Kg Of Plutonium Research Articles

FAST ROSSI-ALPHA MEASUREMENTS OF PLUTONIUM USING ORGANIC SCINTILLATORS

In this work, Rossi-alpha measurements were simultaneously performed with a 3He-based detection system and an organic scintillator-based detection system. The assembly is 15 kg of plutonium (93 wt% 239Pu) reflected by copper and moderated by lead. The goal of Rossi-alpha measurements is to estimate the prompt neutron decay constant, alpha. Simulations estimate keff = 0:624 and α = 52:3 ± 2:5 ns for the measured assembly. The organic scintillator system estimated α = 47:4 ± 2:0 ns, having a 9.37% error (though the 1.09 standard deviation confidence intervals overlapped). The 3He system estimated α = 37 μs. The known slowing down time of the 3He system is 35-40 μs, which means the slowing down time dominates and obscures the prompt neutron decay constant. Subsequently, the organic scintillator system should be used for assemblies with alpha much less than 35 μs.

Accounting for Unfissioned Plutonium from the Trinity Atomic Bomb Test.

The Trinity test device contained about 6 kg of plutonium as its fission source, resulting in a fission yield of 21 kT. However, only about 15% of the Pu actually underwent fission. The remaining unfissioned plutonium eventually was vaporized in the fireball and after cooling, was deposited downwind from the test site along with the various fission and activation products produced in the explosion. Using data from radiochemical analyses of soil samples collected postshot (most many years later), supplemented by model estimates of plutonium deposition density estimated from reported exposure rates at 12 h postshot, we have estimated the total activity and geographical distribution of the deposition density of this unfissioned plutonium in New Mexico. A majority (about 80%) of the unfissioned plutonium was deposited within the state of New Mexico, most in a relatively small area about 30-100 km downwind (the Chupadera Mesa area). For most of the state, the deposition density was a small fraction of the subsequent deposition density of Pu from Nevada Test Site tests (1951-1958) and later from global fallout from the large US and Russian thermonuclear tests (1952-1962). The fraction of the total unfissioned Pu that was deposited in New Mexico from Trinity was greater than the fraction of fission products deposited. Due to plutonium being highly refractory, a greater fraction of the Pu was incorporated into large particles that fell out closer to the test site as opposed to more volatile fission products (such as Cs and I) that tend to deposit on the surface of smaller particles that travel farther before depositing. The plutonium deposited as a result of the Trinity test was unlikely to have resulted in significant health risks to the downwind population.

Imaging Special Nuclear Material using a Handheld Dual Particle Imager

A compact radiation imaging system capable of detecting, localizing, and characterizing special nuclear material (e.g. highly-enriched uranium, plutonium…) would be useful for national security missions involving inspection, emergency response, or war-fighters. Previously-designed radiation imaging systems have been large and bulky with significant portions of volume occupied by photomultiplier tubes (PMTs). The prototype imaging system presented here uses silicon photomultipliers (SiPMs) in place of PMTs because SiPMs are much more compact and operate at low power and voltage. The SiPMs are coupled to the ends of eight stilbene organic scintillators, which have an overall volume of 5.74 × 5.74 × 7.11 cm3. The prototype dual-particle imager’s capabilities were evaluated by performing measurements with a 252Cf source, a sphere of 4.5 kg of alpha-phase weapons-grade plutonium known as the BeRP ball, a 6 kg sphere of neptunium, and a canister of 3.4 kg of plutonium oxide (7% 240Pu and 93% 239Pu). These measurements demonstrate neutron spectroscopic capabilities, a neutron image resolution for a Watt spectrum of 9.65 ± 0.94° in the azimuthal direction and 22.59 ± 5.81° in the altitude direction, imaging of gamma rays using organic scintillators, and imaging of multiple sources in the same field of view.

Effects of aging on PuO2·xH2O particle size in alkaline solution

AbstractBetween 1944 and 1989, 54.5 metric tons of the United States´ weapons-grade plutonium and an additional 12.9 metric tons of fuels-grade plutonium were produced in and separated from irradiated uranium metal fuel at the Hanford Site. Acidic high-activity wastes containing around 600 kg of plutonium were made alkaline and discharged to underground storage tanks from separations, isolation, and recycle processes to yield average plutonium concentration of about 0.003 g/L (or ∼ 0.0002 wt. %) in the ∼ 200 million liter tank waste volume. The plutonium is largely associated with low-solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements (e.g., iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of hydrated plutonium oxide, PuO2·xH2O, could undergo sufficient crystal growth through dissolution and reprecipitation in the alkaline tank waste to potentially become separable from neutron absorbing constituents by settling or sedimentation. Thermodynamic considerations and laboratory studies of systems chemically analogous to tank waste show that the plutonium, precipitated in the alkaline tank waste by neutralization from acid solution, probably entered as 2–5-nm PuO2·xH2O crystallite particles that, because of the low concentration of the neutral Pu(IV) dissolved species and opposition from radiolytic processes, grow from that point at exceedingly slow rates.

