Malaysian monazite contains about 184.5 tonnes of thorium. Although thorium can become a major radiological problem to our environment, but with the significant deposit of thorium in Malaysian monazite, it has a prospect as a future alternative fuel in nuclear technology. This paper will discuss the thorium issues in Malaysia especially its long term radiological risks to public health and environment at storage and disposal stages, the prospect of exploring and producing high purity thorium from our rare earth elements minerals for future thorium based reactor. This paper also highlights the holistic approach in thorium recovery from Malaysian rare earth element production residue to reduce its radioactivity and extraction of thorium and rare earth elements from the minerals with minimum radiological impact to health and environment.

The radiological risks associated with the thorium fuelled HTGR fuel cycle. A comparative risk evaluation

International Nuclear Information System (INIS)

Dodd, D.H.; Hienen, J.F.A. van.

1995-10-01

This report presents the results of task B.3 of the 'Technology Assessment of the High Temperature Reactor' project. The objective of task B.3 was to evaluate the radiological risks to the general public associated with the sustainable HTGR cycle. Since the technologies to be used at several stages of this fuel cycle are still in the design phase and since a detailed specification of this fuel cycle has not yet been developed, the emphasis was on obtaining a global impression of the risk associated with a generic thorium-based HTGR fuel cycle. This impression was obtained by performing a comparative risk analysis on the basis of data given in the literature. As reference for the comparison a generic uranium fuelled LWR cycle was used. The major benefit with respect to the radiological rsiks of basing the fuel cycle around modular HTGR technology instead of the LWR technology is the increase in reactor safety. The design of the modular HTGR is expected to prevent the release of a significant amount of radioactive material to the environment, and hence early deaths in the surrounding population, during accident conditions. This implies that there is no group risk as defined in the Dutch risk management policy. The major benefit of thorium based fuel cycles over uranium based fuel cycles is the reduction in the radiological risks from unraium mining and milling. The other stages of the nuclear fuel cycle which make a significant contribution to the radiological risks are electricity generation, reprocessing and final disposal. The risks associated with the electricity generation stage are dominated by the risks from fission products, activated corrosion products and the activation products tritium and carbon-14. The risks associated with the reprocessing stage are determined by fission and activation products (including actinides). (orig./WL)

The radiological risks associated with the thorium fuelled HTGR fuel cycle. A comparative risk evaluation

Energy Technology Data Exchange (ETDEWEB)

Dodd, D.H.; Hienen, J.F.A. van

1995-10-01

This report presents the results of task B.3 of the `Technology Assessment of the High Temperature Reactor` project. The objective of task B.3 was to evaluate the radiological risks to the general public associated with the sustainable HTGR cycle. Since the technologies to be used at several stages of this fuel cycle are still in the design phase and since a detailed specification of this fuel cycle has not yet been developed, the emphasis was on obtaining a global impression of the risk associated with a generic thorium-based HTGR fuel cycle. This impression was obtained by performing a comparative risk analysis on the basis of data given in the literature. As reference for the comparison a generic uranium fuelled LWR cycle was used. The major benefit with respect to the radiological rsiks of basing the fuel cycle around modular HTGR technology instead of the LWR technology is the increase in reactor safety. The design of the modular HTGR is expected to prevent the release of a significant amount of radioactive material to the environment, and hence early deaths in the surrounding population, during accident conditions. This implies that there is no group risk as defined in the Dutch risk management policy. The major benefit of thorium based fuel cycles over uranium based fuel cycles is the reduction in the radiological risks from unraium mining and milling. The other stages of the nuclear fuel cycle which make a significant contribution to the radiological risks are electricity generation, reprocessing and final disposal. The risks associated with the electricity generation stage are dominated by the risks from fission products, activated corrosion products and the activation products tritium and carbon-14. The risks associated with the reprocessing stage are determined by fission and activation products (including actinides). (orig./WL).

Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

Energy Technology Data Exchange (ETDEWEB)

Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio [Universidad Simón Bolívar, Nuclear Physics Laboratory, Apdo 89000, Caracas 1080A (Venezuela, Bolivarian Republic of); Davila, Jesus [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

2015-07-23

During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

Science.gov (United States)

Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

2015-07-01

During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

Indian Academy of Sciences (India)

These disadvantages of thorium fuel cycle were seemingly the reasons why that ... According to the data of figure 2, maximum (equilibrium) content of 233U in ..... Self-sufficient mode is related with rather big effort in the extraction of isotopes of.

A data base for PHW reactor operating on a once-through, low enriched uranium-thorium cycle

International Nuclear Information System (INIS)

Lungu, S.

1984-04-01

The study of a detailed data base for a new once-through uranium-thorium cycle using low enriched uranium (4 and 5,5% wt. U-235) and distinct UO 2 and ThO 2 fuel channels has been performed. With reference to a standard 638 MWe CANDU-type PHWR with 380 channels, evaluation of economics, fuel behaviour and safety has been performed. The Feinberg-Galanin method (code FEINGAL) has been used for calculation of axial flux distribution. All parameters have been provided by LATREP code following up the irradiation history. Economical assessment has shown that this fuel cycle is competitive with the natural uranium fuel cycle for 1979-based values of the parameters. Fuel behaviour and safety features modelling has shown that core behaviour of the uranium-thorium reactor under abnormal and accident conditions would be at least as good as that of the standard natural uranium reactor

Enhancing TRU burning and Am transmutation in Advanced Recycling Reactor

International Nuclear Information System (INIS)

Ikeda, Kazumi; Kochendarfer, Richard A.; Moriwaki, Hiroyuki; Kunishima, Shigeru

2011-01-01

Research highlights: → This ARR is an oxide fueled sodium cooled reactor based on innovative technologies to destruct TRU. → TRU burning core is designed to burn TRU at 28 kg/TW th h, adding moderator pins of B 4 C (Enriched B-11). → Am transmutation core can transmute Am at 34 kg/TW th h, adding uranium free AmN blanket to TRU burning core. → The TRU burning core improves TRU burning by 40-50% than the previous core. → The Am transmutation core can transmute Am effectively, keeping the void reactivity acceptable. - Abstract: 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 th h, by adding moderator pins of B 4 C (Enriched B-11) and the Am transmutation core will be able to transmute Am at 34 kg/TW th h, 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.

Actinide and fission product separation and transmutation

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-07-01

The second international information exchange meeting on actinide and fission product separation and transmutation, took place in Argonne National Laboratory in Illinois United States, on 11-13 November 1992. The proceedings are presented in four sessions: Current strategic system of actinide and fission product separation and transmutation, progress in R and D on partitioning processes wet and dry, progress in R and D on transmutation and refinements of neutronic and other data, development of the fuel cycle processes fuel types and targets. (A.L.B.)

Description of Transmutation Library for Fuel Cycle System Analyses

International Nuclear Information System (INIS)

Piet, Steven J.; Bays, Samuel E.; Hoffman, Edward A.

