Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

1999-01-01

The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

The plutonium fuel cycles

International Nuclear Information System (INIS)

Pigford, T.H.; Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

A high temperature reactor could be used to eliminate the Russian military plutonium

International Nuclear Information System (INIS)

Foucher, N.

1999-01-01

The GT-MHR reactor (Gas Turbine Modular Helium Reactor) aims the double objective to eliminate the Russian plutonium coming from weapons, ( until 3 tons by year) and to produce a competitive energy from a small-scale power reactor with a nuclear fuel that can be of different type (plutonium or uranium). This reactor has several advantages: a high yield (47%) as every high temperature reactor and to be used in combined cycle, a high level of safety because of its ability to evacuate the residual power in a totally passive way and because of the nature of its fuel that is made of ceramics with a very high melting point that is to say no possibility of core melt. The fission products are contained in the ceramics so that reactor cannot disseminate radioactivity in its structure and consequently does not induce irradiation for the personnel. (N.C.)

International collaborations about fuel studies for reactor recycling of military quality plutonium

International Nuclear Information System (INIS)

Bernard, H.; Chaudat, J.P.

1997-01-01

In November 1992, an agreement was signed between the French and Russian governments to use in Russia and for pacific purposes the plutonium recovered from the Russian nuclear weapons dismantling. This plutonium will be transformed into mixed oxide fuels (MOX) for nuclear power production. The French Direction of Military Applications (DAM) of the CEA is the operator of the French-Russian AIDA program. The CEA Direction of Fuel Cycle (DCC) and Direction of Nuclear Reactors (DRN) are involved in the transformation of metallic plutonium into sinterable oxide powder for MOX fuel manufacturing. The Russian TOMOX (Treatment of MOX powder Metallic Objects) and DEMOX (MOX Demonstration) plants will produce the MOX fuel assemblies for the 4 VVER 1000 reactors of Balakovo and the fast BN 600 reactor. The second part of the program will involve the German Siemens and GRS companies for the safety studies of the reactors and fuel cycle plants. The paper gives also a brief analysis of the US policy concerning the military plutonium recycling. (J.S.)

Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

International Nuclear Information System (INIS)

1995-07-01

The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D ampersand D) and to reduce the cost of maintaining the facilities prior to D ampersand D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor's fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered

Recovery of weapon plutonium as feed material for reactor fuel

International Nuclear Information System (INIS)

Armantrout, G.A.; Bronson, M.A.; Choi, Jor-Shan

1994-01-01

This report presents preliminary considerations for recovering and converting weapon plutonium from various US weapon forms into feed material for fabrication of reactor fuel elements. An ongoing DOE study addresses the disposition of excess weapon plutonium through its use as fuel for nuclear power reactors and subsequent disposal as spent fuel. The spent fuel would have characteristics similar to those of commercial power spent fuel and could be similarly disposed of in a geologic repository

Economic considerations of plutonium utilization in the nuclear power strategy of Finland

International Nuclear Information System (INIS)

Silvennoinen, P.; Tusa, E.; Routti, J.T.

1977-01-01

Based on the current and predicted share of nuclear power in the national energy supply strategy, an optimal programme is developed for the exploitation of plutonium in both light-water and fast reactor systems. Assuming cost trends beyond the year 2000 for uranium, plutonium, uranium enrichment, fuel fabrication and assessing the availability of plutonium from domestic power plants and from abroad, the nuclear construction programme is optimized economically in view of the estimated development in the investment costs of various plant types. Given the expected nuclear share of the energy procurement this sector is covered by the alternative production schemes, i.e. light-water reactors with and without plutonium recycle, and fast reactors. Defining the objective function in terms of minimized revenue requirement in plant amortization and operation the generated scenarios are screened off and they finally converge to the optimal policy of nuclear power construction up to the year 2000. The established technology is associated with a larger share of the domestic manufacturing and the introduction of a new fuel or reactor type is taken to correspond to a reduced domestic investment share. In the investment costs the domestic fraction is regarded competitive up to a certain marginal excess. Plutonium recycle is seen to be competitive from 1985 or as soon as the required amount of fuel has been reprocessed. The domestic accumulation of plutonium will be able to support the introduction of the LMFBR in 1997. Owing to the uncertainties prevailing in the forecasts, sensitivity studies are performed as functions of the major economic parameters and their temporal development. (author)

Weapons-grade plutonium dispositioning. Volume 3: A new reactor concept without uranium or thorium for burning weapons-grade plutonium

International Nuclear Information System (INIS)

Ryskamp, J.M.; Schnitzler, B.G.; Fletcher, C.D.

1993-06-01

The National Academy of Sciences (NAS) requested that the Idaho National Engineering Laboratory (INEL) examine concepts that focus only on the destruction of 50,000 kg of weapons-grade plutonium. A concept has been developed by the INEL for a low-temperature, low-pressure, low-power density, low-coolant-flow-rate light water reactor that destroys plutonium quickly without using uranium or thorium. This concept is very safe and could be designed, constructed, and operated in a reasonable time frame. This concept does not produce electricity. Not considering other missions frees the design from the paradigms and constraints used by proponents of other dispositioning concepts. The plutonium destruction design goal is most easily achievable with a large, moderate power reactor that operates at a significantly lower thermal power density than is appropriate for reactors with multiple design goals. This volume presents the assumptions and requirements, a reactor concept overview, and a list of recommendations. The appendices contain detailed discussions on plutonium dispositioning, self-protection, fuel types, neutronics, thermal hydraulics, off-site radiation releases, and economics

Civil plutonium in the world: an estimate by the code REACTOR

International Nuclear Information System (INIS)

Braet, J.; Carchon, R.; Van der Meer, K.

1996-11-01

The computer code REACTOR that was developed by the Belgian Nuclear Research Centre SCK/CEN to study the built-up of plutonium stockpiles in the world is described. The code consists of a central database, containing general information about most commercial civil nuclear facilities. Using this code, an overview is given of the evolution of the nuclear energy production in the world, in the past and the medium term future. The nuclear energy production results in the accumulation of spent fuel stocks, containing vast amounts of energy enclosed in the plutonium. The presence and built-up of large stockpiles of spent fuel and separated plutonium originating from the civil fuel cycle is estimated. In this report several possible scenarios are considered for the use of that plutonium, with the aim of minimizing those stocks. According to the different national policies, scenarios such as open fuel cycle, thermal reactors or fast reactor cycle with the burning of plutonium in fast reactors are envisaged

Plutonium recycling and the problem of nuclear proliferation

International Nuclear Information System (INIS)

Albright, D.; Feiveson, H.S.

1988-01-01

A typical 1-gigawatt light water reactor (LWR), the dominant commercial power reactor type today, operating at 70% capacity factor, generates approximately 250 kilograms of plutonium annually. This plutonium, which is produced in the reactor through neutron capture by uranium-238, is then discharged from the reactor along with the other constituents of the spent fuel. About 70% of the plutonium, or 175 kilograms, consists of fissile (odd-numbered) plutonium isotopes. As long as the plutonium discharged from the reactor is left intermixed with the highly radioactive fission products also contained in the spent fuel, it cannot readily be used for power or for weapons. However, upon chemical separation from the radioactive fission products and other components of the spent reactor fuel, the plutonium produced each year in a gigawatt reactor could be used, either in recycled fuel (to replace about 175 kilograms of U-235 in a power reactor) or to provide the fissile material for more than 25 nuclear warheads. Commercial separation of plutonium and the introduction of nuclear fuel cycles using recycled plutonium, which are now impending in several countries, force one to balance the probable increased risks of nuclear proliferation due to these activities against various economic and other motives that have been forwarded in their defense. The authors undertake an assessment of this balancing in this article

Economic considerations of plutonium utilization in the nuclear power strategy of Finland

International Nuclear Information System (INIS)

Silvennoinen, P.; Tusa, E.; Routti, J.T.

1977-01-01

Based on the current and prospected share of nuclear power in the national energy supply strategy an optimal programme is developed for exploitation of plutonium in both light water and fast reactor systems. Assuming cost trends until and beyond the year 2000 for uranium, plutonium, uranium enrichment, fuel fabricaton and assessing the availability of plutonium from the domestic power plants and from abroad the nuclear construction programme is optimized economically in view of the estimated development in the investment costs of various plant types. Given the expected nuclear share of the energy procurement this sector is covered by the alternative production schemes, i.e. light water reactors with and without plutonium recycle and fast reactors. The plant sizes are allowed to be either 500 MWe or 1000 MWe. The installation dates are fixed manually with a minor flexibility of time but with all the three degrees of freedom in the plant types. Defining the objective function in terms of minimized revenue requirement in plant amortization and operation the generated scenarios are screened off and they finally converge to the optimal policy of nuclear power construction up to the year 2000. Special attention is placed on the constraints which eliminate excessive proliferation of reactor types. This is mainly implemented by the criterion of increasing the domestic share in the investments. The established technology is associated with a larger share of the Finnish manufacturing and the introduction of new fuel or reactor type is taken to correspond to a reduced domestic investment share. The results yield the time schedule and installed capacity of the three different production means. Due to the uncertainties prevailing in the forecasts sensitivity studies are performed as functions of the major economic parameters and their temporal development

How much plutonium does North Korea really have?

International Nuclear Information System (INIS)

Dreicer, J.S.

1997-01-01

In a previous study, as part of the Global Nuclear Material Control Model effort, the author estimated the maximum quantity of plutonium that could be produced in thermal research reactors in the potential nuclear weapon states (including North Korea), based on their declared power level. D. Albright has estimated the amount of plutonium the North Koreans may have produced since 1986 in the 5-megawatt-electric power reactor at Yongbon. Albright provided an upper-bound estimate of 53 kilograms of weapon-grade plutonium produced cumulatively if the gas-graphite (magnox) reactor had achieved a load factor of 0.80. This cumulative estimate of 53 kilograms ignores the potential plutonium production in the 8-megawatt-thermal research reactor, IRT-DPRK. To better quantify the cumulative North Korean production, the author conducted a study to estimate the amount of plutonium that could have been produced in the IRT-DPRK research reactor operating at the declared power level during the entire period it has operated, including a period it was not safeguarded. The author estimates that, at most, an additional 6 to 33 kilograms of plutonium could have been produced cumulatively in the research reactor operating at the declared power level during the entire period it has operated, including a 12-year period it was not safeguarded, resulting in a total of 13 to 47 kilograms of plutonium possibly produced in both the research and power reactors

Plutonium

Energy Technology Data Exchange (ETDEWEB)

Miner, William N

1964-01-01

This pamphlet discusses plutonium from discovery to its production, separation, properties, fabrication, handling, and uses, including use as a reactor fuel and use in isotope power generators and neutron sources.

Feasibility study of a magnetic fusion production reactor

Science.gov (United States)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells

Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington

International Nuclear Information System (INIS)

1992-12-01

The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE's responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which contains additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period

Status of power reactor fuel reprocessing in India

International Nuclear Information System (INIS)

Kansra, V.P.

1999-01-01

Spent fuel reprocessing in India started with the commissioning of the Trombay Plutonium Plant in 1964. This plant was intended for processing spent fuel from the 40 MWth research reactor CIRUS and recovering plutonium required for the research and development activities of the Indian Atomic Energy programme. India's nuclear energy programme aims at the recycle of plutonium in view of the limited national resources of natural uranium and abundant quantities of thorium. This is based on the approach which aims at separating the plutonium from the power reactor spent fuel, use it in the fast reactors to breed 233 U and utilise the 233 U generated to sustain a virtually endless source of power through thorium utilisation. The separated plutonium is also being utilised to fabricate MOX fuel for use in thermal reactors. Spent fuel treatment and extracting plutonium from it makes economic sense and a necessity for the Indian nuclear power programme. This paper describes the status and trends in the Indian programme for the reprocessing of power reactor fuels. The extraction of plutonium can also be seen as a far more positive approach to long-term waste management. The closed cycle approach visualised and pursued by the pioneers in the field is now steadily moving India towards the goal of a sustainable source of power through nuclear energy. The experience in building, operating and refurbishing the reprocessing facilities for uranium and thorium has resulted in acquiring the technological capability for designing, constructing, operating and maintaining reprocessing plants to match India's growing nuclear power programme. (author)

Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

International Nuclear Information System (INIS)

Rubini, L.A.

1978-02-01

An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

Possibility of plutonium burning out and minor actinides transmutation in CANDU type reactor

International Nuclear Information System (INIS)

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

2000-01-01

The possibility of power or weapon-grade plutonium use as nuclear fuel in CANDU type reactor with simultaneous minor actinides burn-out is studied. Total thermal power is 1900 MW. The fuel lifetime makes 0.24 years, neutron flux density 10 14 neutr/cm 2 s. About 40-45 % of plutonium is incinerated during fuel lifetime. If weapon-grade plutonium is used in fuel channels instead of power one, its consumption is 40% lower. (author)

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

Trans-Uranium Doping Utilization for Increasing Protected Plutonium Proliferation of Small Long Life Reactor

Energy Technology Data Exchange (ETDEWEB)

Permana, Sidik [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Nuclear and Biophysics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Suud, Zaki [Nuclear and Biophysics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Suzuki, Mitsutoshi [Japan Atomic Energy Agency, Nuclear Non-proliferation Science and Technology Center, 2-4 Shirane Shirakata, Tokai-mura, Ibaraki, 319-1195 (Japan)

2009-06-15

for different power output and core operation time. The two dimensional core optimization analysis (SRAC-CITATION) has been used with nuclear data library (JENDL-33). Higher plutonium production rate of Pu-238 can be achieved by Neptunium (NP)-237 doping compared with Minor Actinide (MA) doping. However, higher plutonium production rate of Pu-240 can be obtained by MA doping. Pu-238 production decreases and Pu-240 production increases with increasing core operation time for both MA and NP-237 doping materials. In case of different doping rate of MA, it shows Pu-238 and Pu-240 productions are strongly depends on the content of MA doping and different core operation time. Finally, some plutonium characterizations will be shown based on several basic criteria which have been adopted in this study. (authors)

Fast Thorium Molten Salt Reactors Started with Plutonium

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

2006-01-01

One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

Safety aspects of Particle Bed Reactor plutonium burner system

International Nuclear Information System (INIS)

Powell, J.R.; Ludewig, H.; Todosow, M.

1993-01-01

An assessment is made of the safety aspects peculiar to using the Particle Bed Reactor (PBR) as the burner in a plutonium disposal system. It is found that a combination of the graphitic fuel, high power density possible with the PBR and engineered design features results in an attractive concept. The high power density potentially makes it possible to complete the plutonium burning without requiring reprocessing and remanufacturing fuel. This possibility removes two hazardous steps from a plutonium burning complex. Finally, two backup cooling systems depending on thermo-electric converters and heat pipes act as ultimate heat removal sinks in the event of accident scenarios which result in loss of fuel cooling

Prospects for plutonium recycling in light-water reactors in the European Communities

International Nuclear Information System (INIS)

Haijtink, B.

1977-01-01

On the basis of a forecast on nuclear power installation programmes and spent nuclear fuel reprocessing capacities in the European Community, an assessment has been made of the developmnt of the plutonium handling industry up to the year 1990. Data on plutonium recovered in the reprocessing plants, requirements for fuelling the fast-breeder reactors and the remaining stockpile have been evaluated. A possible strategy for recycling the plutonium surplus in light-water reactors has been developed and its effects as regards saving on requirements for natural uranium and enrichment services have been estimated

Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

International Nuclear Information System (INIS)

Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

1978-01-01

The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

International Nuclear Information System (INIS)

Chodak, P. III

1996-05-01

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO 2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239 Pu and ≥90% total Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

Energy Technology Data Exchange (ETDEWEB)

Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

1996-05-01

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO₂ assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the ²³⁹Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

Neutronic design of a plutonium-thorium burner small nuclear reactor

International Nuclear Information System (INIS)

Hartanto, Donny

2010-02-01

A small nuclear reactor using thorium and plutonium fuel has been designed from the neutronic point of view. The thermal power of the reactor is 150 MWth and it is proposed to be used to supply electricity in an island in Indonesia. Thorium and plutonium fuel was chosen because in recent years the thorium fuel cycle is one of the promising ways to deal with the increasing number of plutonium stockpiles, either from the utilization of uranium fuel cycle or from nuclear weapon dismantling. A mixed fuel of thorium and plutonium will not generate the second generation of plutonium which will be a better way to incinerate the excess plutonium compared with the MOX fuel. Three kinds of plutonium grades which are the reactor grade (RG), weapon grade (WG), and spent fuel grade (SFG) plutonium, were evaluated as the thorium fuel mixture in the 17x17 Westinghouse PWR Fuel assembly. The evaluated parameters were the multiplication factor, plutonium depletion, fissile buildup, neutron spectrum, and temperature reactivity feedback. An optimization was also done to increase the plutonium depletion by changing the Moderator to Fuel Ratio (MFR). The computer codes TRITON (coupled NEWT and ORIGEN-S) in SCALE version 6 were used as the calculation tool for this assembly level. From the evaluation and optimization of the fuel assembly, the whole core was designed. The core was consisted of 2 types of thorium fuel with different plutonium grade and it followed the checkerboard loading pattern. A new concept of enriched burnable poison was also introduced to the core. The core life is 6.4 EFPY or 75 GWd/MTHM. It can burn up to 58% of its total mass of initial plutonium. VENTURE was used as the calculation tool for the core level

Physics of Plutonium Recycling in Thermal Reactors

International Nuclear Information System (INIS)

Kinchin, G.H.

1967-01-01

A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of 240 Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

Physics of Plutonium Recycling in Thermal Reactors

Energy Technology Data Exchange (ETDEWEB)

Kinchin, G. H. [Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1967-09-15

A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of {sup 240}Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

Multiple microprocessor based nuclear reactor power monitor

International Nuclear Information System (INIS)

Lewis, P.S.; Ethridge, C.D.

1979-01-01

The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a 3 He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection

Optimisation of deep burn incineration of reactor waste plutonium in a PBMR DPP-400 core

International Nuclear Information System (INIS)

Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik

2014-01-01

In this article an original set of coupled neutronics and thermo-hydraulic simulation results for the VSOP 99/05 diffusion code are presented for advanced fuel cycles for the incineration of weapons-grade plutonium, reactor-grade plutonium and reactor-grade plutonium with its associated Minor Actinides in the 400 MW th Pebble Bed Modular Reactor Demonstration Power Plant. These results are also compared to those of the standard 9.6 wt% enriched 9 g/fuel sphere U/Pu fuel cycle. The weapons-grade and reactor-grade plutonium fuel cycles produced good burn-ups. However, the addition of the Minor Actinides to the reactor-grade plutonium caused a large decrease in the burn-up and thus an unacceptable increase in the heavy metal (HM) content in the spent fuel, which was intended for direct disposal in a deep geological repository, without chemical reprocessing. All the plutonium fuel cycles failed the adopted safety limits used in the PBMR400 in that either the maximum fuel temperature of 1130 °C during normal operation, or the maximum power density of 4.5 kW/sphere was exceeded. All the plutonium fuel cycles also produced positive uniform temperature reactivity coefficients, i.e. the reactivity coefficient where the temperatures of the fuel and the graphite moderator in the fuel spheres were varied together. These unacceptable positive coefficients were experienced at low temperatures, typically below 700 °C. This was due to the influence of the thermal fission cross-section resonances of 239 Pu and 241 Pu. Weapons-grade plutonium produced the worst safety performance. The safety performance of the reactor-grade plutonium also deteriorated when the HM loading was reduced from 3 g/sphere to 2 g or 1 g

Optimisation of deep burn incineration of reactor waste plutonium in a PBMR DPP-400 core

Energy Technology Data Exchange (ETDEWEB)

Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School for Mechanical and Nuclear Engineering, North West University, PUK-Campus, Private Bag X6001, Internal Post Box 360, Potchefstroom 2520 (South Africa); Mulder, Eben J. [School for Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)

2014-05-01

In this article an original set of coupled neutronics and thermo-hydraulic simulation results for the VSOP 99/05 diffusion code are presented for advanced fuel cycles for the incineration of weapons-grade plutonium, reactor-grade plutonium and reactor-grade plutonium with its associated Minor Actinides in the 400 MW{sub th} Pebble Bed Modular Reactor Demonstration Power Plant. These results are also compared to those of the standard 9.6 wt% enriched 9 g/fuel sphere U/Pu fuel cycle. The weapons-grade and reactor-grade plutonium fuel cycles produced good burn-ups. However, the addition of the Minor Actinides to the reactor-grade plutonium caused a large decrease in the burn-up and thus an unacceptable increase in the heavy metal (HM) content in the spent fuel, which was intended for direct disposal in a deep geological repository, without chemical reprocessing. All the plutonium fuel cycles failed the adopted safety limits used in the PBMR400 in that either the maximum fuel temperature of 1130 °C during normal operation, or the maximum power density of 4.5 kW/sphere was exceeded. All the plutonium fuel cycles also produced positive uniform temperature reactivity coefficients, i.e. the reactivity coefficient where the temperatures of the fuel and the graphite moderator in the fuel spheres were varied together. These unacceptable positive coefficients were experienced at low temperatures, typically below 700 °C. This was due to the influence of the thermal fission cross-section resonances of {sup 239}Pu and {sup 241}Pu. Weapons-grade plutonium produced the worst safety performance. The safety performance of the reactor-grade plutonium also deteriorated when the HM loading was reduced from 3 g/sphere to 2 g or 1 g.

Non-fertile fuels for burning weapons plutonium in thermal fission reactors

International Nuclear Information System (INIS)

Lombardi, C.; Mazzola, A.; Vettraino, F.

1996-01-01

In the last few years, the excess plutonium disposition has become ever more a topical and critical issue. As a matter of fact, more than 200 MT of plutonium coming from spent fuel reprocessing have been already stockpiled and over the next decade, under the already ratified agreements, another about 200 MT of weapon-grade plutonium are expected to be available from nuclear weapons dismantlement. On this basis, an ever growing plutonium production is no longer the goal and the already stored quantities should be burnt in power reactors by taking care that no new plutonium is generated under irradiation. This new outlook in considering plutonium has led many designers to reassess the Fast Breeder Reactors (FBR) role and shifting from breeder to burner machines perspective. Several solutions for burning plutonium have been so far proposed and discussed from the safeguards, proliferation resistance, environmental safety, technological background, economy and time schedule standpoint. A proposal for plutonium burning in commercial Pressurized Water Reactors (PWR) by using a non-fertile oxide-type fuel consisting of PuO 2 diluted in an inert matrix is reported hereafter. This solution appears to receive an ever growing interest in the nuclear community. In order not to produce new plutonium during irradiation an innovative U-free fuel is being researched, based on an inert matrix which will consist in a mixed compound of inert oxides, such as ZrO 2 , Al2O 3 , MgO, CeO 2 where the plutonium oxide is dispersed in. The matrix will fulfill the following requirements: good chemical compatibility, acceptable thermal conductivity, good nuclear properties, good stability under irradiation, good dissolution resistance. The plutonium relative content will be comparable to that used in MOX fuel. The fuel is expected to be characterized by a high chemical stability (rock-like fuel), so that after discharge from reactor and adequate cooling time, it can be considered a High Level

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

Study of plutonium recycling physics in light water reactors

International Nuclear Information System (INIS)

Reuss, Paul

1979-10-01

A stock of plutonium from the reprocessing of thermal neutron reactor fuel is likely to appear in the next few years. The use of this plutonium as fuel replacing 235 U in thermal reactors is probably more interesting than simple stock-piling storage: immobilization of a capital which moreover would deteriorate by radioactive decay of isotope 241 also fissile and present to an appreciable extend in plutonium from reprocessing (half-life 15 years); recycling, on the other hand, will supply energy without complete degradation of the stock for fast neutron reactor loads, the burned matter having been partially renewed by conversion; furthermore the use of plutonium will meet the needs created by a temporary pressure on the naturel and/or enriched uranium market. For these two reasons the recycling of plutonium in thermal neutron reactors is being considered seriously today. The present work is confined to neutronic aspects and centres mainly on pressurized water-moderated reactors, the most highly developed at present in France. Four aspects of the problem are examined: 1. the physics of a plutonium-recycling reactor special features of neutronic phenomena with respect to the 'conventional' scheme of the 235 U burning reactor; 2. calculation of a plutonium-recycling reactor: adaptation of standard methods; 3. qualification of these calculations from the viewpoint of both data and inevitable approximations; 4. the fuel cycle and particularly the equivalence of fissile matters [fr

Strategies for plutonium recycle in a system of pressurized water reactors

International Nuclear Information System (INIS)

Leaver, D.E.W.

1976-01-01

A methodology is developed to allow a utility fuel manager to determine economic strategies for recycling plutonium in a system of light water reactors. One possible plutonium recycle strategy would be self-generated recycle, in which plutonium discharged from a reactor is recycled back to that same reactor as soon as possible. Another possible strategy is to recycle all the plutonium discharged from several reactors into one reactor. Such a strategy might be advantageous if the reactor receiving the plutonium were of a type that utilized plutonium more effectively than other reactors in the system. There are several considerations which affect the economics of recycling a batch of plutonium to one reactor or cycle vs. another, or which would favor a special recycling strategy. Among these are cycle energy, length of time that plutonium is stored prior to recycle, and isotopes of the plutonium. The methodology developed is used to quantitatively illustrate the effect on recycle strategy of these parameters. The problem of choosing the plutonium recycle strategy which results in the minimum fuel cost is formulated as a mathematical programming problem. The objective function for this problem is the total discounted fuel cost for the reactor system over a specified planning period. The savings of an optimal recycle strategy over self-generated recycle would be typically one million dollars per year for a utility with several large PWRs

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Physics studies of weapons plutonium disposition in the Integral Fast Reactor closed fuel cycle

International Nuclear Information System (INIS)

Hill, R.N.; Wade, D.C.; Liaw, J.R.; Fujita, E.K.

1995-01-01

The core performance impact of weapons plutonium introduction into the Integral Fast Reactor (IFR) closed fuel cycle is investigated by comparing three disposition scenarios: a power production mode, a moderate destruction mode, and a maximum destruction mode, all at a constant heat rating of 840 MW(thermal). For each scenario, two fuel cycle models are evaluated: cores using weapons material as the sole source of transuranics in a once-through mode and recycle cores using weapons material only as required for a makeup feed. In addition, the impact of alternative feeds (recycled light water reactor or liquid-metal reactor transuranics) on burner core performance is assessed. Calculated results include mass flows, detailed isotopic distributions, neutronic performance characteristics, and reactivity feedback coefficients. In general, it is shown that weapons plutonium does not have an adverse effect on IFR core performance characteristics; also, favorable performance can be maintained for a wide variety of feed materials and fuel cycle strategies

The use of plutonium in Swedish reactors

International Nuclear Information System (INIS)

Forsstroem, H.

1982-09-01

The report deals with the utilization of plutonium in Swedish nuclear power plants. The plutonium content of the mixed oxide fuel will normally be 3-7 per cent. The processing of spent nuclear fuel will produce about 6 ton plutonium. The use of mixed oxide fuel in Forsmark 3 and Oskarshamn 3 is discussed. The fuel cycle will start with the manufacturing of the fuel elements abroad and proceeds with transport and utilization, storing of spent fuel about 40 years in Sweden followed by direct disposal. The manufacture and use of mixed oxide (MOX) fuel is based on well-known techniques. Approximately 20 000 MOX fuel rods have been irradiated and the fuel is essentially equivalent to uranium oxide fuel. 30-50 per cent of the core may be composed of MOX-fuel without any effect on the operation and safety of the reactor which has been originally designed for uranium fuel. The evaluation of international fuel cycle (INFCE) states that the proliferation risks are very small. The recycling of plutonium will reduce demand for enriched uranium and the calculations show that 6.3 ton plutonium will replace the enrichment of 600 ton natural uranium. (G.B.)

Fuel balance in nuclear power with fast reactors without a uranium blanket

International Nuclear Information System (INIS)

Naumov, V.V.; Orlov, V.V.; Smirnov, V.S.

1994-01-01

General aspects related to replacing the uranium blanket of a lead-cooled fast reactor burning uranium-plutonium nitride fuel with a more efficient lead reflector are briefly discussed in the article. A study is very briefly summarized, which showed that a breeding ratio of about 1 and electric power of about 300 MW were achievable. A nuclear fuel balance is performed to estimate the increased consumption of uranium to produce power and the gains achievable by eliminating the uranium blanket. Elimination of the uranium blanket has the advantages of simplifying and improving the fast reactor and eliminating the production of weapons quality plutonium. 3 figs

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

Plutonium Proliferation: The Achilles Heel of Disarmament

International Nuclear Information System (INIS)

Leventhal, Paul

2001-01-01

Plutonium is a byproduct of nuclear fission, and it is produced at the rate of about 70 metric tons a year in the world's nuclear power reactors. Concerns about civilian plutonium ran high in the 1970s and prompted enactment of the Nuclear Non-Proliferation Act of 1978 to give the United States a veto over separating plutonium from U.S.-supplied uranium fuel. Over the years, however, so-called reactor-grade plutonium has become the orphan issue of nuclear non-proliferation, largely as a consequence of pressures from plutonium-separating countries. The demise of the fast breeder reactor and the reluctance of utilities to introduce plutonium fuel in light-water reactors have resulted in large surpluses of civilian, weapons-usable plutonium, which now approach in size the 250 tons of military plutonium in the world. Yet reprocessing of spent fuel for recovery and use of plutonium proceeds apace outside the United States and threatens to overwhelm safeguards and security measures for keeping this material out of the hands of nations and terrorists for weapons. A number of historical and current developments are reviewed to demonstrate that plutonium commerce is undercutting efforts both to stop the spread of nuclear weapons and to work toward eliminating existing nuclear arsenals. These developments include the breakdown of U.S. anti-plutonium policy, the production of nuclear weapons by India with Atoms-for-Peace plutonium, the U.S.-Russian plan to introduce excess military plutonium as fuel in civilian power reactors, the failure to include civilian plutonium and bomb-grade uranium in the proposed Fissile Material Cutoff Treaty, and the perception of emerging proliferation threats as the rationale for development of a ballistic missile defense system. Finally, immobilization of separated plutonium in high-level waste is explored as a proliferation-resistant and disarmament-friendly solution for eliminating excess stocks of civilian and military plutonium.

Development of a fresh plutonium fuel container for a prototype fast breeder reactor

International Nuclear Information System (INIS)

Ohtake, T.; Takahashi, S.; Mishima, T.; Kurakami, J.; Yamamoto, Y.; Ohuchi, Y.

1989-01-01

Japan gives a good deal of encouragement to development of a fast breeder reactor (which is considered as the most likely candidate for nuclear power generation) to secure long-term energy source. And, following an experimental fast breeder reactor Joyo, a prototype fast breeder reactor Monju is now under vigorous construction. Related to development of the prototype fast breeder reactor, it is necessary and important to develop transport container which is used for transporting fresh fuel assemblies from Plutonium Fuel Production Facility to the Monju power plant. Therefore, the container is now being developed by Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, shipment and vibration tests, handling performance tests, shielding performance tests and prototype container tests are executed with prototype containers fabricated according to a final design, in order to experimentally confirm soundness of transport container and its contents, and propriety of design technique. This paper describes the summary of general specifications and structures of this container and the summary of preliminary safety analysis of package

The use of plutonium rapides surregenerateurs aspects techniques et economiques

International Nuclear Information System (INIS)

Guillet, H.; Delayre, R.; Mougniot, J.C.; Ferrari, A.

