97 Matching Results

Search Results

Advanced search parameters have been applied.

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

Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals.

Description: This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel alloys that were irradiated at 288 C in helium to 0.4, 1.3, and 3.0 dpa. Fractographic analysis was conducted to determine the fracture surface morphology. Microchemical analysis by Auger electron spectroscopy was performed on BWR neutron absorber tubes to characterize grain-boundary segregation of important elements under BWR conditions. At 0.4 and 1.4 dpa, transgranular fracture was mixed with intergranular fracture. At 3 dpa, transgranular cracking was negligible, and fracture surface was either dominantly intergranular, as in field-cracked core internals, or dominantly ductile or mixed. This behavior indicates that percent intergranular stress corrosion cracking determined at {approx}3 dpa is a good measure of IASCC susceptibility. At {approx}1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At {approx}3 dpa, however, such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to {approx}3 dpa for eight heats of Types 304, 316, and 348 steel that contain very low concentrations of S. Susceptibility of Types 304 and 316 steels that contain >0.003 wt.% S increased drastically. This indicates that a sulfur related critical phenomenon plays an important role in IASCC. A sulfur content of <0.002 wt.% is the primary material factor necessary to ensure good resistance to IASCC. However, for Types 304L and 316L steel and their high-purity counterparts, a sulfur content of <0.002 wt.% alone is not a sufficient condition to ensure good resistance to IASCC. This is in distinct contrast to the behavior of their high-C ...
Date: January 31, 2006
Creator: Chung, H. M.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Fabrication of control rods for the High Flux Isotope Reactor

Description: The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.
Date: March 1, 1998
Creator: Sease, J.D.
Partner: UNT Libraries Government Documents Department

Waste Package Neutron Absorber, Thermal Shunt, and Fill Gas Selection Report

Description: Materials for neutron absorber, thermal shunt, and fill gas for use in the waste package were selected using a qualitative approach. For each component, selection criteria were identified; candidate materials were selected; and candidates were evaluated against these criteria. The neutron absorber materials evaluated were essentially boron-containing stainless steels. Two candidates were evaluated for the thermal shunt material. The fill gas candidates were common gases such as helium, argon, nitrogen, carbon dioxide, and dry air. Based on the performance of each candidate against the criteria, the following selections were made: Neutron absorber--Neutronit A978; Thermal shunt--Aluminum 6061 or 6063; and Fill gas--Helium.
Date: January 28, 2000
Creator: Pasupathi, V.
Partner: UNT Libraries Government Documents Department

Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

Description: This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor.
Date: March 1, 1997
Partner: UNT Libraries Government Documents Department

Critical experiments on an enriched uranium solution system containing periodically distributed strong thermal neutron absorbers

Description: A series of 62 critical and critical approach experiments were performed to evaluate a possible novel means of storing large volumes of fissile solution in a critically safe configuration. This study is intended to increase safety and economy through use of such a system in commercial plants which handle fissionable materials in liquid form. The fissile solution`s concentration may equal or slightly exceed the minimum-critical-volume concentration; and experiments were performed for high-enriched uranium solution. Results should be generally applicable in a wide variety of plant situations. The method is called the `Poisoned Tube Tank` because strong neutron absorbers (neutron poisons) are placed inside periodically spaced stainless steel tubes which separate absorber material from solution, keeping the former free of contamination. Eight absorbers are investigated. Both square and triangular pitched lattice patterns are studied. Ancillary topics which closely model typical plant situations are also reported. They include the effect of removing small bundles of absorbers as might occur during inspections in a production plant. Not taking the tank out of service for these inspections would be an economic advantage. Another ancillary topic studies the effect of the presence of a significant volume of unpoisoned solution close to the Poisoned Tube Tank on the critical height. A summary of the experimental findings is that boron compounds were excellent absorbers, as expected. This was true for granular materials such as Gerstley Borate and Borax; but it was also true for the flexible solid composed of boron carbide and rubber, even though only thin sheets were used. Experiments with small bundles of absorbers intentionally removed reveal that quite reasonable tanks could be constructed that would allow a few tubes at a time to be removed from the tank for inspection without removing the tank from production service.
Date: September 30, 1996
Creator: Rothe, R.E.
Partner: UNT Libraries Government Documents Department

Fluid dynamics, particulate segregation, chemical processes, and natural ore analog discussions that relate to the potential for criticality in Hanford tanks

