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At least 253 records · Page 14
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DOE OSTI2011

High Energy Protons for Remote Standoff Detection of Special Nuclear Materials

High energy protons can be a powerful probe for standoff detection of Special Nuclear Materials (SNM). In particular, the use of protons of 500 MeV and up with modest beam currents, 10 to 100 {mu}A, can be far more efficient in the production of fission neutrons than conventional bremsstrahlung based electron linacs. Such beams can provide both the range and, more importantly, the sensitivity to fissile materials required for remote detection at distances of hundreds of meters or more. The practical application of this technique has been limited by the requirements of producing such high energy protons using conventional proton accelerators--typically very large, fixed installations. The on-going development at MIT of a new generation of very compact superconducting cyclotrons makes a compelling case for examining this approach for a practical standoff detection system where large numbers of neutrons can be produced using an accelerator with {approx}1 GeV energy and modest beam currents in the 1 to 100 {mu}A range. We will describe current work in the design of a 250 MeV, 1 mA superconducting accelerator and its extension to higher energies, and make estimates of sensitivity to reasonable amounts of shielded fissile material.

43 PARTICLE ACCELERATORS↗
DOE OSTI2012

Spent fuel utilization in a compact traveling wave reactor

In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Nuclear programs in India and Pakistan

India and Pakistan launched their respective nuclear programs in the 1940s and 1950s with considerable foreign technical support, especially from the United States Atoms for Peace Program. The technology and training that was acquired served as the platform for later nuclear weapon development efforts that included nuclear weapon testing in 1974 and in 1998 by India, and also in 1998 by Pakistan - which had illicitly acquired uranium enrichment technology especially from Europe and received assistance from China. As of 2013, both India and Pakistan were continuing to produce fissile material for weapons, in the case of India also for nuclear naval fuel, and were developing a diverse array of ballistic and cruise missiles. International efforts to restrain the South Asian nuclear build-up have been largely set aside over the past decade as Pakistani support became central for the U.S. war in Afghanistan and as U.S. geopolitical and economic interests in supporting the rise of India, in part as a counter to China, led to India being exempted both from U.S non-proliferation laws and international nuclear trade guidelines. In the absence of determined international action and with Pakistan blocking the start of talks on a fissile material cutoff treaty, nuclear weapon programs in South Asia are likely to keep growing for the foreseeable future.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2015

Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Variants of closing the nuclear fuel cycle

Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Tritium accountancy in fusion systems

The US Department of Energy (DOE) has clearly defined requirements for nuclear material control and accountability (MCA) of tritium whereas the International Atomic Energy Agency (IAEA) does not since tritium is not a fissile material. MCA requirements are expected for tritium fusion machines and will be dictated by the host country or regulatory body where the machine is operated. Material Balance Areas (MBA) are defined to aid in the tracking and reporting of nuclear material movements and inventories. Material sub-accounts (MSA) are established along with key measurement points (KMP) to further subdivide a MBA to localize and minimize uncertainties in the inventory difference (ID) calculations for tritium accountancy. Fusion systems try to minimize tritium inventory which may require continuous movement of material through the MSA. The ability of making meaningful measurements of these material transfers is described in terms of establishing the MSA structure to perform and reconcile ID calculations. For fusion machines, changes to the traditional ID equation will be discussed which includes breeding, burn-up, and retention of tritium in the fusion device. The concept of 'net' tritium quantities consumed or lost in fusion devices is described in terms of inventory taking strategies and how it is used to track the accumulation of tritium in components or fusion machines. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