1977-01-01

Nuclear energy production utilizing U 235 and U 238 inevitably results in the formation of plutonium. Some of this is directly used by the reactor in power production. Some reactors, e.g. the Candu type, burn most of their plutonium ''in situ''. However the surplus quantity of plutonium produced is increasing, and by 1990 the world stock of plutonium is predicted to be about 1000 tons (300 tons in Europe and 400 in the USA). This represents approximately 0.1 Q of potential power, where Q=10 21 joules. Proposals for dealing with this plutonium include its storage, either as irradiated fuel or as a refined substance (plutonium nitrate or oxide); its use in thermal reactors as a substitute for U 235 ; its use in fast breeder reactors which can act as consumers of plutonium as well as producers and which can therefore regulate the world's plutonium stocks. Leaving aside the question of storage, certain technical conditions must be fulfilled. It is essential that reprocessing plants should be operational and available in adequate numbers. An industry able to make use of plutonium should be developed with large capacity units. There seem to be no problems with plutonium use in reactors: conclusive experiments have proved the accuracy of calculations regarding plutonium recycling in thermal reactors and a number of fast breeders have proved the possibility of using plutonium in them. Experience acquired during the past ten years permits us to deal with the safety and safeguards problem of using plutonium. Many economic questions remain, however, including questions affecting the utilization of plutonium, such as the cost of regenerating irradiated fuel and of making fuel, and questions affecting the strategy of using plutonium in different ways, such as the price at which it should be sold. Also the cost of producing electricity using plutonium is not yet clear. It is unavoidable that plutonium will be used in nuclear power development. The technical and economic problems

Hanford B Reactor Building Hazard Assessment Report

International Nuclear Information System (INIS)

Griffin, P. W.

1999-01-01

The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

Advanced fuels for plutonium management in pressurized water reactors

International Nuclear Information System (INIS)

Vasile, A.; Dufour, Ph.; Golfier, H.; Grouiller, J.P.; Guillet, J.L.; Poinot, Ch.; Youinou, G.; Zaetta, A.

2003-01-01

Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1 . More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

Plutonium

International Nuclear Information System (INIS)

Watson, G.M.

1976-01-01

Discovery of the neutron made it easy to create elements which do not exist in nature. One of these is plutonium, and its isotope with mass number 239 has nuclear properties which make it both a good fuel for nuclear power reactors and a good explosive for nuclear weapons. Since it was discovered during a war the latter characteristic was put to use, but it is now evident that use of plutonium in a particular kind of nuclear reactor, the fast breeder reactor, will allow the world's resources of uranium to last for millennia as a major source of energy. Plutonium is very radiotoxic, resembling radium in this respect. Therefore the widespread introduction of fast breeder reactors to meet energy demands can be contemplated only after assurances on two points; that adequate control of the radiological hazard resulting from the handling of very large amounts of plutonium can be guaranteed, and that diversion of plutonium to illicit use can be prevented. The problems exist to a lesser degree already, since all types of nuclear reactor produce some plutonium. Some plutonium has already been dispersed in the environment, the bulk of it from atmospheric tests of nuclear weapons. (author)

A World made of Plutonium?

International Nuclear Information System (INIS)

Broda, E.

1976-01-01

This lecture by Engelbert Broda was written for the 26th Pugwash Conference in Mühlhausen, Germany, 26 – 31 August 1976: Public doubts about nuclear energy are generally directed at the problems of routine emissions of radionuclides, of catastrophic accidents, and of terminal waste disposal. Curiously, the most important problem is not being given sufficient attention: The use of plutonium from civilian reactors fpr weapons production. According to current ideas about a nuclear future, 5000 tons (order of magnitude) of plutonium are to be made annually by year 2000, and about 10 000 tons will all the time be in circulation (transport, reprocessing, reproduction of fuel elements, etc.). It is a misconception that plutonium from power reactors is unsuitable as a nuclear explosive. 5000 tons are enough for several hundred thousand (!) of bombs, Nagasaki type. By the year 2000 maybe 40 – 50 countries will have home-made plutonium. Plutonium production and proliferation are the most serious problems in a nuclear world. (author)

Experience on Russian military origin plutonium conversion into fast reactor nuclear fuel

International Nuclear Information System (INIS)

Grachev, A.F.; Skiba, O.V.; Bychkov, A.V.; Mayorshin, A.A.; Kisly, V.A.; Bobrov, D.A.; Osipenko, A.G.; Babikov, L.G.; Mishinev, V.B.

2001-01-01

According to the Concept of Russian Minatom on military plutonium excess utilization, the State Scientific Center of Russian Federation ''Research Institute of Atomic Reactors'' (Dimitrovgrad) has begun study on possibility of technological processing of the metal military plutonium into MOX fuel. The Program and the stages of its realization are submitted in the paper. During 1998-2000 the first stage of the Program was fulfilled and 50 kg of military origin metallic plutonium was converted to MOX fuel for the BOR-60 and BN-600 reactor. The plutonium conversion into MOX fuel is carried out under the original technology developed by SSC RIAR. It includes pyro-electrochemical process for production of fuel on the domestic equipment with the subsequent fuel pins manufacturing for the fast reactors by the vibro-packing method. The produced MOX fuel is purified from alloy additives (Ga) and corresponds to the vibro-packed fuel standard for fast reactors. The fuel pins manufacturing for BOR-60 and BN-600 reactors are carried out by the vibro-packing method on a standard procedure, which is used in SSC RIAR more than 20 years. (author)

Testing plutonium fuel assembly production for fast-neutron reactors

International Nuclear Information System (INIS)

Nougues, B.; Benhamou, A.; Bertothy, G.; Lepetit, H.

1975-01-01

The main characteristics of plutonium fuel elements for fast breeder reactors justify specific test procedures and special techniques. The specific tests relating to the Pu content consist of Pu enrichment and distribution tests, determination of the O/M ratio and external contamination tests. The specific tests performed on fuel configuration are: testing of sintered pellet diameter, testing of pin welding and checking of internal assmbly [fr

Plutonium in uranium deposits

International Nuclear Information System (INIS)

Curtis, D.; Fabryka-Martin, J.; Aguilar, R.; Attrep, M. Jr.; Roensch, F.

1992-01-01

Plutonium-239 (t 1/2 , 24,100 yr) is one of the most persistent radioactive constituents of high-level wastes from nuclear fission power reactors. Effective containment of such a long-lived constituent will rely heavily upon its containment by the geologic environment of a repository. Uranium ore deposits offer a means to evaluate the geochemical properties of plutonium under natural conditions. In this paper, analyses of natural plutonium in several ores are compared to calculated plutonium production rates in order to evaluate the degree of retention of plutonium by the ore. The authors find that current methods for estimating production rates are neither sufficiently accurate nor precise to provide unambiguous measures of plutonium retention. However, alternative methods for evaluating plutonium mobility are being investigated, including its measurement in natural ground waters. Preliminary results are reported and establish the foundation for a comprehensive characterization of plutonium geochemistry in other natural environments

Utilization of plutonium in HTGR and its actinide production

International Nuclear Information System (INIS)

Karin, S.; Brogli, R.; Lefler, W.; Nordheim, L.

1976-01-01

The HTGR is a potential plutonium consumer. In this function it would burn plutonium, produce electricity and the valuable fissile isotope U-233. The advantages of this concept are discussed but particular attention is given to the production and the destruction of the higher actinides due to the high burnup achievable in such a system. The presence of the strong resonances in the plutonium isotopes demanded an extension of the methods for evaluation of self-shielding factors, a different structure for broad groups, and the adaptation of the reactor codes to these changes. Specifications for coated plutonium particles were developed. Also procedures were determined to evaluate the alpha ray and neutron emission rates of the actinide nuclides. First cycle calculations were carried out to establish in detail the characteristics of the plutonium reactors and their results are given

Physics of plutonium recycling: volume V. Plutonium recycling in fast reactors

International Nuclear Information System (INIS)

1996-01-01

As part of a programme proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed. In this report, the multi-recycle performance of the metal-fuelled benchmark is evaluated. Benchmark results assess the reactor performance and toxicity behaviour in a closed nuclear fuel cycle for a parametric variation of the conversion ratio between 0.5 and 1.0. Results indicate that a fast burner reactor closed fuel cycle can be utilised to significantly reduce the radiotoxicity originating in the LWR cycle which would otherwise be destined for burial. (Author). tabs., figs., refs

Accident investigation board report on the May 14, 1997, chemical explosion at the Plutonium Reclamation Facility, Hanford Site,Richland, Washington - final report

International Nuclear Information System (INIS)

Gerton, R.E.

1997-01-01

On May 14, 1997, at 7:53 p.m. (PDT), a chemical explosion occur-red in Tank A- 109 in Room 40 of the Plutonium Reclamation Facility (Facility) located in the 200 West Area of the Hanford Site, approximately 30 miles north of Richland, Washington. The inactive processing Facility is part of the Plutonium Finishing Plant (PFP). On May 16, 1997, Lloyd L. Piper, Deputy Manager, acting for John D. Wagoner, Manager, U.S. Department of Energy (DOE), Richland Operations Office (RL), formally established an Accident Investigation Board (Board) to investigate the explosion in accordance with DOE Order 225. 1, Accident Investigations. The Board commenced its investigation on May 15, 1997, completed the investigation on July 2, 1997, and submitted its findings to the RL Manager on July 26, 1997. The scope of the Board's investigation was to review and analyze the circumstances of the events that led to the explosion; to analyze facts and to determine the causes of the accident; and to develop conclusions and judgments of need that may help prevent a recurrence of the accident. The scope also included the application of lessons learned from similar accidents within DOE. In addition to this detailed report, a companion document has also been prepared that provides a concise summary of the facts and conclusions of this report, with an emphasis on management issues (DOE/RL-97-63)

The plutonium recycle for PWR reactors from brazilian nuclear program

International Nuclear Information System (INIS)

Rubini, L.A.

1978-01-01

The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

The PRISM reactor as a possible option to deal with the british civilian plutonium stockpile

Energy Technology Data Exchange (ETDEWEB)

Fichtlscherer, Christopher [IANUS, TU Darmstadt (Germany); Friess, Friederike [IANUS, TU Darmstadt (Germany); ISR, Universitaet fuer Bodenkultur Wien (Boku) (Austria)

2017-07-01

Dealing with stocks of separated weapon-usable plutonium is a big challenge for our modern society. This work focuses on the British civil plutonium stockpiles, which amount to 103.3 tons. One option is seen in irradiating the plutonium in a fast reactor under development, namely the GE PRISM reactor. The PRISM reactor is a small modular, fast reactor which has a thermal power of 840 MW and an electrical output of 311 MW. It is intended to use MOX fuel and proponents claim, that it thus would be possible to produce clean energy, while making the plutonium proliferation resistant. A MCNP model of the reactor is built and depletion calculations with different target burnups of the fuel were conducted to check whether the burned material would fulfil the Spent-Fuel Standard. Particularly it was checked whether the spent fuel is self protecting, meaning that the dose rate does not fall below a limit of 1 Sv/h in 1 meter distance after a cooling period of 30 years. Based on the reactor model calculations the irradiation time to fulfill this limit for the spent fuel is calculated. Based on the needed target burnup, it can be verified, whether it is possible for the PRISM reactor to render the civil plutonium proliferation resistant in only 20 years as is is claimed by its proponents.

Feasibility study of plutonium recycling in light water reactors

International Nuclear Information System (INIS)

Tabuchi, Hideoto

1979-01-01

The feasibility of plutonium recycling in light water reactors has been studied by the Agency of Natural Resources and Energy, MITI. As the first step of the feasibility study, it was planned to charge two fuel assemblies, containing uranium-plutonium mixed oxide (MO 2 ), in the core of the Tsuruga nuclear power plant (BWR) for testing. The design of fuel the safety of these fuel and the operating characteristics of these special fuel assemblies were evaluated. The specifications of MO 2 fuel pin and fuel assembly are compared to those of present uranium oxide (UO 2 ) fuel. The weight of fissile plutonium in one MO 2 fuel assembly is 2.22 kg. The characteristics of MO 2 fuel assemblies, such as reactivity, control rod worth and power distribution can be kept similar to UO 2 fuel. The plutonium isotope ratio of the MO 2 fuel is assumed as that obtained in the fuel taken out of the Tokai No. 1 gas cooled reactor. The temperature distribution in the fuel pellets is shown, compared to that of UO 2 fuel. The linear power density is 440 w/cm at the beginning of the fuel life and 360 w/cm after the burn-up of 44,000 Mwd/t. The stress in the cladding tubes of MO 2 fuel is not different from that of UO 2 fuel. The pellet-cladding interaction (PCM1) was analyzed, utilizing the FEM code, FEAST. Concerning the calculation of resonance absorption, the space dependence of thermal neutron spectra and the nuclear behavior of hollow pellets the methods of design calculation were checked up. It was recognized that regarding the nuclear characteristics of MO 2 fuel, no special technical question remains. (Nakai, Y.)

Use of zero power plutonium reactor measurements as a support of criticality prediction for the SNR-300

International Nuclear Information System (INIS)

Pilate, S.; de Wouters, R.; Wehmann, U.; Helm, F.; Scholtyssek, W.

1978-01-01

Evaluations of criticality measurements performed in various SNEAK and Zero Power Plutonium Reactor (ZPPR) cores are compared. The best available methods of calculations (including transport theory) are used. The ZPPR results support well the trend indicated by the SNEAK evaluations for clean cores and for cores with followers; for cores with absorbers partially inserted, the agreement is only rough. Evaluations of control rod worth measurements are therefore also compared, using the routine method of calculation for SNR-300 (diffusion theory). The control rod worths are largely underestimated in SNEAK (C/E = 0.89), but only slightly underestimated in the ZPPR (C/E = 0.97). The difference in the nature of core fuel (uranium in SNEAK, plutonium in the ZPPR) could be at the origin of this discrepancy

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Malkoske, G.R.; Norton, J.L.; Slack, J.

2002-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

Improved plutonium consumption in a pressurised water reactor

International Nuclear Information System (INIS)

Puill, A.; Bergeron, J.

1995-01-01

Our goal is to improve plutonium consumption in a dedicated PWR while limiting the production of minor actinides. For lack of proving the system's reliability, we stay in reasonable configurations in which power capacity is maintained. Three ways are investigated in determining the fuel assembly: (a) standard geometry with mixed oxide in enriched uranium base; (b) standard geometry with plutonium oxide included in an inert matrix; (c) new geometry with special all-plutonium consumption varies from 50 kg/TWeh (a) up to 140 kg/TWeh (b) (upper point). The new geometry with special all plutonium rods mixed with standard uranium rods appears promising with a burning rate of 92 kg/TWeh for a production of minor actinides of 10 kg/TWeh. (authors). 3 refs., 3 figs., 4 tabs

Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

Energy Technology Data Exchange (ETDEWEB)

Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

2012-09-01

This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

The use of plutonium

International Nuclear Information System (INIS)

Marshall, W.

1980-01-01

The use of plutonium as a vital energy source producing maximum economic benefit with minimum proliferation risks is discussed. Having considered the production of plutonium, several possible plutonium fuel cycle options are identified and the economic value to be attached to plutonium for each examined. It is shown how the use of plutonium in fast reactors gives an opportunity for a non-proliferation policy not available when plutonium is used only in thermal reactors. From the technical considerations reviewed concerning plutonium and fast reactors it is shown that an economic regime involving international trade in spent thermal reactor fuel is possible which benefits equally those countries with fast reactors and those without and also assists in avoiding the proliferation of nuclear weapons. (U.K.)

Uranium-plutonium fuel for fast reactors

International Nuclear Information System (INIS)

Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

1996-01-01

Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

Decision model for evaluating reactor disposition of excess plutonium

International Nuclear Information System (INIS)

Edmunds, T.

1995-02-01

The US Department of Energy is currently considering a range of technologies for disposition of excess weapon plutonium. Use of plutonium fuel in fission reactors to generate spent fuel is one class of technology options. This report describes the inputs and results of decision analyses conducted to evaluate four evolutionary/advanced and three existing fission reactor designs for plutonium disposition. The evaluation incorporates multiple objectives or decision criteria, and accounts for uncertainty. The purpose of the study is to identify important and discriminating decision criteria, and to identify combinations of value judgments and assumptions that tend to favor one reactor design over another

Separation of Plutonium from Irradiated Fuels and Targets

Energy Technology Data Exchange (ETDEWEB)

Gray, Leonard W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Holliday, Kiel S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murray, Alice [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thompson, Major [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thorp, Donald T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Yarbro, Stephen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Venetz, Theodore J. [Hanford Site, Benton County, WA (United States)

2015-09-30

Spent nuclear fuel from power production reactors contains moderate amounts of transuranium (TRU) actinides and fission products in addition to the still slightly enriched uranium. Originally, nuclear technology was developed to chemically separate and recover fissionable plutonium from irradiated nuclear fuel for military purposes. Military plutonium separations had essentially ceased by the mid-1990s. Reprocessing, however, can serve multiple purposes, and the relative importance has changed over time. In the 1960’s the vision of the introduction of plutonium-fueled fast-neutron breeder reactors drove the civilian separation of plutonium. More recently, reprocessing has been regarded as a means to facilitate the disposal of high-level nuclear waste, and thus requires development of radically different technical approaches. In the last decade or so, the principal reason for reprocessing has shifted to spent power reactor fuel being reprocessed (1) so that unused uranium and plutonium being recycled reduce the volume, gaining some 25% to 30% more energy from the original uranium in the process and thus contributing to energy security and (2) to reduce the volume and radioactivity of the waste by recovering all long-lived actinides and fission products followed by recycling them in fast reactors where they are transmuted to short-lived fission products; this reduces the volume to about 20%, reduces the long-term radioactivity level in the high-level waste, and complicates the possibility of the plutonium being diverted from civil use – thereby increasing the proliferation resistance of the fuel cycle. In general, reprocessing schemes can be divided into two large categories: aqueous/hydrometallurgical systems, and pyrochemical/pyrometallurgical systems. Worldwide processing schemes are dominated by the aqueous (hydrometallurgical) systems. This document provides a historical review of both categories of reprocessing.

Pulsed reactors: A dissenting view

International Nuclear Information System (INIS)

Ganev, I.Kh.; Orlov, V.V.

1995-01-01

The preceding article, by G.A. Ivanov et al., contains interesting estimates of the expanded production of plutonium in thermonuclear explosions initiated by plutonium charges. It must be noted that more than 40 years of efforts, despite some technical successes, have not led to a fast-reactor technology suitable for large-scale power production. This explains the incessant search for a nuclear technology for the future and the renewed interest in accelerator, hybrid, and explosive approaches to plutonium production. The success of such efforts will depend largely on the formulation of goals and the choice of the principal criteria. It is appropriate to discuss these issues here because the adoption of the rate of plutonium production or the plutonium doubling time as the principal criterion sets the stage for the repetition of previous errors. However, as a preliminary, I would like to question some categorical assertions that were made by Ivanov et al. without the presentation of adequate supporting data (the assertions that open-quotes the creation of an power industry on the basis of ordinary breeder reactors is practically impossibleclose quotes and that open-quotes adequate power generation in the 21st centuryclose quotes is impossible). In fact, it is simple to calculate that, given a realistic doubling time for fast reactors of ∼10 years and the plutonium produced by thermal reactors (around 10 12 W), it would be possible, if so desired, to introduce power far exceeding 10 14 W in the 21st century

Accelerator-based conversion (ABC) of reactor and weapons plutonium

Energy Technology Data Exchange (ETDEWEB)

Jensen, R.J.; Trapp, T.J.; Arthur, E.D.; Bowman, C.D.; Davidson, J.W.; Linford, R.K.

1993-06-01

An accelerator-based conversion (ABC) system is presented that is capable of rapidly burning plutonium in a low-inventory sub-critical system. The system also returns fission power to the grid and transmutes troublesome long-lived fission products to short lived or stable products. Higher actinides are totally fissioned. The system is suited not only to controlled, rapid burning of excess weapons plutonium, but to the long range application of eliminating or drastically reducing the world total inventory of plutonium. Deployment of the system will require the successful resolution of a broad range of technical issues introduced in the paper.

Accelerator-based conversion (ABC) of reactor and weapons plutonium

International Nuclear Information System (INIS)

Jensen, R.J.; Trapp, T.J.; Arthur, E.D.; Bowman, C.D.; Davidson, J.W.; Linford, R.K.

1993-01-01

An accelerator-based conversion (ABC) system is presented that is capable of rapidly burning plutonium in a low-inventory sub-critical system. The system also returns fission power to the grid and transmutes troublesome long-lived fission products to short lived or stable products. Higher actinides are totally fissioned. The system is suited not only to controlled, rapid burning of excess weapons plutonium, but to the long range application of eliminating or drastically reducing the world total inventory of plutonium. Deployment of the system will require the successful resolution of a broad range of technical issues introduced in the paper

Plutonium use - present status and perspectives

International Nuclear Information System (INIS)

Dievoet, J. van; Fossoul, E.; Jonckheere, E.; Bemden, E. van den

1977-01-01

Plutonium is being produced in increasing quantities in the so-called proven reactors, which are mostly of the light-water type. Evaluation of this production on a world scale shows that it would be theoretically possible to construct a large number of breeders and thus to make the best use of the intrinsic qualities of plutonium as a fissionable material, while considerably reducing the consumption of uranium. This source of plutonium is nevertheless dependent on an essential stage of the fuel cycle, namely reprocessing of irradiated fuel. The long delays in installing an adequate world reprocessing capacity are substantially weakening the prospects for the introduction of breeders. Furthermore, the critical situation as regards reprocessing may delay the development of complementary reprocessing methods for fuels with a high plutonium content and high burn-up. When it is recalled that fast reactors themselves may suffer some delay in their technological development, if only because of the intention to build power plants of very high unit capacity immediately, it must be concluded that another use will have to be considered for the plutonium available in future -use in thermal reactors, i.e. recycling. The recycling of plutonium is a well-known technique today and the objections which could be raised against it hardly stand up to analysis. Utilization of plutonium offers an appreciable saving in terms of uranium and separative work units, the consumption being of a low order of magnitude in comparison with the total amount of plutonium needed for the eventual fabrication of the first fast reactor cores. It can also be shown that immediate reprocessing of the recycling fuel is not essential for the economics of the concept. Temporary storage of recycled fuel has the advantage of concentrating plutonium in a particularly safe form, namely in irradiated plutonium-bearing fuel assemblies. Lastly, recycling offers such flexibility that it does not in practice represent

Quantities of actinides in nuclear reactor fuel cycles

International Nuclear Information System (INIS)

Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000 MW reactors of the following types: water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breeder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium, and recycled uranium. The radioactivity levels of plutonium, americium, and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the United States nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium processed in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing and fuel fabrication to eliminate the off-site transport of separated plutonium. (U.S.)

The generation of denatured reactor plutonium by different options of the fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Broeders, C.H.M.; Kessler, G. [Inst. for Neutron Physics and Reactor Technology, Research Center Karlsruhe (Germany)

2006-11-15

Denatured (proliferation resistant) reactor plutonium can be generated in a number of different fuel cycle options. First denatured reactor plutonium can be obtained if, instead of low enriched U-235 PWR fuel, re-enriched U-235/U-236 from reprocessed uranium is used (fuel type A). Also the envisaged existing 2,500 t of reactor plutonium (being generated world wide up to the year 2010), mostly stored in intermediate fuel storage facilities at present, could be converted during a transition phase into denatured reactor plutonium by the options fuel type B and D. Denatured reactor plutonium could have the same safeguards standard as present low enriched (<20% U-235) LWR fuel. It could be incinerated by recycling once or twice in PWRs and subsequently by multi-recycling in FRs (CAPRA type or IFRs). Once denatured, such reactor plutonium could remain denatured during multiple recycling. In a PWR, e.g., denatured reactor plutonium could be destroyed at a rate of about 250 kg/GWey. While denatured reactor plutonium could be recycled and incinerated under relieved IAEA safeguards, neptunium would still have to be monitored by the IAEA in future for all cases in which considerable amounts of neptunium are produced. (orig.)

Fuel qualification issues and strategies for reactor-based surplus plutonium disposition

International Nuclear Information System (INIS)

Cowell, B.S.; Copeland, G.L.; Moses, D.L.

1997-08-01

The Department of Energy (DOE) has proposed irradiation of mixed-oxide (MOX) fuel in existing commercial reactors as a disposition method for surplus plutonium from the weapons program. The burning of MOX fuel in reactors is supported by an extensive technology base; however, the infrastructure required to implement reactor-based plutonium disposition does not exist domestically. This report identifies and examines the actions required to qualify and license weapons-grade (WG) plutonium-based MOX fuels for use in domestic commercial light-water reactors (LWRs)

Estimating plutonium production fron long-lived radionuclides in permanent structural components of production reactor cores

International Nuclear Information System (INIS)

Fetter, S.

1992-01-01

This paper reports that the United States and the Soviet Union face critical decisions about the future of plutonium production for nuclear weapons. Both countries could eliminate the economic burden of rebuilding their production complexes by agreeing to ban the production of plutonium for weapons. Such an agreement could also provide important national-security benefits by reinforcing the Non-Proliferation Treaty and by diminishing the ability of both nations to break out of nuclear-arms reduction agreements at a later time - especially if the plutonium in the warheads eliminated by the arms reduction agreements is put under safeguards. A production cutoff would be verifiable

Control of civilian plutonium inventories using burning in a non-fertile fuel

Energy Technology Data Exchange (ETDEWEB)

Oversby, V.M. [Lawrence Livermore National Lab., CA (United States); McPheeters, C.C. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439-4837 (United States); Degueldre, C. [Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Paratte, J.M. [Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland)

1997-05-01

The increasing inventories of plutonium generated by commercial nuclear power production represent a potential source for proliferation of nuclear weapons. To address this threat we propose separating the plutonium from the other constituents of commercial reactor spent fuel and burning it in a non-fertile fuel based on a zirconium dioxide matrix. The separation can be performed by the Purex process currently in use, but we recommend development of a more compact separation technology that would produce less secondary waste than currently used technology and would allow for more stringent accounting of plutonium inventories. The non-fertile fuel is designed for use in conventional light water power reactors and does not require development of new reactor technology. (orig.).

Control of civilian plutonium inventories using burning in a non-fertile fuel

Science.gov (United States)

Oversby, V. M.; McPheeters, C. C.; Degueldre, C.; Paratte, J. M.

1997-05-01

The increasing inventories of plutonium generated by commercial nuclear power production represent a potential source for proliferation of nuclear weapons. To address this threat we propose separating the plutonium from the other constituents of commercial reactor spent fuel and burning it in a non-fertile fuel based on a zirconium dioxide matrix. The separation can be performed by the Purex process currently in use, but we recommend development of a more compact separation technology that would produce less secondary waste than currently used technology and would allow for more stringent accounting of plutonium inventories. The non-fertile fuel is designed for use in conventional light water power reactors and does not require development of new reactor technology.

Disposition of weapons-grade plutonium in Westinghouse reactors

International Nuclear Information System (INIS)

Alsaed, A.A.; Adams, M.

1998-03-01

The authors have studied the feasibility of using weapons-grade plutonium in the form of mixed-oxide (MOX) fuel in existing Westinghouse reactors. They have designed three transition Cycles from an all LEU core to a partial MOX core. They found that four-loop Westinghouse reactors such as the Vogtle power plant are capable of handling up to 45 percent weapons-grade MOX loading without any modifications. The authors have also designed two kinds of weapons-grade MOX assemblies with three enrichments per assembly and four total enrichments. Wet annular burnable absorber (WABA) rods were used in all the MOX feed assemblies, some burned MOX assemblies, and some LEU feed assemblies. Integral fuel burnable absorber (IFBA) was used in the rest of the LEU feed assemblies. The average discharge burnup of MOX assemblies was over 47,000 MWD/MTM, which is more than enough to meet the open-quotes spent fuel standard.close quotes One unit is capable of consuming 0.462 MT of weapons-grade plutonium per year. Preliminary analyses showed that important reactor physics parameters for the three transitions cycles are comparable to those of LEU cores including boron levels, reactivity coefficients, peaking factors, and shutdown margins. Further transient analyses will need to be performed

Perspective on plutonium

International Nuclear Information System (INIS)

Sun, L.S.

1993-01-01

This paper is intended as a brief overview on the element plutonium. Plutonium is the first primarily man-made element to play a significant role not only in technological development, but also in the economic growth of many countries. The importance of plutonium centers around its enormous energy making it ideal for wide-scale use in reactors, while the nuclear industry continues to work toward improving safety and efficiency of plutonium as a reactor fuel politicians and the public still debate over the safety and benefits of nuclear power. (30 refs.)

The benefits of an advanced fast reactor fuel cycle for plutonium management

International Nuclear Information System (INIS)

Hannum, W.H.; McFarlane, H.F.; Wade, D.C.; Hill, R.N.

1996-01-01

The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium and long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a 'focus area' for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed

PLUTONIUM METALLIC FUELS FOR FAST REACTORS

Energy Technology Data Exchange (ETDEWEB)

STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

2007-02-07

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

Radionuclide characterization of graphite stacks from plutonium production reactors of the Siberian group of chemical enterprises

International Nuclear Information System (INIS)

Bushuev, A.V.; Verzilov, Yu.M.; Zubarev, V.N.

2001-01-01

The residual radionuclide concentrations and distributions in graphite from moderator stack of plutonium production reactors at Tomsk-7 have been investigated. It was found that the dominant activity of graphite is 14 C. To gain information on surface and volume contamination of graphite blocks from the moderator stack, the special sets of samples were collected and assayed. The schemes are proposed for evaluation of individual radionuclide inventories together with results of the evaluations performed. (author)

Project management plan for the 105-C Reactor interim safe storage project. Revision 1

International Nuclear Information System (INIS)

Miller, R.L.

1997-01-01

In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

A new concept of fast reactors, the potentialities of burning in them of actinoid and weapon-grade plutonium

International Nuclear Information System (INIS)

Murogov, V.M.; Troyanov, M.F.; Ilyunin, V.G.; Rudneva, V.Ya.

1992-01-01

The approach to a possible solution of the problem of peaceful utilization of weapon-grade plutonium released in the result of nuclear disarmament in Russia is given in the repot. As the most safe, ecologically acceptable and economically effective way of the plutonium utilization is the usage of such plutonium as a fuel for atomic power station. It is proposed to decide the problem on the basis of BN-600 and BN-800 reactors. In the approach, thorium could be used as a fertile material. The secondary nuclear fuel U-233 is expedient to use in light-water reactors of new generation. (author)

Plutonium Consumption Program, CANDU Reactor Project final report

Energy Technology Data Exchange (ETDEWEB)

1994-07-31

DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

Plutonium Consumption Program, CANDU Reactor Project final report

International Nuclear Information System (INIS)

1994-01-01

DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

Directory of Open Access Journals (Sweden)

Salahuddin Asif

2013-01-01

Full Text Available Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington. Addendum (Final Environmental Impact Statement)

Energy Technology Data Exchange (ETDEWEB)

1992-12-01

The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE`s responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which contains additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period.

Safety issues at the defense production reactors

International Nuclear Information System (INIS)

1987-01-01

The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors endash the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. The report is necessarily based on a limited review of the defense production reactors. It does not address whether any of the reactors are ''safe,'' because such an analysis would involve a determination of acceptable risk endash a matter of obvious importance, but one that was beyond the purview of the committee. It also does not address whether the safety of the production reactors is comparable to that of commercial nuclear power stations, because even this narrower question extended beyond the charge to the committee and would have involved detailed analyses that the committee could not undertake

Metal plutonium conversion to components of nuclear reactor fuel

International Nuclear Information System (INIS)

Subbotin, V.G.; Panov, A.V.; Mashirev, V.P.