2010-01-01

This report documents the Transmutation Library that is used in Fuel Cycle System Analyses. This version replaces the 2008 version.(Piet2008) The Transmutation Library has the following objectives: (1) Assemble past and future transmutation cases for system analyses. (2) For each case, assemble descriptive information such as where the case was documented, the purpose of the calculation, the codes used, source of feed material, transmutation parameters, and the name of files that contain raw or source data. (3) Group chemical elements so that masses in separation and waste processes as calculated in dynamic simulations or spreadsheets reflect current thinking of those processes. For example, the CsSr waste form option actually includes all Group 1A and 2A elements. (4) Provide mass fractions at input (charge) and output (discharge) for each case. (5) Eliminate the need for either ''fission product other'' or ''actinide other'' while conserving mass. Assessments of waste and separation cannot use ''fission product other'' or ''actinide other'' as their chemical behavior is undefined. (6) Catalog other isotope-specific information in one place, e.g., heat and dose conversion factors for individual isotopes. (7) Describe the correlations for how input and output compositions change as a function of UOX burnup (for LWR UOX fuel) or fast reactor (FR) transuranic (TRU) conversion ratio (CR) for either FR-metal or FR-oxide. This document therefore includes the following sections: (1) Explanation of the data set information, i.e., the data that describes each case. In no case are all of the data presented in the Library included in previous documents. In assembling the Library, we return to raw data files to extract the case and isotopic data, into the specified format. (2) Explanation of which isotopes and elements are tracked. For example, the transition metals are tracked via the following: two Zr isotopes, Zr-other, Tc99, Tc-other, two Mo-Ru-Rh-Pd isotopes, Mo

Method to evaluate covariance data for the thorium-uranium fuel cycle

International Nuclear Information System (INIS)

Kawano, T.; Chadwick, M.B.

2003-01-01

This power point presentation gives an overview about the evaluation strategy for the experimental data for the thorium-uranium fuel cycle. Uncertainties, error propagation and calculation methods are outlined. Covariance evaluation tools and computer codes have been developed and results are presented

Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

2014-02-15

Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

Transmutation of radioactive wastes: how and why?

International Nuclear Information System (INIS)

Patarin, L.

2004-01-01

After having evoked the natural or spontaneous transmutation of natural or artificial radioactive atoms, the author describes how this transmutation is technically obtained, indicates the two main families of atoms present in a used nuclear fuel and for which transmutation is to be investigated (long-lived fission residues or products, and transuranium elements) and of which the behaviour in neutron fluxes must be explored. He discusses the industrial means required for artificial transmutation. He discusses the interest of performing such a transmutation

Calculations of different transmutation concepts. An international benchmark exercise

International Nuclear Information System (INIS)

2000-01-01

In April 1996, the NEA Nuclear Science Committee (NSC) Expert Group on Physics Aspects of Different Transmutation Concepts launched a benchmark exercise to compare different transmutation concepts based on pressurised water reactors (PWRs), fast reactors, and an accelerator-driven system. The aim was to investigate the physics of complex fuel cycles involving reprocessing of spent PWR reactor fuel and its subsequent reuse in different reactor types. The objective was also to compare the calculated activities for individual isotopes as a function of time for different plutonium and minor actinide transmutation scenarios in different reactor systems. This report gives the analysis of results of the 15 solutions provided by the participants: six for the PWRs, six for the fast reactor and three for the accelerator case. Various computer codes and nuclear data libraries were applied. (author)

Comparative study for minor actinide transmutation in various fast reactor core concepts

International Nuclear Information System (INIS)

Ohki, S.

2001-01-01

A comparative evaluation of minor actinide (MA) transmutation property was performed for various fast reactor core concepts. The differences of MA transmutation property were classified by the variations of fuel type (oxide, nitride, metal), coolant type (sodium, lead, carbon dioxide) and design philosophy. Both nitride and metal fuels bring about 10% larger MA transmutation amount compared with oxide fuel. The MA transmutation amount is almost unchanged by the difference between sodium and lead coolants, while carbon dioxide causes a reduction by about 10% compared with those. The changes of MA transmutation property by fuel and coolant types are comparatively small. The effects caused by the difference of core design are rather significant. (author)

High flux transmutation of fission products and actinides

International Nuclear Information System (INIS)

Gerasimov, A.; Kiselev, G.; Myrtsymova, L.