Transmutation Strategy Using Thorium-Reprocessed Fuel ADS for Future Reactors in Vietnam

Nuclear power is believed to be a key to the energy security for a developing country like Vietnam where the power demanding increases rapidly every year. Nevertheless, spent nuclear fuel from nuclear power plants is the source of radiotoxic and proliferation risk. A conceptual design of ADS utilizing thorium fuel as a based fuel and reprocessed fuel as a seed for nuclear waste transmutation and energy production is proposed as one of the clean, safe, and economical solutions for the problem. In the design, 96 seed assemblies and 84 blanket assemblies were inserted into the core to make a heterogeneous subcritical core configuration. Introducing thorium fuel into the core offers an effective way to transmute plutonium and minor actinide (MA) and gain energy from this process. Transmutation rate as a function of burnup is estimated using MCNPX 2.7.0 code. Results show that by using the seed-blanket designed ADS, at 40&#x2009;GWd/t burnup, 192&#x2009;kg of plutonium and 156&#x2009;kg of MA can be eliminated. Equivalently, 1&#x2009;<svg style="vertical-align:-3.39069pt;width:39.462502px;" id="M1" height="15.65" version="1.1" viewBox="0 0 39.462502 15.65" width="39.462502" xmlns:xlink="http://www.w3.org/1999/xlink" xmlns="http://www.w3.org/2000/svg"> <g transform="matrix(.017,-0,0,-.017,.062,11.362)"><path id="x47" d="M692 302v-28q-45 -6 -54.5 -17t-9.5 -55v-86q0 -62 7 -95q-7 -2 -32 -7.5l-37 -8l-35.5 -7.5t-40.5 -7t-38.5 -4t-40.5 -2q-165 0 -266 93t-101 243q0 84 32 151t86 108t121.5 63t142.5 22q29 0 62.5 -4t52 -7.5t47.5 -10.5t31 -7l15 -156l-28 -6q-17 87 -66.5 121.5&#xA;t-128.5 34.5q-117 0 -191 -80.5t-74 -216.5q0 -56 17 -110t49.5 -100.5t87.5 -74.5t124 -28q53 0 87 17q17 9 24.5 23t7.5 43v85q0 54 -15.5 67.5t-89.5 18.5v28h254z" /></g><g transform="matrix(.017,-0,0,-.017,12.217,11.362)"><path id="x57" d="M956 650v-28q-52 -7 -69 -22.5t-39 -77.5q-60 -175 -165 -533h-31l-175 499h-2l-163 -499h-31l-164 535q-18 60 -36 76.5t-65 21.5v28h243v-28q-53 -7 -63 -18.5t3 -49.5l122 -412h2l162 499h29l175 -500h2q87 277 113 401q12 45 0.5 59t-68.5 21v28h220z" /></g> <g transform="matrix(.012,-0,0,-.012,28.725,15.45)"><path id="x74" d="M298 36l-68 -35q-25 -13 -42 -13q-98 0 -98 119v290h-69l-4 11l28 29h45v72l65 67l14 -2v-137h113q8 -8 6 -21t-12 -19h-107v-256q0 -96 59 -96q29 0 61 17z" /></g><g transform="matrix(.012,-0,0,-.012,32.414,15.45)"><path id="x68" d="M513 0h-211v26q44 5 55 17t11 60v166q0 123 -98 123q-59 0 -104 -51v-238q0 -48 10 -60.5t55 -16.5v-26h-213l1 26q46 4 57 16t11 61v477q0 48 -9.5 58.5t-58.5 14.5v24q63 9 147 35v-336q73 73 143 73q64 0 101 -42t37 -129v-175q0 -49 11 -61t55 -16v-26z" /></g> </svg> ADS can be able to transmute the transuranic (TRU) waste from 2&#x2009;<svg style="vertical-align:-3.39069pt;width:39.462502px;" id="M2" height="15.65" version="1.1" viewBox="0 0 39.462502 15.65" width="39.462502" xmlns:xlink="http://www.w3.org/1999/xlink" xmlns="http://www.w3.org/2000/svg"> <g transform="matrix(.017,-0,0,-.017,.062,11.362)"><use xlink:href="#x47"/></g><g transform="matrix(.017,-0,0,-.017,12.217,11.362)"><use xlink:href="#x57"/></g> <g transform="matrix(.012,-0,0,-.012,28.725,15.45)"><use xlink:href="#x74"/></g><g transform="matrix(.012,-0,0,-.012,32.414,15.45)"><use xlink:href="#x68"/></g> </svg> LWRs. 14 units of ADS would be required to eliminate TRUs from the future reactors to be constructed in Vietnam.