2000-01-01

Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

Metal plutonium conversion to components of nuclear reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Subbotin, V.G.; Panov, A.V. [Russian Federal Nuclear Center, ALL-Russian Science and Research, Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

2000-07-01

Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

Fast reactors: potential for power

International Nuclear Information System (INIS)

1983-02-01

The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

Light water breeder reactor using a uranium-plutonium cycle

International Nuclear Information System (INIS)

Radkowsky, A.; Chen, R.

1990-01-01

This patent describes a light water receptor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: a prebreeder section having plutonium fuel containing a Pu-241 component, the prebreeder section being operable to produce enriched plutonium having an increased Pu-241 component; and a breeder section for receiving the enriched plutonium from the prebreeder section, the breeder section being operable for breeding fissile material from the enriched plutonium fuel. This patent describes a method of operating a light water nuclear reactor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: operating the prebreeder to produce enriched plutonium fuel having an increased Pu-241 component; fueling a breeder section with the enriched plutonium fuel to breed the fissile material

Seismic qualification of safety class components in non-reactor nuclear facilities at Hanford site

International Nuclear Information System (INIS)

Ocoma, E.C.

1989-01-01

This paper presents the methods used during the walkdowns to compile as-built structural information to seismically qualify or verify the seismic adequacy of safety class components in the Plutonium Finishing Plant complex. The Plutonium finishing Plant is a non-reactor nuclear facility built during the 1950's and was designed to the Uniform Building Code criteria for both seismic and wind events. This facility is located at the US Department of Energy Hanford Site near Richland, Washington

Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 2

International Nuclear Information System (INIS)

1976-08-01

This environmental statement assesses the impacts of the implementation of plutonium recycle in the LWR industry. It is based on assumptions that are intended to reflect conservatively an acceptable level of the application of current technology. It is not intended to be a representation of the ''as low as reasonably achievable'' (ALARA) philosophy. This generic environmental statement discusses the anticipated effects of recycling plutonium in light water nuclear power reactors. It is based on about 30 years of experience with the element in the context of a projected light water nuclear power industry that is already substantial. A background perspective on plutonium, its safety, and its recycling as a reactor fuel is presented

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Transmutation of plutonium in pebble bed type high temperature reactors

International Nuclear Information System (INIS)

Bende, E.E.

1997-01-01

The pebble bed type High Temperature Reactor (HTR) has been studied as a uranium-free burner of reactor grade plutonium. In a parametric study, the plutonium loading per pebble as well as the type and size of the coated particles (CPs) have been varied to determine the plutonium consumption, the final plutonium burnup, the k ∞ and the temperature coefficients as a function of burnup. The plutonium loading per pebble is bounded between 1 and 3 gr Pu per pebble. The upper limit is imposed by the maximal allowable fast fluence for the CPs. A higher plutonium loading requires a longer irradiation time to reach a desired burnup, so that the CPs are exposed to a higher fast fluence. The lower limit is determined by the temperature coefficients, which become less negative with increasing moderator-actinide ratio. A burnup of about 600 MWd/kgHM can be reached. With the HTR's high efficiency of 40%, a plutonium supply of 1520 kg/GW e a is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GW e a, respectively. (author)

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

Energy Technology Data Exchange (ETDEWEB)

GERBER MS

2009-04-28

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

Expert system for estimating LWR plutonium production

International Nuclear Information System (INIS)

Sandquist, G.M.

1988-01-01

An Artificial Intelligence-Expert System called APES (Analysis of Proliferation by Expert System) has been developed and tested to permit a non proliferation expert to evaluate the capability and capacity of a specified LWR reactor and PUREX reprocessing system for producing and separating plutonium even when system information may be limited and uncertain. APES employs an expert system coded in LISP and based upon an HP-RL (Hewlett Packard-Representational Language) Expert System Shell. The user I/O interface communicates with a blackboard and the knowledge base which contains the quantitative models required to describe the reactor, selected fission product production and radioactive decay processes, Purex reprocessing and ancillary knowledge

Plutonium-enriched thermal fuel production experience in Belgium

International Nuclear Information System (INIS)

LeBlanc, J.M.

1983-01-01

Taking into account the strategic aspects of nuclear energy such as availability and sufficiency of resources and independence of energy supply, most countries planning to use plutonium look mainly to its use in fast reactors. However, by recycling the recovered uranium and plutonium in light water reactors, the saving of the uranium that would otherwise be required could already be higher than 35%. Therefore, until fast reactors are introduced, for macro- or microeconomic reasons, the plutonium recycle option seems to be quite valuable for countries having the plutonium technology. In Belgium, Belgonucleaire has been developing the plutonium technology for more than 20 yr and has operated a mixed oxide fuel fabrication plant since 1973. The past ten years of plant operation have provided for many improvements and relevant new documented experiences establishing a basis for new modifications that will be beneficial to the intrinsic quality, overall safety, and economy of the fuel

233-S Plutonium Concentration Facility data quality objectives

International Nuclear Information System (INIS)

Encke, D.B.

1996-08-01

This document is a summary of the decision-making associated with the Data Quality Objective process that pertains to the characterization activities in the 233-S Plutonium Concentration Facility at the Hanford Site in Richland, Washington. The 233-S Plutonium Concentration Facility is located adjacent to, and north of, the REDOX Plant. The facility was used to concentrate the plutonium nitrate product solution from the REDOX facility. The 233-S Pipe Gallery, Control Room, SWP Change Room, Toilet, Equipment Room and the Electrical Cubicle are currently scheduled for decontamination and cleanout to support future demolition (D and D). Identification of the radiological contamination and presence of hazardous materials is needed to allow for disposal of the D and D debris

A preliminary analysis of the reactor-based plutonium disposition alternative deployment schedules

International Nuclear Information System (INIS)

Zurn, R.M.

1997-09-01

This paper discusses the preliminary analysis of the implementation schedules of the reactor-based plutonium disposition alternatives. These schedule analyses are a part of a larger process to examine the nine decision criteria used to determine the most appropriate method of disposing of U.S. surplus weapons plutonium. The preliminary analysis indicates that the mission durations for the reactor-based alternatives range from eleven years to eighteen years and the initial mission fuel assemblies containing surplus weapons-usable plutonium could be loaded into the reactors between nine and fourteen years after the Record of Decision

Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

International Nuclear Information System (INIS)

1978-01-01

Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

Decommissioning the Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I)

International Nuclear Information System (INIS)

Harper, J.R.; Garde, R.

1981-11-01

The Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I) was decommissioned at the Los Alamos National Laboratory, Los Alamos, New Mexico, in 1980. The LAMPRE I was a sodium-cooled reactor built to develop plutonium fuels for fast breeder applications. It was retired in the mid-1960s. This report describes the decommissioning procedures, the health physics programs, the waste management, and the costs for the operation

Kr-85 signatures for various plutonium production schemes

Energy Technology Data Exchange (ETDEWEB)

Stanoszek, Paul [Carl Friedrich von Weizsaecker-Centre for Science and Peace Research (Germany)

2009-07-01

Kr-85 is considered to be the best atmospheric indicator of unreported weapon-grade material production. This fact is based on the half-life of 10.76 years of Kr-85 and its chemical inactivity, which makes it even detectable after extended periods of cooling time. Kr-85 is produced as fission product during nuclear reactor operation and remains in the fuel until reprocessing starts. In order to determine the detectability of plutonium production the Kr-85 source term has to be assessed. The important issue of this presentation is the question on the minimum signal that an inspector can expect under the assumption that a proliferator minimizes his Kr-85 generation in order to circumvent a Kr-85 detection. A further assumption is that for nuclear weapon production a burn-up of typically around 2 MWd/kg is used. In addition, if clandestine plutonium production takes place, the source term might be used to estimate the amount of separated plutonium. The methodology of this study is based on a linkage between MCNPX and MATLAB. All results for actinide concentrations and Kr-85 are evaluated for different enrichments of U-235 and compared to known literature data. The Kr-85 source term per kilogram plutonium depends on the enrichments. As a result the lowest Kr-85 source term is found for depleted uranium.

Development of small and medium reactors for power and heat production

International Nuclear Information System (INIS)

Becka, J.

1978-01-01

Data are given on the current state of development of small and medium-power reactors designed mainly for electric power production in small power grids, for heat production for small- and medium-power desalination plants with possible electric power generation, for process steam production and heat development for district heating systems, again combined with electric power generation, and for propelling big and fast passenger ships. A diagram is shown of the primary system of an integrated PWR derived from the Otto Hahn reactor. The family is listed of the standard sizes of the integral INTERATOM company pressurized water reactors. Also listed are the specifications and design of CAS 2CG and AS 3G type reactors used mainly for long-distance heating systems. (J.B.)

Reduction of worldwide plutonium inventories using conventional reactors and advanced fuels: A systems study

International Nuclear Information System (INIS)

Krakowski, R.A.; Bathke, C.G.; Chodak, P. III

1997-01-01

The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs of a variety of mixed-oxide forms is examined by means of a cost-based plutonium-flow systems model that includes an approximate measure of proliferation risk. The impact of plutonium recycle in a number of forms is examined, including the introduction of nonfertile fuels into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo-reactor plutonium inventories

Prospects for the establishment of plutonium recycle in thermal reactors in the Foratom countries. Status and assessment

International Nuclear Information System (INIS)

Chamberlain, A.; Melches, C.

1977-01-01

The paper reviews the technical status of plutonium recycle in thermal reactors in the Foratom countries and assesses the prospect for it becoming established in the future with the implicit assumptions that uranium oxide reprocessing capacity will be installed commensurate with the projected programmes for thermal reactor installation and that there will be no insuperable environmental, security or safeguards obstacles to the use of plutonium as a fuel. It is argued that the feasibility of using plutonium as an alternative to 235 U as the fuel for thermal reactors, particularly LWRs, has been extensively demonstrated by a number of Foratom countries and the main problem areas are fuel fabrication and fuel reprocessing. Mixed-oxide fuel fabrication has been well established on the prototype plant scale using low-irradiation plutonium, but it is recognized that the future design of production-scale plants will need to cater for the significantly higher radiation levels from high burnup plutonium and meet stricter environmental requirements on operator dosage and waste arisings. The main constraint on the establishment of recycle up to now has been the lack of available plutonium owing to the absence of significant uranium-oxide fuel reprocessing capacity. An assessment of the plutonium arisings in Europe, based on the projected uranium-oxide reprocessing capacity, shows that by 1990 plutonium, surplus to FBR requirements, should be accumulating by about 10t/a, sufficient to fuel about 8000MW(e) of LWRs. A further constraint would then be the availability and technical problems of mixed-oxide reprocessing, which is one of the areas identified for international collaboration. It is concluded that whilst there is unlikely to be substantial recycle of plutonium in thermal reactors in the Foratom countries before the early 1990s, an incentive could possibly arise about that time. The strength of this incentive will depend on a number of factors including the status of

Radiological safety aspects of handling plutonium

International Nuclear Information System (INIS)

Sundararajan, A.R.

2016-01-01

Department of Atomic Energy in its scheme of harnessing the nuclear energy for electrical power generation and strategic applications has given a huge role to utilization of plutonium. In the power production programme, fast reactors with plutonium as fuel are expected to play a major role. This would require establishing fuel reprocessing plants to handle both thermal and fast reactor fuels. So in the nuclear fuel cycle facilities variety of chemical, metallurgical, mechanical operations have to be carried out involving significant inventories of "2"3"9 Pu and associated radionuclides. Plutonium is the most radiotoxic radionuclide and therefore any facility handling it has to be designed and operated with utmost care. Two problems of major concern in the protection of persons working in plutonium handling facilities are the internal exposure to the operating personnel from uptake of plutonium and transplutonic nuclides as they are highly radiotoxic and the radiation exposure of hands and eye lens during fuel fabrication operations especially while handling recycled high burn up plutonium. In view of the fact that annual limit for intake is very small for "2"3"9Pu and its radiation emission characteristics are such that it is a huge challenge for the health physicists to detect Pu in air and in workers. This paper discusses the principles and practices followed in providing radiological surveillance to workers in plutonium handling areas. The challenges in protecting the workers from receiving exposures to hands and eye lens in handling high burn up plutonium are also discussed. The sites having Pu fuel cycle facilities should have trained medical staff to handle cases involving excessive intake of plutonium. (author)

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Slack, J.; Norton, J.L.; Malkoske, G.R.

2003-01-01

therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

On the Burning of Plutonium Originating from Light Water Reactor Use in a Fast Molten Salt Reactor—A Neutron Physical Study

Directory of Open Access Journals (Sweden)

Bruno Merk

2015-11-01

Full Text Available An efficient burning of the plutonium produced during light water reactor (LWR operation has the potential to significantly improve the sustainability indices of LWR operations. The work offers a comparison of the efficiency of Pu burning in different reactor configurations—a molten salt fast reactor, a LWR with mixed oxide (MOX fuel, and a sodium cooled fast reactor. The calculations are performed using the HELIOS 2 code. All results are evaluated against the plutonium burning efficiency determined in the Consommation Accrue de Plutonium dans les Réacteurs à Neutrons RApides (CAPRA project. The results are discussed with special view on the increased sustainability of LWR use in the case of successful avoidance of an accumulation of Pu which otherwise would have to be forwarded to a final disposal. A strategic discussion is given about the unavoidable plutonium production, the possibility to burn the plutonium to avoid a burden for the future generations which would have to be controlled.

Programme of research and development on plutonium recycling in light-water reactors

International Nuclear Information System (INIS)

1979-01-01

This is the third annual progress report concerning the programme on plutonium recycling in light-water reactors (indirect action) of the Commission of the European Communities. It covers the year 1978 and follows the annual reports for 1977 (EUR 6002 EN) and 1976 (EUR 5780). The preliminary results obtained under the 1975-79 programme indicate that: (a) assuming that plutonium recycling in light-water reactors is industrially developed by the end of the century, the foreseeable radiological impact on both workers and the general public can be maintained within the limits of current radiation protection standards; (b) on the whole, there is a good knowledge and mastery of the specific aspects involved in the plutonium recycling in light-water reactors and in particular they indicate that plutonium fuels have a similar behaviour to uranium fuels

Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity [O1] - Fundamentals of P{sup 3} Mechanism and Methodology Development for Plutonium Categorization

Energy Technology Data Exchange (ETDEWEB)

Saito, Masaki [Research Laboratory for Nuclear Reactor, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro, Tokyo, 1528550 (Japan)

2009-06-15

'Protected Plutonium Production (P{sup 3})' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MAs). Doping the small amount of MAs such as {sup 237}Np or {sup 241}Am with large neutron capture cross-section into the uranium fuel to enhance the production of {sup 238}Pu or {sup 242}Pu, which have high spontaneous fission neutron source or also high decay heat to makes the process of the nuclear weapon manufacture and maintenance technologically difficult, can be effective for improving the isotopic barrier of proliferation resistance of the plutonium in thermal reactors. Super weapon grade plutonium could be produced in the blanket of a conventional FBR. However, by increasing the {sup 238}Pu or {sup 242}Pu ratio in the total plutonium by MAs doping into the fresh blanket, the protected plutonium with high proliferation-resistance can be bred. A new evaluation function, 'attractiveness', defined as a ratio of potential of fission yield to the technological difficulties of nuclear explosive device, has been proposed to evaluate the proliferation resistance of Pu based on the nuclear material property for Plutonium Categorization. In the conference, the fundamentals of P{sup 3} mechanism by transmutation of MA, and the comparison of the 'attractiveness' of the Pu produced in advanced reactors based on P{sup 3} mechanism and in the conventional reactors will be presented. Instead of the geological disposal or just their burning of MAs by the fission reaction, they should be treated as valuable fertile materials to enhance the proliferation resistance of plutonium produced in the thermal and fast breeder reactors for peace and sustainable prosperity in future. Acknowledgement: Some parts of this work have been supported by the Ministry of Education, Culture, Sports, Science and Technology in Japan. (authors)

Economic analysis of self-generated plutonium recycling in light water reactor

International Nuclear Information System (INIS)

Deguchi, Morimoto; Hirabayashi, Fumio; Yumoto, Ryozo

1978-01-01

This paper describes on the economics of plutonium recycle to light water reactors (LWRs). In the situation that plutonium market does not exist, it is realistic for utilities to recycle the self-generated plutonium to their own reactors. The economic incentive to recycle self-generated plutonium, plutonium fuel fabrication penalty, and the dependence of fuel cycle cost on fuel cycle cost parameters are considered. In recycling self-generated plutonium, two alternatives for fuel element design are feasible. Those are the all-plutonium design and the island design. In the present analysis, the all-plutonium design was chosen for PWRs. The calculation of reactivity variation along with burnup for both uranium fuel and plutonium fuel was done with LASER-PNC code. Plutonium inventory and other nuclear data were calculated with CHAIN code. It is expected that equilibrium composition is reached after 5 or 6 times of recycling. For the calculation of fuel cycle cost, MITCOST code was used. The recent increase in the prices of uranium ore, enrichment and reprocessing services was taken into account. The fuel cycle cost of plutonium recycle is lower than that of uranium fuel cycle within a certain limit of plutonium fabrication penalty. It is shown that the fabrication penalty of about 1250 dollar/kgHM for each plutonium successive recycle reduces the cost difference to zero. The change in other cost components affects break-even fabrication penalty, in which the fuel cycle cost of plutonium recycle is equal to that of uranium cycle. (Kato, T.)

Reactor-Based Plutonium Disposition: Opportunities, Options, and Issues

International Nuclear Information System (INIS)

Greene, S.R.

1999-01-01

The end of the Cold War has created a legacy of surplus fissile materials (plutonium and highly enriched uranium) in the United States (U.S.) and the former Soviet Union. These materials pose a danger to national and international security. During the past few years, the U.S. and Russia have engaged in an ongoing dialog concerning the safe storage and disposition of surplus fissile material stockpiles. In January 1997, the Department of Energy (DOE) announced the U. S. would pursue a dual track approach to rendering approximately 50 metric tons of plutonium inaccessible for use in nuclear weapons. One track involves immobilizing the plutonium by combining it with high-level radioactive waste in glass or ceramic ''logs''. The other method, referred to as reactor-based disposition, converts plutonium into mixed oxide (MOX) fuel for nuclear reactors. The U.S. and Russia are moving ahead rapidly to develop and demonstrate the technology required to implement the MOX option in their respective countries. U.S. MOX fuel research and development activities were started in the 1950s, with irradiation of MOX fuel rods in commercial light water reactors (LWR) from the 1960s--1980s. In all, a few thousand MOX fuel rods were successfully irradiated. Though much of this work was performed with weapons-grade or ''near'' weapons-grade plutonium--and favorable fuel performance was observed--the applicability of this data for licensing and use of weapons-grade MOX fuel manufactured with modern fuel fabrication processes is somewhat limited. The U.S. and Russia are currently engaged in an intensive research, development, and demonstration program to support implementation of the MOX option in our two countries. This paper focuses on work performed in the U.S. and provides a brief summary of joint U.S./Russian work currently underway

Reactor-based plutonium disposition: Opportunities, options, and issues

International Nuclear Information System (INIS)

Greene, S.

2000-01-01

The end of the Cold War has created a legacy of surplus fissile materials (plutonium and highly enriched uranium) in the United States (U.S.) and the former Soviet Union. These materials pose a danger to national and international security. During the past few years, the U.S. and Russia have engaged in an ongoing dialog concerning the safe storage and disposition of surplus fissile material stockpiles. In January 1997, the Department of Energy (DOE) announced the U.S. would pursue a dual track approach to rendering approximately 50 metric tons of plutonium inaccessible for use in nuclear weapons. One track involves immobilizing the plutonium by combining it with high-level radioactive waste in glass or ceramic ''logs''. The other method, referred to as reactor-based disposition, converts plutonium into mixed oxide (MOX) fuel for nuclear reactors. The U.S. and Russia are moving ahead rapidly to develop and demonstrate the technology required to implement the MOX option in their respective countries. U.S. MOX fuel research and development activities were started in the 1950s with irradiation of MOX fuel rods in commercial light water reactors (LWR) from the 1960s-1980s. In all, a few thousand MOX fuel rods were successfully irradiated. Though much of this work was performed with weapons-grade or ''near'' weapons-grade plutonium - and favorable fuel performance was observed - the applicability of this data for licensing and use of weapons-grade MOX fuel manufactured with modem fuel fabrication processes is somewhat limited. The U.S. and Russia are currently engaged in an intensive research, development, and demonstration program to support implementation of the MOX option in our two countries. This paper focuses on work performed in the U.S. and provides a brief summary of joint U.S./Russian work currently underway. (author)

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

Science.gov (United States)

Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

2017-12-01

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

Small size modular fast reactors in large scale nuclear power

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Dragunov, U.G.; Stepanov, V.S.; Klimov, N.N.; Kopytov, I.I.; Krushelnitsky, V.N.

2005-01-01

The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

On detection of the possible use of VVERs for unreported production of plutonium. Final report for the period July 1988 - December 1989

International Nuclear Information System (INIS)

Simov, R.; Nelov, N.; Stoyanova, I.; Kovachev, N.; Yonchev, P.

1989-01-01

The study includes an analysis of the feasibility of unreported production of plutonium-239 in VVER-440 reactors. It is shown that for VVER-440 reactors 36 natural uranium oxide fuel assemblies in the peripheral region of the core need to be loaded to produce 8 kg of extra plutonium in one cycle. Substituting the peripheral fuel assemblies with natural uranium oxide fuel assemblies, the changes in the power peaking are negligible and do not affect reactor safety. Unreported production outside the core is not practical due to physical and mechanical constraints, low flux level, etc. The feasibility of unreported removal of irradiated material in spent fuel cask has been also assessed. After about a month cooling time, still within the refueling period, the irradiated natural uranium fuel assemblies could be removed off-site without significant health hazard to the workers. To improve the effectiveness of the safeguards objectives, additional inspection activities are suggested. 10 figs

Assessment and reduction of proliferation risk of reactor-grade plutonium regarding construction of ‘fizzle bombs’ by terrorists

International Nuclear Information System (INIS)

Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik

2014-01-01

The approximately 23.7 wt% 240 Pu in reactor-grade plutonium denatures the 239 Pu to the extent that it cannot fuel high yield nuclear weapons. 240 Pu has a high spontaneous fission rate, which increases the spontaneous neutron flux within the fuel. When such a nuclear weapon is triggered, these neutrons cause the nuclear fission chain reaction to pre-detonate which blows the imploding fuel shell apart before the designed level of compression and reactivity could be attained, thereby greatly reducing the average energy yield of such “fizzle” bombs. Therefore reactor-grade plutonium is normally viewed as highly proliferation resistant. In this article the literature on the proliferation resistance of reactor-grade plutonium and on the mechanism and effect of fizzle bombs is reviewed in order to test this view. It is shown that even very low yield fizzle bombs, exploded in urban areas, would still cause serious blast damage as well as radioactive contamination. Combined with the high levels of induced terror, fizzle bombs might thus be attractive psychological weapons for terrorists. Therefore reactor-grade plutonium may not be sufficiently proliferation resistant against nuclear terrorism. However, denaturisation with more than 9% 238 Pu produces high levels of decay heat which will melt or explode the high explosives around uncooled implosion type weapons, rendering them useless. Unfortunately, reactor-grade Pu contains only 2.7% 238 Pu and is thus not sufficiently proliferation resistant in this respect. It is also shown that the associated neptunium poses a substantial proliferation risk. In the present study strong improvement of the proliferation resistance was demonstrated by simulation of incineration of reactor-grade plutonium in the 400 MW th Pebble Bed Modular Reactor Demonstration Power Plant. Results for modified fuel cycles, aimed at transmutating 237 Np to 238 Pu are also reported. However, these modifications increased the disloaded heavy metal

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

Study on material attractiveness aspect of spent nuclear fuel of LWR and FBR cycles based on isotopic plutonium production

International Nuclear Information System (INIS)

Permana, Sidik; Suzuki, Mitsutoshi; Saito, Masaki; Novitrian,; Waris, Abdul; Suud, Zaki

2013-01-01

Highlights: • The paper analyzes the plutonium production of recycling nuclear fuel option. • To evaluate material attractiveness based on intrinsic feature of material barrier. • Evaluation based on isotopic plutonium composition of spent fuel LWR and FBR. • Even mass number of plutonium gives a significant contribution to material barrier, in particular Pu-238 and Pu-240. • Doping MA in FBR blanket is effective to increase material barrier from weapon grade plutonium to more than MOX fuel grade. - Abstract: Recycling minor actinide (MA) as well as used uranium and plutonium can be considered to reduce nuclear waste production as well as to increase the intrinsic aspect of nuclear nonproliferation as doping material. Plutonium production as a significant aspect of recycling nuclear fuel option, gives some advantages and challenges, such as fissile material utilization of plutonium as well as production of some even mass number plutonium. The study intends to evaluate the material attractiveness based on the intrinsic feature of material barrier such as plutonium composition, decay heat and spontaneous fission neutron components from spent fuel (SF) light water reactor (LWR) and fast breeder reactor (FBR) cycles. A significant contribution has been shown by decay heat (DH) and spontaneous fission neutron (SFN) of even mass number of plutonium isotopes to the total DH and SFN of plutonium element, in particular from isotopic plutonium Pu-238 and Pu-240 contributions. Longer decay cooling time and higher burnup are effective to increase the material barrier (DH and SFN) level from reactor grade plutonium level to MOX grade plutonium level. Material barrier of plutonium element from spent fuel (SF) FBR in the core regions has similarity to the material barrier profile of SF LWR which can be categorized as MOX fuel grade plutonium. Plutonium compositions, DH and SFN components are categorized as weapon grade plutonium level for FBR blanket regions with no

Nuclear fuel: modelling the advanced plutonium assembly

International Nuclear Information System (INIS)

Kaoua, Th.; Lenain, R.

2004-01-01

The benefits of modeling in the nuclear sector are illustrated by the example of the design study for a new plutonium fuel assembly, APA, capable of ensuring maximum consumption of this fuel in pressurized-water reactors. Beyond the physical design of the assembly and its integration into the reactor, this serves for the working out of a complete materials flow and assists in modeling production from the entire inventory of nuclear power stations. (authors)

Nuclear fuel: modelling the advanced plutonium assembly

International Nuclear Information System (INIS)

N'kaoua, Th.; Lenain, R.

2002-01-01

The benefits of modeling in the nuclear sector are illustrated by the example of the design study for a new plutonium fuel assembly, APA, capable of ensuring maximum consumption of this fuel in pressurized-water reactors. Beyond the physical design of the assembly and its integration into the reactor, this serves for the working out of a complete materials flow and assists in modeling production from the entire inventory of nuclear power stations. (authors)

Medium-Power Lead-Alloy Reactors: Missions for This Reactor Technology

International Nuclear Information System (INIS)

Todreas, Neil E.; MacDonald, Philip E.; Hejzlar, Pavel; Buongiorno, Jacopo; Loewen, Eric P.

2004-01-01

A multiyear project at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology investigated the potential of medium-power lead-alloy-cooled technology to perform two missions: (1) the production of low-cost electricity and (2) the burning of actinides from light water reactor (LWR) spent fuel. The goal of achieving a high power level to enhance economic performance simultaneously with adoption of passive decay heat removal and modularity capabilities resulted in designs in the range of 600-800 MW(thermal), which we classify as a medium power level compared to the lower [∼100 MW(thermal)] and higher [2800 MW(thermal)] power ratings of other lead-alloy-cooled designs. The plant design that was developed shows promise of achieving all the Generation-IV goals for future nuclear energy systems: sustainable energy generation, low overnight capital cost, a very low likelihood and degree of core damage during any conceivable accident, and a proliferation-resistant fuel cycle. The reactor and fuel cycle designs that evolved to achieve these missions and goals resulted from study of the following key trade-offs: waste reduction versus reactor safety, waste reduction versus cost, and cost versus proliferation resistance. Secondary trade-offs that were also considered were monolithic versus modular design, active versus passive safety systems, forced versus natural circulation, alternative power conversion cycles, and lead versus lead-bismuth coolant.These studies led to a selection of a common modular design with forced convection cooling, passive decay heat removal, and a supercritical CO 2 power cycle for all our reactor concepts. However, the concepts adopt different core designs to optimize the achievement of the two missions. For the low-cost electricity production mission, a design approach based on fueling with low enriched uranium operating without costly reprocessing in a once-through cycle was pursued to achieve a

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

International Nuclear Information System (INIS)

Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

1996-09-01

The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

Energy Technology Data Exchange (ETDEWEB)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-02-26

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment

Calculating the plutonium in spent fuel elements

International Nuclear Information System (INIS)

Barnham, Keith

1992-01-01

Many members of the public are concerned about plutonium. They are worried about its environmental, health and proliferation risks. Fundamental to all such considerations are two related questions: how much plutonium do nuclear reactors produce ? and how accurately do the relevant authorities know these production figures ? These two questions have been studied with particular reference to the UK civil Magnox reactors. In 1990 these were still the only UK civil reactors whose spent fuel had been reprocessed to extract plutonium in routine production. It has not been possible to conclude that the relevant government industry and safeguard authorities are aware of how much plutonium these reactors produce and that the figures are known to the highest achievable accuracy. To understand why, this chapter will outline some of the history of the attempts to get answers to these two questions. (author)

Storage of plutonium and nuclear power plant actinide waste in the form of critical-mass-free ceramics containing neutron poisons

Energy Technology Data Exchange (ETDEWEB)

Nadykto, B.A. [RFNC-VNIIEF, Nizhni Novgorod Region (Russian Federation)

2001-07-01

The nuclear weapons production has resulted in accumulation of a large quantity of plutonium and uranium highly enriched with uranium-235 isotope (many tons). The work under ISTC Project 332B-97 treated the issues of safe plutonium storage through making critical-mass-free plutonium oxide compositions with neutron poisons. This completely excludes immediate utilization (without chemical reprocessing) of retained plutonium in nuclear devices. It is therewith possible to locate plutonium most compactly in the storage facility, which would allow reduction in required storage areas and costs. The issues of the surplus weapon-grade plutonium management and utilization have been comprehensively studied in the recent decade. The issues are treated in multiple scientific publications, conferences, and seminars. At the same time, issues of nuclear power engineering actinide waste storage are studied no less extensively. The general issues are material radioactivity and energy release and nuclear accident hazards due to critical mass generation. Plutonium accumulated in nuclear power plant spent fuel is more accessible than weapon-grade plutonium and can become of higher and higher interest with time as its activity reduces, including as material for nuclear devices. The urgency of plutonium management is presently related not only to accumulation of surplus weapon-grade plutonium, but also to the fact that it is high time to decide what has to be done regarding reactor plutonium. Presently, the possibility of actinide separation from NPP spent nuclear fuel and compact underground burial separately from other (mainly fragment) activity is being considered. Actinide and neutron poison base critical-mass-free ceramic materials (similar to plutonium ceramics) may be useful for this burial method. (author)

Storage of plutonium and nuclear power plant actinide waste in the form of critical-mass-free ceramics containing neutron poisons

International Nuclear Information System (INIS)

Nadykto, B.A.

