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Evaluation of strontium-90 radial concentration profiles in Peach Bottom HTGR Core 2 fuel elements. HTGR base technology program, HTGR chemistry studies (189a 01329). [483 2000]

Description: Radial concentration profiles of /sup 90/Sr were evaluated for four fuel elements of the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) Core 2. A transport model utilizing a single effective diffusion coefficient was able to reproduce the strontium concentraion profile in graphite at temperatures above about 960/sup 0/C. Below this temperature, ''tails'' were observed in the experimental concentration profile and the profiles could not be completely reproduced. Diffusion coefficients determined from the study were slightly below those used in current reference HTGR design calculations. Experimental data exhibited a relatively high concentration at the graphite surfaces. Data and analyses are presented which indicate this may be due to a spatial variation in graphite properties resulting from an inhomogeneous distribution of the graphite impregnant. Unlike earlier Peach Bottom Core 1 studies the /sup 90/Sr precursor contribution to the total /sup 90/Sr concentration profile was determined to be negligible for these Core 2 data, apparently due to the much lower fuel partial coating failure fraction in Core 2. The calculated /sup 90/Sr released from the core is less than that measured, but the amount released is well below the amount established by design criteria.
Date: February 1, 1979
Creator: Haire, M.J.
Partner: UNT Libraries Government Documents Department

Computation of Normal Conducting and Superconducting Linear Accelerator (LINAC) Availabilities

Description: A brief study was conducted to roughly estimate the availability of a superconducting (SC) linear accelerator (LINAC) as compared to a normal conducting (NC) one. Potentially, SC radio frequency cavities have substantial reserve capability, which allows them to compensate for failed cavities, thus increasing the availability of the overall LINAC. In the initial SC design, there is a klystron and associated equipment (e.g., power supply) for every cavity of an SC LINAC. On the other hand, a single klystron may service eight cavities in the NC LINAC. This study modeled that portion of the Spallation Neutron Source LINAC (between 200 and 1,000 MeV) that is initially proposed for conversion from NC to SC technology. Equipment common to both designs was not evaluated. Tabular fault-tree calculations and computer-event-driven simulation (EDS) computer computations were performed. The estimated gain in availability when using the SC option ranges from 3 to 13% under certain equipment and conditions and spatial separation requirements. The availability of an NC LINAC is estimated to be 83%. Tabular fault-tree calculations and computer EDS modeling gave the same 83% answer to within one-tenth of a percent for the NC case. Tabular fault-tree calculations of the availability of the SC LINAC (where a klystron and associated equipment drive a single cavity) give 97%, whereas EDS computer calculations give 96%, a disagreement of only 1%. This result may be somewhat fortuitous because of limitations of tabular fault-tree calculations. For example, tabular fault-tree calculations can not handle spatial effects (separation distance between failures), equipment network configurations, and some failure combinations. EDS computer modeling of various equipment configurations were examined. When there is a klystron and associated equipment for every cavity and adjacent cavity, failure can be tolerated and the SC availability was estimated to be 96%. SC availability decreased as increased separation distance ...
Date: July 11, 2000
Creator: Haire, M.J.
Partner: UNT Libraries Government Documents Department

DUF6 Materials Use Roadmap

Description: The U.S. government has {approx}500,000 metric tons (MT) of surplus depleted uranium (DU) in various chemical forms stored at U.S. Department of Energy (DOE) sites across the United States. This DU, most of which is DU hexafluoride (DUF{sub 6}) resulting from uranium enrichment operations, is the largest amount of nuclear material in DOE's inventory. On July 6, 1999, DOE issued the ''Final Plan for the Conversion of Depleted Uranium Hexafluoride as required by Public Law 105-204'', in which DOE committed to develop a ''Depleted Uranium Hexafluoride Materials Use Roadmap'' in order to establish a strategy for the products resulting from conversion of DUF{sub 6} to a stable form. This report meets the commitment in the Final Plan by providing a comprehensive roadmap that DOE will use to guide any future research and development activities for the materials associated with its DUF{sub 6} inventory. The Roadmap supports the decision presented in the ''Record of Decision for Long-Term Management and Use of Depleted Uranium Hexafluoride'', namely to begin conversion of the DUF{sub 6} inventory as soon as possible, either to uranium oxide, uranium metal, or a combination of both, while allowing for future uses of as much of this inventory as possible. In particular, the Roadmap is intended to explore potential uses for the DUF{sub 6} conversion products and to identify areas where further development work is needed. It focuses on potential governmental uses of DUF{sub 6} conversion products but also incorporates limited analysis of using the products in the private sector. The Roadmap builds on the analyses summarized in the recent ''Programmatic Environmental Impact Statement for Alternative Strategies for the Long-Term Management and Use of Depleted Uranium Hexafluoride''. It also addresses other surplus DU, primarily in the form of DU trioxide and DU tetrafluoride. The DU-related inventory considered here includes the ...
Date: September 4, 2002
Creator: Haire, M.J.
Partner: UNT Libraries Government Documents Department

