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The RBU Reactor-Burnup Code: Formulation and Operation Procedures

Description: Report discussing the computer program RBU, which calculates the neutron, reactivity, and isotopic history of a nuclear reactor in such a way as to facilitate the predictions of fuel costs and reactor performance. This report documents RBU's various calculations and operating procedures.
Date: July 1961
Creator: Triplett, J. R.; Merrill, E. T. & Burr, J. R.
Partner: UNT Libraries Government Documents Department

The Absorption and Translocation of Several Fission Elements by Russian Thistle

Description: Abstract: An investigation was conducted to determine the absorption and translocation of fission products by Russian thistle from a localized spot of contaminated soil. The amount and identity of the radioactive elements absorbed and translocated by the Russian thistle is given along with the location of these elements in the plants. Beta radioactivity to the amounts of 10 microcuries per gram caused no visible effects on the growth habits of sectioned material are included.
Date: June 8, 1950
Creator: Selders, A. A.
Partner: UNT Libraries Government Documents Department

Source Strength Information for Shielding and Stack Effluent Calculations: a Standard Practices Guide

Description: Report presenting the curves and methods used "for determining gross fission product gamma energy in Mev/sec-watt, gross fission product beta decay in curies/sec-watt, delayed uranium fission neutrons in neutrons/sec-watt and N16 and N17 decay in photons/sec-cc of water" (p. 1).
Date: October 3, 1958
Creator: Jones, L. H.
Partner: UNT Libraries Government Documents Department

EBR-2 Fisson-Product-Source Test No. 1

Description: A fission-product source (FPS) was irradiated in EBR-II to provide data for calibrating the facility's fuel-element rupture detector (FERD), which is a delayed-neutron monitor, and germanium-lithium argon-scanning system (GLASS), a fission-gas-activity monitor. A metal alloy source, Ni-3.2 wt.% uranium, provided quantitative recoil release of the fission-product nuclides. The source alloy, in tubular form, was irradiated as core-region segments of 18 capsules in the FPS subassembly. The irradiation showed that the response of the FERD was linear with increasing reactor power. The magnitude of the FERD signal was dependent on local fission rate for the FPS and the flow path of the sodium carrying the delayed-neutron emitters. The relatively high fission-gas activity released by the FPS allowed accurate calibration of the GLASS under several modes of operation and provided data for verifying a gas-release model for the reactor.
Date: August 1978
Creator: Strain, R. V.; Fogle, G. L.; Thresh, H. R.; Heinrich, R. R.; Freyer, R. M.; So, B. Y. C. et al.
Partner: UNT Libraries Government Documents Department

Report on the Possible Effects on the Surrounding Population of an Assumed Release of Fission Products into the Atmosphere from a 300 Megawatt Nuclear Reactor Located at Lagoona Beach, Michigan

Description: Report issued by the APDA over studies conducted on the release of radioactive particles near Lagoona Beach, Michigan in 1955. "The possible effects on the surrounding population of a release of fission products at the Lagoona Beach site" (p. 1) are discussed. This report includes tables, maps, and illustrations.
Date: July 1957
Creator: Gomberg, H. J.; Bassett, Thomas; Velez, Carlos & Donnell, Alton P.
Partner: UNT Libraries Government Documents Department

Calculations on U235 Fission Product Decay Chains

Description: Report of equations for calculating decay of U235. The introduction states" Calculations have been made on the U235 fission product decay schemes. The results for a typical example, that of a reactor operating at 1000 kilowatts for 180 days, have been tabulated and graphed. General formulae have been used so that the results can be applied for any power level and any time of irradiation" (p. 2).
Date: May 1952
Creator: Faller, I. L.; Chapman, T. S. & West, J. M.
Partner: UNT Libraries Government Documents Department

REACTOR CHEMISTRY DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING JANUARY 31, 1962

Description: Separate abstracts were prepared for thirty-one of the thirty-three sections. Of the sections not abstracted, the one entitled Fission Product Transport'' contained no information, the other, Transport of Noble Gases in Graphite'' is available in a more complete form as ORNLTM-I35 (NSA 16: 9209) (J.R.D.)
Date: May 11, 1962
Partner: UNT Libraries Government Documents Department