2001-01-01

The nuclear weapons production has resulted in accumulation of a large quantity of plutonium and uranium highly enriched with uranium-235 isotope (many tons). The work under ISTC Project 332B-97 treated the issues of safe plutonium storage through making critical-mass-free plutonium oxide compositions with neutron poisons. This completely excludes immediate utilization (without chemical reprocessing) of retained plutonium in nuclear devices. It is therewith possible to locate plutonium most compactly in the storage facility, which would allow reduction in required storage areas and costs. The issues of the surplus weapon-grade plutonium management and utilization have been comprehensively studied in the recent decade. The issues are treated in multiple scientific publications, conferences, and seminars. At the same time, issues of nuclear power engineering actinide waste storage are studied no less extensively. The general issues are material radioactivity and energy release and nuclear accident hazards due to critical mass generation. Plutonium accumulated in nuclear power plant spent fuel is more accessible than weapon-grade plutonium and can become of higher and higher interest with time as its activity reduces, including as material for nuclear devices. The urgency of plutonium management is presently related not only to accumulation of surplus weapon-grade plutonium, but also to the fact that it is high time to decide what has to be done regarding reactor plutonium. Presently, the possibility of actinide separation from NPP spent nuclear fuel and compact underground burial separately from other (mainly fragment) activity is being considered. Actinide and neutron poison base critical-mass-free ceramic materials (similar to plutonium ceramics) may be useful for this burial method. (author)

World status report: plutonium

International Nuclear Information System (INIS)

Dircks, W.

1992-01-01

In a recent speech in Japan, the Deputy Director General of the International Atomic Energy Agency (IAEA) said that the economic case for reprocessing spent nuclear fuel had been severely eroded. An edited version of the speech is given. The changed prospects for nuclear energy is given as the reason why the demand for plutonium has declined sharply. The oil crisis of the 1970s reduced the demand for electric power and the economic justification for the use of recycled plutonium. The stockpile of isolated plutonium is growing rapidly giving rise to worries about its security. From this point of view, isolated plutonium is best kept in reactor fuel not separated out. In this connection the IAEA has offered to help in the storage of plutonium so that vigorous safety and security requirements are met. In Japan there is a debate about the plutonium which is dependent on the future of the fast breeder reactor programme. (UK)

Study on hydrogen production using the fast breeder reactors (FBR)

International Nuclear Information System (INIS)

Kani, Yoshio

2003-01-01

As the fast breeder reactor (FBR) can effectively convert uranium-238 difficult to carry out nuclear fission at thermal neutron reactors to nuclear fissionable plutonium-239 to use it remarkable upgrading of application on uranium can be performed, to be expected for sustainable energy source. And, by reuse minor actinides of long half-life nuclides in reprocessed high level wasted solutions for fuels of nuclear reactors, reduction of radioactive poison based on high level radioactive wastes was enabled. As high temperature of about 800 centigrade was required on conventional hydrogen production, by new hydrogen production technique even at operation temperature of sodium-cooled FBR it can be enabled. Here were described for new hydrogen production methods applicable to FBR on palladium membrane hydrogen separation method carrying out natural gas/steam modification at reaction temperature of about 500 centigrade, low temperature thermo-chemical method expectable simultaneous simplification of production process, and electrolysis method expected on power load balancing. (G.K.)

A comparison of nuclear power systems for Brazil using plutonium and binary cycles

International Nuclear Information System (INIS)

Ishiguro, Y.; Fernandes, J.E.

1985-01-01

Nuclear power systems based on plutonium cycle and binary cycle are compared taking into account natural uranium demand and reactor combination. The systems start with PWR type reactors (U5/U8) and change to systems composed exclusively of FBR type reactors or PWR-FBR symbiotic systems. Four loading modes are considered for the PWR and two for the FBR. The FBR is either a LMFBR loaded with PU/U or a LMFBR loaded the binary way. A linear and a non-linear capacity growth and two different criteria for the FBR introduction are considered. The results show that a 100 GWe permanent system can be established in 50 years in all cases, based on 300000 t of natural uranium and in case of delay in the FBR introduction and if a thermal-fast symbiotic system is chosen, a binary cycle could be more advantageous than a plutonium cycle. (F.E.) [pt

Advanced concept of reduced-moderation water reactor (RMWR) for plutonium multiple recycling

International Nuclear Information System (INIS)

Okubo, T.; Iwamura, T.; Takeda, R.; Yamamoto, K.; Okada, H.

2001-01-01

An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle. (author)

Reactivity margins in heavy water moderated production reactors

International Nuclear Information System (INIS)

Benton, F.D.

1981-11-01

The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

The 'overlooked trio' of hypothetical terrorist nuclear weapons - reactor grade plutonium, neptunium-237 and tritium

International Nuclear Information System (INIS)

Sholly, S.

2002-01-01

-related complications for ad hoc reprocessing than freshly discharged fuel. There is a need for stringent controls on these three overlooked materials due to their potential usefulness in hypothetical terrorist nuclear weapons. This need is further emphasized by the increasing attention given to reactor-grade plutonium recycle (which places tonne quantities of separated reactor-grade plutonium in commerce), the construction and operation of large fusion research and power production machines requiring and/or producing kilogram quantities of tritium, and to a variety of schemes intended for partitioning and transmutation of spent commercial reactor fuel (which must necessarily separate Np-237 if the residual half-life of non-transmuted material is to meet disposal guidelines of the order of isolation for 10,000 years or less). If very stringent controls are placed on WGPu and HEU, but not on RGPu, Np-237, and tritium, it is perhaps predictable that terrorists seeking to construct nuclear devices will follow the path of least resistance. Thus, the overlooked trio requires near-term attention of the same order as that accorded to WGPu and HEU. (author)

Weapons-grade plutonium dispositioning. Volume 1: Executive summary

International Nuclear Information System (INIS)

Parks, D.L.; Sauerbrun, T.J.

1993-06-01

The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate dispositioning options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) assisted NAS in this evaluation by investigating the technical aspects of the dispositioning options and their capability for achieving plutonium annihilation levels greater than 90%. Additionally, the INEL investigated the feasibility of using plutonium fuels (without uranium) for disposal in existing light water reactors and provided a preconceptual analysis for a reactor specifically designed for destruction of weapons-grade plutonium. This four-volume report was prepared for NAS to document the findings of these studies. Volume 2 evaluates 12 plutonium dispositioning options. Volume 3 considers a concept for a low-temperature, low-pressure, low-power-density, low-coolant-flow-rate light water reactor that quickly destroys plutonium without using uranium or thorium. This reactor concept does not produce electricity and has no other mission than the destruction of plutonium. Volume 4 addresses neutronic performance, fabrication technology, and fuel performance and compatibility issues for zirconium-plutonium oxide fuels and aluminum-plutonium metallic fuels. This volumes gives summaries of Volumes 2--4

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.

Weapons grade plutonium disposition in PWR, CANDU and FR

International Nuclear Information System (INIS)

Deplech, M.; Tommasi, J.; Zaetta, A.

2000-01-01

In the frame work of the AIDA/MOX phase I/I/ program (1994-1997) between France and Russia, the disposition of plutonium in reactors was studied. The LWR (Light Water Reactor), FR (Fast reactors), CANDU (Heavy Water Reactors), HTR (High Temperature Reactors) options for using excess dismantled weapons plutonium for peaceful commercial nuclear power generating purposes offer some advantages over the remaining options (storage). The AIDA/MOX phase 1 program covers different topics, among which are the neutronic aspects of loading reactors with weapons-grade plutonium. The conclusions are that the weapon plutonium consumption is similar in the different type of reactors. However, the use of inert matrices allows to increase the mass balance for a same denaturing level. The use of Thorium as a matrix or special isotopes to increase the proliferation resistance prove to be insufficient. (author)

Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992

Energy Technology Data Exchange (ETDEWEB)

Carlisle, R.P.; Zenzen, J.M.

1994-01-01

This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story of fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.

Assessment and reduction of proliferation risk of reactor-grade plutonium regarding construction of ‘fizzle bombs’ by terrorists

Energy Technology Data Exchange (ETDEWEB)

Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School for Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001 (Internal Post Box 360), Potchefstroom 2520 (South Africa); Mulder, Eben J. [School for Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)

2014-05-01

The approximately 23.7 wt% {sup 240}Pu in reactor-grade plutonium denatures the {sup 239}Pu to the extent that it cannot fuel high yield nuclear weapons. {sup 240}Pu has a high spontaneous fission rate, which increases the spontaneous neutron flux within the fuel. When such a nuclear weapon is triggered, these neutrons cause the nuclear fission chain reaction to pre-detonate which blows the imploding fuel shell apart before the designed level of compression and reactivity could be attained, thereby greatly reducing the average energy yield of such “fizzle” bombs. Therefore reactor-grade plutonium is normally viewed as highly proliferation resistant. In this article the literature on the proliferation resistance of reactor-grade plutonium and on the mechanism and effect of fizzle bombs is reviewed in order to test this view. It is shown that even very low yield fizzle bombs, exploded in urban areas, would still cause serious blast damage as well as radioactive contamination. Combined with the high levels of induced terror, fizzle bombs might thus be attractive psychological weapons for terrorists. Therefore reactor-grade plutonium may not be sufficiently proliferation resistant against nuclear terrorism. However, denaturisation with more than 9% {sup 238}Pu produces high levels of decay heat which will melt or explode the high explosives around uncooled implosion type weapons, rendering them useless. Unfortunately, reactor-grade Pu contains only 2.7% {sup 238}Pu and is thus not sufficiently proliferation resistant in this respect. It is also shown that the associated neptunium poses a substantial proliferation risk. In the present study strong improvement of the proliferation resistance was demonstrated by simulation of incineration of reactor-grade plutonium in the 400 MW{sub th} Pebble Bed Modular Reactor Demonstration Power Plant. Results for modified fuel cycles, aimed at transmutating {sup 237}Np to {sup 238}Pu are also reported. However, these

Study of plutonium disposition using existing GE advanced Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

1994-06-01

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

Study of plutonium disposition using existing GE advanced Boiling Water Reactors

International Nuclear Information System (INIS)

1994-01-01

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ''Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs

Plutonium and minor actinides recycle in equilibrium fuel cycles of pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Waris, A.; Sekimoto, H. [Research Lab. for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

2001-07-01

A study on plutonium and minor actinides (MA) recycle in equilibrium fuel cycles of pressurized water reactors (PWR) has been performed. The calculation results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined plutonium and MA when uranium is discharged from the reactor. However, when uranium is totally confined, the enrichment becomes extremely high. The recycle of plutonium and MA together with discharging uranium can reduce the radio-toxicity of discharged heavy metal (HM) waste to become less than that of loaded uranium. (author)

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K.G.; Tretyakov, A.A.; Sorokin, Y.P.; Bondin, V.V.; Manakova, L.F.; Jardine, L.J.

2001-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is incineration

Strategies for denaturing the weapons-grade plutonium stockpile

International Nuclear Information System (INIS)

Buckner, M.R.; Parks, P.B.

1992-10-01

In the next few years, approximately 50 metric tons of weapons-grade plutonium and 150 metric tons of highly-enriched uranium (HEU) may be removed from nuclear weapons in the US and declared excess. These materials represent a significant energy resource that could substantially contribute to our national energy requirements. HEU can be used as fuel in naval reactors, or diluted with depleted uranium for use as fuel in commercial reactors. This paper proposes to use the weapons-grade plutonium as fuel in light water reactors. The first such reactor would demonstrate the dual objectives of producing electrical power and denaturing the plutonium to prevent use in nuclear weapons

The first metallurgical tests on plutonium

International Nuclear Information System (INIS)

Grison, E.; Abramson, R.; Anselin, F.; Monti, H.

1958-01-01

Metallic plutonium was first prepared in France in January 1956, as soon as we had access to quantities of the order of several grams of plutonium, which had been extracted from the rods of the pile EL2 at Saclay. Since up to the present this reactor, of thermal power 2 000 kW, has been our only source of plutonium, we have so far only worked on experimental quantities sufficient for the basic tests but not for tests on a scale of possible applications. It is this work, carried out during this phase of preliminary research, which is described below. With the starting up of the plutonium extraction plant at Marcoule, where the reactor G1 has been operating at power for more than a year, we shall go on next to a another order of magnitude which will allow the manufacture and experimentation of prototype fuel elements. (author) [fr

Plutonium use - Present status and prospects

International Nuclear Information System (INIS)

Dievoet, J. van; Fossoul, E.; Jonckheere, E.; Bemden, E. van den

1977-01-01

The use of plutonium in thermal and fast reactors is a demonstrated, if not proven, technology. Moreover, plutonium is being produced in increasing quantities. Evaluation of this production on a world scale shows that it would be theoretically possible to construct numerous breeders and thus to make the best use of plutonium, while considerably reducing uranium consumption. This source of plutonium is nevertheless dependent on the reprocessing of irradiated fuel. Long delays in installing and adequate world reprocessing capacity are weakening the prospects for introducing breeders. Furthermore, the critical situation regarding reprocessing may delay the development of complementary reprocessing methods for fuels with a high plutonium content and high burnup. The recycling of plutonium is now a well-known technique and any objections to it hardly bear analysis. Utilization of plutonium offers an appreciable saving in terms of uranium and separative work units; and it can also be shown that immediate reprocessing of the recycling fuel is not essential for the economics of the concept. Temporary storage of recycled fuel is a particularly safe form of concentrating plutonium, namely in irradiated plutonium-bearing fuel assemblies. Finally, recycling offers such flexibility that it represents no obstacle to fuel management at power plants with light-water reactors. These strategic considerations imply that the technology of using plutonium for fabricating thermal or fast reactor fuels is both technically reliable and economically viable. The methods used in industrial facilities are fully reassuring in this respect. Although various unsolved problems exist, none seems likely to impede current developments, while the industrial experience gained has enabled the economics and reliability of the methods to be improved appreciably. Apart from the techno-economic aspects, the plutonium industry must face extremely important problems in connection with the safety of personnel

DOE plutonium disposition study: Analysis of existing ABB-CE Light Water Reactors for the disposition of weapons-grade plutonium

International Nuclear Information System (INIS)

1994-01-01

Core reactivity and basic fuel management calculations were conducted on the selected reactors (with emphasis on the System 80 units as being the most desirable choice). Methods used were identical to those reported in the Evolutionary Reactor Report. From these calculations, the basic mission capability was assessed. The selected reactors were studied for modification, such as the addition of control rod nozzles to increase rod worth, and internals and control system modifications that might also be needed. Other system modifications studied included the use of enriched boric acid as soluble poison, and examination of the fuel pool capacities. The basic geometry and mechanical characteristics, materials and fabrication techniques of the fuel assemblies for the selected existing reactors are the same as for System 80+. There will be some differences in plutonium loading, according to the ability of the reactors to load MOX fuel. These differences are not expected to affect licensability or EPA requirements. Therefore, the fuel technology and fuel qualification sections provided in the Evolutionary Reactor Report apply to the existing reactors. An additional factor, in that the existing reactor availability presupposes the use of that reactor for the irradiation of Lead Test Assemblies, is discussed. The reactor operating and facility licenses for the operating plants were reviewed. Licensing strategies for each selected reactor were identified. The spent fuel pool for the selected reactors (Palo Verde) was reviewed for capacity and upgrade requirements. Reactor waste streams were identified and assessed in comparison to uranium fuel operations. Cost assessments and schedules for converting to plutonium disposition were estimated for some of the major modification items. Economic factors (incremental costs associated with using weapons plutonium) were listed and where possible under the scope of work, estimates were made

Accident investigation board report on the May 14, 1997, chemical explosion at the Plutonium Reclamation Facility, Hanford Site,Richland, Washington - summary report

International Nuclear Information System (INIS)

Gerton, R.E.

1997-01-01

This report is a summary of the Accident Investigation Board Report on the May 14, 1997, Chemical Explosion at the Plutonium Reclamation Facility, Hanford Site, Richland, Washington (DOE/RL-97-59). The referenced report provides a greater level of detail and includes a complete discussion of the facts identified, analysis of those facts, conclusions derived from the analysis, identification of the accident's causal factors, and recommendations that should be addressed through follow-up action by the U.S. Department of Energy and its contractors. This companion document provides a concise summary of that report, with emphasis on management issues. Evaluation of emergency and occupational health response to, and radiological and chemical releases from, this accident was not within the scope of this investigation, but is the subject of a separate investigation and report (see DOE/RL-97-62)

Viability of inert matrix fuel in reducing plutonium amounts in reactors

International Nuclear Information System (INIS)

2006-08-01

Reactors worldwide have produced more than 2000 tonnes of plutonium, contained in spent fuel or as separated forms through reprocessing. Disposition of fissile materials has become a primary concern of nuclear non-proliferation efforts. There is a significant interest in IAEA Member States to develop proliferation resistant nuclear fuel cycles for incineration of plutonium such as inert matrix fuels (IMFs). The present report summarises R and D work on inert matrix fuel for plutonium and (to a lesser extent) minor actinide stock-pile reduction, and discusses the possible strategies to include inert matrix fuel approaches to the nuclear fuel cycle. The publication reviews the status of potential IMF candidates and describes several identified candidate materials for both fast and thermal reactors: MgO, ZrO2, SiC, Zr alloy, SiAl, ZrN; some of these have undergone test irradiations and post-irradiation examination. Also discussed are modelling of IMF fuel performance and safety analysis. System studies have identified strategies for both implementation of IMF fuel as homogeneous or heterogeneous phases, as assemblies or core loadings and in existing reactors in the shorter term, as well as in new reactors in the longer term

Combining a gas turbine modular helium reactor and an accelerator and for near total destruction of weapons grade plutonium

Energy Technology Data Exchange (ETDEWEB)

Baxter, A.M.; Lane, R.K.; Sherman, R. [General Atomics, San Diego, CA (United States)

1995-10-01

Fissioning surplus weapons-grade plutonium (WG-Pu) in a reactor is an effective means of rendering this stockpile non-weapons useable. In addition the enormous energy content of the plutonium is released by the fission process and can be captured to produce valuable electric power. While no fission option has been identified that can accomplish the destruction of more than about 70% of the WG-Pu without repeated reprocessing and recycling, which presents additional opportunities for diversion, the gas turbine modular helium-cooled reactor (GT-MHR), using an annular graphite core and graphite inner and outer reflectors combines the maximum plutonium destruction and highest electrical production efficiency and economics in an inherently safe system. Accelerator driven sub-critical assemblies have also been proposed for WG-Pu destruction. These systems offer almost complete WG-Pu destruction, but achieve this goal by using circulating aqueous or molten salt solutions of the fuel, with potential safety implications. By combining the GT-MHR with an accelerator-driven sub-critical MHR assembly, the best features of both systems can be merged to achieve the near total destruction of WG-Pu in an inherently safe, diversion-proof system in which the discharged fuel elements are suitable for long term high level waste storage without the need for further processing. More than 90% total plutonium destruction, and more than 99.9% Pu-239 destruction, could be achieved. The modular concept minimizes the size of each unit so that both the GT-MHR and the accelerator would be straightforward extensions of current technology.

Design characteristics of research zero power fast reactor Lasta

International Nuclear Information System (INIS)

Milosevic, M.; Stefanovic, D.; Pesic, M.; Nikolic, D.; Antic, D.; Zavaljevski, N.; Popovic, D.

1990-01-01

LASTA is a flexible zero power reactor with uranium and plutonium fuel designed for research in the neutron physics and in the fast reactor physics. Safety considerations and experimental flexibility led to the choice of a fixed vertical assembly with two safety blocks as the main safety elements, so that safety devices would be operated by gravity. The neutron and reactor physics, the control and safety philosophy adopted in our design, are described in this paper. Developed computer programs are presented. (author)

Storage for the Fast Flux Test Facility unirradiated fuel in the Plutonium Finishing Plant Complex, Hanford Site, Richland, Washington

International Nuclear Information System (INIS)

1992-01-01

This Environmental Assessment evaluates the proposed action to relocate and store unirradiated Fast Flux Test Facility fuel in the Plutonium Finishing Plant Complex on the Hanford Site, Richland, Washington. The US Department of Energy has decided to cease fuel fabrication activities in the 308 Building in the 300 Area. This decision was based on a safety concern over the ability of the fuel fabrication portion of the 308 Building to withstand a seismic event. The proposed action to relocate and store the fuel is based on the savings that could be realized by consolidating security costs associated with storage of the fuel. While the 308 Building belowgrade fuel storage areas are not at jeopardy by a seismic event, the US Department of Energy is proposing to cease storage operations along with the related fabrication operations. The US Department of Energy proposes to remove the unirradiated fuel pins and fuel assemblies from the 308 Building and store them in Room 192A, within the 234-5Z Building, a part of the Plutonium Finishing Plant Complex, located in the 200 West Area. Minor modifications to Room 192A would be required to accommodate placement of the fuel. The US Department of Energy estimates that removing all of the fuel from the 308 Building would save $6.5 million annually in security expenditures for the Fast Flux Test Facility. Environmental impacts of construction, relocation, and operation of the proposed action and alternatives were evaluated. This evaluation concluded that the proposed action would have no significant impacts on the human environment

Innovative inert matrix-thoria fuels for in-reactor plutonium disposition

International Nuclear Information System (INIS)

Vettraino, F.; Padovani, E.; Luzzi, L.; Lombardi, C.; Thoresen, H.; Oberlander, B.; Iversen, G.; Espeland, M.

1999-01-01

The present leading option for plutonium disposition, either civilian or weapons Pu, is to burn it in LWRs after having converted it to MOX fuel. However, among the possible types of fuel which can be envisaged to burn plutonium in LWRs, innovative U-free fuels such as inert matrix and thoria fuel are novel concept in view of a more effective and ultimate solution from both security and safety standpoint. Inert matrix fuel is an non-fertile oxide fuel consisting of PuO 2 , either weapon-grade or reactor-grade, diluted in inert oxides such as for ex. stabilized ZrO 2 or MgAl 2 O 4 , its primary advantage consisting in no-production of new plutonium during irradiation, because it does not contain uranium (U-free fuel) whose U-238 isotope is the departure nuclide for breeding Pu-239. Some thoria addition in the matrix (thoria-doped fuel) may be required for coping with reactivity feedback needs. The full thoria-plutonia fuel though still a U-free variant cannot be defined non-fertile any more because the U-233 generation. The advantage of such a fuel option consisting basically on a remarkable already existing technological background and a potential acceleration in getting rid of the Pu stocks. All U-free fuels are envisaged to be operated under a once-through cycle scheme being the spent fuel outlooked to be sent directly to the final disposal in deep geological formations without requiring any further reprocessing treatment, thanks to the quality-poor residual Pu and a very high chemical stability under the current fuel reprocessing techniques. Besides, inert matrix-thoria fuel technology is suitable for in-reactor MAs transmutation. An additional interest in Th containing fuel refers to applicability in ADS, the innovative accelerated driven subcritical systems, specifically aimed at plutonium, minor actnides and long lived fission products transmutation in a Th-fuel cycle scheme which enables to avoid generations of new TRUs. A first common irradiation experiment

Use of plutonium and minor actinides as fuel in high temperature pebble bed reactors for waste minimization

International Nuclear Information System (INIS)

Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenther

2009-01-01

Energy production by nuclear fission gives rise to longlived radionuclides, such as plutonium and americium. The ''PuMA'' (Plutonium and Minor Actinides Waste Management) research project within the 6th Framework Program of the European Union serves to minimize waste arisings and transmute plutonium and minor actinides from spent LWR fuel elements by means of modular high-temperature reactors (HTR). Coating the fuel, which consists of kernels approx. 250 μm in radius and surrounded by graphite as the moderator material, allows very high operating and accident temperatures and very high burnups. One point examined is whether the inherent safety characteristics known for uranium oxide also exist for (PuO 2 + MAO 2 ) fuel. On the basis of a reference reactor similar to the South African PBMR-400, various loading strategies at maximum burnup are considered with a view to the inherent safety of the HTR. (orig.)

The influence of thorium on the temperature reactivity coefficient in a 400 MWth pebble bed high temperature plutonium incinerating reactor

International Nuclear Information System (INIS)

Richards, Guy A.; Serfontein, Dawid E.

2014-01-01

This article investigates advanced fuel cycles containing thorium and reactor grade plutonium (Pu(PWR)) in a 400 MW th Pebble Bed Modular Reactor (PBMR) Demonstration Power Plant. Results presented were determined from coupled neutronics and thermo-hydraulic simulations of the VSOP 99/05 diffusion codes. In a previous study impressive burn-ups (601 MWd/kg heavy metal (HM)) and thus plutonium destruction rates (69.2 %) were obtained with pure plutonium fuel with mass loadings of 3 g Pu(PWR)/fuel sphere or less. However the safety performance was poor in that the limit on the maximum fuel temperature during equilibrium operation was exceeded and positive Uniform Temperature Reactivity Coefficients (UTCs) were obtained. In the present study fuel cycles containing mixtures of thorium and plutonium achieved negative maximum UTCs. Plutonium only fuel cycles also achieved negative maximum UTCs, provided that much higher mass loadings are used. It is proposed that the lower thermal neutron flux was responsible for this effect. The plutonium only fuel cycle with 12 g Pu(PWR)/fuel sphere also achieved the adopted safety limits for the PBMR DPP-400 in that the maximum fuel temperature and the maximum power density did not exceed 1130°C or 4.5 kW/sphere respectively. This design would thus be licensable and could potentially be economically feasible. However the burn-up was much lower at 181 MWd/kgHM and thus the plutonium destruction fraction was also much lower at 24.5%, which may be sub-optimal with respect to proliferation and waste disposal objectives and therefore further optimisation studies are proposed. (author)

Safety issues at the defense production reactors

International Nuclear Information System (INIS)

1987-01-01

The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors - the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. This report examines the safety objective established by the Department of Energy for the production reactors and the process the Department of its contractors use to implement the objective; focuses on a variety of uncertainties concerning the production reactors, particularly those related to potential vulnerabilities to severe accidents; and identifies ways in which the DOE approach to management of the safety of the production reactors can be improved

Breeding and plutonium characterization analysis on actinides closed water-cooled thorium reactor

International Nuclear Information System (INIS)

Permana, Sidik; Sekimoto, Hiroshi; Takaki, Naoyuki

2009-01-01

Higher difficulties (barrier) or more complex design of nuclear weapon, material fabrication and handling and isotopic enrichment can be achieved by a higher isotopic barrier. The isotopic material barrier includes critical mass, heat-generation rate, spontaneous neutron generation and radiation. Those isotopic barriers in case of plutonium isotope is strongly depend on the even mass number of plutonium isotope such as 238 Pu, 240 Pu and 242 Pu and for 233 U of thorium cycle depends on 232 U. In this present study, fuel sustainability as fuel breeding capability and plutonium characterization as main focus of proliferation resistance analysis have been analyzed. Minor actinide (MA) is used as doping material to be loaded into the reactors with thorium fuel. Basic design parameters are based on actinide closed-cycle reactor cooled by heavy water. The evaluation use equilibrium burnup analysis coupled with cell calculation of SRAC and nuclear data library is JENDL.32. Parametrical survey has been done to analyze the effect of MA doping rate, different moderation ratio for several equilibrium burnup cases. Plutonium characterization which based on plutonium isotope composition is strongly depending on MA doping concentration and different moderation conditions. Breeding condition can be achieved and high proliferation resistance level can be obtained by the present reactor systems. Higher isotopic plutonium composition of Pu-238 (more than 40%) can be obtained compared with other plutonium isotopes. In addition, higher moderation ratio gives the isotope composition of 238 Pu increases, however, it obtains lower composition when MA doping is increased and it slightly lower composition for higher burnup. Meanwhile, higher 240 Pu composition can be achieved by higher MA doping rate as well as for obtaining higher breeding capability. (author)

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

Energy Technology Data Exchange (ETDEWEB)

Cowell, B.S.; Fisher, S.E.

1999-02-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

International Nuclear Information System (INIS)

Cowell, B.S.; Fisher, S.E.

1999-01-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

Evaluation of utilizing spent fuel and plutonium by optimization model for nuclear fuel cycle

International Nuclear Information System (INIS)

Yoshida, Naoto; Fujii, Yasumasa; Komiyama, Ryoichi

2016-01-01

The nuclear power generation has played an important role in power generation mix as a base load power supply. On the other hand, increasing spent fuel and separated plutonium is a long-standing problem. It is expected that advanced fast reactor and high temperature gas reactor could reduce nuclear waste and effectively consume it as valuable resources. Specific scenarios about spent fuel and the gross weight of plutonium are assumed in this study, and the installable potential of fuel cycle and the most suitable reactor mix are analyzed. The model is formulated as liner programing. The model identifies the best strategy of mix of nuclear reactor types to minimize the present value of total cost in a forecast period. As a result, Fast Breeder Reactor and High Temperature Gas Reactor reduce stored spent fuel and increase the consumptions of plutonium. (author)

Plutonium multi-recycling in increased moderating ratio reactors (IMR)

International Nuclear Information System (INIS)

Barbrault, P.; Larderet, P.

1998-01-01

The large core of the future jointly defined European PWR (EPR), would be compatible with an increased Moderating Ratio (MR) enabling better plutonium burnout. The purpose of current work on the subject is to assess plutonium multi-recycling possibilities in IMR reactors. What additional operating constraints would be involved under normal and accidental conditions and are they acceptable? The conclusion is that Plutonium multi-recycling in a PWR of the type envisaged for the EPR raises no major problems under the following conditions: use of an IMR MOX core, enhancing both plutonium burnout and absorber efficiency; use of enriched boron in both the primary coolant soluble boron and the B4C boron carbide in the control rods. Deeper investigation should be performed concerning the partial or total core drain-out, in view of the high total Pu concentrations involved (13%) and the types of core considered (100% MOX). (author)

Transportation requirements for the disposition of excess weapon plutonium by burning in fission reactors

International Nuclear Information System (INIS)

Hovingh, J.; Walter, C.E.

1996-01-01

Both the US and Russia are planning to dispose of about 50 Mg of excess weapon plutonium over a 25-year period. One option is to transfer the plutonium to Advanced Light Water (power) Reactors (ALWRs) for use as fuel. Subsequent disposal would then be considered commercial spent fuel. This disposition option, like others, involves the transportation of plutonium in various material forms as it proceeds through various points in the recovery operation. This paper examines both the disposition option and the issues surrounding the transportation of 50 Mg of excess plutonium within the US under current regulatory and infrastructure constraints. Transportation issues include criticality control, shielding, and containment of the contents. Allowable limits on each of these issues are specified by the applicable (or selected) regulation. The composition and form of the radioactive materials to be transported will determine, in part, the applicable portions of the regulations as well as the packaging design. The regulations and the packaging design, along with safeguard and security issues, will determine the quantity of plutonium or fuel assemblies per package as well as the number of packages per shipment and the type of highway carrier. For the disposition of 50 Mg of weapon plutonium using ALWRs in a 25-year campaign, the annual shipment rates are determined for the various types of carriers

Civil plutonium management

International Nuclear Information System (INIS)

Sicard, B.; Zaetta, A.

2004-01-01

During 1960 and 1970 the researches on the plutonium recycling in fast neutrons reactors were stimulated by the fear of uranium reserves diminishing. At the beginning of 1980, the plutonium mono-recycling for water cooled reactors is implementing. After 1990 the public opinion concerning the radioactive wastes management and the consequences of the disarmament agreements between Russia and United States, modified the context. This paper presents the today situation and technology associated to the different options and strategical solutions of the plutonium management: the plutonium use in the world, the neutronic characteristics, the plutonium effect on the reactors characteristics, the MOX behavior in the reactors, the MOX fabrication and treatment, the possible improvements to the plutonium use, the concepts performance in a nuclear park. (A.L.B.)

Analysis of fuel cycles with natural uranium, Phase I, Economic analysis of plutonium recycling in BHWR; Analiza gorivnih ciklusa sa prirodnim uranom, II faza - Ekonomska analiza recikliranja plutonijuma u BHWR reaktorima

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Bosevski, T [Institute of Nuclear Sciences Boris Kidric, Laboratorija za fiziku i dinamiku reaktora, Vinca, Beograd (Serbia and Montenegro)

1965-11-15

The objective of this analysis was establishing a method for determination of the fuel price fraction in the total cost of nuclear power production. Special attention was devoted to recycling of plutonium in natural uranium reactors, plutonium to be used in the same reactor type. The adopted method would enable economic comparison of different types of fuel cycles for different reactors.