ALARA Review of the Spallation Neutron Source Accumulator Ring and Transfer Lines

Description: The Spallation Neutron Source (SNS) is designed to meet the growing need for new tools that will deepen our understanding in materials science, life science, chemistry, fundamental and nuclear physics, earth and environmental sciences, and engineering sciences. The SNS is an accelerator-based neutron-scattering facility that when operational will produce an average beam power of 2 MW at a repetition rate of 60 Hz. The accelerator complex consists of the front-end systems, which will include an ion source; a 1-GeV full-energy linear accelerator; a single accumulator ring and its transfer lines; and a liquid mercury target. This report documents an as-low-as-reasonably-achievable (ALARA) review of the accumulator ring and transfer lines at their early design stage. An ALARA working group was formed and conducted a review of the SNS ring and transfer lines at the {approx}25% complete design stage to help ensure that ALARA principles are being incorporated into the design. The radiological aspects of the SNS design criteria were reviewed against regulatory requirements and ALARA principles. Proposed features and measures were then reviewed against the SNS design criteria. As part of the overall review, the working group reviewed the design manual; design drawings and process and instrumentation diagrams; the environment, safety, and health manual; and other related reports and literature. The group also talked with SNS design engineers to obtain explanations of pertinent subject matter. The ALARA group found that ALARA principles are indeed being incorporated into the early design stage. Radiation fields have been characterized, and shielding calculations have been performed. Radiological issues are being adequately addressed with regard to equipment selection, access control, confinement structure and ventilation, and contamination control. Radiation monitoring instrumentation for worker and environment protection are also being considered--a good practice at this early design stage. The ring and transfer lines are being designed for hands-on ...
Date: June 30, 2003
Creator: Haire, M.J.
Partner: UNT Libraries Government Documents Department

Refinements to temperature calculations of spent fuel assemblies when in a stagnant gas environment

Description: Undesirably high temperatures are possible in irradiated fuel assemblies because of the radioactive decay of fission products formed while in the reactor. The COXPRO computer code has been used for some time to calculate temperatures in spent fuel when the fuel is suspended in a stagnant gas environment. This code assumed radiation to be the only mode of heat dissipation within the fuel pin bundle. Refinements have been made to include conduction as well as radiation heat transfer within this code. Comparison of calculated and measured temperatures in four separate and independent tests indicate that maximum fuel assembly temperatures can be predicted to within about 6%. 2 references, 5 figures.
Date: January 1, 1984
Creator: Rhodes, C.A. & Haire, M.J.
Partner: UNT Libraries Government Documents Department

Nuclear fuel cycle costs

Description: The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel cycle costs are given for the pressurized water reactor once-through and fuel recycle systems, and for the liquid-metal fast breeder reactor system. These calculations show that fuel cycle costs are a small part of the total power costs. For breeder reactors, fuel cycle costs are about half that of the present once-through system. The total power cost of the breeder reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment.
Date: February 1, 1982
Creator: Burch, W.D.; Haire, M.J. & Rainey, R.H.
Partner: UNT Libraries Government Documents Department

Status of the joint United States/Japanese criticality data development program for the Fast Breeder fuel cycle

Description: An experimental program is described to accumulate the basic criticality data on uranium-plutonium aqueous solution systems, including pin and water systems. This experimental program, managed by the Consolidated Fuel Reprocessing Program, utilized existing critical experiment facilities. However, Japan at least for the near future, does not have a facility for conducting critical experiments where fissible solution can be treated and processed. Therefore, Japan has remoted the existing US experimental program as well as enlarged the program so that a close cooperative relationship may exist in this area. This paper is an update of a description of the program presented in September 1985. 9 refs., 5 figs., 1 tab.
Date: January 1, 1987
Creator: Takeda, H.; Kishimoto, Y.; Matsumoto, T. & Haire, M.J.
Partner: UNT Libraries Government Documents Department

Analysis of fission product behavior in the Saclay Spitfire Loop Test SSL-1. [HTGR]