Implications for Advanced Safeguards Derived from a Proliferation Resistance and Physical Protection Case Study for a Generation IV Nuclear Energy System

The Proliferation Resistance and Physical Protection Evaluation Methodology Working Group of the Generation IV International Forum produced a full-system case study on the Example Sodium Fast Reactor Nuclear Energy System (ESFR-NES). The ESFR-NES is a hypothetical fuel cycle complex consisting of four sodium-cooled fast reactors of medium size collocated with an on-site dry-fuel storage facility and a spent-fuel reprocessing facility based on electrochemical recycling technology. The complex recycles irradiated fuels from two feed streams, oxide fuel from off-site light water reactors and metal fuel from the on-site sodium-cooled fast reactors. Both of these streams are recycled on-site; uranium and transuranics are sent to the electrochemical reprocessing fuel cycle facility. The two streams combine and the fuel cycle facility creates new ESFR-NES metal fuel for the four on-site sodium-cooled fast reactors. The major safeguards concepts driving the safeguards analysis were timeliness goals and material quantity goals. Specifically, the recycled fuel, the in-process material in the fuel reprocessing facility, the off-site light water reactor spent fuel received at the ESFR-NES, and spent fuel from the on-site fast reactors will contain plutonium. The International Atomic Energy Agency defines the material within the ESFR-NES as “direct-use material” with a stringent timeliness goal of 3 months and a material quantity goal of 8 kg of plutonium. Furthermore, the ESFR-NES may have some intrinsic safeguards features if the plutonium and uranium are not separated during reprocessing. This facility would require major modifications to separate the plutonium from other transuranic elements in the reprocessed fuel. The technical difficulty in diverting material from the ESFR-NES is at least as strongly impacted by the adversaries’ overall technical capabilities as it is by the effort required to overcome those barriers intrinsic to the nuclear fuel cycle. The intrinsic proliferation resistance of the ESFR-NES can affect how extrinsic measures in the safeguards approach for the complex will provide overall proliferation resistance.

Detection of anomalous reactor activity using antineutrino count evolution over the course of a reactor cycle

Abstract

Exploring Uranium Resource Constraints on Fissile Material Production in Pakistan

This paper evaluates possible scenarios for Pakistan's uranium enrichment and plutonium production programs since the late 1970s by using Pakistan's supply of natural uranium as a constraint. Since international sanctions have prevented Pakistan from importing uranium for decades, it has had to rely on domestic uranium production—currently estimated as approximately 40 tons a year. The paper divides the development of Pakistan's uranium enrichment and plutonium production programs into three broad periods: from the beginning in the late 1970s until the 1998 nuclear tests; from 1999 to the present; and from the present to 2020; and considers how Pakistan could allocate its domestic uranium between its uranium enrichment and plutonium production programs for each period. This assessment is completed for enrichment capacities ranging from 15,000 to 75,000 separative work units (SWU) and takes into account the construction of the second and third plutonium production reactors at Khushab. The study finds that Pakistan may have sufficient natural uranium to fuel the three reactors, if they are approximately 50 MWt each, but that for some of these enrichment capacities, there will be a shortfall of natural uranium by 2020. The paper considers the impact of alternative sources of enrichment feed such as depleted tails from previous enrichment activity and reprocessed uranium from low-burn-up spent fuel from the Khushab reactors. There are signs Pakistan early on may have enriched some reprocessed uranium, possibly acquired from China. It finds that by 2020, Pakistan could have accumulated approximately 450 kg of plutonium from the Khushab reactors and 2500–6000 kg of highly enriched uranium (HEU) (90 percent enriched) for enrichment capacities ranging from 15,000–75,000 SWU. These stocks would be sufficient for perhaps 100–240 simple fission weapons based on HEU and for 90 plutonium weapons. Pakistan may be able to produce more weapons if it either increases its rate of uranium mining or has more advanced weapon designs requiring less fissile material in each weapon.

