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Overview of tritium fast-fission yields

Description: Tritium production rates are very important to the development of fast reactors because tritium may be produced at a greater rate in fast reactors than in light water reactors. This report focuses on tritium production and does not evaluate the transport and eventual release of the tritium in a fast reactor system. However, if an order-of-magnitude increase in fast fission yields for tritium is confirmed, fission will become the dominant production source of tritium in fast reactors.
Date: March 1, 1981
Creator: Tanner, J.E.
Partner: UNT Libraries Government Documents Department

Feasibility study on the verification of fresh fuel assemblies in shipping containers

Description: The purpose of this study was to examine the feasibility of using various nondestructive measurement techniques to determine the presence of fuel assemblies inside shipping containers and to examine the feasibility of measuring the fissile content of the containers. Passive and active techniques based on both gamma and neutron assay were examined. In addition, some experiments and calculations were performed to evaluate neutron techniques. Passive counting of the 186 keV gamma from {sup 235}U is recommended for use as an attributes measurement technique. Experiments and studies indicated that a bismuth germanate (BGO) scintillator is the preferred detector. A properly designed system based on this detector will provide a compact detector that can selectively verify fuel assemblies within a shipping container while the container is in a stack of similarly loaded containers. Missing fuel assemblies will be readily detected, but gamma counting of assemblies cannot detect changes in the fissile content of the inner rods in an assembly. If a variables technique is required, it is recommended that more extensive calculations be performed and removal of the outer shipping container be considered. Marking (sealing) of the assemblies with a uniquely identifiable transponder was also considered. This would require the development of procedures that would assure proper application and removal of the seal. When change to a metal outer container occurs, the technique will no longer be useful unless a radiolucent window is included in the container. 20 refs., 7 figs., 2 tabs.
Date: September 1, 1990
Creator: Swinth, K.L. & Tanner, J.E.
Partner: UNT Libraries Government Documents Department

Preliminary safety evaluation of a commercial-scale krypton-85 encapsulation facility

Description: This paper demonstrates that a commercial-scale facility for encapsulating krypton-85 in zeolite-5A or glass at a 2000 MTHM per year nuclear fuel reprocessing plant can be designed to contain fragments and the 340 to 850 kCi krypton-85 inventory from an assumed catastrophic failure of the high pressure vessel. The vessel failure was assumed as a worst case and was not based on a detailed design evaluation or operating experience. The process design is based on existing commercial hot isostatic pressing technology operated at up to 40 times the scale required for krypton encapsulation. From the calculated process gas inventory in the pressure vessel and vessel design, the maximum explosive energy of 8.4 kg TNT and resulting vessel plug and fragment velocities were calculated. The facility Containment Cell housing the high pressure vessel was designed to contain the gases, fragments, and the shock wave energy calculated for a hypothetical vessel failure. The Access Cell located directly above the Containment Cell was designed to be a tertiary confinement of krypton-85, should the access hatch be breached. 3 figures, 2 tables.
Date: January 1, 1980
Creator: Christensen, A.B.; Tanner, J.E. & Knecht, D.A.
Partner: UNT Libraries Government Documents Department

Basic criticality relations for gas core design

Description: Minimum critical fissile concentrations are calculated for U-233, U-235, Pu-239, and Am-242m mixed homogeneously with hydrogen at temperatures to 15,000K. Minimum critical masses of the same mixtures in a 1000 liter sphere are also calculated. It is shown that propellent efficiencies of a gas core fizzler engine using Am-242m as fuel would exceed those in a solid core engine as small as 1000L operating at 100 atmospheres pressure. The same would be true for Pu-239 and possibly U-233 at pressures of 1000 atm. or at larger volumes.
Date: May 22, 1992
Creator: Tanner, J. E.
Partner: UNT Libraries Government Documents Department

Improved gas core propulsion model

Description: A thermodynamic, radiation transport model of a gas core nuclear propulsion reactor has been developed in one-dimensional, spherical geometry, which satisfies local energy balance and allows for arbitrary variation of fuel/propellant ratio and flow rate as functions of radius. Initial cases calculated yield specific impulses of about 1150 sec, but very low thrusts ranging 5--10 kN.
Date: October 1, 1993
Creator: Tanner, J. E.
Partner: UNT Libraries Government Documents Department

Determination of the radioactive material and plutonium holdup in ducts and piping in the 325 Building