Description: The behavior of the fission metal cesium and the fission gases krypton and xenon in the Saclay Spitfire Loop SSL-1 test has been compared to that predicted using General Atomic reference data and computer code models. This is the first in a series of analyses planned in order to provide quantitative validation of HTGR fission product design methods. In this analysis, the first attempt to rigorously verify fission product design methods, the FIPERQ code was used to model the diffusion of cesium graphite and release to the coolant stream. The comparisons showed that the cesium profile shape in the graphite web and the partition coefficient between fuel rod matrix material and fuel element graphite were correctly modeled, although the overall release was significantly underpredicted. Uncertainties in the source term (fissile particle failure fraction) and total release to the coolant precluded an accurate appraisal of the validity of FIPERQ. However, several recommendations are presented to improve the applicability of future in-pile test data for the validation of fission metal release codes. The half-life dependence of fission gas release during irradiation was found to be in good agreement with the model used in the reference design materials, providing assurance that this aspect of the fission gas release predictions is properly modeled.
Date: February 1, 1978
Creator: Jensen, D.D.; Haire, M.J. & Ballagny, A.
Partner: UNT Libraries Government Documents Department

System availability top-down apportionment method

Description: A top-down method for apportioning overall manufacturing facility availability among systems and subsystems is presented. Characteristics which influence equipment reliability are defined. Experts, using engineering judgement, score each characteristic for each equipment system whose availability design goal is to be established. Scores for each characteristic are combined into weighting factors. A mathematical model is derived which incorporates these weighting factors. This methodology establishes tradeoffs among facility systems: the method imposes high-availability requirements on those systems in which an incremental increase in availability is easier to attain, and lower requirements on those in which an increase in availability is more difficult and costly. An example application of the method is presented.
Date: January 1, 1985
Creator: Haire, M.J.; Maltese, J.G. & Sohmer, R.G.
Partner: UNT Libraries Government Documents Department

Nuclear-fuel-cycle costs. Consolidated Fuel-Reprocessing Program

Description: The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel-cycle costs are given for the pressurized-water reactor once-through and fuel-recycle systems, and for the liquid-metal fast-breeder-reactor system. These calculations show that fuel-cycle costs are a small part of the total power costs. For breeder reactors, fuel-cycle costs are about half that of the present once-through system. The total power cost of the breeder-reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment.
Date: January 1, 1981
Creator: Burch, W.D.; Haire, M.J. & Rainey, R.H.
Partner: UNT Libraries Government Documents Department

Fuel-cycle costs for alternative fuels

Description: This paper compares the fuel cycle cost and fresh fuel requirements for a range of nuclear reactor systems including the present day LWR without fuel recycle, an LWR modified to obtain a higher fuel burnup, an LWR using recycle uranium and plutonium fuel, an LWR using a proliferation resistant /sup 233/U-Th cycle, a heavy water reactor, a couple of HTGRs, a GCFR, and several LMFBRs. These reactor systems were selected from a set of 26 developed for the NASAP study and represent a wide range of fuel cycle requirements.
Date: January 1, 1980
Creator: Rainey, R.H.; Burch, W.D.; Haire, M.J. & Unger, W.E.
Partner: UNT Libraries Government Documents Department

Depleted uranium oxides as spent-nuclear-fuel waste-package invert and backfill materials

Description: A new technology has been proposed in which depleted uranium, in the form of oxides or silicates, is placed around the outside of the spent nuclear fuel waste packages in the geological repository. This concept may (1) reduce the potential for repository nuclear criticality events and (2) reduce long-term release of radionuclides from the repository. As a new concept, there are significant uncertainties.
Date: July 7, 1997
Creator: Forsberg, C.W. & Haire, M.J.
Partner: UNT Libraries Government Documents Department

Nuclear criticality safety modeling of an LEU deposit

Description: The construction of the Oak Ridge Gaseous Diffusion Plant (now known as the K-25 Site) began during World War H and eventually consisted of five major process buildings: K-25, K-27, K-29, K-31, and K-33. The plant took natural (0.711% {sup 231}U) uranium as feed and processed it into both low-enriched uranium (LEU) and high-enriched uranium (HEU) with concentrations up to {approximately}93% {sup 231}U. The K-25 and K-27 buildings were shut down in 1964, but the rest of the plant produced LEU until 1985. During operation, inleakage of humid air into process piping and equipment caused reactions with gaseous uranium hexafluoride (UF{sub 6}) that produced nonvolatile uranyl fluoride (UO{sub 2}F{sub 2}) deposits. As part of shutdown, most of the uranium was evacuated as volatile UF{sub 6}. The UO{sub 2}F{sub 2} deposits remained. The U.S. Department of Energy has mitiated a program to unprove nuclear criticality safety by removing the larger enriched uranium deposits.
Date: November 1996
Creator: Haire, M. J.; Elam, K. R.; Jordan, W. C. & Dahl, T. L.
Partner: UNT Libraries Government Documents Department

