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Startup data report for NRC/PNL Halden Assembly IFA-513

Description: This report presents data from the first month of operation of IFA-513, which is a heavily instrumented 6-rod test assembly in the Halden Reactor in Norway. The assembly is jointly sponsored by the Halden Project and the Nuclear Regulatory Commission (NRC), and is part of a series of irradiation tests sponsored by the NRC to verify its single-rod fuel modeling computer programs. All the rods in the series are of the basic BWR-6 design with variations in gap size, fuel type, fill gas composition, and fill gas pressure. The first two tests in the series were IFA-431 and IFA-432. These were identical 6-rod assemblies, each containing the same variations of gap size and fuel pellet types, but operating at different power levels and burnups. The present assembly, IFA-513, is the third in the series; its 6 rods are all identical, except for variations in fill gas composition and pressure. The fourth and last assembly, designated IFA-527, is yet to be built, and will study the effects of fuel pellet cracking and relocation. The measurements made in IFA-513 and the earlier tests include: (1) fuel temperature and power (both steady-state and transient), (2) total cladding elongation, and (3) fill gas pressure. The measurements were made on a continuous basis, providing a record of their variation with both power and burnup. Along with the data, this report includes some analysis to put the IFA-513 startup data in perspective to similar data from IFA-431 and IFA-432.
Date: July 1, 1979
Creator: Lanning, D.D. & Cunningham, M.E.
Partner: UNT Libraries Government Documents Department

End-of-irradiation data report for the instrumented fuel assembly (IFA)-527. [PWR; BWR]

Description: This report presents data obtained during the irradiation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of US Nuclear Regulatory Commission (NRC)-sponsored tests to obtain data for the development and verification of steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 230-..mu..m diametral gaps and one rod with similar fuel pellets but with a 60-..mu..m diametral gap. All six rods were xenon-filled to simulate the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. This report presents both pre- and postfailure data for IFA-527.
Date: May 1, 1982
Creator: Cunningham, M.E. & Lanning, D.D.
Partner: UNT Libraries Government Documents Department

Interpretation of fuel centerline thermocouple response to reactor scrams

Description: This report compares and contrasts fuel thermocouple scram responses from the low-burnup assemblies IFA-513 and IFA-505 and the high-burnup assembly IFA-432. Even on a qualitative basis it is observed that the IFA-432 thermocouple responses are sluggish relative to that of low-burnup counterparts in rods of equivalent thermal resistance and design/enrichment. By numerical analysis, it is concluded that this apparent sluggishness is due to thermocouple decalibration, and an estimate of the decalibration is made.
Date: January 1, 1980
Creator: Lanning, D.D. & Cunningham, M.E.
Partner: UNT Libraries Government Documents Department

Storage of LWR spent fuel in air. Volume 3, Results from exposure of spent fuel to fluorine-contaminated air

Description: The Behavior of Spent Fuel in Storage (BSFS) Project has conducted research to develop data on spent nuclear fuel (irradiated U0{sub 2}) that could be used to support design, licensing, and operation of dry storage installations. Test Series B conducted by the BSFS Project was designed as a long-term study of the oxidation of spent fuel exposed to air. It was discovered after the exposures were completed in September 1990 that the test specimens had been exposed to an atmosphere of bottled air contaminated with an unknown quantity of fluorine. This exposure resulted in the test specimens reacting with both the oxygen and the fluorine in the oven atmospheres. The apparent source of the fluorine was gamma radiation-induced chemical decomposition of the fluoro-elastomer gaskets used to seal the oven doors. This chemical decomposition apparently released hydrofluoric acid (HF) vapor into the oven atmospheres. Because the Test Series B specimens were exposed to a fluorine-contaminated oven atmosphere and reacted with the fluorine, it is recommended that the Test Series B data not be used to develop time-temperature limits for exposure of spent nuclear fuel to air. This report has been prepared to document Test Series B and present the collected data and observations.
Date: June 1, 1995
Creator: Cunningham, M.E. & Thomas, L.E.
Partner: UNT Libraries Government Documents Department