Description: This report presents an in-depth review of the potential for nuclear criticality to occur in Hanford defense waste tanks during past, current and future safe storage and maintenance operations. The report also briefly discusses the potential impacts of proposed retrieval activities, although retrieval was not a main focus of scope. After thorough review of fluid dynamic aspects that focus on particle segregation, chemical aspects that focus on solubility and adsorption processes that might concentrate plutonium and/or separate plutonium from the neutron absorbers in the tank waste, and ore-body formation and mining operations, the interdisciplinary team has come to the conclusion that there is negligible risk of nuclear critically under existing storage conditions in Hanford site underground waste storage tanks. Further, for the accident scenarios considered an accidental criticality is incredible.
Date: September 27, 1996
Creator: Barney, G.S.
Partner: UNT Libraries Government Documents Department

Characterization strategy report for the criticality safety issue

Description: High-level radioactive waste from nuclear fuels processing is stored in underground waste storage tanks located in the tank farms on the Hanford Site. Waste in tank storage contains low concentrations of fissile isotopes, primarily U-235 and Pu-239. The composition and the distribution of the waste components within the storage environment is highly complex and not subject to easy investigation. An important safety concern is the preclusion of a self-sustaining neutron chain reaction, also known as a nuclear criticality. A thorough technical evaluation of processes, phenomena, and conditions is required to make sure that subcriticality will be ensured for both current and future tank operations. Subcriticality limits must be based on considerations of tank processes and take into account all chemical and geometrical phenomena that are occurring in the tanks. The important chemical and physical phenomena are those capable of influencing the mixing of fissile material and neutron absorbers such that the degree of subcriticality could be adversely impacted. This report describes a logical approach to resolving the criticality safety issues in the Hanford waste tanks. The approach uses a structured logic diagram (SLD) to identify the characterization needed to quantify risk. The scope of this section of the report is limited to those branches of logic needed to quantify the risk associated with a criticality event occurring. The process is linked to a conceptual model that depicts key modes of failure which are linked to the SLD. Data that are needed include adequate knowledge of the chemical and geometric form of the materials of interest. This information is used to determine how much energy the waste would release in the various domains of the tank, the toxicity of the region associated with a criticality event, and the probability of the initiating criticality event.
Date: June 1, 1997
Creator: Doherty, A.L.; Doctor, P.G.; Felmy, A.R.; Prichard, A.W. & Serne, R.J.
Partner: UNT Libraries Government Documents Department

Quantification of corrosion resistance of a new-class of criticality control materials: thermal-spray coatings of high-boron iron-based amorphous metals - Fe49.7Cr17.7Mn1.9Mo7.4W1.6B15.2C3.8Si2.4

Description: An iron-based amorphous metal, Fe{sub 49.7}Cr{sub 17.7}Mn{sub 1.9}Mo{sub 7.4}W{sub 1.6}B{sub 15.2}C{sub 3.8}Si{sub 2.4} (SAM2X5), with very good corrosion resistance was developed. This material was produced as a melt-spun ribbon, as well as gas atomized powder and a thermal-spray coating. Chromium (Cr), molybdenum (Mo) and tungsten (W) provided corrosion resistance, and boron (B) enabled glass formation. The high boron content of this particular amorphous metal made it an effective neutron absorber, and suitable for criticality control applications. Earlier studies have shown that ingots and melt-spun ribbons of these materials have good passive film stability in these environments. Thermal spray coatings of these materials have now been produced, and have undergone a variety of corrosion testing, including both atmospheric and long-term immersion testing. The modes and rates of corrosion have been determined in the various environments, and are reported here.
Date: March 28, 2007
Creator: Farmer, J C; Choi, J S; Shaw, C K; Rebak, R; Day, S D; Lian, T et al.
Partner: UNT Libraries Government Documents Department

SECOND WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: GENERATION AND EVALUATION OF INTERNAL CRITICIALITY CONFIGURATIONS