Description: This report describes the measurements performed to determine the radionuclide content and mass of Pu in exposed ducts, filters, and piping in the 325 Building at the Hanford Site. This information is needed to characterize facility radiation levels, to verify compliance with criticality safety specifications, and to allow more accurate nuclear material control using nondestructive assay. Gamma assay was used to determine the gamma-emitting isotopes in ducts, filters, and piping. Passive neutron counting was used to estimate the Pu content. A high-purity Ge detector and a neutron slab detector containing 5 {sup 3}He proportional counters were used. Almost all the gamma activity is from {sup 137}Cs and {sup 60}Co. Estimated Pu mass gram equivalents in the basement ductwork and filters are 31 g; the radioactive liquid waste system (RLWS) line has 12 g; the laboratory vacuum system has 2 g equiv. Pu; the retention process sewer has 3 g. Total Pu mass holdup for basement areas range from 48 to 27 g. Estimated Pu mass gram equivalents for all laboratories range from 385 to 581 g. Individual laboratory estimates are tabulated. Total estimated Pu gram equivalent holdup and material in process for the facility is 410 g. In summary, results indicate that no significant Pu levels, from a criticality safety perspective, reside in the ductwork, laboratory vacuum system lines, RLWS pipes, or any one laboratory in the 325 Building.
Date: August 1, 1996
Creator: Haggard, D.L.; Tanner, J.E. & Tomeraasen, P.L.
Partner: UNT Libraries Government Documents Department

Determination of curie content and {sup 134/137}cesium ratios by gamma spectroscopy of high burnup plutonium-aluminum fuel assemblies

Description: Nondestructive assay (NDA) gamma spectroscopy techniques were used to measure {sup 134/137}Cs ratios on nine PuAl Mark 42 fuel assemblies. The purpose of the ratio measurement was to confirm theoretical burnup calculations. {sup 134/137}Cs ratios were determined from the measured activity based on corrected net peak area counts for the 605 keV peak from {sup 134}Cs and the 662 keV peak from {sup 137}Cs/{sup 137m}Ba. Assembly No. 2 {sup 134/137}Cs ratio measured on 4-15-92 was 0.19. The measured {sup 134/137}Cs ratio was decay corrected to be 2.11 on 8-1-84 based on the half lives of {sup 134}Cs and {sup 137}Cs. The measured {sup 134/137}Cs ratio range was 1.90--2.14 for all nine assemblies. These measured values compare to a theoretical ratio of 1.7 on 8-1-84 determined by burnup calculations. Total cesium curie content was also requested and determined using the NDA direct measurements. Gamma spectral data were measured on the nine sectioned Mark 42 fuel assemblies. Measured cesium curie content, decay corrected to 8-1-84, ranged from 18170--24480 curies of {sup 134}Cs and 8620--11646 curies of {sup 137}Cs. Theoretical cesium curie content of 8-1-84 was 15200 curies {sup 134}Cs and 8973 curies {sup 137}Cs. Direct assay cesium ratio is 12% to 26% higher than the predicted ratio of 1.7. The measured {sup 134}Cs data indicate between 20%--61% more activity than that predicted by the burnup code, whereas the measured {sup 137}Cs activity is between 4% less to 30% more than the predicted activity. This information may be used to address issues concerning criticality safety, storage, and shipping of this type of material.
Date: June 1, 1997
Creator: Haggard, D. L. & Tanner, J. E.
Partner: UNT Libraries Government Documents Department

Determination of the radioactive material and plutonium holdup in ducts and piping in the 324 Building

Description: This report describes the measurements Performed to determine the radionuclide content and mass of plutonium in exposed ducts, filters, and piping in the 324 Building at the US Department of Energy Hanford Site in Washington State. This information is needed to characterize facility radiation levels, to verify compliance with criticality safety specifications, and to allow more accurate nuclear material control using nondestructive assay (NDA) methods. Gamma assay techniques typically employed for NDA analysis were used to determine the gamma-emitting isotopes in the ducts, filters, and piping. Passive neutron counting was selected to estimate -the plutonium content because high gamma levels from fission and activation products effectively mask any gamma emissions from plutonium. A high-purity gamma-ray detector Was used to measure the mixed fission and activation radionuclides. A neutron slab detector containing five {sup 3}He proportional counters was used to determine the neutron emission rates and estimate the mass of plutonium present. Both measurement systems followed the methods and procedures routinely used for nuclear waste assay and safeguards measurements.
Date: January 1, 1996
Creator: Haggard, D.L.; Brackenbush, L.W. & Tanner, J.E.
Partner: UNT Libraries Government Documents Department

Calculated and measured depth dose profiles in a phantom exposed to neutron radiation fields