Vacuum Brazing of Beryllium Copper Components for the National Ignition Facility

Description: A process for vacuum brazing beryllium copper anode assemblies was required for the Plasma Electrode Pockels Cell System, or PEPC, a component for the National Ignition Facility (NIF). Initial problems with the joint design and wettability of the beryllium copper drove some minor design changes. Brazing was facilitated by plating the joint surface of the beryllium copper rod with silver 0.0006 inch thick. Individual air sampling during processing and swipe tests of the furnace interior after brazing revealed no traceable levels of beryllium.
Date: June 4, 2002
Creator: Tyhurst, C.C. & Cunningham, M.A.
Partner: UNT Libraries Government Documents Department

Effects of fill gas composition and pellet eccentricity: comparison between instrumented fuel assemblies IFA-431 and IFA-432

Description: Two 6-rod, instrumented test assemblies, designated IFA-431 and IFA-432, were designed to study the effects of fabricated fuel - cladding gap size, fuel density and stability, fill gas composition, linear heat rate, and burnup. Rod 4 of each assembly was designed to study the effects of fuel - cladding geometry and minimum gas thermal conductivity upon heat transfer across the gap between the fuel and cladding. To accomplish this, mechanical restraint was used to form a concentric fuel - cladding geometry in the upper end of the rod, and an eccentric geometry in the lower end of the rod; the rod was then backfilled with xenon gas. The eccentric fuel region of the xenon rod was observed to have lower fuel temperatures, for equal power, than the concentric fuel region. It is concluded that fuel - cladding eccentricity enhances the azimuthal average gap conductance, thus reducing the fuel centerline temperature. The xenon-filled rods were compared to rods backfilled with helium. As expected, the xenon rods had higher fuel temperatures than the helium-filled rods in the assemblies, although not as high as was initially predicted. It is concluded that the reduced temperatures (relative to the predicted temperatures) were the result of a greater decrease in fuel - cladding gap size than is predicted to occur from thermal expansion alone. The ratio of the gap conductances for a xenon rod and a helium rod with equal operating thermal radial gaps, is equal to the ratio of the gas thermal conductivities for the two rods.
Date: April 1, 1979
Creator: Cunningham, M.E.; Williford, R.E. & Hann, C.R.
Partner: UNT Libraries Government Documents Department

Final data report for the instrumented fuel assembly (IFA)-432

Description: This report presents the in-reactor data collected during the irradiation of the six-rod instrumented fuel assembly (IFA)-432 in the Halden Boiling Water Reactor (HBWR) from June 1980 through June 1981. This Pacific Northwest Laboratory (PNL)-designed assembly was one of a series of US Nuclear Regulatory Commission (NRC)-sponsored tests to obtain data for the development and verification of steady-state fuel performance computer codes. IFA-432 operated from December 1975 until June 1981, when it was removed from the reactor. Two of the rods were removed for examination, and the assembly was reinserted in December 1981 to obtain additional data. Fuel centerline temperatures, cladding elongations, internal fuel rod pressures, and local powers at thermocouple positions were monitored during the irradiation of IFA-432; and the resulting data are presented in this report.
Date: June 1, 1982
Creator: Bradley, E.R.; Cunningham, M.E. & Lanning, D.D.
Partner: UNT Libraries Government Documents Department

Resource Management plan for the Oak Ridge Reservation. Volume 28, Wetlands on the Oak Ridge Reservation

Description: A survey of wetlands on the Oak Ridge Reservation (ORR) was conducted in 1990. Wetlands occurring on ORR were identified using National Wetlands Inventory (NWI) maps and field surveys. More than 120 sites were visited and 90 wetlands were identified. Wetland types on ORR included emergent communities in shallow embayments on reservoirs, emergent and aquatic communities in ponds, forested wetland on low ground along major creeks, and wet meadows and marshes associated with streams and seeps. Vascular plant species occurring on sites visited were inventoried, and 57 species were added to the checklist of vascular plants on ORR. Three species listed as rare in Tennessee were discovered on ORR during the wetlands survey. The survey provided an intensive ground truth of the wetlands identified by NWI and offered an indication of wetlands that the NWI remote sensing techniques did not detect.
Date: December 1, 1991
Creator: Cunningham, M. & Pounds, L.
Partner: UNT Libraries Government Documents Department