Description: This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having sonic or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment. The ultimate objective of this analysis is to augment the information gained from the Initial Waste Package Probabilistic Criticality Analyses (Ref. 5.8 and 5.9, hereafter referred to as IPA) to a degree which will support preliminary waste package design recommendations intended to reduce the risk of waste package criticality and the risk to total repository system performance posed by the consequences of any criticality. The IPA evaluated the criticality potential under the assumption that the waste package basket retained its structural integrity, so that the assemblies retained their initial separation, even when the neutron absorbers had been leached from the basket. This analysis is based on the more realistic condition that removal of the neutron absorbers is a consequence of the corrosion of the steel in which they are contained, which has the additional consequence of reducing the structural support between assemblies. The result is a set of more reactive configurations having a smaller spacing between assemblies, or no inter-assembly spacing at all. Another difference from the IPA is the minimal attention to probabilistic evaluation given in this study. Although the IPA covered a time horizon to 100,000 years, the lack of consideration of basket degradation modes made it primarily applicable to the first 10,000 years. In contrast, this study, by focusing on the ...
Date: March 27, 1996
Creator: P. Gottlieb, J.R. Massari, J.K. McCoy
Partner: UNT Libraries Government Documents Department

Criticality Expermints with Subcritical Clusters of 2.35 Wt% and 4.31 Wt% 2.35U Enriched UO2 Rods in Water at a Water-to-Fuel Volume Ratio of 1.6

Description: The fourth in a series of Nuclear Regulatory Commission funded criticality experiments have provided data for 2.35 wt% and 4.31 wt% {sup 235}U enriched U0{sub 2} rods at a water-to-fuel volume ratio of 1.6. The results from some 147 critical experiments are presented. They include for each enrichment: {sm_bullet}The critical size of single lattices or clusters of fuel {sm_bullet}The critical separation between sub-critical clusters of fuel {sm_bullet}The critical separation between sub-critical clusters of fuel having fixed neutron absorbers between the fuel clusters {sm_bullet}The isolation distance between fuel clusters {sm_bullet}The critical size of fuel clusters containing water holes and voids {sm_bullet}The critical size of fuel clusters separated by flux traps The fixed neutron absorbers for which data were obtained include 304-L steel, borated 304-L steel, copper, copper containing 1 wt% cadmium, cadmium, aluminium, zirconium and two trade name materials containing boron (Boral and Borofl ex).
Date: July 1, 1980
Creator: Bierman, SR & Clayton, ED
Partner: UNT Libraries Government Documents Department

Coupled moderator neutronics

Description: Optimizing the neutronic performance of a coupled-moderator system for a Long-Pulse Spallation Source is a new and challenging area for the spallation target-system designer. For optimal performance of a neutron source, it is essential to have good communication with instrument scientists to obtain proper design criteria and continued interaction with mechanical, thermal-hydraulic, and materials engineers to attain a practical design. A good comprehension of the basics of coupled-moderator neutronics will aid in the proper design of a target system for a Long-Pulse Spallation Source.
Date: December 1, 1995
Creator: Russell, G.J.; Pitcher, E.J. & Ferguson, P.D.
Partner: UNT Libraries Government Documents Department

Melt-Dilute Form of AI-Based Spent Nuclear Fuel Disposal Criticality Summary Report

Description: Criticality analysis of the proposed melt-dilute (MD) form of aluminum-based spent nuclear fuel (SNF), under geologic repository conditions, was performed [1] following the methodology documented in the Disposal Criticality Analysis Methodology Topical Report [2]. This methodology evaluates the potential for nuclear criticality for a waste form in a waste package. Criticality calculations show that even with waste package failure, followed by degradation of material within the waste package and potential loss of neutron absorber materials, sub-critical conditions can be readily demonstrated for the MD form of aluminum-based SNF.
Date: August 26, 2002
Creator: Vinson, D. & Serika, A.
Partner: UNT Libraries Government Documents Department

Neutronic moderator design for the Spallation Neutron Source (SNS)

Description: Neutronics analyses are now in progress to support the initial selection of moderator design parameters for the Spallation Neutron Source (SNS). The results of the initial optimization studies involving moderator poison plate location, moderator position, and premoderator performance for the target system are presented in this paper. Also presented is an initial study of the use of a composite moderator to produce a liquid methane like spectrum.
Date: November 1, 1998
Creator: Charlton, L.A.; Barnes, J.M.; Johnson, J.O. & Gabriel, T.A.
Partner: UNT Libraries Government Documents Department

Applying x-ray digital imaging to the verification of cadmium in fuel-storage components