Economic assumptions for evaluating reactor-related options for managing plutonium

International Nuclear Information System (INIS)

Rothwell, G.

1996-01-01

This paper discusses the economic assumptions in the U.S. National Academy of Sciences' report, Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options (1995). It reviews the Net Present Value approach for discounting and comparing the costs and benefits of reactor-related options. It argues that because risks associated with the returns to plutonium management are unlikely to be constant over time, it is preferable to use a real risk-free rate to discount cash flows and explicitly describe the probability distributions for costs and benefits, allowing decision makers to determine the risk premium of each option. As a baseline for comparison, it assumes that one economic benefit of changing the current plutonium management system is a reduction in on-going Surveillance and Maintenance (S and M) costs. This reduction in the present value of S and M costs can be compared with the discounted costs of each option. These costs include direct construction costs, indirect costs, operating costs minus revenues, and decontamination and decommissioning expenses. The paper also discusses how to conduct an uncertainty analysis. It finishes by summarizing conclusions and recommendations and discusses how these recommendations might apply to the evaluation of Russian plutonium management options. (author)

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

The first metallurgical tests on plutonium; Premiers essais metallurgiques sur le plutonium

Energy Technology Data Exchange (ETDEWEB)

Grison, E; Abramson, R; Anselin, F; Monti, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

Metallic plutonium was first prepared in France in January 1956, as soon as we had access to quantities of the order of several grams of plutonium, which had been extracted from the rods of the pile EL2 at Saclay. Since up to the present this reactor, of thermal power 2 000 kW, has been our only source of plutonium, we have so far only worked on experimental quantities sufficient for the basic tests but not for tests on a scale of possible applications. It is this work, carried out during this phase of preliminary research, which is described below. With the starting up of the plutonium extraction plant at Marcoule, where the reactor G1 has been operating at power for more than a year, we shall go on next to a another order of magnitude which will allow the manufacture and experimentation of prototype fuel elements. (author) [French] La premiere elaboration de plutonium metallique en France fut faite en janvier 1956, des que nous pumes disposer de quantites de plutonium de l'ordre de quelques grammes, qui avaient ete retires des barreaux de la pile EL2 de Saclay. Ce reacteur, d'une puissance thermique de 2 000 kW, ayant ete jusqu'a present notre seule source de plutonium, nous n'avons encore travaille que sur des quantites experimentales suffisantes pour les essais de base, mais non pour des essais a l'echelle d'applications possibles. Ce sont les travaux effectues pendant cette phase de recherches preliminaires qui seront evoques ci-dessous. Avec la mise eu route de l'usine d'extraction de plutonium de Marcoule, ou le reacteur G1 fonctionne en puissance depuis plus d'un an, nous allons passer prochainement a un autre ordre de grandeur, qui nous permettra la fabrication et l'experimentation d'elements combustibles prototypes. (auteur)

Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

International Nuclear Information System (INIS)

Popa-Simil, Liviu

2008-01-01

Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages - possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools. (author)

Plutonium release from Fukushima Daiichi fosters the need for more detailed investigations

Science.gov (United States)

Schneider, Stephanie; Walther, Clemens; Bister, Stefan; Schauer, Viktoria; Christl, Marcus; Synal, Hans-Arno; Shozugawa, Katsumi; Steinhauser, Georg

2013-10-01

The contamination of Japan after the Fukushima accident has been investigated mainly for volatile fission products, but only sparsely for actinides such as plutonium. Only small releases of actinides were estimated in Fukushima. Plutonium is still omnipresent in the environment from previous atmospheric nuclear weapons tests. We investigated soil and plants sampled at different hot spots in Japan, searching for reactor-borne plutonium using its isotopic ratio 240Pu/239Pu. By using accelerator mass spectrometry, we clearly demonstrated the release of Pu from the Fukushima Daiichi power plant: While most samples contained only the radionuclide signature of fallout plutonium, there is at least one vegetation sample whose isotope ratio (0.381 +/- 0.046) evidences that the Pu originates from a nuclear reactor (239+240Pu activity concentration 0.49 Bq/kg). Plutonium content and isotope ratios differ considerably even for very close sampling locations, e.g. the soil and the plants growing on it. This strong localization indicates a particulate Pu release, which is of high radiological risk if incorporated.

DOE plutonium disposition study: Analysis of existing ABB-CE Light Water Reactors for the disposition of weapons-grade plutonium. Final report

Energy Technology Data Exchange (ETDEWEB)

1994-06-01

Core reactivity and basic fuel management calculations were conducted on the selected reactors (with emphasis on the System 80 units as being the most desirable choice). Methods used were identical to those reported in the Evolutionary Reactor Report. From these calculations, the basic mission capability was assessed. The selected reactors were studied for modification, such as the addition of control rod nozzles to increase rod worth, and internals and control system modifications that might also be needed. Other system modifications studied included the use of enriched boric acid as soluble poison, and examination of the fuel pool capacities. The basic geometry and mechanical characteristics, materials and fabrication techniques of the fuel assemblies for the selected existing reactors are the same as for System 80+. There will be some differences in plutonium loading, according to the ability of the reactors to load MOX fuel. These differences are not expected to affect licensability or EPA requirements. Therefore, the fuel technology and fuel qualification sections provided in the Evolutionary Reactor Report apply to the existing reactors. An additional factor, in that the existing reactor availability presupposes the use of that reactor for the irradiation of Lead Test Assemblies, is discussed. The reactor operating and facility licenses for the operating plants were reviewed. Licensing strategies for each selected reactor were identified. The spent fuel pool for the selected reactors (Palo Verde) was reviewed for capacity and upgrade requirements. Reactor waste streams were identified and assessed in comparison to uranium fuel operations. Cost assessments and schedules for converting to plutonium disposition were estimated for some of the major modification items. Economic factors (incremental costs associated with using weapons plutonium) were listed and where possible under the scope of work, estimates were made.

The Minatom concept of surplus weapons plutonium utilization in Russia

International Nuclear Information System (INIS)

Yegorov, N.N.; Bogdan, V.V.; Kagramanian, V.S.

1996-01-01

The fuel cycle industry in Russia has necessary basis and experience to begin solving problems of ensuring safe utilisation of weapons plutonium. Russian concept of plutonium management (both civil and military) is based on the fuel cycle closing in the nuclear power industry to increase the efficiency of the fuel use and decrease the activity of the long lived waste. Short term program of plutonium management in Russia includes safe and reliable storage of weapons and separated civil plutonium until they are used in reactors. Further studies are needed concerning optimal use of MOX fuel in fast BN reactors as well as in WWER type reactors having in mind non-proliferation aspects, nuclear radiation safety, economics and ecology

Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

CERN Document Server

International Electrotechnical Commission. Geneva

1998-01-01

This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

Production of Plutonium Metal from Aqueous Solutions

Energy Technology Data Exchange (ETDEWEB)

Orth, D.A.

2003-01-16

The primary separation of plutonium from irradiated uranium by the Purex solvent extraction process at the Savannah River Plant produces a dilute plutonium solution containing residual fission products and uranium. A cation exchange process is used for concentration and further decontamination of the plutonium, as the first step in the final preparation of metal. This paper discusses the production of plutonium metal from the aqueous solutions.

Implications of plutonium and americium recycling on MOX fuel fabrication

International Nuclear Information System (INIS)

Renard, A.; Pilate, S.; Maldague, Th.; La Fuente, A.; Evrard, G.

1995-01-01

The impact of the multiple recycling of plutonium in power reactors on the radiation dose rates is analyzed for the most critical stage in a MOX fuel fabrication plant. The limitation of the number of Pu recycling in light water reactors would rather stem from reactor core physics features. The case of recovering americium with plutonium is also considered and the necessary additions of shielding are evaluated. A comparison between the recycling of Pu in fast reactors and in light water reactors is presented. (author)

Achievement and future plan on plutonium use in Japan

International Nuclear Information System (INIS)

Uematsu, Kunihiko

1998-05-01

The plutonium recycle option has been supported from the early stage of nuclear development by Japanese Government because of the very weak structure of energy supplying system. Power Reactor and Nuclear Fuel Development Corporation (PNC) has been developing the plutonium recycling technology such as fast breeder reactor, MOX fuel fabrication and reprocessing technologies based on the decision of the government. The paper describes the obtained result of technical development and the future R and D program in the field of plutonium recycle. (J.P.N.)

Nonproliferation and safeguards aspects of the Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Hannum, W.H.; Wade, D.C.; McFarlane, H.F.; Hill, R.N.

1997-01-01

The National Academy of Sciences (NAS) has declared that the large and growing stocks of plutonium from weapons dismantlement in the U.S. and the former Soviet Union FSU are a ''clear and present danger'' to peace and security. Moreover, the opinion of some experts that plutonium of any isotopic blend is a proliferation threat has been well publicized, heightening the concern that plutonium produced in the civilian fuel cycle is itself a proliferation threat. Assuring that separated plutonium, from dismantled warheads as well as from civilian power programs, is under effective control has (again) become a high priority of U.S. diplomacy. One pole of the debate on how to manage this material is to declare it to be a waste, and to search for some way to dispose of it safely, securely, and permanently. The other pole is to view it as an energy resource and to safeguard it against diversion, putting it into active use in the civilian power program. The ultimate choice cannot be separated from the long-term strategy for use of peaceful nuclear power. Continued use of a once-through fuel cycle will lead to an ever-increasing quantity of excess plutonium-requiring safeguarding. Alternatively, recycling the world's stocks of plutonium in fast reactors, contrary to common misconception, will cap the world supply of plutonium and hold it in working inventories for generating power. Transition from the current-generation light water cooled reactors (LWRs) to a future fast-reactor-based nuclear energy supply under international safeguards would, henceforth, limit world plutonium inventories to the amount necessary and useful for power generation, with no further excess production. (author)

Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options

International Nuclear Information System (INIS)

Brownson, D.A.; Hanson, D.J.; Blackman, H.S.

1993-06-01

The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition

Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options

Energy Technology Data Exchange (ETDEWEB)

Brownson, D.A.; Hanson, D.J.; Blackman, H.S. [and others

1993-06-01

The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition.

Process control and safeguards system plutonium inventory conrol for MOX fuel facility

International Nuclear Information System (INIS)

Mishima, T.; Aoki, M.; Muto, T.; Amanuma, T.

1979-01-01

The plutonium inventory control (PINC) system is a real-time material accountability control system that is expected to be applied to a new large-scale plutonium fuel production facility for both fast breeder reactor and heavy water reactor at the Power Reactor and Nuclear Development Corporation. The PINC is basically a system for material control but is expected to develop into a whole facility control system, including criticality control, process control, quality control, facility protection, and so forth. Under PINC, every process and storage area is divided into a unit area, which is the smallest unit for both accountability and process control. Item and material weight automatically are accounted for at every unit area, and data are simultaneously treated by a computer network system. Sensors necessary for the system are being developed. 9 figures

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

2002-01-01

The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

International Nuclear Information System (INIS)

D'Hondt, P.; Gehin, J.; Na, B.C.; Sartori, E.; Wiesenack, W.

2001-01-01

One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)

Plutonium contamination in soils and sediments at Mayak PA, Russia.

Science.gov (United States)

Skipperud, Lindis; Salbu, Brit; Oughton, Deborah H; Drozcho, Eugeny; Mokrov, Yuri; Strand, Per

2005-09-01

The Mayak Production Association (Mayak PA) was established in the late 1940's to produce plutonium for the Soviet Nuclear Weapons Programme. In total, seven reactors and two reprocessing plants have been in operation. Today, the area comprises both military and civilian reactors as well as reprocessing and metallurgical plants. Authorized and accidental releases of radioactive waste have caused severe contamination to the surrounding areas. In the present study, [alpha]-spectrometry and inductively coupled plasma-mass spectrometry (ICP-MS) have been used to determine plutonium activities and isotope ratios in soil and sediment samples collected from reservoirs of the Techa River at the Mayak area and downstream Techa River. The objective of the study was to determine the total inventory of plutonium in the reservoirs and to identify the different sources contributing to the plutonium contamination. Results based on [alpha]-spectrometry and ICP-MS measurements show the presence of different sources and confirmed recent reports of civilian reprocessing at Mayak. Determination of activity levels and isotope ratios in soil and sediment samples from the Techa River support the hypothesis that most of the plutonium, like other radionuclides in the Techa River, originated from the very early waste discharges to the Techa River between 1949 and 1951. Analysis of reservoir sediment samples suggest that about 75% of the plutonium isotopes could have been released to Reservoir 10 during the early weapons production operation of the plant, and that the majority of plutonium in Reservoir 10 originates from discharges from power production or reprocessing. Enhanced 240Pu/239Pu atom ratios in river sediment upper layers (0-2 cm) between 50 and 250 km downstream from the plant indicate a contribution from other, non-fallout sources.

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

Phenomenology of the behavior of nuclear fuels containing plutonium in the cycles of water reactors. Development of a model on the equivalence of Plutonium

International Nuclear Information System (INIS)

Azzoug, D.

1990-05-01

In the scope of fuel recycling, in nuclear reactors with water cooling systems, a model concerning the plutonium equivalence and adapted to the thermal spectra is proposed. The physical phenomena involving the plutonium isotopes are studied. A method based on the sensitivity analysis allows the understanding of the plutonium isotope behavior. An equivalence model of plutonium for thermal spectre is established. The validity of the model for different cycle lengths and supports is proved [fr

Hydrogen production with fully integrated fuel cycle gas and vapour core reactors

International Nuclear Information System (INIS)

Anghaie, S.; Smith, B.

2004-01-01

-end of the gas core reactors' fuel cycle are fission fragments that are continuously separated from the fuel. As a result the G/VCR systems do not require spent fuel storage or reprocessing. Due to very low fuel inventory and continuous burning and transmutation of actinides, gas core reactors minimise the environmental impact and stewardship burden. G/VCR systems also feature outstanding proliferation resistance characteristics and minimum vulnerability to external threats. Even for comparable spectral characteristic, gas core reactors produce fissile plutonium two orders of magnitude less than light water reactors (LWRs). In addition, the continuous recycling and burning of actinides further reduces the quality of the fissile plutonium inventory. Results of the study indicate that due to high temperature operation of G/VCR systems an overall combined electricity and hydrogen production efficiencies of well over 50% are feasible. The high production efficiency is demonstrated for combined electric power generation and hydrogen production using all three schemes for wide range of output ratio. (author)

Integral measurement of fission products capture in fast breeder reactors

International Nuclear Information System (INIS)

Martin Deidier, Loick.

1979-12-01

For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr

Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 4

International Nuclear Information System (INIS)

1976-08-01

Information concerning the use of plutonium recycle in water cooled reactors is presented under the following chapter headings: probable adverse environmental effects that cannot be avoided; means for mitigating adverse environmental effects; alternative dispositions of plutonium; relationship between local short term uses of man's environment and the maintenance and enhancement of long term productivity; irreversible and irretrievable commitments of resources; and economic analysis and cost-benefit balancing

Management of Russian military plutonium

International Nuclear Information System (INIS)

Zaleski, C.P.

1996-01-01

The objective of this paper is to propose and discuss a solution which enables storing as quickly as possible all weapons-grade plutonium from Russian military program in a way which would prevent diversion. Two main conditions apply to this solution. First, it should be achieved in a manner acceptable to Russian government, notably by preserving plutonium for possible future energy production, and second, the economics of the total system should be good enough to ensure no charge or limited charge for the storage of plutonium. A proposal is made to store plutonium in a specially designed fast reactor or specially designed reactor core. This solution could be favorable in comparison to other solutions applying the above mentioned goal and conditions. Additionally the proposed solution would have the following side advantages: utilizing available personnel and installations of the Russian nuclear complex; providing possible basis for decommissioning of older and less safe Russian reactors; giving experience of construction and operation of a series of sodium-cooled fast reactors. The major problem however is the need for large capital investment with the risk of getting no adequate return on investment due to difficult political and economic situation in Russia

Management of Hanford Site non-defense production reactor spent nuclear fuel, Hanford Site, Richland, Washington

International Nuclear Information System (INIS)

1997-03-01

The US Department of Energy (DOE) needs to provide radiologically, and industrially safe and cost-effective management of the non-defense production reactor spent nuclear fuel (SNF) at the Hanford Site. The proposed action would place the Hanford Site's non-defense production reactor SNF in a radiologically- and industrially-safe, and passive storage condition pending final disposition. The proposed action would also reduce operational costs associated with storage of the non-defense production reactor SNF through consolidation of the SNF and through use of passive rather than active storage systems. Environmental, safety and health vulnerabilities associated with existing non-defense production reactor SNF storage facilities have been identified. DOE has determined that additional activities are required to consolidate non-defense production reactor SNF management activities at the Hanford Site, including cost-effective and safe interim storage, prior to final disposition, to enable deactivation of facilities where the SNF is now stored. Cost-effectiveness would be realized: through reduced operational costs associated with passive rather than active storage systems; removal of SNF from areas undergoing deactivation as part of the Hanford Site remediation effort; and eliminating the need to duplicate future transloading facilities at the 200 and 400 Areas. Radiologically- and industrially-safe storage would be enhanced through: (1) removal from aging facilities requiring substantial upgrades to continue safe storage; (2) utilization of passive rather than active storage systems for SNF; and (3) removal of SNF from some storage containers which have a limited remaining design life. No substantial increase in Hanford Site environmental impacts would be expected from the proposed action. Environmental impacts from postulated accident scenarios also were evaluated, and indicated that the risks associated with the proposed action would be small

Plutonium recovery from spent reactor fuel by uranium displacement

Science.gov (United States)

Ackerman, J.P.

1992-03-17

A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

Plutonium recovery from spent reactor fuel by uranium displacement

International Nuclear Information System (INIS)

Ackerman, J.P.

1992-01-01

A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished

How not to reduce plutonium stocks. The danger of MOX-fuelled nuclear reactors

International Nuclear Information System (INIS)

1999-01-01

Plutonium is a radioactive by-product of nuclear reactor operation and one of the most toxic substances known. The world would be a safer place if the governments of countries with stocks of it, including Britain, would adopt effective policies for reducing and managing them. Two recent authoritative reports recommend that the British government take urgent action to reduce its 'civil' plutonium stock - currently one quarter of the world's total and set to rise to about two-thirds by the year 2010. The March 1999 House of Lords report, Management of Nuclear Waste, concludes that British government policy on plutonium 'should be the maintenance of the minimum strategic stock, and the declaration of the remainder as waste'. A report from the Royal Society, Britain's main learned society, meanwhile states that: 'In addition to disposing of some of the plutonium already in the stockpile, steps should be taken to reduce the amount added to it each year, primarily by reducing the amount of reprocessing carried out'. The government's reply to the House of Lords is expected to be followed by a public consultation before changes in legislation are proposed. But, at the same time, the government is considering an application from British Nuclear Fuels Limited (BNFL), the government-owned company which separates plutonium from spent nuclear fuel rods, for a licence to operate a new plant at Sellafield in Cumbria to produce mixed-oxide (MOX) nuclear fuel from its plutonium stockpile. The nuclear industry justifies the Sellafield MOX plant as one way of reducing plutonium stocks. But critics point out that this is not a rational way to manage plutonium. This briefing aims to contribute to an informed debate during the current flurry of British government nuclear policymaking by explaining why. (author)

Systems work for Plutonium Fuel Production Facility (PFPF) near-real-time accounting

International Nuclear Information System (INIS)

Picard, R.R.; Hafer, J.F.; Pillay, K.K.S.; Takahashi, S.; Ohtani, T.; Eguchi, K.; Seya, M.

1990-01-01

A joint effort by the Los Alamos National Laboratory and the Power Reactor and Nuclear Fuel Development Corporation of Japan examines materials accounting for the Plutonium Fuel Production Facility. A unique feature of the systems work is a sophisticated data generator. This software follows individual items throughout the process, creating detailed data files for variance propagation. The data generator deals with user-specified process operations and handles related accounting problems, such as the tracking of individual measurements through numerous blending and splitting procedure, frequent decay correction (important for large inventories), scrap recovery, and automated determination of static inventory. There is no need to rely on simplified assumptions regarding process operation and material measurement. Also, the joint study applies recent theoretical work on stratified inspection of nonhomogeneous inventories and sequential analysis of MUF -- D. 4 refs

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

Science.gov (United States)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

Effect of the plutonium isotopic composition on the performance of fast reactors; Effet de la composition isotopique du plutonium sur le rendement de reacteurs a neutrons rapides; Vliyanie izotopnogo sostava plutoniya na rabotu reaktorov na bystrykh nejtronakh; Efectos de la composicion isotopica del plutonio sobre el funcionamiento de los reactores rapidos

Energy Technology Data Exchange (ETDEWEB)

Yiftah, S [Israel Atomic Energy Commission (Israel)

1962-03-15

The isotopic composition of plutonium to be used as fuel for fast reactors will depend on the source of plutonium. In principle three different sources are possible: (a) production reactors; (6) thermal power reactors (using natural uranium or enriched uranium as fuel); (c) fast reactor blankets. In general, source (a) and to some extent source (c) will provide relatively 'clean' plutonium, that is mostly Pu{sup 239}, while plutonium from source (6) will be 'dirty' plutonium, that is plutonium rich in Pu{sup 240}, Pu{sup 241}, and Pu{sup 242}. The degree of 'dirtiness' will depend on the kind of reactor, amount of burn-up and in general on the irradiation history of the fuel. The question then arises, can one use as fuel for fast reactors any kind of plutonium? To investigate the effect of different isotopic composition of the plutonium fuel, in the metallic, oxide and carbide form, on the performance of fast reactors, a limited series of spherical geometry 16-group diffusion theory calculations were performed, using the 16-group cross-section set developed recently by Yiftah, Okrent and Moldauer and taking three different kinds of plutonium, starting with pure Pu{sup 239} and increasing the amount of higher isotopes. For the systems studied-800, 1500 and 2500-l core-volumes, which are typical for large fast power reactors-the result is, when one takes into account only the thermally fissionable isotopes Pu{sup 239} arid Pu{sup 241}, that the 'dirtier' the plutonium, the smaller the critical mass and the higher the breeding ratio. For the 1500-l reactor, taken as an example, it is further found that in the metallic, oxide and carbide plutonium fuels the reactivity change upon removal of 40% of the sodium initially present in the core is made more negative (or less positive) when the plutonium is richer in higher isotopes. (author) [French] La composition isotopique du plutonium qui doit etre utilise comme combustible dans des reacteurs a neutrons rapides depend de

Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

International Nuclear Information System (INIS)

Vallet, V.

2012-01-01

Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233 U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233 U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233 U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author) [fr

Assessment studies on plutonium recycle in CANDU reactors

International Nuclear Information System (INIS)

1978-11-01

This paper describes the CANDU reactor system in detail and goes on to explore the potential for using the system with plutonium recycle fuelling to improve fuel utilisation and to meet the long-term challenge of economic supplies of nuclear fuel. The paper includes comments on costs and non-proliferation aspects. It concludes that: recycle fuelling is feasible with little modification to the reactor design and no degradation of safety, and could offer over 50% savings in uranium requirements. However, recycle fuelling costs do not appear competitive with natural uranium in the CANDU system under current economic conditions

Neutronics benchmark of a MOX assembly with near-weapons-grade plutonium

International Nuclear Information System (INIS)

Difilippo, F.C.; Fisher, S.E.

1998-01-01

One of the proposed ways to dispose of surplus weapons-grade plutonium (Pu) is to irradiate the high-fissile material in light-water reactors in order to reduce the Pu enrichment to the level of spent fuels from commercial reactors. Considerable experience has been accumulated about the behavior of mixed-oxide (MOX) uranium and plutonium fuels for plutonium recycling in commercial reactors, but the experience is related to Pu enrichments typical of spent fuels quite below the values of weapons-grade plutonium. Important decisions related to the kind of reactors to be used for the disposition of the plutonium are going to be based on calculations, so the validation of computational algorithms related to all aspects of the fuel cycle (power distributions, isotopics as function of the burnup, etc.), for weapons-grade isotopics is very important. Analysis of public domain data reveals that the cycle-2 irradiation in the Quad cities boiling-water reactor (BWR) is the most recent US destructive examination. This effort involved the irradiation of five MOX assemblies using 80 and 90% fissile plutonium. These benchmark data were gathered by General Electric under the sponsorship of the Electric Power Research Institute. It is emphasized, however, that global parameters are not the focus of this benchmark, since the five bundles containing MOX fuels did not significantly affect the overall core performance. However, since the primary objective of this work is to compare against measured post-irradiation assembly data, the term benchmark is applied here. One important reason for performing the benchmark on Quad Cities irradiation is that the fissile blends (up to 90%) are higher than reactor-grade and, quite close to, weapons-grade isotopics

Final report on production of Pu-238 in commercial power reactors: target fabrication, postirradiation examination, and plutonium and neptunium recovery

International Nuclear Information System (INIS)

Pobereskin, M.; Langendorfer, W.; Lowry, L.; Farmelo, D.; Scotti, V.; Kruger, O.

1975-01-01

Considerable interest has been generated in more extensive applications of radioisotope thermoelectric generator (RTG) systems. This raises questions concerning the availability of 238 Pu to supply an expanding demand. The development of much of this demand will depend upon a considerable reduction in cost of 238 Pu. Two neptunia--zirconia--fuel target rods, containing four sections each of different NpO 2 concentrations, were irradiated in the Connecticut Yankee Reactor for approximately one year. Following irradiation both target rods were subjected to nondestructive examination. One rod was chosen for destructive testing and analysis. Post-irradiation chemical analyses included total Pu and Np, ppM 236 Pu/ 238 Pu, and Pu isotopic abundance. The results of these analyses and of electron microprobe analysis which provided the relative Pu concentration across the pellet diameters are tabulated. It was concluded that the feasibility of all operations involved in the production of 238 Pu by irradiation of 237 NpO 2 targets in commercial nuclear power reactors was demonstrated and that the demonstration should be extended to a pilot-scale leading to installation of a full production capacity. (U.S.)

[Present conceptions of the C.E.A. concerning] the development of fast neutron reactors in France

International Nuclear Information System (INIS)

Vendryes, G.; Gaussens, J.

1964-01-01

1 - The position of fast neutron reactors in the French nuclear energy program. In developing a program based on natural uranium, France will have an important stock of plutonium rich in higher isotopes. The existence of this plutonium and of the depleted uranium arising from the same reactors, has, as a logical consequence, the use of both in fast neutron reactors. Justified by this short term interest, the achievement of fast neutron reactors does, moreover, provide for a future necessity. 2 - Description of a fast neutron central power station of 1000 MWe. We indicate the characteristics of a future fast neutron central power station, plutonium fuelled, and sodium cooled. However uncertain these characteristics may be, they constitute a necessary guide in the orientation of our work. 3 - Studies carried out up to the present time. We give an outline of those studies, often very preliminary, which have given the characteristics cited above. The principal technical areas taken up are the following: - Neutronics (critical masses, breeding ratios, enrichments, flattening of the neutron flux, coefficients of reactivity, reactivity changes as a function of irradiation). - Dynamics, control, and safety. - Technology (design of the core and vessel, of the sodium system, and of the fuel handling mechanisms). These technical studies are complemented by economic considerations. The choice of the optimum characteristics is related to the existence of power production programs, and, in these programs, to the existence of plutonium producing thermal reactors. It is shown how, in this context, the existence of plutonium should be taken into account, and, in addition which mechanisms relate the economics of this plutonium to the choice of the most important parameters of the breeder reactors. 4 - Prototype reactor. The interest in an intermediate stage consisting of a reactor of a power level of about 80 MWe is justified. Its essential characteristics are briefly presented

Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-04-30

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

Preconceptual design for separation of plutonium and gallium by ion exchange

International Nuclear Information System (INIS)

DeMuth, S.F.

1997-01-01

The disposition of plutonium from decommissioned nuclear weapons, by incorporation into commercial UO 2 -based nuclear reactor fuel, is a viable means to reduce the potential for theft of excess plutonium. This fuel, which would be a combination of plutonium oxide and uranium oxide, is referred to as a mixed oxide (MOX). Following power generation in commercial reactors with this fuel, the remaining plutonium would become mixed with highly radioactive fission products in a spent fuel assembly. The radioactivity, complex chemical composition, and large size of this spent fuel assembly, would make theft difficult with elaborate chemical processing required for plutonium recovery. In fabricating the MOX fuel, it is important to maintain current commercial fuel purity specifications. While impurities from the weapons plutonium may or may not have a detrimental affect on the fuel fabrication or fuel/cladding performance, certifying the effect as insignificant could be more costly than purification. Two primary concerns have been raised with regard to the gallium impurity: (1) gallium vaporization during fuel sintering may adversely affect the MOX fuel fabrication process, and (2) gallium vaporization during reactor operation may adversely affect the fuel cladding performance. Consequently, processes for the separation of plutonium from gallium are currently being developed and/or designed. In particular, two separation processes are being considered: (1) a developmental, potentially lower cost and lower waste, thermal vaporization process following PuO 2 powder preparation, and (2) an off-the-shelf, potentially higher cost and higher waste, aqueous-based ion exchange (IX) process. While it is planned to use the thermal vaporization process should its development prove successful, IX has been recommended as a backup process. This report presents a preconceptual design with material balances for separation of plutonium from gallium by IX

ZPR-9 airborne plutonium monitoring system

International Nuclear Information System (INIS)

Rusch, G.K.; McDowell, W.P.; Knapp, W.G.

1975-01-01

An airborne plutonium monitoring system which is installed in the ZPR-9 (Zero Power Reactor No. 9) facility at Argonne National Laboratory is described. The design and operational experience are discussed. This monitoring system utilizes particle size and density discrimination, alpha particle energy discrimination, and a background-subtraction techique operating in cascade to separate airborne-plutonium activity from other, naturally occurring, airborne activity. Relatively high sensitivity and reliability are achieved

A Mixed-Oxide Assembly Design for Rapid Disposition of Weapons Plutonium in Pressurized Water Reactors

International Nuclear Information System (INIS)

Alonso, Gustavo; Adams, Marvin L.

2002-01-01

We have created a new mixed-oxide (MOX) fuel assembly design for standard pressurized water reactors (PWRs). Design goals were to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which we call MIX-33, offers some advantages for the disposition of weapons-grade plutonium; it increases the disposition rate by 8% while increasing the worth of control material, compared to a previous Westinghouse design. The MIX-33 design is based upon two ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the addition of water holes in the assembly. The main result of this paper is that both of these ideas are effective at increasing Pu throughput and increasing the worth of control material. With this new design, according to our analyses, we can transition smoothly from a full low-enriched-uranium (LEU) core to a full MIX-33 core while meeting the operational and safety requirements of a standard PWR. Given an interruption of the MOX supply, we can transition smoothly back to full LEU while meeting safety margins and using standard LEU assemblies with uniform pinwise enrichment distribution. If the MOX supply is interrupted for only one cycle, the transition back to a full MIX-33 core is not as smooth; high peaking could cause power to be derated by a few percent for a few weeks at the beginning of one transition cycle

Considerations in the design of a high power medical isotope production reactor

International Nuclear Information System (INIS)

Ball, Russell M.; Nordyke, William H.; Brown, Roy

2002-01-01

For the low enriched aqueous homogeneous reactor to be economic in the production of medical isotopes, such as Mo-99 and Sr-89, the power level should be of the order of 100 kWth. This is double the earlier designs and this paper discusses the design changes which must be considered to meet this goal. The topics considered are: 1. Heat removal from the reactor solution; 2. Recombination of radiolytic gases; 3. Adequate radiation shielding; 4. Stability of reactor power with fluctuating reactivity; 5. Adequate cooling of the reflector; 6. Independent shutdown mechanisms; 7. Required volume of the reactor; 8. Economic implementation. (author)

Power unit with GT-MHR reactor plant for electricity production and district heating

International Nuclear Information System (INIS)

Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

2000-01-01

Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

International Nuclear Information System (INIS)

Hill, R.N.; Wade, D.C.; Palmiotti, G.