DISSOLUTION OF 3013-DE SAMPLE 10-16

The HB-Line Facility has a long-term mission to dissolve and disposition legacy fissile materials. HB-Line dissolves plutonium dioxide (PuO{sub 2}) from K-Area parting support of the 3013 Destructive Examination (DE) program. The PuO{sub 2}-bearing solids originate from a variety of unit operations and processing facilities, but all of the material is assumed to be high-fired (i.e., calcined in air for a minimum of two hours at {ge} 750 C). The Savannah River National Laboratory (SRNL) conducted dissolution flowsheet studies on 3013 DE Sample 10-16 (can R610826), which contains weapons-grade plutonium (Pu) as the fissile material. The dissolution flowsheet study was performed for 4 hours at 108 C on unwashed material using 12 M nitric acid (HNO{sub 3}) containing 0.20 M potassium fluoride (KF). After 4 hours at 108 C, the {sup 239}Pu Equivalent concentration was 32.5 g/L (gamma, 5.0% uncertainty). The insoluble residue comprised 9.88 wt % of the initial bulk weight, and contained 5.31-5.95 wt % of the initial Pu. The residue contained Pu in the highest concentration, followed by tungsten (W). Analyses detected 2,770 mg/L chloride (Cl{sup -}) in the final dissolver solution (3.28 wt %), which is significantly lower than the amount of Cl{sup -} detected by prompt gamma (9.86 wt %) and the 3013 DE Surveillance program (14.7 wt %). A low bias in chloride measurement is anticipated due to volatilization during the experiment. Gas generation studies found approximately 60 mL of gas per gram of sample produced during the first 30 minutes of dissolution. Little to no gas was produced after the first 30 minutes. Hydrogen gas (H{sub 2}) was not detected in the sample. Based on detection limits and accounting for dilution, the generated gas contained < 0.12 vol % H{sub 2}, which is well below the 4.0 vol % flammability limit for H{sub 2} in air. Filtration of the dissolver solution occurred readily. When aluminum nitrate nonahydrate (ANN) was added to the filtered dissolver solution at a 3:1 Al:F molar ratio, and stored at room temperature for 24 hours, the solution filtered approximately 6 times slower than when filtered 30 minutes after ANN addition, requiring 6 minutes to complete compared to 55 seconds for the first filtration. It is likely that the ambient-temperature solubility for ANN was exceeded. A 4-hour dissolution time at a temperature of 108 C in 12 M HNO{sub 3}/0.2 M KF is recommended for dissolution of this material.

08 HYDROGEN↗
DOE OSTI2011

CF Neutron Time of Flight Transmission for Material Identification for weapons Trainers

The neutron transmission, elastic scattering, and non-elastic reactions can be used to distinguish various isotopes. Neutron transmission as a function of energy can be used in some cases to identify materials in unknown objects. A time tagged californium source that provides a fission spectrum of neutrons is a useful source for neutron time-of-flight (TOF) transmission measurements. Many nuclear weapons trainer units for a particular weapons system (no fissile, but of same weight and center of gravity) in shipping containers were returned to the National Nuclear Security Administration Y-12 National Security Complex in the mid 1990s. Nuclear Materials Identification System (NMIS) measurements with a time tagged californium neutron source were used to verify that these trainers did not contain fissile material. In these blind tests, the time distributions of neutrons through the containers were measured as a function of position to locate the approximate center of the trainer in the container. Measurements were also performed with an empty container. TOF template matching measurements were then performed at this location for a large number of units. In these measurements, the californium source was located on one end of the container and a proton recoil scintillator was located on the other end. The variations in the TOF transmission for times corresponding to 1 to 5 MeV were significantly larger than statistical. Further examination of the time distribution or the energy dependence revealed that these variations corresponded to the variations in the neutron cross section of aluminum averaged over the energy resolution of the californium TOF measurement with a flight path of about 90 cm. Measurements using different thicknesses of aluminum were also performed with the source and detector separated the same distance as for the trainer measurements. These comparison measurements confirmed that the material in the trainers was aluminum, and the total thickness of aluminum through the trainers was determined. This is an example of how californium transmission TOF measurements can be used to identify materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Metal Matrix Microencapsulated (M3) fuel neutronics performance in PWRs