2001-01-01

Long-lived fission products and minor actinides accumulated in spent nuclear fuel of power reactors comprise the major part of high level radwaste. Their incineration is important from the point of view of radwaste management. Transmutation of these nuclides by means of neutron irradiation can be performed either in conventional nuclear reactors, or in specialized transmutation reactors, or in ADS facilities with subcritical reactor and neutron source with application of proton accelerator. Different types of transmutation nuclear facilities can be used in order to insure optimal incineration conditions for radwaste. The choice of facility type for optimal transmutation should be based on the fundamental data in the physics of nuclide transformations. Transmutation of minor actinides leads to the increase of radiotoxicity during irradiation. It takes significant time compared to the lifetime of reactor facility to achieve equilibrium without effective transmutation. High flux nuclear facilities allow to minimize these draw-backs of conventional facilities with both thermal and fast neutron spectrum. They provide fast approach to equilibrium and low level of equilibrium mass and radiotoxicity of transmuted actinides. High flux facilities are advantageous also for transmutation of long-lived fission products as they provide short incineration time

Data base for a CANDU-PHW operating on the thorium cycle

International Nuclear Information System (INIS)

1979-07-01

This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on various thorium cycles. In the reference cycle, the requirements for externally supplied fissile material are met using U-235, with the feed adjusted to provide a fuel burnup of approximately 30 000 MW.d/t(U). Two versions of the reference cycle are treated. In one, the U-235 is supplied in a highly enriched form (93 percent U-235 in uranium); in the other, the U-235 is supplied at a lower enrichment, such that the uranium present in the feed fuel is 'denatured'. The effects of varying the fuel burnup and the recycle delay time are discussed. Data are also given for thorium cycles using plutonium instead of U-235 to meet requirements for externally-supplied fissile material. The special case of 'self-sufficient equilibrium thorium cycles', which require no external source of fissile material for equilibrium operation, is also treated. (author)

Scenarios for minor actinides transmutation in the framework of the French Act on Waste Management

International Nuclear Information System (INIS)

Coquelet-Pascal, C.; Meyer, M.; Tiphine, M.; Girieud, R.; Eschbach, R.; Chabert, C.; Garzenne, C.; Barbrault, P.; Van Den Durpel, L.; Caron-Charles, M.; Favet, D.; Arslan, M.; Caron-Charles, M.; Carlier, B.; Lefevre, J.C.

2013-01-01

In the framework of the French Act on Waste Management, options of minor actinides (MA) transmutation are studied, based on several scenarios of sodium fast reactor deployment. Basically, one of these scenarios considers the deployment of a 60 GWe SFR fleet in two steps (20 GWe from 2040 to 2050 and 40 GWe, as well as, from 2080 to 2100). For this scenario, the advantages and drawbacks of different transmutation options are evaluated: - transmutation of all minor actinides or only of americium; - transmutation in homogeneous mode (MA bearing fuel in all the core or just in the outer core) or in heterogeneous mode (MA bearing radial blankets). Scenarios have been optimised to limit the impacts of MA transmutation on the cycle: - reduction of the initial MA content in the core in the case of transmutation in homogeneous mode to reduce the impact on reactivity coefficients; - reduction of the number of rows of blankets and fuel decay heat in the case of transmutation in heterogeneous mode. The sensitivity of transmutation options to cycle parameters such as the fuel cooling time before transportation is also assessed. Thus, the transmutation of only americium in one row of radial blankets containing initially 10 pc % Am and irradiated during the same duration as the standard fuel assemblies appears to be a suitable solution to limit the transmutation impacts on fuel cycle and facilities. A comparison of results obtained with MA transmutation in dedicated systems is also presented with a symbiotic scenario considering ADS (accelerator-driven system) deployment to transmute MA together with a SFR fleet to produce energy. The MA inventory within the cycle is higher in the case of transmutation in ADS than in the case of transmutation in SFR. Considering the industrial feasibility of MA transmutation, it appears important to study 'independently' SFR deployment and MA transmutation. Consequently, scenarios of progressive introduction of MA options are assessed

The future role of thorium in assuring CANDU fuel supplies

International Nuclear Information System (INIS)

Slater, J.B.