Description: An accurate evaluation of doses caused by external sources of neutron radiation depends on knowledge of the transport of radiation inside the human body. Health physicists use two primary methods for studying this radiation transport: computer calculations and measurements. Both computer calculations and measurements were performed under well controlled, nearly identical conditions to determine the extent of their agreement. A comparison of the dose profiles predicted by both measurements and calculations was thus possible. The measurements were performed in a cylindrical phantom made of tissue equivalent plastic. The phantom size, 61 cm high and 30 cm in diameter, was chosen to approximate the human torso and to match the dimensions of cylindrical phantoms used by previous calculations. Holes were drilled down through the phantom to accommodate small tissue equivalent proportional counters (TEPCs) at various depths in the phantom. These counters were used to measure the neutron dose inside the phantom when it was exposed to various sources of neutrons. The holes in the phantom could also accommodate miniature Geiger-Mueller detectors to measure the gamma component of the dose. Neutron and gamma dose profiles were measured for two different sources of neutrons: an unmoderated /sup 252/Cf source and a 733-keV neutron beam generated by a Van de Graaff accelerator. 14 refs., 13 figs., 11 tabs.
Date: May 1, 1989
Creator: Scherpelz, R.I.; Tanner, J.E.; Sigalla, L.A. & Hadlock, D.E.
Partner: UNT Libraries Government Documents Department

Verification of an effective dose equivalent model for neutrons

Description: Since the effective dose equivalent, based on the weighted sum of organ dose equivalents, is not a directly measurable quantity, it must be estimated with the assistance of computer modeling techniques and a knowledge of the radiation field. Although extreme accuracy is not necessary for radiation protection purposes, a few well-chosen measurements are required to confirm the theoretical models. Neutron measurements were performed in a RANDO phantom using thermoluminescent dosemeters, track etch dosemeters, and a 1/2-in. (1.27-cm) tissue equivalent proportional counter in order to estimate neutron doses and dose equivalents within the phantom at specific locations. The phantom was exposed to bare and D{sub 2}O-moderated {sup 252}Cf neutrons at the Pacific Northwest Laboratory's Low Scatter Facility. The Monte Carlo code MCNP with the MIRD-V mathematical phantom was used to model the human body and calculate organ doses and dose equivalents. The experimental methods are described and the results of the measurements are compared to the calculations. 8 refs., 3 figs., 3 tabs.
Date: October 1, 1991
Creator: Tanner, J.E.; Piper, R.K.; Leonowich, J.A. & Faust, L.G.
Partner: UNT Libraries Government Documents Department

Loading and leakage of krypton immobilized in zeolites and glass

Description: Krypton-85 is formed in nuclear power reactors and remains trapped until the fuel is reprocessed. Federal regulations limit the release of /sup 85/Kr to the environment, requiring recovery and storage of 85% of the /sup 85/Kr produced in commercial light-water reactors after January 1, 1983. One of the long-term storage options involves encapsulating /sup 85/Kr in zeolites or glasses at high pressure and temperature. This paper presents experimental results for krypton encapsulation and leakage in sodalite, zeolite 5A, and Vycor Thristy glass. The results show that all three materials are feasible for /sup 85/Kr immobilization and long-term storage, although zeolite 5A and Thirsty Vycor are preferable due to lower leakage rates.
Date: January 1, 1980
Creator: Christensen, A.B.; Del Debbio, J.A.; Knecht, D.A. & Tanner, J.E.
Partner: UNT Libraries Government Documents Department

Krypton-85 health risk assessment for a nuclear fuel reprocessing plant

Description: The risks involved in the routine release of /sup 85/Kr from nuclear fuel reprocessing operations to the environment were compared to those resulting from the capture and storage of /sup 85/Kr. Instead of releasing the /sup 85/Kr to the environment when fuel is reprocessed, it can be captured, immobilized and stored. Two alternative methods of capturing /sup 85/Kr (cryogenic distillation and fluorocarbon absorption) and one method of immobilizing the captured gas (ion implantation/sputtering) were theoretically incorporated into a representative fuel reprocessing plant, the Barnwell Nuclear Fuel Plant, even though there are no known plans to start up this facility. Given the uncertainties in the models used to generate lifetime risk numbers (0.02 to 0.027 radiation induced fatal cancers expected in the occupational workforce and 0.017 fatal cancers in the general population), the differences in total risks for the three situations, (i.e., no-capture and two-capture alternatives) cannot be considered meaningful. It is possible that no risks would occur from any of the three situations. There is certainly no reason to conclude that risks from /sup 85/Kr routinely released to the environment are greater than those that would result from the other two situations considered. Present regulations mandate recovery and disposal of /sup 85/Kr from the off gases of a facility reprocessing spent fuel from commercial sources. Because of the lack of a clear-cut indication that recovery woud be beneficial, it does not seem prudent to burden the facilities with a requirement for /sup 85/Kr recovery, at least until operating experience demonstrates the incentive. The probable high aging of the early fuel to be processed and the higher dose resulting from the release of the unregulated /sup 3/H and /sup 14/C also encourage delaying implementation of the /sup 85/Kr recovery in the early plants.
Date: August 1, 1984
Creator: Mellinger, P.J.; Brackenbush, L.W.; Tanner, J.E. & Gilbert, E.S.
Partner: UNT Libraries Government Documents Department