Metal Matrix Microencapsulated (M3) fuel consists of TRISO or BISO coated fuel particles directly dispersed in a matrix of zirconium metal to form a solid rod (Fig. 1). In this integral fuel concept the cladding tube and the failure mechanisms associated with it have been eliminated. In this manner pellet-clad-interactions (PCI), thin tube failure due to oxidation and hydriding, and tube pressurization and burst will be absent. M3 fuel, given the high stiffness of the integral rod design, could as well improve grid-to-rod wear behavior. Overall M3 fuel, compared to existing fuel designs, is expected to provide greatly improved operational performance. Multiple barriers to fission product release (ceramic coating layers in the coated fuel particle and te metal matrix) and the high thermal conductivity zirconium alloy metal matrix contribute to the enhancement in fuel behavior. The discontinuous nature of fissile material encapsulated in coated particles provides additional assistance; for instance if the M3 fuel rod is snapped into multiple pieces, only the limited number of fuel particles at the failure cross section are susceptible to release fission products. This is in contrast to the conventional oxide fuel where the presence of a small opening in the cladding provides the pathway for release of the entire inventory of fission products from the fuel rod. While conventional metal fuels (e.g. U-Zr and U-Mo) are typically expected to experience large swelling under irradiation due to the high degree of damage from fission fragments and introduction of fission gas into the lattice, this is not the case for M3 fuels. The fissile portion of the fuel is contained within the coated particle where enough room is available to accommodate fission gases and kernel swelling. The zirconium metal matrix will not be exposed to fission products and its swelling is known to be very limited when exposed solely to neutrons. Under design basis RIA and LOCA, fuel performance will be superior to the conventional oxide fuel since PCMI and rod burst, respectively, have been mitigated. Under beyond design basis accident scenarios where the fuel is exposed to high temperature steam for prolonged periods, larger inventory of zirconium metal in the core could negatively affect the accident progression. A thin steam resistant layer (e.g. alumina forming alloy steel), integrated into the solid rod during fabrication by co-extrusion or hot-isostatic-pressing, offers the potential to provide additional fuel protection from steam interaction: blanketing under a range of boiling regimes and under severe accident conditions up to high temperatures well beyond what is currently possible in the conventional fuel. A crucial aspect to the viability of M3 fuel in light water reactors is the reduced heavy metal load compared to standard pellet fuel. This study evaluated the design requirements to operate a Pressurized Water Reactor (PWR) with M3 fuel in order to obtain fuel cycle length, reactivity coefficients, and power peaking factors comparable to that of standard fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Distinguishing Pu Metal From Pu Oxide Using Fast Neutron Counting

We describe a method for simultaneously determining the {alpha}-ratio and k{sub eff} for fissile materials using fast neutrons. Our method is a generalization of the Hage-Cifarrelli method for determining k{sub eff} for fissile assemblies which utilizes the shape of the fast neutron spectrum. In this talk we illustrate the method using Monte Carlo simulations of the fast neutrons generated in PuO{sub 2} to calculate the fast neutron spectrum and Feynman correlations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Creating NDA working standards through high-fidelity spent fuel modeling

The Next Generation Safeguards Initiative (NGSI) is developing advanced non-destructive assay (NDA) techniques for spent nuclear fuel assemblies to advance the state-of-the-art in safeguards measurements. These measurements aim beyond the capabilities of existing methods to include the evaluation of plutonium and fissile material inventory, independent of operator declarations. Testing and evaluation of advanced NDA performance will require reference assemblies with well-characterized compositions to serve as working standards against which the NDA methods can be benchmarked and for uncertainty quantification. To support the development of standards for the NGSI spent fuel NDA project, high-fidelity modeling of irradiated fuel assemblies is being performed to characterize fuel compositions and radiation emission data. The assembly depletion simulations apply detailed operating history information and core simulation data as it is available to perform high fidelity axial and pin-by-pin fuel characterization for more than 1600 nuclides. The resulting pin-by-pin isotopic inventories are used to optimize the NDA measurements and provide information necessary to unfold and interpret the measurement data, e.g., passive gamma emitters, neutron emitters, neutron absorbers, and fissile content. A key requirement of this study is the analysis of uncertainties associated with the calculated compositions and signatures for the standard assemblies; uncertainties introduced by the calculation methods, nuclear data, and operating information. An integral part of this assessment involves the application of experimental data from destructive radiochemical assay to assess the uncertainty and bias in computed inventories, the impact of parameters such as assembly burnup gradients and burnable poisons, and the influence of neighboring assemblies on periphery rods. This paper will present the results of high fidelity assembly depletion modeling and uncertainty analysis from independent calculations performed using SCALE and MCNP. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Nuclear Naval Propulsion: A Feasible Proliferation Pathway?