1985-01-01

Atomic Energy of Canada Limited (AECL), in partnership with Canadian industry and power utilities, has developed the CANDU reactor as a safe, reliable and economic means of transforming nuclear fuel into useable power. The use of thorium/uranium-233 recycle gives the possibility of a many-fold increase in energy yield over that which can be obtained from the use of uranium in once-through cycles. The neutronic properties of uranium-233 combine with the inherent neutron economy of the CANDU reactor to offer the possibility of near-breeder cycles in which there is no net consumption of fissile material under equilibrium fuelling conditions. Use of thorium cycles in CANDU will limit the impact of higher uranium prices. When combined with the potential for significant reductions in CANDU capital costs, then the long-term prospect is for generating costs near to current levels. Development of thorium cycles in CANDU will safeguard against possible uranium shortages in the next century, and will maintain and continue the commercial viability of CANDU as a long-term energy technology. (author)

Outline of research program on thorium fuel supported by grant-in-aid for energy research of ministry of education, science and culture

International Nuclear Information System (INIS)

Shibata, Toshikazu

1984-01-01

Since 1980, the Research Program on Thorium Fuel has been performed under the support of Grant-in-Aid for Energy Research of the Ministry of Education, Science and Culture of Japanese Government on the university basis including several tens professors. The main results have been published in the English-written report, ''Research on Thorium Fuel (SPEY-9, 1984)''. This report describes the outline and review of the symposium held on January 31, 1984. It consists of nuclear data, reactor physics, thorium fuel, irradiation of thorium, down-stream, biological effect, molten salt reactor engineering and others. It has been the first trial to perform such a big systematic cooperative studies in nuclear field on the university basis in Japan. (author)

Minor actinide transmutation on PWR burnable poison rods

International Nuclear Information System (INIS)

Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

2015-01-01

Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

Partitioning and transmutation: Radioactive waste management option

International Nuclear Information System (INIS)

Stanculescu, A.

2005-01-01

Growing world population with increasing energy needs, especially in the developing countries, Threat of global warming due to CO 2 emissions demands non-fossil electricity production. Nuclear will have to be part of a sustainable mix of energy production options Figures show that 350 GWe worldwide capacity is 'nuclear'. Present worldwide spent fuel (containing high Pu inventory) and HLW would need large repositories. In view of the previous facts this lecture deals Partitioning and transmutation as radioactive waste management option. Partitioning and transmutation (P and T) is a complex technology i.e. advanced reprocessing, and demand transuranics fuel fabrication plants, as well as innovative and/or dedicated transmutation reactors. In addition to U, Pu, and 129 I, 'partitioning' extracts from the liquid high level waste the minor actinides (MA) and the long-lived fission products (LLFP) 99-Tc, 93-Zr, 135-Cs, 107-Pd, and 79-Se). 'Transmutation' requires fully new fuel fabrication plants and reactor technologies to be developed and implemented on industrial scale. Present LWRs are not suited for MA and LLFP transmutation (safety consideration, plant operation, poor incineration capability). Only specially licensed LWRs can cope with MOX fuel; for increased Pu loadings (up to 100%), special reactor designs (e.g., ABB80+) are required; a combination of these reactor types could allow Pu inventory stabilization. Long-term waste radiotoxicity can be effectively reduced only if transuranics are 'incinerated' through fission with very hard neutron spectra. New reactor concepts (dedicated fast reactors, Accelerator Driven Systems (ADS), fusion/fission hybrid reactors) have been proposed as transmuters/incinerators. Significant Pu+MAs incineration rates can be achieved in symbiotic scenarios: LWR-MOX and dedicated fast reactors; fast neutron spectrum ADS mainly for MA incineration; very high thermal flux ADS concepts could also provide a significant transuranics

Thorium utilization in a small long-life HTR. Part II: Seed-and-blanket fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Ding, Ming, E-mail: dingming@hrbeu.edu.cn [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