Beneficial Uses of Depleted Uranium

Description: Naturally occurring uranium contains 0.71 wt% {sup 235}U. In order for the uranium to be useful in most fission reactors, it must be enriched the concentration of the fissile isotope {sup 235}U must be increased. Depleted uranium (DU) is a co-product of the processing of natural uranium to produce enriched uranium, and DU has a {sup 235}U concentration of less than 0.71 wt%. In the United States, essentially all of the DU inventory is in the chemical form of uranium hexafluoride (UF{sub 6}) and is stored in large cylinders above ground. If this co-product material were to be declared surplus, converted to a stable oxide form, and disposed, the costs are estimated to be several billion dollars. Only small amounts of DU have at this time been beneficially reused. The U.S. Department of Energy (DOE) has begun the Beneficial Uses of DU Project to identify large-scale uses of DU and encourage its reuse for the primary purpose of potentially reducing the cost and expediting the disposition of the DU inventory. This paper discusses the inventory of DU and its rate of increase; DU disposition options; beneficial use options; a preliminary cost analysis; and major technical, institutional, and regulatory issues to be resolved.
Date: August 1, 1997
Creator: Brown, C.; Croff, A.G. & Haire, M. J.
Partner: UNT Libraries Government Documents Department

Storage array reflection considerations

Description: The assumptions used for reflection conditions of single containers are fairly well established and consistently applied throughout the industry in nuclear criticality safety evaluations. Containers are usually considered to be either fully water-reflected (i.e. surrounded by 6 to 12 in. of water) for safety calculations or reflected by 1 in. of water for nominal (structural material and air) conditions. Tables and figures are usually available for performing comparative evaluations of containers under various loading conditions. Reflection considerations used for evaluating the safety of storage arrays of fissile material are not as well established.
Date: August 1, 1997
Creator: Haire, M.J.; Jordan, W.C. & Taylor, R.G.
Partner: UNT Libraries Government Documents Department

Depleted Uranium in Repositories

Description: For uranium to be useful in most fission nuclear reactors, it must be enriched (i.e. the concentration of the fissile isotope 235U must be increased). Therefore, depleted uranium (DU)-uranium which has less than naturally occurring concentrations of 235U-is a co-product of the enrichment process. Four to six tons of DU exist for every ton of fresh light water reactor fuel. There were 407,006 MgU 407,000 metric tons (t) of DU stored on U.S. Department of Energy (DOE) sites as of July 1993. If this DU were to be declared surplus, converted to a stable oxide form, and emplaced in a near surface disposal facility, the costs are estimated to be several billion dollars. However, the U.S. Nuclear Regulatory Commission has stated that near surface disposal of large quantities of DU tails is not appropriate. Thus, there is the possibility that disposition via disposal will be in a deep geological repository. One alternative that may significantly reduce the cost of DU disposition is to use it beneficially. In fact, DOE has begun the Beneficial Uses of DU Project to identify large scale uses of DU and to encourage its reuse. Several beneficial uses, many of which involve applications in the repository per se or in managing the wastes to go into the repository, are discussed in this report.
Date: December 31, 1997
Creator: Haire, M.J. & Croff, A.G.
Partner: UNT Libraries Government Documents Department

Nuclear criticality safety calculations for a K-25 site vacuum cleaner

Description: A modified Nilfisk model GSJ dry vacuum cleaner is used throughout the K-25 Site to collect dry forms of highly enriched uranium (HEU). When vacuuming, solids are collected in a cyclone-type separator vacuum cleaner body. Calculations were done with the SCALE (KENO V.a) computer code to establish conditions at which a nuclear criticality event might occur if the vacuum cleaner was filled with fissile solution. Conditions evaluated included full (12-in. water) reflection and nominal (1-in. water) reflection, and full (100%) and 20% {sup 235}U enrichment. Validation analyses of SCALE/KENO and the SCALE 27-group cross sections for nuclear criticality safety applications indicate that a calculated k{sub eff} + 2{sigma} < 0.9605 may be considered safely subcritical. Thus, a system with a calculated k{sub eff} + 2{sigma} {ge} 0.9605 is considered unsafe and may be critical. Critical conditions were calculated to be 70 g U/L for 100% {sup 235}U and full 12-in. water reflection. This corresponds to a minimum critical mass of approximately 1,400 g {sup 235}U for the approximate 20.0-L volume of the vacuum cleaner. The actual volume of the vacuum cleaner is smaller than the modeled volume because some internal materials of construction were assumed to be fissile solution. The model was an overestimate, for conservatism, of fissile solution occupancy. At nominal reflection conditions, the critical concentration in a vacuum cleaner full of UO{sub 2}F{sub 2} solution was calculated to be 100 g{sup 235}U/L, or 2,000 g mass of 100% {sup 235}U. At 20% {sup 235}U for the 20.0-L volume of the vacuum cleaner. At 15% {sup 235}U enrichment and full reflection, critical conditions were not reached at any possible concentration of uranium as a uranyl fluoride solution. At 17.5% {sup 235}U enrichment, criticality was reached at approximately 1,300 g U/L which is beyond saturation at 25 C.
Date: February 1, 1997
Creator: Shor, J. T. & Haire, M. J.
Partner: UNT Libraries Government Documents Department