1995-01-01

As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted

Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

International Nuclear Information System (INIS)

1996-09-01

Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU's. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work

Effects of natural phenomena on the Babcock and Wilcox Co. Plutonium Fabrication Plant at the Parks Township site, Leechburg, Pennsylvania. Docket No. 70-364

International Nuclear Information System (INIS)

1979-03-01

The proposed action is to issue a renewal to the full-term Special Nuclear Material License No. SNM-414 (Docket No. 70-364) authorizing the Nuclear Material Division of the Babcock and Wilcox Company (BandW) to operate nuclear-fuel-fabrication facilities located in Leechburg, Pennsylvania. The plutonium fuel facility is presently being used to fabricate fuel for the fast test reactor under construction at the Hanford Reservation near Richland, Washington. Implicit in Sections 70.22 and 70.23 of 10CFR70 is a requirement that existing plutonium fabrication plants be examined with the objective of improving, to the extent practicable, their abilities to withstand adverse natural phenomena without loss of capability to protect the public. In accordance with these regulations, an analysis was initiated of the effects of natural phenomena on the BandW Plutonium Fabrication Plant. Following completion of the analysis, a condensation was prepared of the effects of natural phenomena on the facility

Evolutionary/advanced light water reactor data report

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-09

The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

Evolutionary/advanced light water reactor data report

International Nuclear Information System (INIS)

1996-01-01

The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (''burned'') in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ''evolutionary'' or ''advanced'' designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ''evolutionary'' LWR alternative

Study of plutonium multi-recycle in high moderation LWR cores

International Nuclear Information System (INIS)

Iwata, Yutaka; Yamamoto, Toru; Ueji, Masao; Hibi, Koki; Aoyama, Motoo; Sakurada, Koichi

2000-01-01

Nuclear Power Engineering Corporation (NUPEC) has been studying advanced cores that are dedicated to enhance the plutonium consumption per recycling for effective use of plutonium. In this study, a fissile plutonium consumption rate is adopted as an index of the effective use of plutonium, which is defined as a ratio of consumption to loading of fissile plutonium in a core. High moderation core concepts have been studied in order to increase this index based on full MOX cores in the latest designs of LWRs in Japan that are the Advanced Boiling Water Reactor (ABWR) and the Advanced Pressurized Water Reactor (APWR). As a part of this study, core performance in the case of plutonium multi-recycling has been surveyed with these higher moderation cores aiming further effective use of plutonium. The design and analyses for equilibrium cores show that nuclear and thermal hydraulics parameters satisfy design criteria, and a fissile plutonium consumption rate increases up to 20% for ABWRs and 30% for APWRs even in plutonium multi-recycling condition. It was confirmed that the high moderation cores are feasible from a viewpoint of nuclear and thermal hydraulics, safety and plutonium consumption in the condition of plutonium multi-recycling. (author)

Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11

Energy Technology Data Exchange (ETDEWEB)

ULLAH, M K

2001-02-26

The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.

The nuclear question at the start of the '80s: the breeder reactor

International Nuclear Information System (INIS)

Owen, R.; Svensson, B.

1980-01-01

The four newspaper articles and the letter cover the following matters: general introduction about breeder reactors and the situation in Swedish politics; visit to Dounreay to discuss breeder reactors (breeding, safety, plutonium production, radiation protection); PuO 2 -UO 2 mixed fuel; description of breeder reactors; efficiency in use of U-235; DFR and PFR; breeder reactors in Swedish politics (arguments for and against nuclear power in general, breeder reactors in particular); discussion of the future of nuclear power in Sweden. (U.K.)

Passive neutron survey of the 233-S Plutonium Concentration Facility

International Nuclear Information System (INIS)

1996-08-01

A passive neutron survey was performed at the 233-S Plutonium Concentration Facility (located at the Hanford Site in Richland, Washington) during late 1994 and early 1995. Four areas were surveyed: an abandoned filter box and pipe trench, column laydown trench, load-out hood, and process hood. The primary purpose of the survey was to identify locations that had plutonium to help direct decontamination and decommissioning activities. A secondary purpose of the survey was to determine the quantity of material when its presence was identified

Effects of various moderators on the critical mass of plutonium

International Nuclear Information System (INIS)

Doherty, A.L.

1986-01-01

The fissile material storage tanks in the Hanford Plutonium Critical Mass Laboratory (CML) in Richland, Washington, are presently being upgraded. During the design and planning phase of this modification, criticality analysis was necessary to compare potential moderator/absorber materials used as isolators between tanks. A parameter study was performed to assist in determining the appropriate moderator material to be used in the plutonium nitrate storage tank system in the mix room at the CML. Four moderator/absorber materials were identified as providing adequate isolation between the tanks

Qualitative chemical analysis of plutonium by Alpha spectroscopy

International Nuclear Information System (INIS)

Ramirez G, J Qumica.J.

1994-01-01

In this work the separation and purification of plutonium from irradiated uranium was done. The plutonium, produced by the irradiation of uranium in a nuclear reactor and the β decay of 239 Np, was stabilized to Pu +4 with sodium nitrite. Plutonium was separated from the fission products and uranium by ion exchange using the resin Ag 1 X 8. It was electrodeposited on stainless steel discs and the alpha radioactivity of plutonium was measured in a surface barrier detector. The results showed that plutonium was separated with a radiochemical purity higher than 99 %. (Author)

The manufacture of plutonium fuels for light water reactors

International Nuclear Information System (INIS)

Lebastard, G.

1985-01-01

This paper describes the agreement concluded between COGEMA and BELGONUCLEAIRE, reflected in the creation of the COMMOX group which has been made reponsible for promoting and marketing plutonium fuel rods for light water reactors. One then analyses the main aspects of manufacturing this type of fuel and the resources deployed. Finally one indicates the sales prospects scheduled to meet requirements (MELOX plant) [fr

Analysis of boiling water reactors capacities for the 100% MOX fuel recycling

International Nuclear Information System (INIS)

Knoche, Dietrich

1999-01-01

The electro-nuclear park exploitation leads to plutonium production. The plutonium recycling in boiling water reactors performs a use possibility. The difference between the neutronic characteristics of the uranium and the plutonium need to evaluate the substitution impact of UOX fuel by MOX fuel on the reactor operating and safety. The analysis of the main points reached to the following conclusions: the reactivity coefficients are negative, during a cooling accident the re-divergence depends on the isotopic vector of the used plutonium, the efficiency lost of control cross resulting from the plutonium utilization can be compensate by the increase of the B 4C enrichment by 10 B and the change of the steel structure by an hafnium structure, the reactivity control in evolution can be obtained by the fuel poisoning (gadolinium, erbium) and the power map control by the plutonium content monitoring. (A.L.B.)

Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

International Nuclear Information System (INIS)

PARMA JR, EDWARD J.; PICKARD, PAUL S.; SUO-ANTTILA, AHTI JORMA

2003-01-01

The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept

Method of operating water cooled reactor with blanket

International Nuclear Information System (INIS)

Suzuki, Katsuo.

1988-01-01

Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

Methodology development for plutonium categorization and enhancement of proliferation resistance by P3 mechanism

Energy Technology Data Exchange (ETDEWEB)

Saito, M.; Kimura, Y.; Sagara, H.; Han, C. Y. [Tokyo Institute of Technology, Tokyo (Japan); Koyama, S. [Japan Atomic Energy Agency, Ibaraki (Japan)

2012-03-15

'Protected Plutonium Production (P3)' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MA). For example, adding the small amount of Minor Actinides such as {sup 237}Np or {sup 241}Am with large neutron capture cross-section to the uranium fuel to enhance the production of {sup 238}Pu, which has high spontaneous fission neutron rate do deteriorate the quality of the nuclear weapon manufacture and maintenance technologically difficult, is very effective for improving the isotopic barrier for the proliferation of plutonium. To demonstrate the P3 mechanism experimentally, U samples with 2, 5 and 10% {sup 237}Np doping were irradiated in Advanced Thermal Reactor (ATR) of INL. The fuel test samples were removed from the core at 100, 200 and 300 effective full power days (EFPD), and then post irradiation examination was completed at Chemical Lab. in Idaho National Laboratory(INL). The theoretical results of P3 mechanism predict the experimental ones quite well. The evaluation function, 'Attractiveness', was introduced as the ratio of function of Rossi-alpha to the 'Technical Difficulties for Fission Explosive Device Use. 'Rossi-alpha defined as the ratio of super-criticality to prompt neutron lifetime is the meaningful feature of the explosive yield. The Technical Difficulties for Fission Explosive Device Use can be expressed by the function of specific decay heat , spontaneous fission neutron rate and radiation of plutonium metal. Original methodology to evaluate Attractiveness of Plutonium has been improved by considering the effect of the compression of Plutonium isotope and also pre-detonation probability due to spontaneous fission neutron ate, which was applied for the categorization of the plutonium from the conventional reactors and the innovative reactors based on P3 mechanism. In the present paper, the fundamentals of P3 mechanism, the experimental demonstration of P3

Methodology development for plutonium categorization and enhancement of proliferation resistance by P3 mechanism

International Nuclear Information System (INIS)

Saito, M.; Kimura, Y.; Sagara, H.; Han, C. Y.; Koyama, S.

2012-01-01

'Protected Plutonium Production (P3)' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MA). For example, adding the small amount of Minor Actinides such as 237 Np or 241 Am with large neutron capture cross-section to the uranium fuel to enhance the production of 238 Pu, which has high spontaneous fission neutron rate do deteriorate the quality of the nuclear weapon manufacture and maintenance technologically difficult, is very effective for improving the isotopic barrier for the proliferation of plutonium. To demonstrate the P3 mechanism experimentally, U samples with 2, 5 and 10% 237 Np doping were irradiated in Advanced Thermal Reactor (ATR) of INL. The fuel test samples were removed from the core at 100, 200 and 300 effective full power days (EFPD), and then post irradiation examination was completed at Chemical Lab. in Idaho National Laboratory(INL). The theoretical results of P3 mechanism predict the experimental ones quite well. The evaluation function, 'Attractiveness', was introduced as the ratio of function of Rossi-alpha to the 'Technical Difficulties for Fission Explosive Device Use. 'Rossi-alpha defined as the ratio of super-criticality to prompt neutron lifetime is the meaningful feature of the explosive yield. The Technical Difficulties for Fission Explosive Device Use can be expressed by the function of specific decay heat , spontaneous fission neutron rate and radiation of plutonium metal. Original methodology to evaluate Attractiveness of Plutonium has been improved by considering the effect of the compression of Plutonium isotope and also pre-detonation probability due to spontaneous fission neutron ate, which was applied for the categorization of the plutonium from the conventional reactors and the innovative reactors based on P3 mechanism. In the present paper, the fundamentals of P3 mechanism, the experimental demonstration of P3 mechanism in ATR of INL and the methodology

Safe disposal of surplus plutonium

Science.gov (United States)

Gong, W. L.; Naz, S.; Lutze, W.; Busch, R.; Prinja, A.; Stoll, W.

2001-06-01

About 150 tons of weapons grade and weapons usable plutonium (metal, oxide, and in residues) have been declared surplus in the USA and Russia. Both countries plan to convert the metal and oxide into mixed oxide fuel for nuclear power reactors. Russia has not yet decided what to do with the residues. The US will convert residues into a ceramic, which will then be over-poured with highly radioactive borosilicate glass. The radioactive glass is meant to provide a deterrent to recovery of plutonium, as required by a US standard. Here we show a waste form for plutonium residues, zirconia/boron carbide (ZrO 2/B 4C), with an unprecedented combination of properties: a single, radiation-resistant, and chemically durable phase contains the residues; billion-year-old natural analogs are available; and criticality safety is given under all conceivable disposal conditions. ZrO 2/B 4C can be disposed of directly, without further processing, making it attractive to all countries facing the task of plutonium disposal. The US standard for protection against recovery can be met by disposal of the waste form together with used reactor fuel.

System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

International Nuclear Information System (INIS)

Ellis, R.J.

2000-01-01

The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

Safety problems relating to plutonium recycling in light water reactors

International Nuclear Information System (INIS)

Devillers, C.; Frison, J.M.; Mercier, J.P.; Revais, J.P.

1991-01-01

This paper describes the specific nature, as regards safety, of the mixed oxide (MOX) fuel cycle, with the exception of safety problems relating to the operation of nuclear power plants. These specific characteristics are due mainly to the presence of plutonium in fresh fuel and to the higher plutonium and transuranic element content in spent fuel assemblies. The fuel cycle steps analysed here are the transport of plutonium oxide, the manufacture of MOX fuel assemblies, the transport of fresh and spent fuel assemblies and the processing of spent fuel assemblies

Safety problems relating to plutonium recycling in light water reactors

International Nuclear Information System (INIS)

Devillers, C.; Frison, J.M.; Mercier, J.P.; Revais, J.P

1991-01-01

This paper describes the specific nature, as regards safety, of the mixed oxide (MOX) fuel cycle, with the exception of safety problems relating to the operation of nuclear power plants. These specific characteristics are due mainly to the presence of plutonium in fresh fuel and to the higher plutonium and transuranic element content in spent fuel assemblies. The fuel cycle steps analysed here are the transport of plutonium oxide, the manufacture of MOX fuel assemblies, the transport of fresh and spent fuel assemblies and the processing of spent fuel assemblies. (author) [fr

Disposition of plutonium-239 via production of fission molybdenum-99

Energy Technology Data Exchange (ETDEWEB)

Mushtaq, A., E-mail: muahtaq_a1953@hotmail.co [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

2011-04-15

A heritage of physical consequences of the U.S.-Soviet arms race has accumulated, the weapons-grade plutonium (WPu), which will become excess as a result of the dismantlement of the nuclear weapons under the arms reduction agreements. Disposition of Pu has been proposed by mixing WPu with high-level radioactive waste with subsequent vitrification into large, highly radioactive glass logs or fabrication into mixed oxide fuel with subsequent irradiation in existing light water reactors. A potential option may be the production of medical isotope molybdenum-99 by using Pu-239 targets.

Review of fast reactor activities in India (1983-84)

International Nuclear Information System (INIS)

Paranjpe, S.R.

1984-01-01

The last year was very significant for the Indian Nuclear Energy Programme as the first indigeneously built heavy water moderated natural uranium reactor called Madras Atomic Power Plant Unit-I was made operational and connected to the grid. The power level has been gradually increased and the reactor has been operating at a power level of 200 MWe (temporarily limited by Plutonium build up during approach to equilibrium core loading). The 'plutonium peak' will be crossed shortly clearing the way for raising the reactor to the full power of 235 MWe gross. The second unit of MAPP, is well advanced and barring unforeseen difficulties, is expected to become operational during this financial year. This has been a big morale booster for the programme in general and the Fast Reactor Programme in particular as plutonium produced in these reactors is expected to be the inventory for Prototype Fast Breeder Reactors. It may be recalled that in the last report to this group, a reference was made to initiation of some preliminary design studies for such a reactor

In-core fuel management for the course on operational physics of power reactors

International Nuclear Information System (INIS)

Levine, S.H.

1982-01-01

The heart of a nuclear power station is the reactor core producing power from the fissioning of uranium or plutonium fuel. Expertise in many different technical fields is required to provide fuel for continuous economical operation of a nuclear power plant. In general, these various technical disciplines can be dichotomized into ''Out-of-core'' and ''In-core'' fuel management. In-core fuel management is concerned, as the name implies, with the reactor core itself. It entails calculating the core reactivity, power distribution, and isotopic inventory for the first and subsequent cores of a nuclear power plant to maintain adequate safety margins and operating lifetime for each core. In addition, the selection of reloading schemes is made to minimize energy costs

Using the Integral Fast Reactor (IFR) to dispose of excess weapons plutonium

International Nuclear Information System (INIS)

Hannum, W.H.; Wade, D.C.

1997-01-01

Plutonium is a man-made radioactive element with a long half-life. The only way to dispose of plutonium permanently is by causing it to fission. The fission process is efficient only in a fast neutron spectrum, and multiple recycle with a minimal loss is required to approach complete destruction. To be consistent with nonproliferation objectives, the process should be compatible with rigorous safeguards, and should not involve handling separated plutonium; The Integral Fast Reactor (IFR) meets all of these requirements. In addition, several near-term denaturing options are available that are fully compatible with complete destruction. When coupled with electrical generation, ample revenues would be available to cover all handling, operating and safeguards costs, with a substantial residual net return on the investment. (author)

Radiological safety aspects in the fabrication of mixed oxide fuel elements. [Derived working limits in air and water for plutonium, enriched uranium and their mixture

Energy Technology Data Exchange (ETDEWEB)

Krishnamurthi, T.N.; Janardhanan, S.; Soman, S.D. (Bhabha Atomic Research Centre, Bombay (India). Health Physics Div.)

The problems of radiological safety in the fabrication of (U, Pu)O/sub 2/ fuel assemblies for fast reactors utilising high exposure plutonium are discussed. Derived working limits for plutonium as a function of the burn-up of RAPS (Rajasthan Atomic Power Station) fuel, external gamma and neutron exposures from feed product batches, finished fuel pins and assemblies are presented. Shielding requirements for the various glove box operations are also indicated. In general, high exposure plutonium handling calls for remote fabrication and automation at various stages would play a key role in minimising exposures to personnel in a large production plant.

Experience and activities in the field of plutonium recycling in civilian nuclear power plants in the European Union

International Nuclear Information System (INIS)

Decressin, A.; Gambier, D.J.; Lehmann, J.-P.; Nietzold, D.E.

1996-01-01

The European Union industry has established a world-wide leadership position in manufacturing and exploiting plutonium bearing fuel (MOX). About 15 to 20 tons of plutonium have been manufactured in the MOX fuel fabrication plants of E.U. companies. The current capacity of about 60 tons of MOX fuel per year is being upgraded to reach 400 tons/year by the year 2000. As a result, the excess amounts of separated plutonium, presently stored in the European Union, should no longer raise but should steadily decrease to converge to zero. Studies by the European Commission have indicated that the best use at present of weapons-grade and reactor-grade plutonium is to burn it in operating and future planned nuclear reactors. Disposing of plutonium by blending it with fission products or immobilising it into synthetic matrices appears to be far from being an industrially viable option. Following this path would mean to continue storing the excess plutonium of both military and civilian origin for an unknown, but very long period of time. For these and other reasons, the European Commission is striving to foster international cooperation between the European Union companies, having a long industrial experience accumulated in the field of recycling plutonium, and, so far, the Russian Federation and the Newly Independent States. This cooperation is aiming at supporting projects that could be mutually beneficial to all parties involved. To meet this objective, several programmes have been established either bilaterally or multilaterally, in particular within the framework of the International Science and Technology Centre (I.S.T.C.) in Moscow. Some examples of such collaborations will be described. (author)

Nuclear power and the proliferation issue

International Nuclear Information System (INIS)

Marshall, W.

1978-02-01

The purpose of the lecture is to discuss nuclear proliferation, analyse which problems are real and which are a misapprehension, and to suggest a way forward which retains the benefits of nuclear power while providing a more certain protection against undesirable proliferation. After an introductory section the lecture continues under the following headings: plutonium production and accessibility; the use of plutonium; fast reactor fuel; the interim period; conclusions. (U.K.)

Cycle downstream: the plutonium question; Aval du cycle la question du plutonium

Energy Technology Data Exchange (ETDEWEB)

Zask, G [Electricite de France, EDF/DAC, 75 - Paris (France); Rome, M [Electricite de France, EDF, Service Etudes et Projets Thermiques et Nucleaires, 92 - Courbevoie (France); Delpech, M [CEA Cadarache, Dept. d' Etudes des Reacteurs/SPRC, 13 - Saint-Paul-lez-Durance (France); and others

1998-06-29

This day, organized by the SFEN, took place at Paris the 4 june 1998. Nine papers were presented. They take stock on the plutonium physics and its utilization as a nuclear fuel. This day tried to bring information to answer the following questions: do people have to keep the plutonium in the UOX fuel or in the MOX fuel in order to use it for future fast reactors? Do people have to continue obstinately the plutonium reprocessing in the MOX for the PWR type reactors? Will it be realized a underground disposal? Can it be technically developed plutonium incinerators and is it economically interesting? The plutonium physics, the experimental programs and the possible solutions are presented. (A.L.B.)

Life cycle costs for the domestic reactor-based plutonium disposition option

International Nuclear Information System (INIS)

Williams, K.A.

1999-01-01

Projected constant dollar life cycle cost (LCC) estimates are presented for the domestic reactor-based plutonium disposition program being managed by the US Department of Energy Office of Fissile Materials Disposition (DOE/MD). The scope of the LCC estimate includes: design, construction, licensing, operation, and deactivation of a mixed-oxide (MOX) fuel fabrication facility (FFF) that will be used to purify and convert weapons-derived plutonium oxides to MOX fuel pellets and fabricate MOX fuel bundles for use in commercial pressurized-water reactors (PWRs); fuel qualification activities and modification of facilities required for manufacture of lead assemblies that will be used to qualify and license this MOX fuel; and modification, licensing, and operation of commercial PWRs to allow irradiation of a partial core of MOX fuel in combination with low-enriched uranium fuel. The baseline cost elements used for this document are the same as those used for examination of the preferred sites described in the site-specific final environmental impact statement and in the DOE Record of Decision that will follow in late 1999. Cost data are separated by facilities, government accounting categories, contract phases, and expenditures anticipated by the various organizations who will participate in the program over a 20-year period. Total LCCs to DOE/MD are projected at approximately $1.4 billion for a 33-MT plutonium disposition mission

Studies on components for a molten salt reactor

International Nuclear Information System (INIS)

Nejedly, M.; Matal, O.

2003-01-01

The aim is contribute to a design of selected components of molten salt reactors with fuel in the molten fluoride salt matrix. Molten salt reactors (MSRs) permit the utilization of plutonium and minor actinides as new nuclear fuel from a traditional nuclear power station with production of electric energy. Results of preliminary feasibility studies of an intermediate heat exchanger, a small power molten salt pump and a modular conception of a steam generator for a demonstration unit of the MSR (30 MW) are summarized. (author)

Some aspects of a technology of processing weapons grade plutonium to nuclear fuel

International Nuclear Information System (INIS)

Bibilashvili, Y.; Glagovsky, E.M.; Zakharkin, B.S.; Orlov, V.K.; Reshetnikov, F.G.; Rogozkin, B.G.; Soloni-N, M.I.

2000-01-01

The concept by Russia to use fissile weapons-grade materials, which are being recovered from nuclear pits in the process of disarmament, is based on an assessment of weapons-grade plutonium as an important energy source intended for use in nuclear power plants. However, in the path of involving plutonium excessive from the purposes of national safety into industrial power engineering there are a lot of problems, from which effectiveness and terms of its disposition are being dependent upon. Those problems have political, economical, financial and environmental character. This report outlines several technology problems of processing weapons-grade metallic plutonium into MOX-fuel for reactors based on thermal and fast neutrons, in particular, the issue of conversion of the metal into dioxide from the viewpoint of fabrication of pelletized MOX-fuel. The processing of metallic weapons-grade plutonium into nuclear fuel is a rather complicated and multi-stage process, every stage of which is its own production. Some of the stages are absent in production of MOX-fuel, for instance the stage of the conversion, i.e. transferring of metallic plutonium into dioxide of the ceramic quality. At this stage of plutonium utilization some tasks must be resolved as follows: I. As a result of the conversion, a material purified from ballast and radiogenic admixtures has to be obtained. This one will be applied to fabricate pelletized MOX-fuel going from morphological, physico-mechanical and technological properties. II. It is well known that metallic gallium, which is used as an alloying addition in weapons-grade plutonium, actively reacts with multiple metals. Therefore, an important issue is to study the effect of gallium on the technology of MOX-fuel production, quality of the pellets, as well as the interaction of gallium oxide with zirconium and steel shells of fuel elements depending upon the content of gallium in the fuel. The rate of the interaction of gallium oxide

Ceramics as nuclear reactor fuels

International Nuclear Information System (INIS)

Reeve, K.D.

1975-01-01

Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

Disposing of the world's excess plutonium

International Nuclear Information System (INIS)

McCormick, J.M.; Bullen, D.B.

1998-01-01

The authors undertake three key objectives in addressing the issue of plutonium disposition at the end of the Cold War. First, the authors estimate the total global inventory of plutonium both from weapons dismantlement and civil nuclear power reactors. Second, they review past and current policy toward handling this metal by the US, Russia, and other key countries. Third, they evaluate the feasibility of several options (but especially the vitrification and mixed oxide fuel options announced by the Clinton administration) for disposing of the increasing amounts of plutonium available today. To undertake this analysis, the authors consider both the political and scientific problems confronting policymakers in dealing with this global plutonium issue. Interview data with political and technical officials in Washington and at the International Atomic Energy Agency in Vienna, Austria, and empirical inventory data on plutonium from a variety of sources form the basis of their analysis

Fuel supply demand balances for future FBR commercialization: impacts on plutonium pricing and reactor design

International Nuclear Information System (INIS)

Braun, C.; Zebroski, E.L.

1985-01-01

Plutonium supply and demand balances for fast breeder reactor (FBR) commercialization post-2000 were computed to determine: (a) the maximum supportable number of FBRs that could be installed based on plutonium availability considerations and (b) the feasibility of a reasonable FBR capacity growth case assuming slow introduction post-2010 and rapid capacity growth post-2035. The purpose of the analysis was to determine the outer limitation on the maximum future FBR introduction, or the bounds of a possible plutonium-limited introduction rate, and to estimate the reasonableness of a more limited capacity growth case

DOE Richland readiness review for PUREX

International Nuclear Information System (INIS)

Zamorski, M.J.

1984-01-01

For ten months prior to the November 1983 startup of the Plutonium and URanium EXtraction (PUREX) Plant, the Department of Energy's Richland Operations Office conducted an operational readiness review of the facility. This review was performed consistent with DOE and RL Order 5481.1 and in accordance with written plans prepared by the program and safety divisions. It involved personnel from five divisions within the office. The DOE review included two tasks: (1) overview and evaluation of the operating contractor's (Rockwell Hanford) readiness review for PUREX, and (2) independent assessment of 25 significant aspects of the startup effort. The RL reviews were coordinated by the program division and were phased in succession with the contractor's readiness review. As deficiencies or concerns were noted in the course of the review they were documented and required formal response from the contractor. Startup approval was given in three steps as the PUREX Plant began operation. A thorough review was performed and necessary documentation was prepared to support startup authorization in November 1983, before the scheduled startup date

Plutonium

International Nuclear Information System (INIS)

Anon.

1995-01-01

Plutonium, which was obtained and identified for the first time in 1941 by chemist Glenn Seaborg - through neutron irradiation of uranium 238 - is closely related to the history of nuclear energy. From the very beginning, because of the high radiotoxicity of plutonium, a tremendous amount of research work has been devoted to the study of the biological effects and the consequences on the environment. It can be said that plutonium is presently one of the elements, whose nuclear and physico-chemical characteristics are the best known. The first part of this issue is a survey of the knowledge acquired on the subject, which emphasizes the sanitary effects and transfer into the environment. Then the properties of plutonium related to energy generation are dealt with. Fissionable, like uranium 235, plutonium has proved a high-performance nuclear fuel. Originally used in breeder reactors, it is now being more and more widely recycled in light water reactors, in MOX fuel. Reprocessing, recycling and manufacturing of these new types of fuel, bound of become more and more widespread, are now part of a self-consistent series of operations, whose technical, economical, industrial and strategical aspects are reviewed. (author)

Plutonium cores of zenith

Energy Technology Data Exchange (ETDEWEB)

Barclay, F R; Cameron, I R; Drageset, A; Freemantle, R G; Wilson, D J

1965-03-15

The report describes a series of experiments carried out with plutonium fuel in the heated zero power reactor ZENITH, with the aim of testing current theoretical methods, with particular reference to excess reactivity, temperature coefficients, differential spectrum and reaction rate distributions. Two cores of widely different fissile/moderator atom ratios were loaded in order to test the theory under significantly varied spectrum conditions.

Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

International Nuclear Information System (INIS)

1980-03-01

The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans

Disposition of plutonium as non-fertile fuel for water reactors

International Nuclear Information System (INIS)

Chidester, K.; Eaton, S.L.; Ramsey, K.B.

1998-01-01

This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The original intent of this project was to investigate the possible use of a new fuel form as a means of dispositioning the declared surplus inventory of weapons-grade plutonium. The focus soon changed, however, to managing the larger and rapidly growing inventories of plutonium arising in commercial spent nuclear fuel through implementation of a new fuel form in existing nuclear reactors. LANL embarked on a parallel path effort to study fuel performance using advanced physics codes, while also demonstrating the ability to fabricate a new fuel form using standard processes in LANL's Plutonium Facility. An evolutionary fuel form was also examined which could provide enhanced performance over standard fuel forms, but which could be implemented in a much shorter time frame than a completely new fuel form. Recent efforts have focused on implementation of results into global energy models and development of follow-on funding to continue this research

Non-proliferation issues with weapons-usable plutonium

International Nuclear Information System (INIS)

Gray, L.W.

2000-01-01

In this paper author deals with the plutonium produced in power reactors and with their using. Excess plutonium, mineralized in a ceramic matrix and incised in HLW glass, is a less attractive target for terrorist groups than either aged, irradiated weapons grade MOX fuel, or aged, U oxide spent fuel. This is especially true after the Russian and United States' Pu Disposition Programs have been completed, until the material (spent MOX fuel or the immobilized form) is stored in a sealed, repository. (authors)

Plutonium Finishing Plant (PFP) Treatment and Storage Unit Waste Analysis Plan

International Nuclear Information System (INIS)

PRIGNANO, A.L.

2000-01-01

The purpose of this waste analysis plan (WAP) is to document waste analysis activities associated with the Plutonium Finishing Plant Treatment and Storage Unit (PFP Treatment and Storage Unit) to comply with Washington Administrative Code (WAC) 173-303-300(1), (2), (4)(a) and (5). The PFP Treatment and Storage Unit is an interim status container management unit for plutonium bearing mixed waste radiologically managed as transuranic (TRU) waste. TRU mixed (TRUM) waste managed at the PFP Treatment and Storage Unit is destined for the Waste Isolation Pilot Plant (WIPP) and therefore is not subject to land disposal restrictions [WAC 173-303-140 and 40 CFR 268]. The PFP Treatment and Storage Unit is located in the 200 West Area of the Hanford Facility, Richland Washington (Figure 1). Because dangerous waste does not include source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of this documentation. The information on radionuclides is provided only for general knowledge

Safety considerations in the fast reactor fuel cycle

International Nuclear Information System (INIS)

Baker, A.R.; Burton, W.R.; Taylor, H.A.