A SURVEY AND EVALUATION OF U$sup 233$ FISSION YIELD DATA

Description: A survey of the pertinent literature was made to ascertain the status of data on U/sup 233/ fission-product yields. The various experimental determinations were evaluated, and the most recent mass-spectrometric results were used as a basis for deriving a set of preferred yields. These yields were compared with values reported in two other recent compilations, and for yields >1%, the three setrs agreed with each other to an average precision of <5%. It was concluded that recent measurements have somewhar improved the reliability of U/sup 233/ fission yield data, but some recommendations for additional experimental work were made (auth)
Date: July 13, 1962
Creator: Ferguson, R.L. & O'Kelley, G.D.
Partner: UNT Libraries Government Documents Department

Production Separations of Fission-Product Groups for the Radioisotope Program

Description: Report issued by the Oak Ridge National Laboratory discussing the production separation for the radioisotope program. As stated in the abstract, "a general description is given of five years' experience in routine production of fission products of high concentration and high activity levels for the radioisotope program. Details of construction and production processes are given for two systems which were built on ion-exchange principle" (p. 2). This report includes illustrations, and photographs.
Date: August 13, 1952
Creator: Schallert, P. O.
Partner: UNT Libraries Government Documents Department

Xe-135 Production from Cf-252

Description: 135Xe is a good indicator that fission has occurred and is a valuable isotope that helps enforce the Comprehensive Test Ban Treaty. Due to its rather short half life and minimal commercial interest, there are no known sources where 135Xe can be purchased. Readily available standards of this isotope for calibrating collection and analytical techniques would be very useful. 135Xe can be produced in the fissioning of actinide isotopes, or by neutron capture on 134Xe. Since the neutron capture cross section of 134Xe is 3 mB, neutron capture is a low yield, though potentially useful, production route. 135Xe is also produced by spontaneous fission of 252Cf. 252Cf has a spontaneous fission rate of about 6 x 1011 s-1g-1. The cumulative yield from the spontaneous fission of 252Cf is 4.19%; and the competing neutron capture reaction that depletes 135Xe in thermal reactor systems is negligible because the neutron capture cross-section is low for fast fission neutrons. At the INL, scientists have previously transported fission products from an electroplated 252Cf thin source for the measurement of nuclear data of short-lived fission products using a technique called He-Jet collection. We have applied a similar system to the collection of gaseous 135Xe, in order to produce valuable standards of this isotope.
Date: March 1, 2012
Creator: McGrath, C. A.; Houghton, T. P.; Pfeiffer, J. K. & Hague, R. K.
Partner: UNT Libraries Government Documents Department

Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

Description: This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.
Date: September 1, 2013
Creator: Hiruta, Hikaru & Youinou, Gilles
Partner: UNT Libraries Government Documents Department

Terminal Status Report for the Processing Refabrication Experiment

Description: Introduction: A low-capacity, low-decontamination plant can be built, as part of a power reactor complex, to avoid long distance transfer of fuel to a high-capacity aqueous processing plant. Activation of such a complex, with the processing plant adjacent to the reactor it serves, could decrease the cost of the integrated fuel cycle. The study of this concept is a major objective of the Processing Refabrication Experiment (PRE).
Date: November 15, 1959
Creator: Sinizer, D. I.; Mattern, K. L. & Kendall, E. G.
Partner: UNT Libraries Government Documents Department

The Metabolic Properties of the Fission Products and Actinide Elements

Description: An investigation of the assimilation, distribution, retention, an excretion of the fission products and actinide elements in the rat has been conducted at the Crocker Radiation Laboratory, University of California, Berkeley, California. These studies were initiated October 15, 1942, and are continuing at the present time. An extensive survey has been made of the metabolism of twenty-two different radio elements in the rat.
Date: March 1, 1948
Creator: Hamilton M.D., J.G.
Partner: UNT Libraries Government Documents Department