2014-02-15

Highlights: • Seed-and-blanket (S and B) fuel blocks are proposed for a small block-type HTR. • S and B fuel blocks consist of a seed region (UO{sub 2}) and a blanket region (ThO{sub 2}). • The neutronic performance of S and B fuel blocks are analyzed using SCALE 6. • Three S and B fuel blocks with a reactivity swing of 0.1 Δk are recommended. • S and B fuel blocks are compared with thorium MOX fuel blocks. - Abstract: In order to utilize thorium in high temperature gas-cooled reactors (HTRs), the concept of seed-and-blanket (S and B) fuel block is introduced into the U-Battery, which is a long-life block-type HTR with a thermal power of 20 MWth. A S and B fuel block consists of a seed region with uranium in the center, and a blanket region with thorium. The neutronic performance, such as the multiplication factor, conversion ratio and reactivity swing, of a typical S and B fuel block was investigated by SCALE 6.0 by parametric analysis of the composition parameters and geometric parameters of the fuel block for the U-Battery application. Since the purpose of U-235 in the S and B fuel block is to ignite the fission reactions in the fuel block, 20% enriched uranium is recommended for the S and B fuel block. When the ratio of the number of carbon to heavy metal atoms changes with the geometric parameters of the fuel block in the range of 200–250, the reactivity swing reaches very small values. Furthermore, for a reactivity swing of 0.1 Δk during 10 effective full power years, three configurations with 36, 54 and 78 UO{sub 2} fuel rods are recommended for the application of the U-Battery. The comparison analysis of the S and B fuel block with the Th/U MOX fuel block shows that the former has a longer lifetime and a lower reactivity swing.

The French partitioning-transmutation programme, assets and prospects

International Nuclear Information System (INIS)

Viala, M.; Salvatores, M.; Mouney, H.

1997-01-01

Partitioning-transmutation studies are covered by the 1991 French law concerning radioactive waste management. The programme is progressing with a dual approach: - What can be done in partitioning-transmutation? At what cost? In what timescale? - How can long-term gains and short-term disadvantages be qualified and quantified? The first approach concerns technical know-how. The studies based on today's technologies are continuing (reactors, fuels and targets, separation of radionuclides by solvents). The second approach involves an assessment activity, based firstly on studies of scenarios. Pertinent assessment criteria must be brought out. (authors)

The role of Z-pinch fusion transmutation of waste in the nuclear fuel cycle

International Nuclear Information System (INIS)

Smith, James Dean; Drennen, Thomas E.; Rochau, Gary Eugene; Martin, William Joseph; Kamery, William; Phruksarojanakun, Phiphat; Grady, Ryan; Cipiti, Benjamin B.; Wilson, Paul Philip Hood; Mehlhorn, Thomas Alan; Guild-Bingham, Avery; Tsvetkov, Pavel Valeryevich

2007-01-01

The resurgence of interest in reprocessing in the United States with the Global Nuclear Energy Partnership has led to a renewed look at technologies for transmuting nuclear waste. Sandia National Laboratories has been investigating the use of a Z-Pinch fusion driver to burn actinide waste in a sub-critical reactor. The baseline design has been modified to solve some of the engineering issues that were identified in the first year of work, including neutron damage and fuel heating. An on-line control feature was added to the reactor to maintain a constant neutron multiplication with time. The transmutation modeling effort has been optimized to produce more accurate results. In addition, more attention was focused on the integration of this burner option within the fuel cycle including an investigation of overall costs. This report presents the updated reactor design, which is able to burn 1320 kg of actinides per year while producing 3,000 MWth

Partitioning and Transmutation: IAEA Activities

International Nuclear Information System (INIS)

Basak, U.; Monti, S.; )