Description: The High Flux Isotope Reactor utilizes large underwater fuel-storage arrays to stage irradiated fuel before it is shipped from the facility. Cadmium is required as a thermal neutron absorber in these fuel-storage arrays to produce an acceptable margin of nuclear subcriticality during both normal and off-normal operating conditions. Due to incomplete documentation from the time of their fabrication, the presence of cadmium within two stainless-steel parts of fuel-storage arrays must be experimentally verified before they are reused in new fuel-storage arrays. A cadmium-verification program has been developed in association with the Waste Examination and Assay Facility located at the Oak Ridge national Laboratory to nondestructively examine these older shroud assemblies. The program includes the following elements (1) x-ray analog imaging, (2) x-ray digital imaging, (3) prompt-gamma-ray spectroscopy measurements, and (4) neutron-transmission measurements. X-ray digital imaging utilizes an analog-to-digital convertor to record attenuated x-ray intensities observed on a fluorescent detector by a video camera. These x-ray intensities are utilized in expressions for cadmium thickness based upon x-ray attenuation theory.
Date: March 1997
Creator: Dabbs, R. D. & Cook, D. H.
Partner: UNT Libraries Government Documents Department

Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel

Description: Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For criticality safety purposes, current PWR storage and transportation canister designs employ flux traps between assemblies. Credit for fuel burnup will eliminate the need for these flux traps, and thus, significantly increase the PWR assembly capacity (for a fixed canister volume). Increases in assembly capacity of approximately one-third are expected. In contrast, current BWR canister designs do not require flux traps for criticality safety, and thus, are already at their maximum capacity in terms of physical storage. Therefore, benefits associated with burnup credit for BWR storage and transportation casks may be limited to increasing the enrichment capacity and/or decreasing the neutron ...
Date: October 12, 2000
Creator: Wagner, J.C.
Partner: UNT Libraries Government Documents Department

A users guide for the REBUS-PC code, version 1.4.

Description: The Reduced Enrichment Research and Test Reactor (RERTR) Program uses the REBUS-PC computer code to provide reactor physics and core design information such as neutron flux distributions in space, energy, and time, and to track isotopic changes in fuel and neutron absorbers with burnup. REBUS-PC has evolved away from the original REBUS code, which was created starting in the 1960's to study large liquid metal cooled fast breeder reactors. REBUS and REBUS-PC both model the external cycle, and are very general codes with 1D, 2D, and 3D neutronics capabilities, and with complete fuel shuffling capabilities. REBUS-PC has evolved to its present status over the past decade. While it incorporates the same neutronics capabilities from DIF3D 9.0 as does REBUS 9.0 created by the RAE Division of ANL, REBUS-PC has numerous changes and enhancements directed toward the needs of the thermal reactor analyst using WINDOWS or linux-based PC's.
Date: January 30, 2002
Creator: Olson, A. P.
Partner: UNT Libraries Government Documents Department

FY05 HPCRM Annual Report: High-Performance Corrosion-Resistant Iron-Based Amorphous Metal Coatings Evaluation of Corrosion Reistance FY05 HPCRM Annual Report # Rev. 1DOE-DARPA Co-Sponsored Advanced Materials Program

Description: New corrosion-resistant, iron-based amorphous metals have been identified from published data or developed through combinatorial synthesis, and tested to determine their relative corrosion resistance. Many of these materials can be applied as coatings with advanced thermal spray technology. Two compositions have corrosion resistance superior to wrought nickel-based Alloy C-22 (UNS No. N06022) in some very aggressive environments, including concentrated calcium-chloride brines at elevated temperature. Two Fe-based amorphous metal formulations have been found that appear to have corrosion resistance comparable to, or better than that of Ni-based Alloy C-22, based on breakdown potential and corrosion rate. Both Cr and Mo provide corrosion resistance, B enables glass formation, and Y lowers critical cooling rate (CCR). SAM1651 has yttrium added, and has a nominal critical cooling rate of only 80 Kelvin per second, while SAM2X7 (similar to SAM2X5) has no yttrium, and a relatively high critical cooling rate of 610 Kelvin per second. Both amorphous metal formulations have strengths and weaknesses. SAM1651 (yttrium added) has a low critical cooling rate (CCR), which enables it to be rendered as a completely amorphous thermal spray coating. Unfortunately, it is relatively difficult to atomize, with powders being irregular in shape. This causes the powder to be difficult to pneumatically convey during thermal spray deposition. Gas atomized SAM1651 powder has required cryogenic milling to eliminate irregularities that make flow difficult. SAM2X5 (no yttrium) has a high critical cooling rate, which has caused problems associated with devitrification. SAM2X5 can be gas atomized to produce spherical powders of SAM2X5, which enable more facile thermal spray deposition. The reference material, nickel-based Alloy C-22, is an outstanding corrosion-resistant engineering material. Even so, crevice corrosion has been observed with C-22 in hot sodium chloride environments without buffer or inhibitor. Comparable metallic alloys such as SAM2X5 and SAM1651 may also experience crevice corrosion ...
Date: September 19, 2007
Creator: Farmer, J. C.; Haslam, J. J. & Day, S. D.
Partner: UNT Libraries Government Documents Department