Fluence-to-dose equivalent conversion factors for polyethylene-moderated {sup 252}Cf

Description: Neutron measurements and calculations were conducted to characterize the polyethylene-moderated {sup 252}Cf source at Oak Ridge National Laboratory`s Radiation Calibration Laboratory (RADCAL). The 12-inch-diameter polyethylene sphere produces a highly scattered neutron spectrum which is more representative of most radiation fields found in the workplace than the D{sub 2}O-moderated {sup 252}Cf neutron spectrum typically used for dosimeter calibration. However, the energy-dependent fluence and dose equivalent must be well known before using such a source for radiation protection purposes. The measurements and calculations were performed as independent checks of the desired quantities which were the flux, the absorbed dose rate, the dose equivalent rate, and the average energy. These quantities were determined for the polyethylene sphere with and without an outer cadmium shell and compared with a D{sub 2}O-moderated {sup 252}Cf source.
Date: April 1, 1994
Creator: Tanner, J. E.; Soldat, K. L.; Stewart, R. D. & Casson, W. H.
Partner: UNT Libraries Government Documents Department

BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

Description: This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.
Date: February 1, 1986
Creator: McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M. & King, C.E.
Partner: UNT Libraries Government Documents Department

US Army facility for the consolidation of low-level radioactive waste

Description: A preliminary study of a waste consolidation facility for the Department of the Army's low-level radioactive waste was carried out to determine a possible site and perform a cost-benefit analysis. Four sites were assessed as possible locations for such a facility, using predetermined site selection criteria. To assist in the selection of a site, an evaluation of environmental issues was included as part of each site review. In addition, a preliminary design for a waste consolidation facility was developed, and facilities at each site were reviewed for their availability and suitability for this purpose. Currently available processes for volume reduction, as well as processes still under development, were then investigated, and the support and handling equipment and the staff needed for the safe operation of a waste consolidation facility were studied. Using current costs for the transportation and burial of low-level waste, a cost comparison was then made between waste disposal with and without the utilization of volume reduction. Finally, regulations that could affect the operation of a waste consolidation facility were identified and their impact was assessed. 11 references, 5 figures, 16 tables.
Date: December 1, 1983
Creator: Stein, S.L.; Tanner, J.E.; Murphy, B.L.; Gillings, J.C.; Hadley, R.T.; Lyso, O.M. et al.
Partner: UNT Libraries Government Documents Department

An example postclosure risk assessment using the potential Yucca Mountain Site

Description: The risk analysis described in this document was performed for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM) over a 2-year time period ending in June 1988. The objective of Pacific Northwest Laboratory`s (PNL) task was to demonstrate an integrated, though preliminary, modeling approach for estimating the postclosure risk associated with a geologic repository for the disposal of high-level nuclear waste. The modeling study used published characterization data for the proposed candidate site at Yucca Mountain, Nevada, along with existing models and computer codes available at that time. Some of the site data and conceptual models reported in the Site Characterization Plan published in December 1988, however, were not yet available at the time that PNL conducted the modeling studies.
Date: May 1, 1992
Creator: Doctor, P.G.; Eslinger, P.W.; Elwood, D.M.; Engel, D.W.; Freshley, M.D.; Liebetrau, A.M. et al.
Partner: UNT Libraries Government Documents Department

Subcritical measurements of the WINCO slab tank experiment using the source-jerk technique

Description: Subcritical measurements of the WINCO slab tank using the source-jerk technique are presented. This technique determines subcriticality by analyzing the transient response produced by the sudden removal of an extraneous neutron source (i.e., a source jerk). We have found that the technique can provide an accurate means of measuring k in configurations that are close to critical (i.e., 0.90 < k < 1.0). As the system becomes more subcritical (i.e., k < 0.90), spatial effects introduce significant biases depending on the source and detector positions. A comparison between the measurements and Monte Carlo code calculations is also presented. 15 refs., 6 figs., 2 tabs.
Date: January 1, 1989
Creator: Spriggs, G.D.; Hansen, G.E.; Martin, E.R.; Plassmann, E.A.; Pederson, R.A.; Schlesser, J.A. et al.
Partner: UNT Libraries Government Documents Department