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Evaluation of Neutron Poison Materials for DOE SNF Disposal Systems

Description: Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors is being consolidated at the Savannah River Site (SRS) for ultimate disposal in the Mined Geologic Disposal System (MGDS). Most of the aluminum-based fuel material contains highly enriched uranium (HEU) (more than 20 percent 235U), which challenges the preclusion of criticality events for disposal periods exceeding 10,000 years. Recent criticality analyses have shown that the addition of neutron absorbing materials (poisons) is needed in waste packages containing DOE SNF canisters fully loaded with Al-SNF under flooded and degraded configurations to demonstrate compliance with the requirement that Keff less than 0.95. Compatibility of poison matrix materials and the Al-SNF, including their relative degradation rate and solubility, are important to maintain criticality control. An assessment of the viability of poison and matrix materials has been conducted, and an experimental corrosion program has been initiated to provide data on degradation rates of poison and matrix materials and Al-SNF materials under repository relevant vapor and aqueous environments. Initial testing includes Al6061, Type 316L stainless steel, and A516Gr55 in synthesized J-13 water vapor at 50 degrees C, 100 degrees C, and 200 degrees C and in condensate water vapor at 100 degrees C. Preliminary results are presented herein.
Date: September 1, 1998
Creator: Vinson, D.W.; Caskey, G.R. Jr. & Sindelar, R.L.
Partner: UNT Libraries Government Documents Department

Critical mass study of 231 process tanks

Description: An estimated minimum critical mass for each of the process vessels in the 231 Building has been calculated on the basis of critical mass data given in the P-11 Project Document HW-24514. The calculations are made assuming the plutonium to be a homogeneous mixture of precipitate and water with some slight neutron poisoning due to other elements. The precipitate is further assumed to have partially settled making an effectively infinite water reflector above the plutonium and hence reducing the critical mass.
Date: August 19, 1952
Creator: Lanning, D.D.
Partner: UNT Libraries Government Documents Department

Split-target neutronics and the MLNSC spallation target system

Description: The Manuel Lujan, Jr., Neutron Scattering Center (MLNSC) at the Los Alamos National Laboratory is one of four operating Short-Pulse Spallation Sources worldwide. The MLNSC target system (composed of targets, moderators, and reflectors) was first installed in 1985. The target system employs a split tungsten spallation target with a void space in between (the flux-trap gap); this target system will be upgraded in 1998. The ability to efficiently split a spallation target allowed us to introduce the concept of flux-trap moderators and ultimately the notion of backscattering and upstream moderators. The upgraded MLNSC target system will employ both flux-trap and upstream/backscattering moderators to simultaneously service 16 neutron flight paths with high-intensity neutron beams for materials science research.
Date: December 31, 1996
Creator: Russell, G.J.; Ferguson, P.D.; Pitcher, E.J. & Court, J.D.
Partner: UNT Libraries Government Documents Department

Criticality safety when using gadolinium as a neutronic poison with plutonium

Description: The authors have discovered that under special circumstances it is possible to have a criticality accident which can release tons of high explosive energy equivalent. We report a computer calculational, experiment which illustrates this point so that the circumstances which might lead to such an accident can be avoided. Our computer calculation gives the energy release in a criticality event involving a water solution of plutonium where (1) gadolinium has been used as a neutron poison and (2) the mass is then allowed to go slightly super-critical at room temperature. The reactivity has a strong positive temperature coefficient in this system. The result is that the self-heating from a small amount of fissioning of the plutonium in the delayed supercriticality regime raises the temperature to the prompt criticality regime and an explosive release of energy becomes possible. In this computer experiment the release is 19 tonnes (metric tons) of high explosive equivalent for a two-cubic meter sphere of water containing 50 grams of plutonium per liter. Our concern is confined to gadolinium in moderating solutions and does not relate to any other known neutronic poison, because gadolinium is the only isotope having the special absorption characteristics that make possible this type of energy release.
Date: September 1, 1995
Creator: Condit, R.H. & Pettibone, J.S.
Partner: UNT Libraries Government Documents Department

Criticality concerns in cleaning large uranium hexafluoride cylinders

Description: Cleaning large cylinders used to transport low-enriched uranium hexafluoride (UF{sub 6}) presents several challenges to nuclear criticality safety. This paper presents a brief overview of the cleaning process, the criticality controls typically employed and their bases. Potential shortfalls in implementing these controls are highlighted, and a simple example to illustrate the difficulties in complying with the Double Contingency Principle is discussed. Finally, a summary of recommended criticality controls for large cylinder cleaning operations is presented.
Date: June 1, 1995
Creator: Sheaffer, M.K.; Keeton, S.C. & Lutz, H.F.
Partner: UNT Libraries Government Documents Department