Establishing Availability Requirements Using Characteristics Factors and Expert Opinion

Description: System design engineers must translate permitted overall facility downtime into detailed design and operating specifications for numerous systems and subsystems that make up the facility. The process of assigning reliability and maintainability requirements to individual equipment systems to attain a desired overall availability is known as availability apportionment. Apportionment is normally required early in conceptual design when little or no hardware information is available. Apportionment, when coupled with availability prediction, enables the selection of viable alternative configurations, identifies problem areas, and provides redirection of the program into more productive areas as necessary. A method for apportioning, or budgeting, overall facility availability requirements among systems and subsystems is presented. An example of applying this methodology to the Spallation Neutron Source (SNS) facility is given.
Date: June 18, 2000
Creator: Haire, M.J. & Schryver, J.C.
Partner: UNT Libraries Government Documents Department

Spallation Neutron Source Availability Top-Down Apportionment Using Characteristic Factors and Expert Opinion

Description: Apportionment is the assignment of top-level requirements to lower tier elements of the overall facility. A method for apportioning overall facility availability requirements among systems and subsystems is presented. Characteristics that influence equipment reliability and maintainability are discussed. Experts, using engineering judgment, scored each characteristic for each system whose availability design goal is to be established. The Analytic Hierarchy Process (AHP) method is used to produce a set of weighted rankings for each characteristic for each alternative system. A mathematical model is derived which incorporates these weighting factors. The method imposes higher availability requirements on those systems in which an incremental increase in availability is easier to achieve, and lower availability requirements where greater availability is more difficult and costly. An example is given of applying this top-down apportionment methodology to the Spallation Neutron Source (SNS) facility.
Date: October 1, 1999
Creator: Haire, M.J. & Schryver, J.C.
Partner: UNT Libraries Government Documents Department

Practical risk-based decision making: Good decisions made efficiently

Description: The Robotics and Process Systems Division of the Oak Ridge National Laboratory and the Westinghouse Savannah River Company have teamed with JBF Associates, Inc. to address risk-based robotic planning. The objective of the project is to provide systematic, risk-based relative comparisons of competing alternatives for solving clean-up problems at DOE facilities. This paper presents the methodology developed, describes the software developed to efficiently apply the methodology, and discusses the results of initial applications for DOE. The paper also addresses current work in applying the approach to problems in other industries (including an example from the hydrocarbon processing industry).
Date: December 1, 1995
Creator: Haire, M.J.; Guthrie, V.; Walker, D. & Singer, R.
Partner: UNT Libraries Government Documents Department

CHARACTERISTICS OF NEXT-GENERATION SPENT NUCLEAR FUEL (SNF) TRANSPORT AND STORAGE CASKS

Description: The design of spent nuclear fuel (SNF) casks used in the present SNF disposition systems has evolved from early concepts about the nuclear fuel cycle. The reality today is much different from that envisioned by early nuclear scientists. Most SNF is placed in pool storage, awaiting reprocessing (as in Russia) or disposal at a geologic SNF repository (as in the United States). Very little transport of SNF occurs. This paper examines the requirements for SNF casks from today's perspective and attempts to answer this question: What type of SNF cask would be produced if we were to start over and design SNF casks based on today's requirements? The characteristics for a next-generation SNF cask system are examined and are found to be essentially the same in Russia and the United States. It appears that the new depleted uranium dioxide (DUO2)-steel cermet material will enable these requirements to be met. Depleted uranium (DU) is uranium in which a portion of the 235U isotope has been removed during a uranium enrichment process. The DUO2-steel cermet material is described. The United States and Russia are cooperating toward the development of a next-generation, dual-purpose, storage and transport SNF system.
Date: October 3, 2004
Creator: Haire, M. J.; Forsberg, C. W.; Matveev, V. Z. & Shapovalov, V. I.
Partner: UNT Libraries Government Documents Department