2015-01-01

Full text of publication follows: The importance of partitioning and transmutation (P and T) processes for sustaining nuclear energy growth in the world has been realised in several countries across the world. P and T processes aim at separation and recycling of actinides including minor actinides (MAs) from the spent fuel or high-level liquid waste. The objective of these processes include reuse of separated fissile materials from spent nuclear fuels to obtain energy, enhance resource utilisation, reduce the disposal of toxic radio-nuclides and improve long-term performance of geological repositories. R and D programmes have been launched in many of the Member States to develop advanced partitioning process based on either aqueous or pyro to recover MAs along with other actinides as well as automated and remote techniques for manufacturing fuels containing MAs for the purpose of transmuting them either in fast reactors or accelerator driven hybrids. A number of Member States have been also developing such transmutation systems with the aim to construct and operate demo plants and prototypes in the next decade. The International Atomic Energy Agency has a high priority for the activities on partitioning and transmutation and regularly organises conferences, workshops, seminars and technical meetings in the areas of P and T as a part of information exchange and knowledge sharing at the international level. In the recent past, the Agency organised two technical meetings on advanced partitioning processes and actinide recycle technologies with the objective of providing a common platform for the scientists and engineers working in the areas of separation of actinides along with MAs from spent nuclear fuels and manufacturing of advanced fuels containing MAs in order to bridge the technological gap between them. In 2010, the Agency concluded a Coordinated Research Project (CRP) related to Assessment of Partitioning Processes. The Agency also conducted a first CRP on

Advanced nuclear fuel cycles activities in IAEA

International Nuclear Information System (INIS)

Nawada, H.P.; Ganguly, C.

2007-01-01

Full text of publication follows. Of late several developments in reprocessing areas along with advances in fuel design and robotics have led to immense interest in partitioning and transmutation (P and T). The R and D efforts in the P and T area are being paid increased attention as potential answers to ever-growing issues threatening sustainability, environmental protection and non-proliferation. Any fuel cycle studies that integrate partitioning and transmutation are also known as ''advanced fuel cycles'' (AFC), that could incinerate plutonium and minor actinide (MA) elements (namely Am, Np, Cm, etc.) which are the main contributors to long-term radiotoxicity. The R and D efforts in developing these innovative fuel cycles as well as reactors are being co-ordinated by international initiatives such as Innovative Nuclear Power Reactors and Fuel Cycles (INPRO), the Generation IV International Forum (GIF) and the Global Nuclear Energy Partnership (GENP). For these advanced nuclear fuel cycle schemes to take shape, the development of liquid-metal-cooled reactor fuel cycles would be the most essential step for implementation of P and T. Some member states are also evaluating other concepts involving the use of thorium fuel cycle or inert-matrix fuel or coated particle fuel. Advanced fuel cycle involving novel partitioning methods such as pyrochemical separation methods to recover the transuranic elements are being developed by some member states which would form a critical stage of P and T. However, methods that can achieve a very high reduction (>99.5%) of MA and long-lived fission products in the waste streams after partitioning must be achieved to realize the goal of an improved protection of the environment. In addition, the development of MA-based fuel is also an essential and crucial step for transmutation of these transuranic elements. The presentation intends to describe progress of the IAEA activities encompassing the following subject-areas: minimization of

Modelling the inventory and impact assessment of partitioning and transmutation approaches to spent nuclear fuel management

International Nuclear Information System (INIS)

Hoggett-Jones, C.; Robbins, C.; Gettinby, G.; Blythe, S.

2002-01-01

An inventory modelling and impact assessment system to investigate the potential effects of partitioning and transmutation is proposed. It is founded on a mass based inventory analysis using the principles of basic nuclear physics and the international standards for assessing radiological health effects. It is specific to the back-end of the nuclear fuel cycle and is applied to four alternative spent fuel management strategies. The system accounts for the dynamic nature of post-irradiation scenarios and is being used to develop software for use within the nuclear power industry. Four example waste-disposal options are considered using the method. Impact assessments and parameter sensitivity analyses are presented

Modelling the inventory and impact assessment of partitioning and transmutation approaches to spent nuclear fuel management

Energy Technology Data Exchange (ETDEWEB)