Critical mass studies of plutonium solutions

Description: The chain reacting conditions for plutonium nitrate in water solution have been examined experimentally for a variety of sizes of spheres and cylinders. The effects on the critical mass of the displacement of hydrogen and the addition of poisons to the fuel were measured in water tamped and bare reactors. In this report the data obtained in the investigation is presented graphically and in tables. Some preliminary analysis has been made yielding the results: (i) the absorption cross-section of Pu{sup 240} is 925 {plus_minus} 200 barns and (ii) the minimum critical mass of Pu{sup 239} in water is 510 grams at concentration of about 33 grams per liter.
Date: May 19, 1952
Creator: Kruesi, F.E.; Erkman, J.O. & Lanning, D.D.
Partner: UNT Libraries Government Documents Department

DISSOLUTION OF PLUTONIUM METAL IN 8-10 M NITRIC ACID

Description: The H-Canyon facility will be used to dissolve Pu metal for subsequent purification and conversion to plutonium dioxide (PuO{sub 2}) using Phase II of HB-Line. To support the new mission, the development of a Pu metal dissolution flowsheet which utilizes concentrated (8-10 M) nitric acid (HNO{sub 3}) solutions containing potassium fluoride (KF) is required. Dissolution of Pu metal in concentrated HNO{sub 3} is desired to eliminate the need to adjust the solution acidity prior to purification by anion exchange. The preferred flowsheet would use 8-10 M HNO{sub 3}, 0.015-0.07 M KF, and 0.5-1.0 g/L Gd to dissolve the Pu up to 6.75 g/L. An alternate flowsheet would use 8-10 M HNO{sub 3}, 0.1-0.2 M KF, and 1-2 g/L B to dissolve the Pu. The targeted average Pu metal dissolution rate is 20 mg/min-cm{sup 2}, which is sufficient to dissolve a 'standard' 2250-g Pu metal button in 24 h. Plutonium metal dissolution rate measurements showed that if Gd is used as the nuclear poison, the optimum dissolution conditions occur in 10 M HNO{sub 3}, 0.04-0.05 M KF, and 0.5-1.0 g/L Gd at 112 to 116 C (boiling). These conditions will result in an estimated Pu metal dissolution rate of {approx}11-15 mg/min-cm{sup 2} and will result in dissolution times of 36-48 h for standard buttons. The recommended minimum and maximum KF concentrations are 0.03 M and 0.07 M, respectively. The maximum KF concentration is dictated by a potential room-temperature Pu-Gd-F precipitation issue at low Pu concentrations. The purpose of the experimental work described in this report was two-fold. Initially a series of screening experiments was performed to measure the dissolution rate of Pu metal as functions of the HNO{sub 3}, KF, and Gd or B concentrations. The objective of the screening tests was to propose optimized conditions for subsequent flowsheet demonstration tests. ...
Date: February 21, 2012
Creator: Rudisill, Tracy S. & Pierce, R.
Partner: UNT Libraries Government Documents Department

Caustic Precipitation of Plutonium Using Gadolinium as the Neutron Poison for Disposition to High Level Waste

Description: Nuclear Materials Management Division (NMMD) has proposed that up to 100 kg of the plutonium (Pu) solutions stored in H-Canyon be precipitated with a nuclear poison and dispositioned to H-Area Tank Farm. The use of gadolinium (Gd) as the poison would greatly reduce the number of additional glass logs resulting from this disposition. This report summarizes the characteristics of the precipitation process and addresses criticality concerns in the Nuclear Criticality Safety Evaluation. No problems were found with the nature of the precipitate or the neutralization process.
Date: June 24, 2002
Creator: Bronikowski, M.G.
Partner: UNT Libraries Government Documents Department

Resource Conservation and Recovery Act (RCRA) closure sumamry for the Uranium Treatment Unit

Description: This closure summary has been prepared for the Uranium Treatment Unit (UTU) located at the Y-12 Plant in Oak Ridge, Tennessee. The actions required to achieve closure of the UTU area are outlined in the Closure Plan, submitted to and approved by the Tennessee Department of Environmental and Conservation staff, respectively. The UTU was used to store and treat waste materials that are regulated by the Resource Conservation and Recovery Act. This closure summary details all steps that were performed to close the UTU in accordance with the approved plan.
Date: May 1, 1996
Partner: UNT Libraries Government Documents Department

Metal Poisons in Waste Tanks (U)

Description: Many of the storage tanks with waste from processing fissile materials contain, along with the fissile material, metals which may serve as nuclear criticality poisons. It would be advantageous to the criticality evaluation of these wastes if it can be demonstrated that the poisons remain with the fissile materials and if an always safe poison-to-fissile ratio can be established. The first task, demonstrating that the materials stay together, is the job of the chemist, the second, demonstrating an always safe ratio, is the job of the physicist. The latter task is the object of this paper
Date: October 14, 1996
Creator: Williamson, T.G.
Partner: UNT Libraries Government Documents Department