Plutonium oxide polishing for MOX fuel fabrication

Los Alamos National Laboratory (LANL) successfully polished 120 kg of plutonium from surplus nuclear weapons for the European Mixed Oxide (MOX) Lead Test Assembly (LTA), managed by Duke, COGEMA, and Stone & Webster (DCS). The purified oxide was fabricated into fuel pellets, which comprised the LTAs. The LTAs have been inserted into the Catawba (SC, USA) nuclear reactor, where they are presently being used to generate electricity. The material used in the MOX LTA represents the first plutonium oxide from LANL to be polished under the quality requirements set forth by the Nuclear Regulatory Commission (NRC). In addition, the MOX project has been hailed as the “largest single nonproliferation project in history” by US Ambassador Linton Brooks (23/9/04). The process flow for PuO 2 purification was based on aqueous recovery that included various unit operations (dissolution, ion exchange, oxalate precipitation, and calcination). Data from a variety of chemical and physical analyses demonstrate product quality and process consistency. Further process optimization experiments will be performed prior to polishing an additional 330 kg of PuO 2 for the Mixed Oxide Fuel Fabrication Facility (MFFF) at Savannah River Site. During LTA production in 2004, approximately 500 L of 7 M nitric acid was used for washing impurities from the plutonium in order to meet product purity specifications. The majority of this acid was sent to the LANL Radioactive Liquid Waste Treatment Facility (TA-50) as effluent. Optimization experiments during 2006 and 2007 will investigate the use of recycled HNO 3 for PuO 2 washing and ion exchange. If successful, aqueous polishing processes will re-use almost all of the nitric acid that would otherwise have been sent to TA-50. In addition, an overall reduction in the volume of nitric acid used during ion exchange will be tested. The use of new Inconel calcination boats will also be verified. Additional process optimization activities will be conducted to validate the use of 100% quaternized Reillex ion exchange resin. Qualification exercises will include a series of process runs, followed by chemical analyses to assess product oxide impurity levels. Once the optimized processes have been qualified, the production phase of the project will begin.

Plutonium Oxide Polishing for MOX Fuel Production

:Los Alamos National Laboratory (LANL) has been tasked to produce 120 kg of plutonium as highly purified PuO2 for the European Mixed Oxide (MOX) Lead Test Assembly managed by Duke, COGEMA, Stone & Webster (DCS). To meet stringent and challenging technical requirements for PuO2 production, the LANL aqueous polishing team recently established consistency in generating quality material from weapons-grade PuO2. Polishing was required to remove undesirable quantities of trace-metal impurities, particularly gallium, and to produce a material with appropriate powder characteristics, such as particle size and surface area. The process flow sheet for PuO2 polishing was based on aqueous purification and included the unit operations of dissolution, ion exchange, oxalate precipitation, and calcination.

Conclusions on plutonium separation from atmospheric krypton-85 measured at various distances from the Karlsruhe reprocessing plant

For wide-area atmospheric monitoring, krypton-85 is the best indicator for clandestine plutonium separations. The detection and false alarm rates were determined from weekly samples at five different distances from the Karlsruhe reprocessing plant between 1985 and 1988. The detection rate for the separation of 4 kg of plutonium per week was found to be as high as 80–90% at a distance of less than 1 km, 70% at 5 km, 40% at 39 km, and 15% at 130 km. At distances up to 40 km, the false alarm rate is less than 3.5%. On average, the fuel released 28 TBq krypton-85 per kg plutonium. For weapons-grade plutonium, the krypton signal would be lower by a factor of 2. Hence, the given percentages correspond to the detection probabilities for the separation of a significant quantity (8 kg) of plutonium per weekly sample under the specific meteorological conditions of the WAK. The minimum separation rates that could have been detected are 2 gram of weapons-grade plutonium per week at a distance of less than 1 km, 40 g/week at 5 km, 200 g/week at 39 km, and 1000 g/week at 130 km.

What views and uses of radiation sources in the 21st century?