There is no better time than now to close the loophole in Article IV of the Nuclear Non-proliferation Treaty (NPT) that excludes military uses of fissile material from nuclear safeguards. Several countries have declared their intention to pursue and develop naval reactor technology, including Argentina, Brazil, Iran, and Pakistan, while other countries such as China, India, Russia, and the United States are expanding their capabilities. With only a minority of countries using low enriched uranium (LEU) fuel in their naval reactors, it is possible that a state could produce highly enriched uranium (HEU) under the guise of a nuclear navy while actually stockpiling the material for a nuclear weapon program. This paper examines the likelihood that non-nuclear weapon states exploit the loophole to break out from the NPT and also the regional ramifications of deterrence and regional stability of expanding naval forces. Possible solutions to close the loophole are discussed, including expanding the scope of the Fissile Material Cut-off Treaty, employing LEU fuel instead of HEU fuel in naval reactors, amending the NPT, creating an export control regime for naval nuclear reactors, and forming individual naval reactor safeguards agreements.

export controls↗
DOE OSTI2012

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

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Simultaneous {sup 233}U and{sup 235}U characterization through the assay of delayed neutron temporal behavior

Aqueous solutions containing dissolved uranium-233 and uranium-235 were irradiated for 60's in the SLOWPOKE-2 reactor at the Royal Military College of Canada. The temporal behavior of the delayed neutrons produced was recorded by the Facility's Delayed Neutron Counting (DNC) system. The percentage of uranium-233 as a function of total fissile mass present in each sample ranged from 0 to 100% and was predicted by the DNC system with average absolute errors of {+-} 4%. Future work will upgrade the system electronics and software to reduce both uncertainties in timings and electrical noise. Mixture analysis will also be expanded to include plutonium-239 and fissile materials contained in non-aqueous matrices. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Application of fully ceramic microencapsulated fuels in light water reactors

This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Final Demolition and Disposition of 209-E Critical Mass Laboratory - 12267

The 209-E Critical Mass Laboratory was constructed in 1960 to provide a heavy shielded reactor room where quantities of plutonium or uranium in solution could be brought to near-critical configurations under carefully controlled and monitored conditions. In the late 1980's, the responsible contractor, Pacific Northwest National Laboratory (PNNL), was directed by the Department of Energy (DOE) to prepare the facility for unoccupied status. The facility was demolished under a Removal Action Work Plan pursuant to the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). The funding for this project was provided by the American Recovery and Reinvestment Act (ARRA). The primary rooms of concern with regards to contamination in 209-E facility, which is over 9,000 square feet, are the criticality assembly room (CAR), the mix room, and the change room. The CAR contained two reactor hoods (HO-140 and HO-170), which each had a high efficiency particulate air (HEPA) filter system. The CAR contained 13 tanks ranging from 38 L (10 gal) to 401 L (106 gal). Tanks TK-109 and TK-110 are below grade, and were removed as part of this demolition and disposition remedy. Nonradiological and radiological hazardous substances were removed, decontaminated, or fixed in place, prior to demolition. Except for the removal of below grade tanks TK-109 and TK-110, the facility was demolished to slab-on-grade. PNNL performed stabilization and deactivation activities that included removal of bulk fissile material and chemicals, flushing tanks, stabilizing contamination within gloveboxes and hoods, and packaging and removing waste. The removal of the contaminated plutonium equipment and materials from the 209E facility presented a number of challenges similar in nature to those associated with the inventory reduction and cleanup activities at the Plutonium Finishing Plant. Although there were no bulk fissile materials or chemicals within the facility, there were residual radiological materials (isotopes of plutonium and americium) in the tanks and hoods. The complexity of the remedy was present because of the various configurations of the tanks and hoods, combined with the residual contaminants. Because of the weight and dimensional configuration, size reduction of the slab tanks, as well as removal and disposal of the different material used for moderation and absorption, were two examples of challenges that were resolved to complete the remedy. One of the key methods developed and implemented at the facility was the design and construction of a shroud to allow the cutting of the Pu contaminated tanks. The shroud design, development and implementation at the 209E Project was an example of enhanced work planning and task hazards analysis with worker involvement. This paper will present the lessons learned from the 209E facility inventory reduction activities including the shroud and other methodologies used. The initial Lessons Learned discussion for this project was scheduled for late January 2012. This facility is the first open-air demolition of a highly contaminated plutonium-contaminated facility accomplished by CH2M Hill under the Plateau Remediation Contract. The demolition was completed without spread of contamination to the workers and the surrounding area. As with any project of this complexity, there are significant accomplishments, as well as experience that can be applied to future demolition of plutonium-contaminated facilities on the Hanford Site. These experiences will be documented at a later date. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Terminating Safeguards on Excess Special Nuclear Material: Defense TRU Waste Clean-up and Nonproliferation - 12426