Application of Linear Propagation of Errors to Fuel Rod Temperature and Stored Energy Calculations

Description: Linear propagatlon of errors evaluates modeling uncertainty by approximating a function of interest by first-order Taylor's series expansions and then approximating the variance of the function by the variance of the linear approximation. This report discusses uncertainty analysis for different nuclear fuel rod designs, the process of model validation, and the effect of cracked pellet fuel models upon temperabre uncertainty. Using a postulated power history, the uncertainty for the predicted thermal response of boiling water reactor (BWR) and pressurized water reactor (PWR} fuel rods was evaluated. Beginning-of-life (BOL) relative uncertainty for BWR and PWR fuel rods is approximately the same. while different end-of-fife {EOL} thermal response results in different EOL uncertainty. Determining the validity of modeling relative to reality is discussed in qualitative terms. Validity is dependent upon verifying that the code correctly implements the model and that satisfactory agreement is found between the model and measurements. Fuel modeling codes are now using cracked pellet fuel models, which result in decreased fuel surface temperature. Estimated stored energy is lowered; but its relative uncertainty is increased. In general, however, the absolute upper uncertainty bound for stored energy is lower for a cracked pellet model than for a solid pellet model.
Date: October 1, 1980
Creator: Cunningham, M. E.; Olsen, A. R.; Lanning, D. D. & Willford, R. E.
Partner: UNT Libraries Government Documents Department

Data Report for the NRC/PNL Halden Assembly IFA-432: April 1978-May 1980

Description: This report presents the in-reactor data collected from the U.S. Nuclear Regulatory Commission (NRC)/Pacific Northwest Laboratory (PNL) Halden test assembly IFA-432 for the period from April 1978 through May 1980. The irradiation test is part of an experimental program entitled 11 Experimental Support and Development of Single-Rod Fuel Codes" sponsored by the Fuel Behavior Research Branch of the NRC. The purpose of this program is to reduce the uncertainties of predicting the thermal and mechanical behavior of an operating nuclear fuel rod, Fuel centerline temperatures, cladding elongation, internal fuel rod pressures, and local powers at the thermocouple (TC) positions are shown as a function of time. The local powers were derived from neutron detector readings while the other variables were measured directly. Detailed analysis of the data is not made, but topical reports discussing certain aspects of the data are referenced. Descriptions of the assembly, instrumentation and calibration, and data processing methods are also presented.
Date: April 1, 1981
Creator: Bradley, E. R.; Cunningham, M. E.; Lanning, D. D. & Williford, R. E.
Partner: UNT Libraries Government Documents Department

Data Report for the Instrumented Fuel Assembly IFA-513

Description: This report presents the in-reactor data collected to date from the NRC/PNL Halden test assembly IFA-513. The irradiation test is part of an experimental program entitled 11 Experimental Support and Development of Single Rod Fuel Codes 11 sponsored by the Fuel Behavior Branch of the U.S. Nuclear Regulatory Commission (NRC). The purpose of this program is to reduce the uncertainties of predicting the thermal and mechanical behavior of an operating nuclear fuel rod. Fuel centerline temperatures, cladding elongation, internal fuel rod pressures, and local powers at the thermocouple positions are shown as a function of time for the irradiation period from November 1978 to January 1980. The local powers were derived from neutron detector readings while the other variables were measured directly. The general trends in the data as a function of burnup are presented and discussed. Detailed analysis of the data is not made, but topical reports discussing certain aspects of the data are referenced. Descriptions of the assembly, instrumentation and calibration, and data processing methods are also presented.
Date: August 1, 1981
Creator: Bradley, E. R.; Cunningham, M. E.; Lanning, D. D. & Williford, R. E.
Partner: UNT Libraries Government Documents Department

Observation of Porosity Reduction in a Densification-Prone Test Fuel Rod: Data and Analysis