Considering that in 1899 neither biotechnology nor the electronic revolution were foreseen, some humility might be advisable when one looks into the crystal ball for the future role of radiation sources. In the past 50 years, nuclear medicine, nuclear weapons, and nuclear power have had a huge impact in the world. In the next 50 years, nuclear weapons may be phased out, nuclear power revived, and nuclear medicine may continue, especially for diagnostic purposes. Conflicts between great powers and blocks will no longer be about territorial or ideological domination but about trade, finance, information, and the environment and the weapons used will not be bombs but investments, credits, and control of information. Nuclear power-still based on fission-will be relaunched and get more uses, e.g., to propel ships, to produce heat for industry and for space heating, and perhaps to desalinate water. The public will be more at ease with radiation as it is better educated, as nuclear safety continuously improves and new types of nuclear power plants emerge, as waste sites fail to cause any problems, and as no other energy source is found to deliver so much energy at reasonable cost with negligible impact on climate and environment. One kilogram of oil corresponds to 4 kWh of electricity. One kilogram of uranium fuel corresponds to 50,000 kWh, and 1 kg of plutonium 6,000,000 kWh! In nuclear medicine, radiation may give way to other treatments as the understanding of cancer advances. On the other hand, the extreme ease with which sources of radiation can be identified is unmatched and likely to make them useful tools as tracers and markers in medicine-and other fields-for a long time. For certain uses--perhaps food irradiation--radiation sources, such as cobalt, may be replaced by accelerators which may be switched on and off at will. As more sources are used, registration and control of them must be made very effective around the whole world. Very high natural emissions of radon will continue to call for cautionary measures, but many other nonradiating substances will be identified as hazardous to health and call for vigorous intervention.

Partitioning of plutonium between glass, secondary phases, and contact refractories

AbstractA wide variety of plutonium-containing wastes have been generated from both defense and commercial nuclear programs. The Advanced Mixed Waste Treatment Project (AMWTP) involves the privatized treatment of 65,000 m3 of mixed TRU and low-level waste that is in storage at DOE's INEEL site. BNFL Inc. leads the commercial team including GTS Duratek, Inc., that has been selected to design, construct, and operate a facility to perform that service. Since vitrification is a potential treatment option, the current study was conducted to investigate issues surrounding the incorporation of the approximately 700 kg of plutonium that is present in the AMWTP waste. A series of glasses were prepared with plutonium (0.5 wt%), with cerium (as a potential surrogate), with both, and with neither. Scanning electron microscopy coupled with energy dispersive x-ray spectroscopy (SEM-EDS) and optical analyses confirmed that both elements were completely dissolved into the as-melted glasses. Heat treatments forced the formation of secondary phases, which were analyzed for Pu and Ce incorporation. Plutonium-containing glasses were spiked with sulfates and remelted to force the formation of a sulfate salt phase, which was analyzed to determine partitioning of Pu and Ce. Coupons of Monofrax K-3 glass contact refractory were exposed to melts of these glasses for six days at 1208°C using both standard and modified versions of ASTM procedure C621-84. Thickness loss, microstructural alteration, and the extent to which Pu and Ce were incorporated into the material were then determined by methods employing SEMEDS\ wavelength dispersive x-ray spectroscopy (WDS). Extreme heat treatments were found to lead to the formation of small amounts (&lt; 0.1 vol%) of either (Pu, Ce)O2, or PuO2, depending upon whether Ce was present or absent, respectively. The Pu level in the sulfate phase was at least a factor often below that of the parent glass, and no tendency to concentrate Pu in the K-3 refractory above the level in the parent glass was found.

Separation of Americium, Curium, and Plutonium from Irradiated Targets

Abstract The Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL) processes highly irradiated targets for the Mark 42 program to separate Am, Cm, and Pu. The target feed material for each assembly was 3.3 kg of plutonium (78% 239Pu) that was irradiated at the Savannah River Site to yield approximately 100 g each of 243Am and 244Cm, and 100-g quantities of 242Pu for special DOE projects. The REDC has plans to process ten of these target assemblies over the next few years. The first assembly has been dissolved, and approximately 1/4 of this material has been used to test the processing flowsheet. Various aqueous processes developed at the REDC over the past years were utilized to dissolve the target segments, separate the bulk of the impurities from the transuranics, separate the plutonium from the transplutonium actinides, and separate the rare earth fission products from the Am-Cm. The separation of the Am-Cm products to the desired purity levels presented new proces...