Characterization of Corrosion Products from Static Cell Testing of Al-SNF Forms

Description: Aluminum-based spent nuclear fuel from foreign and domestic research reactors is being consolidated at the Savannah River Site (SRS) for ultimate disposal in the Monitored Geologic Repository. The melt-dilute treatment technology has been developed to consolidate fuel assemblies by a melting/casting process in which depleted uranium is added to reduce enrichment below 20 percent 235-U. The melt-dilute product is essentially a binary uranium-aluminum alloy to which neutron absorber materials may be readily added.
Date: September 17, 2002
Creator: Adams, T.
Partner: UNT Libraries Government Documents Department

Preparation and Properties of Nitrate-Deficient Gadolinium Nitrate Solutions

Description: Because of the high neutron absorption cross sections of some gadolinium isotopes, gadolinium salts in solution are used to control nuclear reactivity in aqueous systems. The present studies concern the preparation and analysis of nitrate-deficient solutions, the effect of time and gamma radiation on their stability, and the determination of the solubility of gadolinium hydroxide in H2O and D2O.
Date: March 15, 2001
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Melt-Dilute Form of Al-Based Spent Nuclear Fuel Disposal Criticality Summary Report

Description: Criticality analysis of the proposed melt-dilute (MD) form of aluminum-based spent nuclear fuel (SNF), under geologic repository conditions, was performed following the methodology documented in the Disposal Criticality Analysis Methodology Topical Report. This methodology evaluates the potential for nuclear criticality for a waste form in a waste package. Criticality calculations show that even with waste package failure, followed by degradation of material within the waste package and potential loss of neutron absorber materials, sub-critical conditions can be readily demonstrated for the MD form of aluminum-based SNF.
Date: September 17, 2002
Creator: Vinson, D.W.
Partner: UNT Libraries Government Documents Department

Experimental critical parameters of enriched uranium solution in annular tank geometries

Description: A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant`s Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all.
Date: April 1, 1996
Creator: Rothe, R.E.
Partner: UNT Libraries Government Documents Department

Disposal criticality analysis for aluminum-based DOE fuels

Description: This paper describes the disposal criticality analysis for canisters containing aluminum-based Department of Energy fuels from research reactors. Different canisters were designed for disposal of highly enriched uranium (HEU) and medium enriched uranium (MEU) fuel. In addition to the standard criticality concerns in storage and transportation, such as flooding, the disposal criticality analysis must consider the degradation of the fuel and components within the waste package. Massachusetts Institute of Technology (MIT) U-Al fuel with 93.5% enriched uranium and Oak Ridge Research Reactor (ORR) U-Si-Al fuel with 21% enriched uranium are representative of the HEU and MEU fuel inventories, respectively. Conceptual canister designs with 64 MIT assemblies (16/layer, 4 layers) or 40 ORR assemblies (10/layer, 4 layers) were developed for these fuel types. Borated stainless steel plates were incorporated into a stainless steel internal basket structure within a 439 mm OD, 15 mm thick XM-19 canister shell. The Codisposal waste package contains 5 HLW canisters (represented by 5 Defense Waste Processing Facility canisters from the Savannah River Site) with the fuel canister placed in the center. It is concluded that without the presence of a fairly insoluble neutron absorber, the long-term action of infiltrating water can lead to a small, but significant, probability of criticality for both the HEU and MEU fuels. The use of 1.5kg of Gd distributed throughout the MIT fuel and the use of carbon steels for the structural basket or 1.1 kg of Gd distributed in the ORR fuel will reduce the probability of criticality to virtually zero for both fuels.
Date: November 1, 1997
Creator: Davis, J.W. & Gottlieb, P.
Partner: UNT Libraries Government Documents Department