OPTIMIZATION OF FISSION FRAGMENT CATCHER FOIL EXPOSURE TIME

Description: The exposure-time for fission fragment catcher foils, used in nuclear reactor power mapping, was arbitrarily set at 20 minutes. Work performed to evaluate this choice and to attempt an optimization of the exposure time is reported. A true optimum was not found. Forty minute runs are suggested, however, as a practical optimization and as an alternative to the 20 minute runs in current usage. (auth)
Date: January 1, 1958
Creator: Renaker, J.N. & Clark, R.G.
Partner: UNT Libraries Government Documents Department

SOLIDS ACCUMULATION AND FISSION HEATING IN THE HRT CHEMICAL PLANT UNDERFLOW POT (CO-OP REPORT, FALL QUARTER, 1958)

Description: A study was conducted to develop equations for calculating fission product heating in the RRT-CP undeflow pot from measured temperatures and to attempt to correlate the rate of solids accumulation in the undeflow pot with flssion heating and reactor power. Using fission heating data calculated from the heat balances developed, several semi-empirical equations relating solids accumulation and heating were tested. In one case an error of no greater thaQ 26% was incurred in the calculation of the total weight of solids collected during chemical plant runs 17-4, 17-5, and 17-6. Further development work will be done on this correlation. (auth)
Date: June 10, 1959
Creator: Dunn, W E
Partner: UNT Libraries Government Documents Department

Solvent Extraction of Tc and Cs from Alkaline Nitrate Wastes

Description: This paper summarizes progress at three collaborating US national laboratories on the extraction of the fission products {sup 99}Tc and {sup 137}Cs from alkaline high-level wastes (HLW). Efficient, economical processes for Tc and Cs extraction (SRTALK and alkaline-side CSEX, respectively) have been developed, and testing has progressed through batch tests on actual wastes and continuous countercurrent centrifugal-contactor tests on simulants.
Date: July 11, 1999
Creator: Bonnesen, P.V.; Conner, C.; Delmau, L.H.; Haverlock, T.J.; Leonard, R.A.; Lumetta, G.J. et al.
Partner: UNT Libraries Government Documents Department

PNNL Review of Proposed Relevant Radionuclide List

Description: A list of fission products and activation products has been proposed for possible adoption as an official table of relevant isotopes for CTBT use. It is our understanding that the purpose of this list is to discriminate Level 4 spectra from Level 5 spectra in the decision logic diagram. The current understanding is that a single short-lived, relevant isotope that is atypical for a station would cause a spectrum to be marked as Level 4. A second relevant isotope would cause a spectrum to be marked as Level 5, which would perhaps require a sample to undergo additional laboratory conflation measurements. The list consists of a very comprehensive set of fission products and activation products. We have examined the list for accuracy and have also flagged potential problems with members of the list. in our opinion, several of these isotopes have serious problems and many have no practical chance of ever being the first or second detected isotopes. We are not arguing whether or not these isotopes might be seen in a large atmospheric test. On the other hand, there may be no harm associated with having a long list. The issue of activation products is different. Some activation products are indicative of the soil or rock composition in the vicinity of an explosion. Others may only be dependent on materials in the weapon or in the support structures. We don't think that a great deal of analysis of these isotopes by the CTBTO should be encouraged. In any case, if particulate activation products are in the atmosphere, fission products should be even more prevalent, thus removing the need for an activation list component.
Date: May 10, 1999
Creator: Miley, HS & Arthur, RJ
Partner: UNT Libraries Government Documents Department

Measurement of the Hydrogen Yield in the Radiolysis of Water by Dissolved Fission Products

Description: Hydrogen from the radiolysis of water by dissolved fission products is stripped from the solution and collected by bubbling carbon dioxide through the solution. Quantitative measurements of the G value for hydrogen show that the yield is essentially the same as would be obtained by external gamma radiolysis of nonradioactive solutions of the same chemical composition. The hydrogen yield can be enhanced by addition of a hydrogen-atom donor, such as formic acid, to the solution. The yield of hydrogen from fission-waste solutions is discussed with respect to the question of whether it represents a significant energy source.
Date: April 1976
Creator: Sauer, M. C.; Hart, E. J.; Flynn, K. F. & Gindler, J. E.
Partner: UNT Libraries Government Documents Department