Spatial Kinetics Calculations of MOX Fueled Core: Variant 22

Description: This work is part of a Joint US/Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, to compare in future point and spatial kinetics models (on the base of a set of selected accidents) in view of eventual creation of RELAP option with 3D kinetics. The document contains input data and results of model operation of three emergency dynamic processes in the VVER-1000 core: central control rod ejection by pressure drop caused by destroying of the moving mechanism cover; overcooling of the reactor core caused by steam line rupture and non-closure of steam generator stop valve; and the boron dilution of coolant in part of the VVER-1000 core caused by penetration of the distillate slug into the core at start up of non-working loop.
Date: January 11, 2001
Creator: Pavlovichev, A.M.
Partner: UNT Libraries Government Documents Department

Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./ Russian Progress Report for Fiscal Year 1997, Volume 4, Part 8 - Neutron Poison Plates in Assemblies Containing Homogeneous Mixtures of Polystyrene-Moderated Plutonium and Uranium Oxides

Description: In the 1970s at the Battelle Pacific Northwest Laboratory (PNL), a series of critical experiments using a remotely operated Split-Table Machine was performed with homogeneous mixtures of (Pu-U)O{sub 2}-polystyrene fuels in the form of square compacts having different heights. The experiments determined the critical geometric configurations of MOX fuel assemblies with and without neutron poison plates. With respect to PuO{sub 2} content and moderation [H/(Pu+U)atomic] ratio (MR), two different homogeneous (Pu-U) O{sub 2}-polystyrene mixtures were considered: Mixture (1) 14.62 wt% PuO{sub 2} with 30.6 MR, and Mixture (2) 30.3 wt% PuO{sub 2} with 2.8 MR. In all mixtures, the uranium was depleted to about O.151 wt% U{sup 235}. Assemblies contained copper, copper-cadmium or aluminum neutron poison plates having thicknesses up to {approximately}2.5 cm. This evaluation contains 22 experiments for Mixture 1, and 10 for Mixture 2 compacts. For Mixture 1, there are 10 configurations with copper plates, 6 with aluminum, and 5 with copper-cadmium. One experiment contained no poison plate. For Mixture 2 compacts, there are 3 configurations with copper, 3 with aluminum, and 3 with copper-cadmium poison plates. One experiment contained no poison plate.
Date: May 1, 1999
Creator: Yavuz, M.
Partner: UNT Libraries Government Documents Department

Criticality safety requirements for transporting EBR-II fuel bottles stored at INTEC

Description: Two carrier/shipping cask options are being developed to transport bottles of EBR-II fuel elements stored at INTEC. Some fuel bottles are intact, but some have developed leaks. Reactivity control requirements to maintain subcriticality during the hypothetical transport accident have been examined for both transport options for intact and leaking bottles. Poison rods, poison sleeves, and dummy filler bottles were considered; several possible poison materials and several possible dummy filler materials were studied. The minimum number of poison rods or dummy filler bottles has been determined for each carrier for transport of intact and leaking bottles.
Date: March 14, 2000
Creator: Lell, R. M. & Pope, C. L.
Partner: UNT Libraries Government Documents Department

Nuclear analysis of the chornobyl fuel containing masses with heterogeneous fuel distribution.

Description: Although significant data has been obtained on the condition and composition of the fuel containing masses (FCM) located in the concrete chambers under the Chernobyl Unit 4 reactor cavity, there is still uncertainty regarding the possible recriticality of this material. The high radiation levels make access extremely difficult, and most of the samples are from the FCM surface regions. There is little information on the interior regions of the FCM, and one cannot assume with confidence that the surface measurements are representative of the interior regions. Therefore, reasonable assumptions on the key parameters such as fuel concentration, the concentrations of impurities and neutron poisons (especially boron), the void fraction of the FCM due to its known porosity, and the degrees of fuel heterogeneity, are necessary to evaluate the possibility of recriticality. The void fraction is important since it introduces the possibility of water moderator being distributed throughout the FCM. Calculations indicate that the addition of 10 to 30 volume percent (v/o) water to the FCM has a significant impact on the calculated reactivity of the FCM. Therefore, water addition must be considered carefully. The other possible moderators are graphite and silicone dioxide. As discussed later in this paper, silicone dioxide moderation does not represent a criticality threat. For graphite, both heterogeneous fuel arrangements and very large volume fractions of graphite are necessary for a graphite moderated system to go critical. Based on the observations and measurements of the FCM compositions, these conditions do not appear creditable for the Chernobyl FCM. Therefore, the focus of the analysis reported in this paper will be on reasonable heterogeneous fuel arrangements and water moderation. The analysis will evaluate a range of fuel and diluent compositions.
Date: October 14, 1998
Creator: Turski, R. B.
Partner: UNT Libraries Government Documents Department