Fate of Plutonium Released from the Nagasaki A-Bomb, Japan

About 15 kg of plutonium was used in the Nagasaki A-bomb, but only 1 kg was fissioned and the rest (14 kg) was released into the environment. The fate of this plutonium was investigated at the east side of Nagasaki city where the local fallout was deposited. The four objectives investigated concerning the unfissioned 239+240Pu were: (1) the geographical distribution up to 100 km from the hypocentre, (2) the vertical movement of the deposited plutonium within a soil core and a reservoir sediment core, (3) the ecological distribution, and (4) an estimation of the total mass deposited as local fallout. The highest concentration of 239+240Pu 64.5 mBq/g, was found at a point 2.8 km from the hypocentre. The local fallout was observed over a limited area up to 18 km east from the hypocentre. Plutonium was mobile within the soils and reservoir sediments. For the former an amount of 10% moved downward during the last four decades and the rest (90%) remained within 10 cm from the surface. A considerable amount of plutonium was found in fish (0.0195 mBq/g dry), freshwater shellfish (0.0278 mBq/g dry), ginger root (0.0366 mBq/g dry), and sweet potato (0.0110 mBq/g dry). An estimation of plutonium deposited as local fallout from the A-bomb was 37.54 grams.

Plutonium Recycling and the Problem of Nuclear Proliferation

A typical 1 gigawatt LWR, the dominant commercial power reactor type today, operating at 70 per cent capacity factor, generates approximately 250kg of plutonium annually. This plutonium, which is produced in the reactor through neutron capture by U-238, is then discharged from the reactor along with the other constituents of the spent fuel. About 0.7 per cent of the plutonium, or 175kg, is fissile (odd) isotopes of plutonium. 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.

Development and Active Demonstration of Acid Digestion of Plutonium-Bearing Waste

At the Eurochemic site, 800 kg of combustible alpha waste containing ∼7 kg of plutonium were treated from March 1983 to July 1985 with the aim of concentrating the plutonium by oxidating the waste and converting it into a soluble form so that the established purification processes could be applied.In a batch process, shredded waste is oxidized with nitric acid in sulfuric acid. The digester content is then kept for several hours at digestion temperature to complete the dissolution of plutonium dioxide. The cold digester content is then filtered and the plutonium-containing filter cake is sent to the plutonium purification system. The off-gases generated are freed from the acids by scrubbing.The process is demonstrated in a plant with a daily throughput of 10 kg of waste. For the oxidation of waste and the dissolution of plutonium dioxide, a ring-type digester made of technical glass is used. The following principal results have been obtained: 1. Complete oxidation of the waste material is achieved within 15 min at a digester acid temperature of 250°C under oxidizing conditions provided by nitric acid. 2. At 250°C and with constant stirring of the digester content, a plutonium oxide to plutonium sulfate conversion rate of up to 99% is obtained within 8 h. 3. The average waste throughput achieved has been 4.1 kg per run (maximum of 10.4 kg). The plutonium decontamination factors were 1010 for the cleaned off-gas and 106 for the liquid secondary waste.

A Novel Non-Power-Producing Fusion-Fission Breeder Reactor

One of the potentially attractive applications of nuclear fusion is to breed fissile fuel for use in fission reactors. A fusion-fission breeder is examined, based on four unique concepts: operations in a non-power-producing mode, a low technology (low pressure and temperature) aqueous self-cooled blanket for breeding fissile fuel, the spherical torus confinement scheme (low-aspect-ratio tokamak), and the catalyzed deuterium-deuterium (D-D) fuel cycle. The breeding of fissile fuel is accomplished by dissolving a uranium salt, i.e., uranyl nitrate, in heavy water that cools both the first wall and blanket. The use of the catalyzed D-D fuel cycle eliminates the need for tritium breeding. The neutron wall loading for this reactor is only --0.5 MW/m/sup 2/, and the fusion power output is --1000 MW(thermal). Analysis of this novel reactor concept indicates a fissile breeding ratio of 1.34 fissile atom/source neutron using a 15-cm beryllium moderator/multiplier region and 7 mol% uranyl nitrate in the heavy water. A typical reactor using this blanket can produce more than 7400 kg of plutonium per operating year. This concept can provide fissile fuel at a cost that is comparable to previous fusion breeder designs but at a capital cost of about one-third that of the previous designs.

Feasibility study of a magnetic fusion production reactor

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 electricity, and (2) there is a risk of not meeting the design goals.

Recovery of Plutonium by Pyroredox Processing *

Using pyrochemical oxidation and reduction, we have developed a process to recover the plutonium in impure scrap with less than 95% plutonium. This plutonium metal was further purified by pyrochemical electrorefining. During development of the procedures, depleted electrorefining anodes were processed, and over 80% of the plutonium was recovered as high-purity metal in one electrorefining cycle. Over 40 kg of plutonium has been recovered from 55 kg of impure anodes with our procedures. 6 refs., 2 figs., 5 tabs.