1977-01-01

The fuel cycle safety problems for fast reactors, as compared with thermal reactors, are enhanced by the higher fissile content and heat rating of the fuel. Additionally recycling leads to the build up of substantial isotopes which contribute to the alpha and neutron hazards. The plutonium arisings in a nuclear power reactor programme extending into the next century are discussed. A requirement is to be able to return the product plutonium to a reactor about 9 months after the end of irradiation and it is anticipated that progress will be made slowly towards this fuel cycle, having regard to the necessity for maintaining safe and reliable operations. Consideration of the steps in the fuel cycle has indicated that it will be best to store the irradiated fuel on the reactor sites while I131 decays and decay heat falls before transporting and a suitable transport flask is being developed. Reprocessing development work is aimed at the key area of fuel breakdown, the inter-relation of the fuel characteristics on the dissolution of the plutonium and a solvent extract cycle leading to a product suitable for a co-located fabrication plant. Because of the high activity of recycled fuel it is considered that fabrication must move to a fully remote operation as is already the case for reprocessing, and a gel precipitation process producing a vibro compacted fuel is under development for this purpose. The waste streams from the processing plants must be minimised, processed for recovery of plutonium where applicable and then conditioned so that the final products released from the processing cycle are acceptable for ultimate disposal. The safety aspects reviewed cover protection of operators, containment of radioactive materials, criticality and regulation of discharges to the environment

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Safeguards and security requirements for weapons plutonium disposition in light water reactors

International Nuclear Information System (INIS)

Thomas, L.L.; Strait, R.S.

1994-10-01

This paper explores the issues surrounding the safeguarding of the plutonium disposition process in support of the United States nuclear weapons dismantlement program. It focuses on the disposition of the plutonium by burning mixed oxide fuel in light water reactors (LWR) and addresses physical protection, material control and accountability, personnel security and international safeguards. The S and S system needs to meet the requirements of the DOE Orders, NRC Regulations and international safeguards agreements. Experience has shown that incorporating S and S measures into early facility designs and integrating them into operations provides S and S that is more effective, more economical, and less intrusive. The plutonium disposition safeguards requirements with which the US has the least experience are the implementation of international safeguards on plutonium metal; the large scale commercialization of the mixed oxide fuel fabrication; and the transportation to and loading in the LWRs of fresh mixed oxide fuel. It is in these areas where the effort needs to be concentrated if the US is to develop safeguards and security systems that are effective and efficient

THE DEACTIVATION, DECONTAMINATION AND DECOMMISSIONING OF THE PLUTONIUM FINISHING PLANT, A FORMER PLUTONIUM PROCESSING FACILITY AT DOE'S HANFORD SITE

International Nuclear Information System (INIS)

CHARBONEAU, S.L.

2006-01-01

The Plutonium Finishing Plant (PFP) was constructed as part of the Manhattan Project during World War II. The Manhattan Project was developed to usher in the use of nuclear weapons to end the war. The primary mission of the PFP was to provide plutonium used as special nuclear material (SNM) for fabrication of nuclear devices for the war effort. Subsequent to the end of World War II, the PFP's mission expanded to support the Cold War effort through plutonium production during the nuclear arms race and later the processing of fuel grade mixed plutonium-uranium oxide to support DOE's breeder reactor program. In October 1990, at the close of the production mission for PFP, a shutdown order was prepared by the Department of Energy (DOE) in Washington,; DC--and issued to the Richland DOE field office. Subsequent to the shutdown order, a team from the Defense Nuclear Facilities Safety Board (DNFSB) analyzed the hazards at PFP associated with the continued storage of certain forms of plutonium solutions and solids. The assessment identified many discrete actions that were required to stabilize the different plutonium forms into stable form and repackage the material in high integrity containers. These actions were technically complicated and completed as part of the PFP nuclear material stabilization project between 1995 and early 2005. The completion of the stabilization project was a necessary first step in deactivating PFP. During stabilization, DOE entered into negotiations with the U.S. Environmental Protection Agency (EPA) and the State of Washington and established milestones for the Deactivation and Decommissioning (DandD) of the PFP. The DOE and its contractor, Fluor Hanford (Fluor), have made great progress in deactivating, decontaminating and decommissioning the PFP at the Hanford Site as detailed in this paper. Background information covering the PFP DandD effort includes descriptions of negotiations with the State of Washington concerning consent

Some questions and answers concerning fast reactors

International Nuclear Information System (INIS)

Marshall, W.

1980-01-01

The theme of the lecture is the place of the fast reactor in an evolving nuclear programme. The whole question of plutonium is first considered, ie its method of production and the ways in which it can be used in the fast reactor fuel cycle. Whether fast reactors are necessary is then discussed. Their safety is examined with particular attention to those design features which are most criticised ie high volumetric power density of the core, and the use of liquid sodium as coolant. Attention is then paid to environmental and safeguard aspects. (U.K.)

Programme of research and development on plutonium recycling in light water reactors

International Nuclear Information System (INIS)

1978-01-01

The state of progress on 31 December 1977 of the work relating to the research and development programme on plutonium recycling in light-water nuclear power stations is presented in this second annual report. Since almost the entire programme is in the process of implementation, the report contains either the technical specifications and the objectives of recently concluded contracts or the initial results obtained. The prime objective of the programme is to facilitate the acceptance of the plutonium industry in the Community. Among the projects necessary to attain this prime objective is a forward analysis of plutonium utilization and of its impact on the environment. Various preliminary projects have been implemented in order to lessen this impact. The second objective is aimed at improving scientific and technical knowledge of the basic neutron physics of the higher isotopes of plutonium and transplutonium elements, of the behavior of the power station (static and dynamic) and of the fuel

A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

International Nuclear Information System (INIS)

Bartram, B.W.; Dougherty, D.K.

1987-01-01

This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs

The structure of atomic power industry with allowance for energy production other than electricity

International Nuclear Information System (INIS)

Aleksandrov, A.P.; Legasov, V.A.; Sidorenko, V.A.; Ponomarev-Stepnoj, N.N.; Protsenko, A.N.; Grebennik, V.N.; Glushkov, E.S.

1977-01-01

capable of providing the appropriate speed of energetics' development by plutonium production with necessary rate. Some aspects of the influence of conceivable advanced development of hydrogen energetics conception on the structure of nuclear power industry are elucidated. In terms of the obtained forecasted relations in the nuclear power industry between electrical energy and heat production, the paper analyzed the possible structural composition of nuclear power industry, the relation between fast (when using them only in the electroenergetics) and thermal reactors, and also physical technical characteristics of reactors providing their necessary relation under conditions of self-provision with nuclear fuel [ru

An Update on the Status of the Supply of Plutonium-238 for Future NASA Missions

Science.gov (United States)

Wham, R. M.

2016-12-01

For more than five decades, Radioisotope Power Systems (RPSs) have enabled space missions to operate in locations where the Sun's intensity is too weak, obscured, or otherwise inadequate for solar power or other conventional power‒generation technologies. The natural decay heat (0.57 W/g) from the radioisotope, plutonium-238 (238Pu), provides the thermal energy source used by an RPS to generate electricity for operation of instrumentation, as well as heat to keep key subsystems warm for missions such as Voyagers 1 and 2, the Cassini mission to Saturn, the New Horizons flyby of Pluto, and the Mars Curiosity rover which were sponsored by the National Aeronautics and Space Administration (NASA). Plutonium-238 is produced by irradiation of neptunium-237 in a nuclear reactor a relatively high neutron flux. The United States has not produced new quantities of 238Pu since the early 1990s. RPS‒powered missions have continued since then using existing 238Pu inventory managed by the U.S. Department of Energy (DOE), including material purchased from Russia. A new domestic supply is needed to ensure the continued availability of RPSs for future NASA missions. NASA and DOE are currently executing a project to reestablish a 238Pu supply capability using its existing facilities and reactors, which are much smaller than the large-scale production reactors and processing canyon equipment used previously. The project is led by the Oak Ridge National Laboratory (ORNL). Target rods, containing NpO2, will be fabricated at ORNL and irradiated in the ORNL High Flux Isotope Reactor and the Advanced Test Reactor at Idaho National Laboratory. Irradiated targets will be processed in chemical separations at the ORNL Radiochemical Engineering Center to recover the plutonium product and unconverted neptunium for recycle. The 238PuO2 product will be shipped to Los Alamos National Laboratory for fabrication of heat source pellets. Key activities, such as transport of the neptunium to ORNL

The plutonium challenge for the future

International Nuclear Information System (INIS)

Gray, L.W.

2000-01-01

In this paper author deal with the weapons-usable plutonium and with the possibilities of their managing. Russia has not disclosed the amount of plutonium produced, but various estimates indicate that the production was about 130 tonnes. Production has been curtailed in Russia; three dual-purpose reactors still produce weapons-grade plutonium - two at Tomsk-7 (renamed Seversk) and one at Krasnoyarsk-26 (renamed Zheleznogorsk Mining and Chemical Combine). In a 1994 United States-Russian agreement that has yet to enter into force, Russia agreed to close the remaining operating reactors by the year 2000. Treaties between the United States and Russia have already cut the number of nuclear warheads from more than 10,000 to about 6,000 under START 1, which has been ratified, and to about 3,500 under START 2, which still awaits approval. If Russia and the United States conclude START 3, that number could drop to between 2,000 and 2,500. On September 2, 1998, the Presidents of the United States and Russia signed the 'Joint statement of principles for Management and Disposition of Plutonium, Designated as No Longer Required for Defense Purposes.' In this joint statement the Presidents affirm the intention of each country to remove by stages approximately 50 metric tons of plutonium and to convert the nuclear weapons programs, and to convert this material so that it can never be used in nuclear weapons. These 100 tonne of plutonium must be managed in proper way such that it becomes neither a proliferation for an environmental risk. The United States has proposed that it manage it's 50 tonnes by a dual approach-once through MOX burning of a portion of the plutonium and immobilization in a ceramic matrix followed by en- casement in high level waste glass. Russia has proposed that it manage its full 50 tonnes by burning in a reactor. The MOX program in the United States would bum the cleaner plutonium metal and residues. Weapons components would be converted to plutonium oxide

Gamma ray NDA assay system for total plutonium and isotopics in plutonium product solutions

International Nuclear Information System (INIS)

Cowder, L.R.; Hsue, S.T.; Johnson, S.S.; Parker, J.L.; Russo, P.A.; Sprinkle, J.K.; Asakura, Y.; Fukuda, T.; Kondo, I.

1979-01-01

A LASL-designed gamma-ray NDA instrument for assay of total plutonium and isotopics of product solutions at Tokai-Mura is currently installed and operating. The instrument is, optimally, a densitometer that uses radioisotopic sources for total plutonium measurements at the K absorption edge. The measured transmissions of additional gamma-ray lines from the same radioisotopic sources are used to correct for self-attenuation of passive gamma rays from plutonium. The corrected passive data give the plutonium isotopic content of freshly separated to moderately aged solutions. This off-line instrument is fully automated under computer control, with the exception of sample positioning, and operates routinely in a mode designed for measurement control. A one-half percent precision in total plutonium concentration is achieved with a 15-minute measurement

Plutonium Plant, Trombay

International Nuclear Information System (INIS)

Yadav, J.S.; Agarwal, K.

2017-01-01

The journey of Indian nuclear fuel reprocessing started with the commissioning of Plutonium Plant (PP) at Trombay on 22"n"d January, 1965 with an aim to reprocess the spent fuel from research reactor CIRUS. The basic process chosen for the plant was Plutonium Uranium Reduction EXtraction (PUREX) process. In seventies, the plant was subjected to major design modifications and replacement of hardware, which later met the additional demand from research reactor DHRUVA. The augmented plutonium plant has been operating since 1983. Experience gained from this plant was very much helpful to design future reprocessing plant in the country

Plutonium and U-233 mines

International Nuclear Information System (INIS)

Milgram, M.S.

1983-08-01

A comparison is made among second generation reactor systems fuelled primarily with fissile plutonium and/or U-233 in uranium or thorium. This material is obtained from irradiated fuel from first generation CANDU reactors fuelled by natural or enriched uranium and thorium. Except for plutonium-thorium reactors, second generation reactors demand similar amounts of reprocessing throughput, but the most efficient plutonium burning systems require a large prior allocation of uranium. Second generation reactors fuelled by U-233 make more efficient use of resources and lead to more flexible fuelling strategies, but require development of first generation once-through thorium cycles and early demonstration of the commercial viability of thorium fuel reprocessing. No early implementation of reprocessing technology is required for these cycles

Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 3

International Nuclear Information System (INIS)

1976-08-01

An assessment is presented of the health, safety and environmental effects of the entire light water reactor fuel cycle, considering the comparative effects of three major alternatives: no recycle, recycle of uranium only, and recycle of both uranium and plutonium. The assessment covers the period from 1975 through the year 2000 and includes the cumulative effects for the entire period as well as projections for specific years. Topics discussed include: the light water reactor with plutonium recycle; mixed oxide fuel fabrication; reprocessing plant operations; supporting uranium fuel cycle; transportation of radioactive materials; radioactive waste management; storage of plutonium; radiological health assessment; extended spent fuel storage; and blending of plutonium and uranium at reprocessing plants

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

International Nuclear Information System (INIS)

Friess, Friederike Renate

2017-01-01

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

Energy Technology Data Exchange (ETDEWEB)

Friess, Friederike Renate

2017-07-12

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Ground testing of an SP-100 prototypic reactor

International Nuclear Information System (INIS)

Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

1988-01-01

SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

Fast flux fluid fuel reactor: A concept for the next generation of nuclear power production

International Nuclear Information System (INIS)

Palmiotti, G.; Feldman, E.E.

1999-01-01

Nuclear energy has not become the preferred method of electrical energy production largely because of economic, safety, and proliferation concerns and challenges posed by nuclear waste disposal. Economies is the most important factor. To reduce the capital costs, the authors propose a compact configuration with a very high power density and correspondingly reduced reactor component sizes. Enhanced efficiency made possible by higher operating temperatures will also improve the economics of the design, and design simplicity will keep capital, operational, and maintenance costs down. The most direct solution to the nuclear waste problem is to eliminate waste production or, at least, minimize its amount and long-term radiotoxicity. This can be achieved by very high burnups, ideally 100%, and by the eventual transmutation of the long-lived fission products in situ. Very high burnups also improve the economics by optimal exploitation of the fuel. Safety concerns can be addressed by an inherently safe reactor design. Because of the intrinsic nature of nuclear materials, there probably is no definitive answer to proliferation concerns for systems that generate neutrons; however, it is important to minimize proliferation risks. The thorium cycle is a promising option because (a) plutonium is produced only in very small quantities, (b) the presence of 232 U makes handling the fuel very difficult and therefore proliferation resistant, and (c) 233 U is a fissile isotope that is less suitable than 239 Pu for making weapons and can be diluted with other uranium isotopes. An additional benefit of the thorium cycle is that it increases nuclear fuel resources by one order of magnitude. A fast flux fluid fuel reactor is a concept that can satisfy all the foregoing requirements. The fluid fuel systems have a very simple structure. Because integrity of the fuel is not an issue, these systems can operate at very high temperatures, can have high power densities, and can achieve very

The Optimum Plutonium Inert Matrix Fuel Form for Reactor-Based Plutonium Disposition

International Nuclear Information System (INIS)

Tulenko, J.S.; Wang, J.; Acosta, C.

2004-01-01

The University of Florida has underway an ongoing research program to validate the economic, operational and performance benefits of developing an inert matrix fuel (IMF) for the disposition of the U.S. weapons plutonium (Pu) and for the recycle of reprocessed Pu. The current fuel form of choice for Pu disposition for the Department of Energy is as a mixed oxide (MOX) (PuO2/UO2). We will show analyses that demonstrate that a Silicon Carbide (SiC) IMF offers improved performance capabilities as a fuel form for Pu recycle and disposition. The reason that UF is reviewing various materials to serve as an inert matrix fuel is that an IMF fuel form can offer greatly reduced Pu and transuranic isotope (TRU) production and also improved thermal performance characteristics. Our studies showed that the Pu content is reduced by an order of magnitude while centerline fuel temperatures are reduced approximately 380 degrees centigrade compared to MOX. These reduced temperatures result in reduced stored heat and thermal stresses in the pellet. The reduced stored heat reduces the consequences of the loss of coolant accident, while the reduced temperatures and thermal stresses yield greatly improved fuel performance. Silicon Carbide is not new to the nuclear industry, being a basic fuel material in gas cooled reactors

The use of the average plutonium-content for criticality evaluation of boiling water reactor mixed oxide-fuel transport and storage packages

International Nuclear Information System (INIS)

Mattera, C.

2003-01-01

Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and (Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by an homogenous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, Cogema Logistics (formerly Transnucleaire) has studied a new calculation method based on the use of the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in the paper. (author)

Plutonium production in a remote cell

International Nuclear Information System (INIS)

Maddux, E.P.; Purcell, J.A.

1987-01-01

Production of 239 Pu metal has traditionally been carried out in glovebox enclosures for protection of operating personnel and the environment. With the advent of more stringent federal regulations for protecting personnel, the environment, and security of special nuclear materials, it is proposed that plutonium be processed in a totally remote environment using automated work stations and robotic materials transfer and equipment maintenance. At the Savannah River Plant, the existing plutonium production facility is a batch-type glovebox operation built in the 1950s. A viable alternative to restoration of the existing glovebox operation is to provide a facility that uses remote cell operation

CANDU fuel - fifteen years of power reactor experience

International Nuclear Information System (INIS)

Fanjoy, G.R.; Bain, A.S.

1977-01-01

CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

Cycle downstream: the plutonium question

International Nuclear Information System (INIS)

Zask, G.; Rome, M.; Delpech, M.

1998-01-01

This day, organized by the SFEN, took place at Paris the 4 june 1998. Nine papers were presented. They take stock on the plutonium physics and its utilization as a nuclear fuel. This day tried to bring information to answer the following questions: do people have to keep the plutonium in the UOX fuel or in the MOX fuel in order to use it for future fast reactors? Do people have to continue obstinately the plutonium reprocessing in the MOX for the PWR type reactors? Will it be realized a underground disposal? Can it be technically developed plutonium incinerators and is it economically interesting? The plutonium physics, the experimental programs and the possible solutions are presented. (A.L.B.)

PRODUCTION OF PLUTONIUM METAL

Science.gov (United States)

Lyon, W.L.; Moore, R.H.

1961-01-17

A process is given for producing plutonium metal by the reduction of plutonium chloride, dissolved in alkali metal chloride plus or minus aluminum chloride, with magnesium or a magnesium-aluminum alloy at between 700 and 800 deg C and separating the plutonium or plutonium-aluminum alloy formed from the salt.

Impact of ADTT concepts on the management of global plutonium inventories

International Nuclear Information System (INIS)

Davidson, J.W.; Krakowski, R.A.; Arthur, E.D.

1996-01-01

The impact of a number of current and future nuclear systems on global plutonium inventories is assessed under realistic forecasts of nuclear power growth. Advanced systems, such as those employing Accelerator Driven Transmutation Technologies (ADTT) and liquid metal reactors, show significant promise for meeting future plutonium management needs. These analyses also indicate requirements for a higher level of detail in the nuclear fuel cycle model and for development of a metric to more quantitatively assess the proliferation risk of plutonium arising from the civilian fuel cycle

Plutonium disposition study phase 1b final report

International Nuclear Information System (INIS)

1993-01-01

This report provides the results of the Westinghouse activities performed as part of the Plutonium Disposition Study Phase 1b. These activities, which took place from May 16, 1993 to September 15, 1993, build upon the work completed in Phase 1a, which concluded on May 15, 1993. In Phase 1a, three Plutonium Disposal Reactor (PDR) options were developed for the disposal of excess weapons grade plutonium from returned and dismantled nuclear weapons. This report documents the results of several tasks that were performed to further knowledge in specific areas leading up to Phase 2 of the PDR Study. The Westinghouse activities for Phase 1b are summarized as follows: (1) resolved technical issues concerning reactor physics including equilibrium cycle calculations, use of gadolinium, moderator temperature coefficient, and others as documented in Section 2.0; (2) analyzed large Westinghouse commercial plants for plutonium disposal; (3) reactor safety issues including the steam line break were resolved, and are included in Section 2.0; (4) several tasks related to the PDR Fuel Cycle were examined; (5) cost and deployment options were examined to determine optimal configuration for both plutonium disposal and tritium production; (6) response to questions from DOE and National Academy of Scientists (NAS) reviewers concerning the PDR Phase 1a report are included in Appendix A

The fast reactor

International Nuclear Information System (INIS)

1980-02-01

The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

Reactor-grade plutonium inventory taking in the RT-1 storage production association Mayak site

International Nuclear Information System (INIS)

Zatorskij, Yu.M.

1999-01-01

General characterization of plutonium physical inventory taking (PIT) system at the RT-1 plant is given in the paper. The proposed PIT procedure is based on measuring indirect parameters of inventory plutonium, the so called attributes of items - containers with plutonium. The method how to assess quality of item attribute measurements is proposed. If any defect is detected, the container is sent back to the place where it been filed and packed in order to perform direct measurements of plutonium mass. Taking into account the above-mentioned techniques and the availability of several access control means it is proposed the period between two PIT procedures be equal to 12 months [ru

Uranium plutonium oxide fuels

International Nuclear Information System (INIS)

Cox, C.M.; Leggett, R.D.; Weber, E.T.

1981-01-01

Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

2010 South Carolina DNR Lidar: Richland County

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Provide high density LiDAR elevation data map of Richland County, SC. Provide Bare Earth DEM (vegetation removal) of Richland County, SC.

Plutonium fuel cycles in the spectral shift controlled reactor

International Nuclear Information System (INIS)

Sider, F.M.; Matzie, R.A.

1980-01-01

The spectral shift controlled reactor (SSCR) controls excess core reactivity during an operating cycle through the use of variable heavy water concentrations in the moderator. With heavy water in the coolant, the neutron spectrum is shifted to higher energy levels, thus increasing fertile conversion. In addition, since heavy water obviates the need for soluble boron, neutron losses to control poison are eliminated. As a result, better resource utilization is obtained in the SSCR employing plutonium fuel cycles compared to similarly fueled pressurized water reactors (PWRs). The SSCR, however, is not competitive with the PWR due to higher capital costs, operation and maintenance costs, and the heavy water costs, which outweigh the fuel cycle cost savings. The SSCR may become an attractive alternative to the PWR if uranium prices increase substantially

Disposition of plutonium from dismantled warheads: Belgonucleaire's proposal

International Nuclear Information System (INIS)

Haas, D.; Vanderborck, Y.; Vandenberg, C.; Vliet, J. van

1996-01-01

Set up in 1957 by Union Miniere, Belgonucleaire (BN) has been working since its origin in the field of plutonium and has accumulated extensive experience in the design and fabrication of mixed-oxide (MOX) fuel, as well as in the fuel management and licensing of MOX cores for fast breeder reactors (FBRs), pressurized water reactors (PWRs), and boiling water reactors (BWRs). Although BN's MOX plant first went into operation in 1973 (it was used for FBRs and light water reactor (LWR) demonstration fuel fabrication), industrial production started in 1986. The MOX fuel produced (280 t heavy metal (HM) until the end of 1995) has been loaded in PWRs and BWRs in four countries in Europe: France, Germany, Switzerland, and Belgium. They propose the development of MOX plants as the means for disposal of plutonium from warhead disassembly

Smart unattended systems for plutonium safeguards

International Nuclear Information System (INIS)

Menlove, H.O.; Abhold, M.; Eccleston, G.; Puckett, J.M.

1996-01-01

Large automated facilities for fabricating plutonium fuel present both difficulties and challenges for improved accounting of nuclear materials. The traditional methods of sample measurements, requiring the transfer of the sample from the production line to the assay measurement station, are not possible in automated facilities. The robotics used for automation require special containers for nuclear material that cannot be easily removed from the production line. Safety and radiation protection considerations also require that the assay instrumentation be installed in the fuel production lines because, in general, personnel cannot be in the fuel-handling area with nuclear material during operations. Such operational constraints are common in many of the modern facilities that have been designed for fabricating and processing plutonium fuel. A bilateral safeguards agreement between the US Department of Energy (DOE) and Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan was signed to develop and implement nondestructive assay (NDA) systems to provide continuous safeguards measurements for material accountancy in the robot-automated Plutonium Fuel Fabrication Facility (PFFF). The PFFF assay systems were required to operate in unattended mode with a size and fuel mass capability to match the robotics fuel manipulators. Unattended assay systems reduce the requirement for inspector's oversight of measurement operations, reduce the inspector's workload, and improve inspection efficiencies. In addition, unattended measurements become essential when facility constraints limit the access of inspectors to the operations area during material processing. Authentication techniques were incorporated into the NDA systems so that data obtained from unattended assays could be used by independent inspectors such as the IAEA

Irradiation Experiments on Plutonium Fuels for Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Frost, B. R.T.; Wait, E. [Atomic Energy Research Establishment Harwell, Berks. (United Kingdom)

1967-09-15

An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O{sub 2} -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650 Degree-Sign C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO{sub 2} and(U,Pu)O{sub 2} particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to {approx} 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent

Strategies for the plutonium utilization

International Nuclear Information System (INIS)

Zouain, D.M.; Lima, J.O.V.; Sakamoto, L.H.

1981-11-01

A review of the activities involving plutonium (its recycle, utilization and technological status and perspectives) is done. These informations are useful for an economic viability study for the plutonium utilization in thermal reactors (recycling) and in fast breeders reactor (FBR), trying to collect the major number of informations about these subjects. The initial phase describes the present status and projections of plutonium accumulation and requirements. Then, the technological process are described and some strategies are analyzed. (E.G.) [pt

TerraPower, Bill Gates' reactor

International Nuclear Information System (INIS)

Guidez, J.

2016-01-01

TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

Plutonium and surrogate fission products in a composite ceramic waste form

International Nuclear Information System (INIS)

Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

1999-01-01

Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known

Improvement of reference nuclear data for commercial power reactor analysis and design. Final report

International Nuclear Information System (INIS)

Levitt, L.; Magurno, B.; Rose, P.

1977-10-01

The National Neutron Cross Section Center (NNCSC) has participated in the Electric Power Research Institute (EPRI) program of improving the nuclear data base to be used as input for the analysis of power reactors. The NNCSC contribution consisted of organizing and participating in specialists' meetings on the Thermal Neutron Cross Sections for the Actinides and Fission Products, participating in a meeting on the recommended value for the 252 Cf anti nu, and keeping abreast of new measurements on the low energy resonance parameters of 238 U. In addition, complete evaluations for seven isotopes of gadolinium have been produced for the next version of the data library. Definitive benchmark calculations using the Monte Carlo code RECAP have been carried out for two homogeneous uranium nitrate critical spheres (ORNL 1 and 2), two water moderated uranium metal lattices (TRX-1 and 2), and three plutonium nitrate spheres (PNL 2, 3, and 5). Results for the uranium spheres gave good agreement with experiments and with previous calculations. The calculation of the TRX lattice parameters, however, casts doubt on the validity of analytic leakage corrections used previously. Eigenvalues of the plutonium nitrate spheres are found to increase with increasing ratio of total 240 Pu/ 239 Pu captures

Conceptual study of axial offset fluctuations upon stepwise power changes in a thorium–plutonium core to improve load-following conditions

International Nuclear Information System (INIS)

Lau, Cheuk Wah; Dykin, Victor; Nylén, Henrik; Björk, Klara Insulander; Sandberg, Urban

2014-01-01

Highlights: • Thorium–plutonium mixed oxide to improve nuclear reactors load-following capability. • SIMULATE-3 was the main calculation tool. • The Ringhals-3 PWR unit in Sweden was used as a reference. • Lower xenon poisoning and shorter reactor dead time. - Abstract: The increased share of renewable energy, such as wind and solar power, will increase the demand for load-following power sources, and nuclear reactors could be one option. However, during rapid load-following events, traditional UOX cores could be restricted by the volatile oscillation of the power distribution. Therefore, a conceptual study on stability properties of Th-MOX PWR concerning axial offset power excursion during load-following events are investigated and discussed. The study is performed in SIMULATE-3 for a realistic PWR core (Ringhals-3) at the end of cycle, where the largest amplitude of the axial offset oscillations is expected. It is shown that the Th-MOX core possesses much better stability characteristics and shorter reactor dead time compared with a traditional UOX core, and the main reasons are the lower sensitivity to perturbations in the neutron spectrum, lower xenon poisoning and lower thermal neutron flux

Design-Only Conceptual Design Report: Plutonium Immobilization Plant

International Nuclear Information System (INIS)

DiSabatino, A.; Loftus, D.

1999-01-01

This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

Some investigations into the behaviour of plutonium in the marine environment

International Nuclear Information System (INIS)

Hetherington, J.A.; Jefferies, D.F.; Lovett, M.B.

1975-01-01

Plutonium and other transuranic elements are produced during the operation of magnox reactors, although they do not usually present any problems of waste management until the fuel reaches the reprocessing stage. At this stage one of the chief aims is to separate and recover the plutonium from the depleted uranium and the fission products, the incentive being the value of plutonium itself as a nuclear fuel particularly for the fast breeder reactors. For this reason the amounts of plutonium appearing in the low activity waste streams for discharge to the environment are usually small, and their radiological impact insignificant compared with that of the more abundant fission products. However, with the prospect of use of large quantities of plutonium in the fast reactors, considerable interest surrounds all aspects of the properties of plutonium, and in an attempt to give a fuller understanding of the behaviour of plutonium in the marine environment use has been made of the small amounts which have been discharged in recent years to the north-east Irish Sea from the fuel reprocessing plant at Windscale in Cumbria. Details of the amounts of 239 Pu and 240 Pu discharged to sea are given and the resultant measured concentrations of the isotopes in the water throughout the Irish Sea are shown. The distribution of plutonium in sea water is shown to follow the same basic pattern as the fission products, details of which have been reported previously. By comparing the concentration gradients with distance from the point of discharge for plutonium and caesium, a value for the rate of loss of plutonium from the water to the other important compartment, namely the sediment, has been deduced. Measurements have been made of 239 Pu and 240 Pu in seabed and estuarine sediments and by combining these with the water data values of the concentration factor for the nuclide in sediment have been calculated. The results of a number of core samples have been used to calculate an

Experimental study of the higher isotopes of plutonium in the coupled fast-thermal assembly Ermine

International Nuclear Information System (INIS)

Bioux, Philippe.

1973-01-01

The most important results so far of the plutonium isotopes programme carried out in Ermine are reviewed. The results were obtained on four networks: one with uranium fuel and three with plutonium. The choice of these four networks was essentially based on the energy range of their nuclear reactions, wide enough to include the large reactor energy range, and yet realistic, with a neutron energy distribution in the 4 networks which is never very far from that encountered in a power reactor. The information relative to the high isotopes of plutonium which were obtained in these networks mostly came from: the fission indices measured in the 4 networks, the study of the balance in the two media at k(infinite)=1, reactivity measurements on cell replacements and on various plutonium samples, carried out on the three plutonium fuel networks [fr

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

Removal of the Plutonium Recycle Test Reactor - 13031

International Nuclear Information System (INIS)

Herzog, C. Brad; Guercia, Rudolph; LaCome, Matt

2013-01-01

The 309 Facility housed the Plutonium Recycle Test Reactor (PRTR), an operating test reactor in the 300 Area at Hanford, Washington. The reactor first went critical in 1960 and was originally used for experiments under the Hanford Site Plutonium Fuels Utilization Program. The facility was decontaminated and decommissioned in 1988-1989, and the facility was deactivated in 1994. The 309 facility was added to Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) response actions as established in an Interim Record of Decision (IROD) and Action Memorandum (AM). The IROD directs a remedial action for the 309 facility, associated waste sites, associated underground piping and contaminated soils resulting from past unplanned releases. The AM directs a removal action through physical demolition of the facility, including removal of the reactor. Both CERCLA actions are implemented in accordance with U.S. EPA approved Remedial Action Work Plan, and the Remedial Design Report / Remedial Action Report associated with the Hanford 300-FF-2 Operable Unit. The selected method for remedy was to conventionally demolish above grade structures including the easily distinguished containment vessel dome, remove the PRTR and a minimum of 300 mm (12 in) of shielding as a single 560 Ton unit, and conventionally demolish the below grade structure. Initial sample core drilling in the Bio-Shield for radiological surveys showed evidence that the Bio-Shield was of sound structure. Core drills for the separation process of the PRTR from the 309 structure began at the deck level and revealed substantial thermal degradation of at least the top 1.2 m (4LF) of Bio-Shield structure. The degraded structure combined with the original materials used in the Bio-Shield would not allow for a stable structure to be extracted. The water used in the core drilling process proved to erode the sand mixture of the Bio-Shield leaving the steel aggregate to act as ball bearings against the

Plutonium and minor actinide transmutation by long irradiation in LWR

International Nuclear Information System (INIS)

Facchini, A.; Sanjust, V.