Description: Instrumented fuel assembly (IFA)-431 was irradiated in the Halden Boiling Water Reactor (HBWR) for the purpose of extending the steady-state data base. Rod 6 of this assembly began irradiation with UO{sub 2} fuel of 92% theoretical density (TD) that was unstable with respect to in-reactor densification. Thermal resintering tests resulted in a final density of 95.3% TD while post-irradiation examination (PIE) indicated a final density of 96.5% TD. Observed microstructural changes were consistent with published densification studies; there was a marked depletion of submicrometer diameter pores and total pore volume. However, grain size increased only slightly, indicating that internal pellet temperatures did not reach the 1875K applied in resintering tests. Oensification was observed to increase the temperatures in rod 6, but temperatures did not become as high as for a sibling rod that simulated instantaneous densification. Temperatures calculated with U.S. Nuclear Regulatory Commission (NRC) fuel performance computer codes were generally higher than observed temperatures.
Date: October 1, 1981
Creator: Cunningham, M. E.; Daniel, J. L. & Lanning, D. D.
Partner: UNT Libraries Government Documents Department

Fuel Performance Annual Report for 1979

Description: This annual report, the second in a series, provides a brief description of fuel performance in commercial nuclear power plants. Brief summaries are given of fuel surveillance programs, fuel performance problems, and fuel design changes. References to additional, more detailed, information and related NRC evaluation are provided.
Date: January 1, 1981
Creator: Tokar, M.; Mailey, W. J. & Cunningham, M. E.
Partner: UNT Libraries Government Documents Department

PSA in America

Description: Although the concept of acceptable risk has always been the foundation of the nuclear industry design, the use of formal PSA (or PRA-probabilistic risk assessment) in the U.S. nuclear power industry has followed an unusual path in arriving at its current level of notability. Prior to 1975, probabilistic evaluations were limited to a few specific applications such as the evaluation of man-made (i.e., airplane crashes) and natural (i.e., earthquakes) hazards. In 1975, the industry was introduced to comprehensive PSA by the Reactor Safety Study (WASH-1400). However, the study languished in relative obscurity until the accident at Three Mile Island 2 (TMI-2) in 1979. This event significantly altered the industry`s view of severe accidents in the U.S. and worldwide. Investigative committees of TMI-2 recommended that PSA techniques be more widely used to augment the traditional deterministic methods of determining nuclear plant safety. This initiated an unprecedented effort by nuclear regulators and licensees worldwide to significantly improve the state of knowledge of severe accidents at nuclear power plants. In the U.S., use of PSA began to increase as evidenced by its application in the anticipated transient without scram and station blackout rulemakings, generic issue prioritization and resolution, risk-based inspection guidelines, backfit policy, and technical specification improvements. However, broad application of probabilistic techniques to the industry as a whole was initiated in 1986 with the publication of Safety Goals for the Operation of Nuclear Power Plant; Policy Statement. This put PSA front and center in the U.S. regulatory arena by {open_quotes}establish[ing] goals that broadly define an acceptable level of radiological risk that might be imposed on the public as a result of nuclear power plant operation.{close_quotes} Both qualitative safety goals and quantitative objectives were articulated in this policy statement.
Date: December 1996
Creator: Linn, M. A.; Cunningham, M. A. & Johnson, D. H.
Partner: UNT Libraries Government Documents Department

Ozone chemiluminescent detection of olefins: Potential applications for real-time measurements of natural hydrocarbon emissions

Description: A chemiluminescence analyzer has been constructed that takes advantage of the temperature dependence of the ozone-hydrocarbon reaction. When operated at a temperature of 170 C, the analyzer functions as a total nonmethane hydrocarbon analyzer with sensitivities 10--1,000 times better than a conventional FID. However, with operation at varying temperatures, the chemiluminescent signal reflects the differences in rates of reaction of the hydrocarbons with ozone. Preliminary studies at room temperature indicated that the relative rates of reaction of isoprene, {alpha}-pinene, {beta}-pinene, and limonene with ozone correlated with the observed chemiluminescence signal. When hydrocarbons are grouped in classes of similar structure, their rates of reaction with electrophilic atmospheric oxidants (e.g., OH, O{sub 3}, NO{sub 3}) can be correlated with each other. By varying the temperature of the reaction chamber, the chemiluminescence analyzer can be tuned to more reactive classes of hydrocarbons. Therefore, the chemiluminescence analyzer has the ability to determine atmospheric hydrocarbon concentrations as a function of class and will also provide a measure of the atmospheric reactivity of the hydrocarbons.
Date: October 1, 1997
Creator: Marley, N.A.; Gaffney, J.S. & Cunningham, M.M.
Partner: UNT Libraries Government Documents Department

Evaluation of the in-pile pressure data from instrumented fuel assemblies IFA-431 and IFA-432.