The Department of Energy (DOE) and the National Nuclear Security Administration (NNSA) manages defense nuclear material that has been determined to be excess to programmatic needs and declared waste. When these wastes contain plutonium, they almost always meet the definition of defense transuranic (TRU) waste and are thus eligible for disposal at the Waste Isolation Pilot Plant (WIPP). The DOE operates the WIPP in a manner that physical protections for attractiveness level D or higher special nuclear material (SNM) are not the normal operating condition. Therefore, there is currently a requirement to terminate safeguards before disposal of these wastes at the WIPP. Presented are the processes used to terminate safeguards, lessons learned during the termination process, and how these approaches might be useful for future defense TRU waste needing safeguards termination prior to shipment and disposal at the WIPP. Also described is a new criticality control container, which will increase the amount of fissile material that can be loaded per container, and how it will save significant taxpayer dollars. Retrieval, compliant packaging and shipment of retrievably stored legacy TRU waste has dominated disposal operations at WIPP since it began operations 12 years ago. But because most of this legacy waste has successfully been emplaced in WIPP, the TRU waste clean-up focus is turning to newly-generated TRU materials. A major component will be transuranic SNM, currently managed in safeguards-protected vaults around the weapons complex. As DOE and NNSA continue to consolidate and shrink the weapons complex footprint, it is expected that significant quantities of transuranic SNM will be declared surplus to the nation's needs. Safeguards termination of SNM varies due to the wide range of attractiveness level of the potential material that may be directly discarded as waste. To enhance the efficiency of shipping waste with high TRU fissile content to WIPP, DOE designed an over-pack container, similar to the pipe component, called the criticality control over-pack, which will significantly enhance the efficiency of disposal. Hundreds of shipments of transuranic SNM, suitably packaged to meet WIPP waste acceptance criteria and with safeguards terminated have been successfully emplaced at WIPP (primarily from the Rocky Flats site clean-up) since WIPP opened. DOE expects that thousands more may eventually result from SNM consolidation efforts throughout the weapons complex. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

ATALANTE research facility-implementation of a rule of fractions for the management of reflecting materials in mass-limited units

ATALANTE, a research unit located in Marcoule, is one of the main Nuclear Facilities of the French CEA. As concerns criticality risk prevention, this facility is divided into work units mostly managed through a mass control mode. The authorized limit is 350 g of fissile material with {sup 239}Pu-H{sub 2}O as reference fissile medium. This mass limit was determined considering a reflection by 20 cm of water. Under these conditions, some neutronic reflectors which are more efficient than water can only be authorized in limited quantities. In ATALANTE the reflecting materials identified as requiring a specific form of management are lead, uranium ({sup 235}U/U{sub total} = 1%), graphite, heavy water and beryllium. Initially a maximum permissible mass was determined for each of these materials taken separately. However, this method requires that when different reflectors are present simultaneously, the sum of the masses of all these reflectors must be less than the limit specified for the most penalizing of them. This rule has proved to be too restrictive for the operators. A new rule has therefore been implemented: the rule of fractions. Criticality calculations were performed to demonstrate that this rule is acceptable. The geometric configurations studied involved a sphere of 350 g of {sup 239}Pu moderated by water and reflected by various masses of 2 to 6 reflectors (successions of concentric shells), followed by 20 cm of water. It was concluded that compliance with the new rule makes it possible to ensure criticality safety in the simultaneous presence of different reflectors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