1993-01-01

An investigation was made on the conceptual possibility of burning in a thermal reactor MOX fuel together with special pins containing plutonium, minor actinides and long lived fission products, recovered from the reprocessing of previously irradiated MOX fuel and mixed with an inter matrix. Preliminary calculations showed that the long term radiotoxicity of the above special pins is reduced to reasonable levels when they are irradiated up to 20 divided-by 30 years, and cooled for some centuries. In particular, during the whole life such a reactor should be able to burn a considerable fraction of plutonium, minor actinides and long lived fission products recovered from the MOX fuel irradiated along the same period of time

PLUTONIUM-ZIRCONIUM ALLOYS

Science.gov (United States)

Schonfeld, F.W.; Waber, J.T.

1960-08-30

A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

International Nuclear Information System (INIS)

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%

Radioactive waste management and plutonium recovery within the context of the development of nuclear energy in Russia

Energy Technology Data Exchange (ETDEWEB)

Kushnikov, V. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation)

1996-05-01

The Russian strategy for radioactive waste and plutonium management is based on the concept of the closed fuel cycle that has been adopted in Russia, and, to a great degree, falls under the jurisdiction of the existing Russian nuclear energy structures. From its very beginning, Russian atomic energy policy was based on finding the most effective method of developing the new fuel direction with the maximum possible utilization of the energy potential from the fission of heavy atoms and the achievement of fuel self-sufficiency through the recycling of secondary fuel. Although there can be no doubt about the importance of economic considerations (for the future), concerns for the safety of the environment are currently of the utmost importance. In this context, spent NPP fuel can be viewed as a waste to be buried only if there is persuasive evidence that such an approach is both economically and environmentally sound. The production of I GW of energy per year is accompanied by the accumulation of up to 800-1000 kg of highly radioactive fission products and approximately 250 kg of plutonium. Currently, spent fuel from the VVER 100 and the RBNK reactors contains approximately 25 tons of plutonium. There is an additional 30 tons of fuel-grade plutonium in the form of purified oxide, separated from spent fuels used in VVER440 reactors and other power production facilities, as well as approximately 100 tons of weapons-grade plutonium from dismantled warheads. The spent fuel accumulates significant amounts of small actinoids - neptunium americium, and curium. Science and technology have not yet found technical solutions for safe and secure burial of non-reprocessed spent fuel with such a broad range of products, which are typically highly radioactive and will continue to pose a threat for hundreds of thousands of years.

Determination of origin and intended use of plutonium metal using nuclear forensic techniques.

Science.gov (United States)

Rim, Jung H; Kuhn, Kevin J; Tandon, Lav; Xu, Ning; Porterfield, Donivan R; Worley, Christopher G; Thomas, Mariam R; Spencer, Khalil J; Stanley, Floyd E; Lujan, Elmer J; Garduno, Katherine; Trellue, Holly R

2017-04-01

Nuclear forensics techniques, including micro-XRF, gamma spectrometry, trace elemental analysis and isotopic/chronometric characterization were used to interrogate two, potentially related plutonium metal foils. These samples were submitted for analysis with only limited production information, and a comprehensive suite of forensic analyses were performed. Resulting analytical data was paired with available reactor model and historical information to provide insight into the materials' properties, origins, and likely intended uses. Both were super-grade plutonium, containing less than 3% 240 Pu, and age-dating suggested that most recent chemical purification occurred in 1948 and 1955 for the respective metals. Additional consideration of reactor modeling feedback and trace elemental observables indicate plausible U.S. reactor origin associated with the Hanford site production efforts. Based on this investigation, the most likely intended use for these plutonium foils was 239 Pu fission foil targets for physics experiments, such as cross-section measurements, etc. Copyright © 2017 Elsevier B.V. All rights reserved.

Recycling : The advanced fuel cycle for existing reactors

International Nuclear Information System (INIS)

Lamorlette, Guy

1994-01-01

In 1993, the Installed capacity of the world's 427 nuclear power plants was over 335 GWe. Additional plants representing 67 GWe were under construction or on order. Taking construction schedules into consideration, their start-up will stretch out over a period of ten years. Nuclear power will therefore increase by 20% at best in ten years, transiting into a relatively modest 2% average annual growth rate. Of these units, about 80% are light water reactors, whether PWR, BWR, or WER. All of these reactors utilize enriched uranium oxide fuel clad with zirconium alloy. From a fuel perspective, these reactors form a pretty homogeneous group. During reactor residence, energy is supplied by fission of three-fourths of the Initial uranium 235, but also by plutonium fission, which is formed in the fuel as soon as it is Irradiated. The plutonium supplies 40% of the generated power. When the fuel is unloaded, it consists of four elements : fission products and structural materials, such as cladding and end-fittings, which are the reel waste, and residual plutonium and uranium, which are energy materials that can be recycled in accordance with French legislation applicable to both non-nuclear and nuclear industries : 'the purpose of this law is to... make use of waste by reusing, recycling or otherwise obtaining reusable material or energy from.'. The nuclear power industry has entered a phase in which most of its capital-intensive projects are behind it. Now, It must depose Itself to ensuring the competitiveness of nuclear energy compared to other sources of power generation, while protecting the environment and respecting safety regulations. Significant gains have been achieved by improving fuel performance : optimization of fuel design, utilization of less neutron-absorbent materials, and increases in fuel burn-up have made it possible to increase the amount of energy derived from one kilogram of natural uranium by more than 50%. Recycling of the fuel in light water reactor

Concerning temporary method for return transport of plutonium

International Nuclear Information System (INIS)

1990-01-01

One of the urgent matters under deliberation in the Working Group for Nuclear Fuel Cycle, Atomic Energy Comission of Japan, is to develop a method for smooth and safe transport of plutonium to be returned from Britain and France, where spent fuel sent from Japan is reprocessed by contract. Return transport of plutonium from these nations will start in 1992. Preparatory work has been conducted in Japan for air transport of plutonium, which is to be used by Power Reactor and Nuclear Fuel Development Corporation. For this, efforts have been made since 1984 to develop containers for air transport of plutonium. Though the corporation will continue research, it will take considerable time until practical containers are developed, so the material have to be transported by sea for the time being. The corporation should play the leading role in conducting the return transport operation in cooperation with other organizations concerned including power companies. To achieve this, the corporation should be active in making preparations including the development of transport plans. (N.K.)

Power Reactor Information System (PRIS)

International Nuclear Information System (INIS)

Spiegelberg, R.

1992-01-01

The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

Plutonium Finishing Plant safety evaluation report

International Nuclear Information System (INIS)

1995-01-01

The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE's independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91

An inherently safe power reactor module

International Nuclear Information System (INIS)

Salerno, L.N.

1985-01-01

General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

Safety and Performance Achievement of Indian Nuclear Power Plant

International Nuclear Information System (INIS)

Kumar, Randhir

2011-01-01

The Nuclear Power Programme in India is based on three stage. The first stage is based on setting up of Pressurized Heavy Water Reactors (PHWRs) using indigenously available natural uranium producing electricity and plutonium. This will be followed by the second stage by plutonium fuelled Fast Breeder Reactors (FBRs) producing electricity and additional quantity of plutonium and also uranium 233 from thorium. The third stage of reactors will be based on thorium uranium 233 cycle.

The development of fast reactors - Effects on the Swedish system of management of spent fuel; Utveckling av snabba reaktorer - Paaverkan paa det svenska systemet foer hantering av anvaent braensle

Energy Technology Data Exchange (ETDEWEB)

Hans Forsstroem, Hans [SKB International AB, Stockholm (Sweden)

2013-09-15

Since the start of the nuclear power era studies have been performed of how to utilise the uranium energy resource in the most effective way. Only about one percent of the energy potential of uranium is utilised in the light water reactors of today. To improve the utilization other types of reactors are needed. With fast reactors theoretically 50-100 times more energy can be extracted from the uranium. This will require reprocessing of the uranium and multiple recycling of the plutonium. Plutonium and uranium can also be recycled in light water reactors, but this will only improve the uranium utilisation by about 20 %. Recycling of plutonium on a routine basis is presently only done in France. The development of fast reactors has been going on since the end of the 1940ies. During the 1970ies the planning was that a large number of fast reactors and their associated fuel cycle facilities would be in operation by the year 2000. The development has, however, for different reasons been much slower than planned. The general assessment today is that fast reactor, if they will be realised, will hardly give an important contribution to energy production until after 2050. Nuclear power production has instead been dominated by light water reactors similar to the ones in use in Sweden. Light water reactors are believed to continue to dominate during the next decades. To start a fast reactor system plutonium (or highly enriched uranium) will be needed. Such plutonium is contained in spent nuclear fuel from light water reactors. This raises the question: Should the spent nuclear fuel be stored so that the potential energy resource in the fuel can be used in the future instead of disposing of it as a waste? The answer to this question will depend on when the material will be useful, i.e. when fast reactors have been introduced on a large scale. It will also depend on the demand for plutonium at this time, i.e. will plutonium be a scarce redundant resource at this point of time

EDF research scenarios for closing the Plutonium cycle

International Nuclear Information System (INIS)

Le Mer, Joël; Garzenne, Claude; Lemasson, David

2013-01-01

Conclusion: → There are various solutions to plutonium fuel closure; → Natural uranium consumption is reduced: • Full generation IV fleet is obviously the most efficient; • Symbiotic fleet makes a better use of its advanced reactors. → Plutonium inventory reaches an equilibrium between 700 tons and 1150 tons. • The multi-recycling of spent MOX fuel must be a long term solution in order to reduce significantly the plutonium inventory. → Spent fuel storage is reduced when MOX spent fuel are reprocessed but sodium pools are challenging. → Fast reactors are not the only solution to use MOX spent fuel: • HCPWR is a roundabout solution: – the reduction of natural uranium is limited; – the high level waste production is high. – The reprocessing plant capacity must be increased during deployment phase → R&D must be continued to improve HCPWR design

Preparation of a primary target for the production of fission products in a nuclear reactor

International Nuclear Information System (INIS)

Arino, H.; Cosolito, F.J.; George, K.D.; Thornton, A.K.

1976-01-01

A primary target for the production of fission products in a nuclear reactor, such as uranium or plutonium fission products, is comprised of an enclosed, cylindrical vessel, preferably comprised of stainless steel, having a thin, continuous, uniform layer of fissionable material, integrally bonded to its inner walls and a port permitting access to the interior of the vessel. A process is also provided for depositing uranium material on to the inner walls of the vessel. Upon irradiation of the target with neutrons from a nuclear reactor, radioactive fission products, such as molybdenum-99, are formed, and thereafter separated from the target by the introduction of an acidic solution through the port to dissolve the irradiated inner layer. The irradiation and dissolution are thus effected in the same vessel without the necessity of transferring the fissionable material and fission products to a separate chemical reactor. Subsequently, the desired isotopes are extracted and purified. Molybdenum-99 decays to technetium-99m which is a valuable medical diagnostic radioisotope. 3 claims, 1 drawing figure

Potentialities of the molten salt reactor concept for a sustainable nuclear power production based on thorium cycle in epithermal spectrum

International Nuclear Information System (INIS)

Nuttin, Alexis

2002-01-01

In the case of a significant nuclear contribution to world energy needs, the problem of present nuclear waste management pose the sustainability of the PWR fuel cycle back into question. Studies on storage and incineration of these wastes should therefore go hand in hand with studies on innovative systems dedicated to a durable nuclear energy production, as reliable, clean and safe as possible. We are here interested in the concept of molten salt reactor, whose fuel is liquid. This particularity allows an online pyrochemical reprocessing which gives the possibility to overcome some neutronic limits. In the late sixties, the MSBR (Molten Salt Breeder Reactor) project of a graphite-moderated fluoride molten salt reactor proved thus that breeding is attainable with thorium in a thermal spectrum, provided that the online reprocessing is appropriate. By means of simulation tools developed around the Monte Carlo code MCNP, we first re-evaluate the performance of a reference system, which is inspired by the MSBR project. The complete study of the pre-equilibrium transient of this 2,500 MWth reactor, started with 232 Th/ 233 U fuel, allows us to validate our reference choices. The obtained equilibrium shows an important reduction of inventories and induced radio-toxicities in comparison with the other possible fuel cycles. The online reprocessing is efficient enough to make the system breed, with a doubling time of about thirty years at equilibrium. From the reference system, we then test different options in terms of neutron economy, transmutation and control of reactivity. We find that the online reprocessing brings most of its flexibility to this system, which is particularly well adapted to power generation with thorium. The study of transition scenarios to this fuel cycle quantifies the limits of a possible deployment from the present French power stock, and finally shows that a rational management of the available plutonium would be necessary in any case. (author)

Activation product transport in fusion reactors

International Nuclear Information System (INIS)

Klein, A.C.; Vogelsang, W.F.

1984-01-01

Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. A computer code, RAPTOR, has been developed to determine the transport of these products in fusion reactor coolant/tritium breeding materials. Without special treatment, it is likely that fusion reactor power plant operators could experience dose rates as high as 8 rem per hour around a number of plant components after only a few years of operation. (orig.)

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

Laboratory directed research and development on disposal of plutonium recovered from weapons. FY1994 final report

International Nuclear Information System (INIS)

Pitts, J.H.; Choi, J.S.

1994-01-01

This research project was conceived as a multi-year plan to study the use of mixed plutonium oxide-uranium oxide (MOX) fuel in existing nuclear reactors. Four areas of investigation were originally proposed: (1) study reactor physics including evaluation of control rod worth and power distribution during normal operation and transients; (2) evaluate accidents focusing upon the reduced control rod worth and reduced physical properties of PuO 2 ; (3) assess the safeguards required during fabrication and use of plutonium bearing fuel assemblies; and (4) study public acceptance issues associated with using material recovered from weapons to fuel a nuclear reactor. First year accomplishments are described. Appendices contain 2 reports entitled: development and validation of advanced computational capability for MOX fueled ALWR assembly designs; and long-term criticality safety concerns associated with weapons plutonium disposition

309 plutonium recycle test reactor ion exchanger vault deactivitation report

International Nuclear Information System (INIS)

Griffin, P.W.

1996-03-01

This report documents the deactivation of the ion exchanger vault at the 309 Plutonium Recycle Test Reactor (PRTR) Facility in the 300 Area. The vault deactivation began in May 1995 and was completed in June 1995. The final site restoration and shipment of the low-level waste for disposal was finished in September 1995. The ion exchanger vault deactivation project involved the removal and disposal of twelve ion exchangers and decontaminating and fixing of residual smearable contamination on the ion exchanger vault concrete surfaces

Fuel Cycle Impacts of Uranium-Plutonium Co-extraction

International Nuclear Information System (INIS)

Taiwo, Temitope; Szakaly, Frank; Kim, Taek-Kyum; Hill, Robert

2008-01-01

A systematic investigation of the impacts of uranium and plutonium co-extraction during fuel separations on reactor performance and fuel cycle has been performed. Proliferation indicators, critical mass and radiation source levels of the separation products or fabricated fuel, were also evaluated. Using LWR-spent-uranium-based MOX fuel instead of natural-uranium-based fuel in a PWR MOX core requires a higher initial plutonium content (∼1%), and results in higher Np-237 content (factor of 5) in the spent fuel, and less consumption of Pu-238 (20%) and Am-241 (14%), indicating a reduction in the effective repository space utilization. Additionally, minor actinides continue to accumulate in the fuel cycle, and thus a separate solution is required for them. Differences were found to be quite smaller (∼0.4% in initial transuranics) between the equilibrium cycles of advanced fast reactor cores using spent and depleted uranium for make-up, in additional to transuranics. The critical masses of the co-extraction products were found to be higher than for weapons-grade plutonium (WG-Pu) and the decay heat and radiation sources of the materials (products) were also found to be generally higher than for WG-Pu in the transuranics content range of 10% to 100% in the heavy-metal. (authors)

Theoretical methods for determination of core parameters in uranium-plutonium lattices

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J.; Bosevski, T.; Matausek, M.; Stefanovic, D.; Strugar, P. [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

1972-07-01

The prediction of plutonium production in power reactors depends essentially on how the change of neutron energy spectra in a reactor cell during burn-up is determined. In the epithermal region, where the build-up of plutonium occurs, the slowing down effects are particularly important, whereas, on the other hand, the thermal neutron spectrum is strongly influenced by the low-lying plutonium resonances. For accurate analysis, multi-group numerical methods are required, which, applied to burn-up prediction, are extremely laborious and time consuming even for large computers. This paper contains a comprehensive review of the methods of core parameter determination in the uranium-plutonium lattices developed in Yugoslavia during the last few years. Faced with the problem of using small computers, the authors had to find new approaches combining physical evidence and mathematical elegance. The main feature of these approaches is the tendency to proceed with analytical treatment as far as possible and then to include suitable numerical improvements. With this philosophy, which is generally overlooked when using large computers, fast and reasonably accurate methods were developed. The methods include original means for adequate treatment of neutron spectra and cell geometry effects,especially suitable for U-Pu systems. In particular, procedures based on the energy dependent boundary conditions, the discrete energy representation, the improved collision probabilities and the Green function slowing down solutions were developed and applied. Results obtained with these methods are presented and compared with those of the experiments and those obtained with other methods. (author)

Theoretical methods for determination of core parameters in uranium-plutonium lattices

International Nuclear Information System (INIS)

Pop-Jordanov, J.; Bosevski, T.; Matausek, M.; Stefanovic, D.; Strugar, P.

1972-01-01

The prediction of plutonium production in power reactors depends essentially on how the change of neutron energy spectra in a reactor cell during burn-up is determined. In the epithermal region, where the build-up of plutonium occurs, the slowing down effects are particularly important, whereas, on the other hand, the thermal neutron spectrum is strongly influenced by the low-lying plutonium resonances. For accurate analysis, multi-group numerical methods are required, which, applied to burn-up prediction, are extremely laborious and time consuming even for large computers. This paper contains a comprehensive review of the methods of core parameter determination in the uranium-plutonium lattices developed in Yugoslavia during the last few years. Faced with the problem of using small computers, the authors had to find new approaches combining physical evidence and mathematical elegance. The main feature of these approaches is the tendency to proceed with analytical treatment as far as possible and then to include suitable numerical improvements. With this philosophy, which is generally overlooked when using large computers, fast and reasonably accurate methods were developed. The methods include original means for adequate treatment of neutron spectra and cell geometry effects,especially suitable for U-Pu systems. In particular, procedures based on the energy dependent boundary conditions, the discrete energy representation, the improved collision probabilities and the Green function slowing down solutions were developed and applied. Results obtained with these methods are presented and compared with those of the experiments and those obtained with other methods. (author)

Plutonium production story at the Hanford site: processes and facilities history

Energy Technology Data Exchange (ETDEWEB)

Gerber, M.S., Westinghouse Hanford

1996-06-20

This document tells the history of the actual plutonium production process at the Hanford Site. It contains five major sections: Fuel Fabrication Processes, Irradiation of Nuclear Fuel, Spent Fuel Handling, Radiochemical Reprocessing of Irradiated Fuel, and Plutonium Finishing Operations. Within each section the story of the earliest operations is told, along with changes over time until the end of operations. Chemical and physical processes are described, along with the facilities where these processes were carried out. This document is a processes and facilities history. It does not deal with the waste products of plutonium production.

Multi-recycling of plutonium and incineration of americium, curium, and technetium in PWRs

International Nuclear Information System (INIS)

Golfier, H.; Bergeron, J.; Puill, A.; Rohart, M.

2000-01-01

The future of nuclear power requires a clear strategy for radwaste and Plutonium management. Pressurized water reactors (PWR) and the associated fuel cycle installations represent the largest part of the French power plants (and are partly paid off). The reactors in service produce an annual 10 tons of Pu, 1.4 tons of minor actinides (MA), and 3.8 tons of long-lived fission products (LLFP). The spent fuel is reprocessed in La Hague plant to recover the energetic elements U and Pu. The latter was initially dedicated to power Fast Breeder Reactors that converted the depleted and reprocessed, thus ensuring a significant part of the French national energy resources. The shut-down of Super-Phenix, the postponement of building of Fast Breeder Reactors (FBR) and the relaxed need for stretching natural U resources raise the issue of Pu management. In fact, the Pu mono-recycling practiced in France since 1987 (St Laurent B1) only slows down the Pu accumulation in spent nuclear fuel, yet it is unable to stabilize the Pu inventory. Beyond the cooperation with its industrial partners, CEA investigates solutions for short and medium term Pu management thus contributing to research required for keeping nuclear power as an energy option. The range of these investigations shall cover both adaptations for light water reactors to facilitate Pu recycling and more innovative solutions concerning reactors, fuel and fuel cycle. The aim of using Pu more efficiently in PWR has led, not only for economic and non-proliferation reasons, but also for considerations related to the optimization of Pu and MA management. The mastery of Pu inventory is a requirement for all long-lived radwaste management methods. In this context, the potential of innovative PWRs has been investigated to control the Pu fluxes and to make them a milestone on the way to clean nuclear power. This paper presents the most recent results related to Pu utilization and MA and LLFP incineration like (Am+Cm) and Tc. To

History of fast reactor development in U.S.A.-I

International Nuclear Information System (INIS)

Ninokata, Hisashi; Sasao, Nobuyki

2007-01-01

History and present state of fast reactor was reviewed in series. As a history of fast reactor development in U.S.A. - I, this third lecture presented the dawn of the fast reactor development in the USA. The first fast reactor was the Clementine reactor with plutonium fuels and mercury coolant. The LAMPRE-1 reactor was the first sodium cooled and molten plutonium reactor. Experimental breeder reactor (EBR-1) was the first reactor to produce electricity and four kinds of fuels were loaded. Zero-power reactors were constructed to conduct reactor physics experiments on fast reactors. Today there are renewed interests in fast reactors due to their ability to fission actinides and reduce radioactive wastes. (T. Tanaka)

An optimization method for parameters in reactor nuclear physics

International Nuclear Information System (INIS)

Jachic, J.

1982-01-01

An optimization method for two basic problems of Reactor Physics was developed. The first is the optimization of a plutonium critical mass and the bruding ratio for fast reactors in function of the radial enrichment distribution of the fuel used as control parameter. The second is the maximization of the generation and the plutonium burnup by an optimization of power temporal distribution. (E.G.) [pt

Inherent protection of plutonium by doping minor actinide in thermal neutron spectra

International Nuclear Information System (INIS)

Peryoga, Yoga; Sagara, Hiroshi; Saito, Masaki; Ezoubtchenko, Alexey

2005-01-01

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% 235 U and 20% 235 U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, 238 Pu, 240 Pu and 242 Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% 235 U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties. (author)

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

In-plant measurements of gamma-ray transmissions for precise K-edge and passive assay of plutonium concentration and isotopic abundance in product solutions at the Tokai Reprocessing Plant

International Nuclear Information System (INIS)

Asakura, Y.; Kondo, I.; Masui, J.; Shoji, K.; Russo, P.A.; Hsue, S.T.; Sprinkle, J.K. Jr.; Johnson, S.S.

1982-01-01

A field test has been carried out for more than 2 years for determination of plutonium concentration by K-edge absorption densitometry and for determination of plutonium isotopic abundance by transmission-corrected passive gamma-ray spectrometry. This system was designed and built at Los Alamos National Laboratory and installed at the Tokai reprocessing plant of the Power Reactor and Nuclear Fuel Development Corporation as a part of the Tokai Advanced Safeguards Technology Exercise (TASTEX). For K-edge measurement of plutonium concentration, the transmissions at two discrete gamma-ray energies are measured using the 121.1- and 122.1-keV gamma rays from 75 Se and 57 Co. Intensities of the plutonium passive gamma rays in the energy regions between 38 and 51 keV and between 129 and 153 keV are used for determination of the isotopic abundances. More than 200 product solution samples have been measured in a timely fashion during these 2 years. The relative precisions and accuracies of the plutonium concentration measurement are shown to be within 0.6% (1 sigma) in these applications, and those for plutonium isotopic abundances are within 3% for 238 Pu, 0.4% for 239 Pu, 1.2% for 240 Pu, 1.3% for 241 Pu, and 7% for 242 Pu. The time required is 10 min for the concentration assay, 10 min for the isotopics assay, and about 15 min for handling procedures in the laboratory

From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

Flexible plutonium management with IFR technology

International Nuclear Information System (INIS)

Hannum, W.H.; Lineberry, M.J.

1993-01-01

From the earliest days of the development of peaceful nuclear power, it has been recognized that efficient utilization of nuclear fuel resources requires a closed fuel cycle (recycle). With a closed cycle, essentially all the energy content of mined uranium can be used, whereas a once-through light water reactor (LWR) cycle uses only ∼0.5%. Since weapons programs have used the PUREX process to extract plutonium, it has further been assumed that this is the appropriate technology for closing the uranium fuel cycle. In the United States, these assumptions were put into question by concerns over commerce in separated plutonium and the threat of diversion of this material for weapons use

Ability to burn plutonium and minor actinides. Interest of accelerator driven system compared to critical reactor

International Nuclear Information System (INIS)

Vergnes, J.; Mouney, H.

1998-01-01

In the frame of the French Act of December 1991, EDF is presently assessing the interest of Acceleration Driven System (ADS) for the Transmutation of the Plutonium and Minor Actinides (MA) produced by its park of nuclear reactors. The studies presented here assess the efficiency of ADS and critical reactors to incinerate Pu and MA (Minor Actinides) and the potential interest of ADS for that purpose. (author)

Experimental results from a reactor monitoring experiment with a cubic meter scale antineutrino detector

International Nuclear Information System (INIS)

Bernstein, A.

2007-01-01

Cubic meter scale antineutrino detectors can stably and no intrusively monitor both plutonium content and reactor power at the few percent level, at a standoff of a few tens of meters. Our Lawrence Livermore National Laboratory/Sandia National Laboratories collaboration has deployed a detector to demonstrate this capability at a 3 GWt pressurized water reactor in Southern California, operating 25 meters from the core center, and acquiring data over an approximate one year period. Such monitoring may be useful for tracking power output and plutonium buildup in nuclear reactors, constraining the fissile content and providing the earliest possible measurement of the amount of plutonium in the reactor core. We present our antineutrino event sample, and show that the observed change in antineutrino rate recorded in our detector over the reactor cycle correlates with plutonium ingrowth according to predictions. We present our current precision and estimate the attainable precision of the method, and discuss the benefits this technology may have for the International Atomic Energy Agency (IAEA) or other safeguards regimes

Study of the Effect of (U0.8Pu0.2O2 Uranium–Plutonium Mixed Fuel Fission Products on a Living Organism

Directory of Open Access Journals (Sweden)

Ayagoz Baimukhanova

2016-08-01

Full Text Available The article describes the results of experiments conducted on pigs to determine the effect of plutonium, which is the most radiotoxic and highly active element in the range of mixed fuel (U0.8Pu0.2O2 fission products, on living organisms. The results will allow empirical prediction of the emergency plutonium radiation dose for various organs and tissues of humans in case of an accident in a reactor running on mixed fuel (U0.8Pu0.2O2.

Plutonium Management, Minor Actinides Partitioning and Transmutation R and D in France

International Nuclear Information System (INIS)

Cavedon, Jean-Marc; Courtois, Charles

2003-01-01

Jean-Marc Cavedon (CEA, France) then presented the developments concerning Plutonium management and minor actinides P and T research and development in France. By the 1991 law on high-level long-lived radioactive waste a research programme was launched in the areas: (i) geological disposal, (ii) conditioning and long-term storage, and (iii) radiotoxicity reduction by P and T. The results of the work in these areas will be presented to the French Government and Parliament in 2006. The control of Plutonium stocks generated by the French PWRs is proposed to increase Plutonium consumption in reactors and minimise radioactive waste production, and requires the recycling of actinides, especially Plutonium. In the long term, CEA intends to develop a new technology based on gas cooled reactors and their associated fuel cycle, including multiple recycling of Plutonium. The advantages of this development consist in the optimisation of the use of natural resources and the concentration of Plutonium in limited quantities of fuel rods. If needed, the minor actinides could also be recycled. The planned CEA developments depend on new fuel types and will lead to novel waste types (light glasses) with a reduction of long-term radiotoxicity. Radiotoxicity reductions by a factor of 3 to 5 are expected for Plutonium recycling scenarios, and by up to a factor of a few hundreds for Plutonium and minor actinides recycling scenarios. This gain is nearly independent on the reactor type used, but needs about 100 years of application to become effective in terms of making a difference in the total waste inventory to be disposed of

Plutonium use in foreign countries (03)

International Nuclear Information System (INIS)

Otagaki, Takao

2004-03-01

European countries and Japan had been implementing the strategy of spent fuel reprocessing in order to use nuclear material to the maximum. Plutonium recovered from reprocessing, however, must be recycle on light water reactors (LWRs) because of considerable delay of fast reactor development. In Europe, much of experiences of plutonium recycling have been accumulated until now. Thus, the status of plutonium recycling up to the end of 2003 in France, Germany, The U.K., Belgium, Switzerland and other countries were studied based on the following scope. (1) Basic policy and present status of plutonium recycling in primary countries of France, Germany, The U.K., Belgium, Switzerland, and Sweden which plans to recycle a part of plutonium: Backend policy and the status of spent fuel management were studied, then integrated analysis and evaluation of the position of plutonium recycling in backend and the status of plutonium recycling development were performed. (2) Plan and experience of Mixed Oxide (MOX) fuel fabrication and reprocessing of spent fuels: The data and information on plan and experience of MOX fuel fabrication and reprocessing in foreign countries were collected. (3) Plutonium inventories: The data and information of plutonium inventories of foreign countries were collected. (author)

In search of plutonium: A nonproliferation journey

Science.gov (United States)

Hecker, Siegfried

2010-02-01

In February 1992, I landed in the formerly secret city of Sarov, the Russian Los Alamos, followed a few days later by a visit to Snezhinsk, their Livermore. The briefings we received of the Russian nuclear weapons program and tours of their plutonium, reactor, explosives, and laser facilities were mind boggling considering the Soviet Union was dissolved only two months earlier. This visit began a 17-year, 41 journey relationship with the Russian nuclear complex dedicated to working with them in partnership to protect and safeguard their weapons and fissile materials, while addressing the plight of their scientists and engineers. In the process, we solved a forty-year disagreement about the plutonium-gallium phase diagram and began a series of fundamental plutonium science workshops that are now in their tenth year. At the Yonbyon reprocessing facility in January 2004, my North Korean hosts had hoped to convince me that they have a nuclear deterrent. When I expressed skepticism, they asked if I wanted to see their ``product.'' I asked if they meant the plutonium; they replied, ``Well, yes.'' Thus, I wound up holding 200 grams of North Korean plutonium (in a sealed glass jar) to make sure it was heavy and warm. So began the first of my six journeys to North Korea to provide technical input to the continuing North Korean nuclear puzzle. In Trombay and Kalpakkam a few years later I visited the Indian nuclear research centers to try to understand how India's ambitious plans for nuclear power expansion can be accomplished safely and securely. I will describe these and other attempts to deal with the nonproliferation legacy of the cold war and the new challenges ahead. )