Description: This report includes results of the examination of the in-pile pressure data from instrumented test assemblies IFA-431 and 432. The pressure data have been used to estimate the fission gas release fraction as a function of fuel burnup. Included are comparisons of the estimated release functions and those predicted by three fission gas release models using the experimental temperature histories of the fuel rods. These comparisons show that fuel temperature is the primary factor in determining fission gas release and that burnup-enhanced fission gas release is not important in UO/sub 2/ fuels irradiated to 1700 GJ/kgU (20,000 MWd/MTM).
Date: October 1, 1979
Creator: Bradley, E.R.; Cunningham, M.E.; Lanning, D.D. & Williford, R.E.
Partner: UNT Libraries Government Documents Department

Control of degradation of spent LWR (light-water reactor) fuel during dry storage in an inert atmosphere

Description: Dry storage of Zircaloy-clad spent fuel in inert gas (referred to as inerted dry storage or IDS) is being developed as an alternative to water pool storage of spent fuel. The objectives of the activities described in this report are to identify potential Zircaloy degradation mechanisms and evaluate their applicability to cladding breach during IDS, develop models of the dominant Zircaloy degradation mechanisms, and recommend cladding temperature limits during IDS to control Zircaloy degradation. The principal potential Zircaloy cladding breach mechanisms during IDS have been identified as creep rupture, stress corrosion cracking (SCC), and delayed hydride cracking (DHC). Creep rupture is concluded to be the primary cladding breach mechanism during IDS. Deformation and fracture maps based on creep rupture were developed for Zircaloy. These maps were then used as the basis for developing spent fuel cladding temperature limits that would prevent cladding breach during a 40-year IDS period. The probability of cladding breach for spent fuel stored at the temperature limit is less than 0.5% per spent fuel rod. 52 refs., 7 figs., 1 tab.
Date: October 1, 1987
Creator: Cunningham, M.E.; Simonen, E.P.; Allemann, R.T.; Levy, I.S. & Hazelton, R.F.
Partner: UNT Libraries Government Documents Department

Post-irradiation data analysis for NRC/PNL Halden assembly IFA-431

Description: Results are presented for the post irradiation examination performed on IFA-431, which was a 6-rod test fuel assembly irradiated in Halden Reactor, Norway, under sponsorship of the Nuclear Regulatory Commission. The irradiation conditions included: peak powers of 33 kW/m; coolant pressure and temperature of 3.3 MPa and 240/sup 0/C, respectively; and peak burnup of 4300 MWd/MTM. IFA-431 included instrumented rods of basic boiling water reactor design, with variations in fill gas composition, gap size, and UO/sub 2/ fuel type. The irradiation was designed to measure the effect of these variations upon fuel rod thermal and mechanical performance. The post irradiation examination assessed the permanent changes to the rods, including induced radioactivity, cladding deformation, fission gas release, and fuel densification.
Date: October 1, 1979
Creator: Nealley, C.; Lanning, D.D.; Cunningham, M.E. & Hann, C.R.
Partner: UNT Libraries Government Documents Department

High Burnup Effects Program

Description: This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the course of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs.
Date: April 1, 1990
Creator: Barner, J. O.; Cunningham, M. E.; Freshley, M. D. & Lanning, D. D.
Partner: UNT Libraries Government Documents Department

Status of the USNRC/PNL Halden test IFA-432. [PWR; BWR]

Description: The NRC/PNL fuel assembly IFA-432 began irradiation in December 1975 and as of January 1980 had reached an assembly average burnup of 2120 GJ/kgU (24.5 GWd/MTM). The rods in this assembly are heavily instrumented and this has allowed a fairly thorough analysis of fuel behavior. The emphasis of the analysis has been to evaluate the observed changes in thermal/mechanical behavior. These changes are related to estimates of fuel density changes, fission gas release, and fuel cracking and relocation. The following conclusions have been reached concerning the irradiation of IFA-432. Based on analysis of steady-state and transient thermal data, the fuel in all rods is extensively cracked and relocated. Significant reductions in the effective fuel thermal conductivity have been deduced from the analysis of the transient thermal data.
Date: January 1, 1980
Creator: Bradley, E.R.; Cunningham, M.E.; Lanning, D.D. & Williford, R.E.
Partner: UNT Libraries Government Documents Department

Two-dimensional computer simulation of hypervelocity impact cratering: some preliminary results for Meteor Crater, Arizona

Description: A computational approach used for subsurface explosion cratering has been extended to hypervelocity impact cratering. Meteor (Barringer) Crater, Arizona, was selected for our first computer simulation because it was the most thoroughly studied. It is also an excellent example of a simple, bowl-shaped crater and is one of the youngest terrestrial impact craters. Shoemaker estimates that the impact occurred about 20,000 to 30,000 years ago (Roddy (1977)). Initial conditions for this calculation included a meteorite impact velocity of 15 km/s. meteorite mass of 1.57E + 08 kg, with a corresponding kinetic energy of 1.88E + 16 J (4.5 megatons). A two-dimensional Eulerian finite difference code called SOIL was used for this simulation of a cylindrical iron projectile impacting at normal incidence into a limestone target. For this initial calculation a Tillotson equation-of-state description for iron and limestone was used with no shear strength. A color movie based on this calculation was produced using computer-generated graphics. Results obtained for this preliminary calculation of the formation of Meteor Crater, Arizona, are in good agreement with Meteor Crater Measurements.
Date: April 1, 1978
Creator: Bryan, J.B.; Burton, D.E.; Cunningham, M.E. & Lettis, L.A. Jr.
Partner: UNT Libraries Government Documents Department

Two-Dimensional Computer Simulation of Hypervelocity Impact Cratering: Some Preliminary Results for Meteor Crater, Arizona

Description: A computational approach used for subsurface explosion cratering was extended to hypervelocity impact cratering. Meteor (Barringer) Crater, Arizona, was selected for the first computer simulation because it is one of the most thoroughly studied craters. It is also an excellent example of a simple, bowl-shaped crater and is one of the youngest terrestrial impact craters. Initial conditions for this calculation included a meteorite impact velocity of 15 km/s, meteorite mass of 1.67 x 10/sup 8/ kg, with a corresponding kinetic energy of 1.88 x 10/sup 16/ J (4.5 megatons). A two-dimensional Eulerian finite difference code called SOIL was used for this simulation of a cylindrical iron projectile impacting at normal incidence into a limestone target. For this initial calculation, a Tillotson equation-of-state description for iron and limestone was used with no shear strength. Results obtained for this preliminary calculation of the formation of Meteor Crater are in good agreement with field measurements. A color movie based on this calculation was produced using computer-generated graphics. 19 figures, 5 tables, 63 references.
Date: June 1978
Creator: Bryan, J. B.; Burton, D. E.; Cunningham, M. E. & Lettis, L. A., Jr.
Location Info:
Partner: UNT Libraries Government Documents Department

Methods for increasing the efficiency of Compton imagers

Description: A Compton scatter camera based on position sensitive, planar Ge and Si(Li) detectors with segmented electrodes is being developed at LLNL. This paper presents various methods that were developed to increase the position resolution of the detectors, the granularity and capability to reconstruct the scattering sequence of the gamma-ray within the detectors. All these methods help to increase the efficiency of the imager, by accepting more photons in the final image. The initial extent and diffusion of charge-carrier clouds inside the semiconductor detectors are found to affect profoundly the fraction of interactions that deposit charge in multiple adjacent electrodes. An accurate identification of these charge-shared interactions is a key factor in correctly reconstructing the position of interactions in the detector.
Date: November 15, 2005
Creator: Mihailescu, L; Vetter, K; Burks, M; Chivers, D; Cunningham, M; Gunter, D et al.
Partner: UNT Libraries Government Documents Department