[en] The modeling described in this report is an extension of previous fate and transport modeling for the Old Radioactive Waste Burial Ground Corrective Measures Study/Feasibility Study. The purpose of this and the previous modeling is to provide quantitative input to the screening of remedial alternatives for the CMS/FS for this site

[en] We show that the discrete Kadomtsev–Petviashvili (KP) equation with sources obtained recently by the “source generalization” method can be incorporated into the squared eigenfunction symmetry extension procedure. Moreover, using the known correspondence between Darboux-type transformations and additional independent variables, we demonstrate that the equation with sources can be derived from Hirota's discrete KP equations but in a space of higher dimension. In this way we uncover the origin of the source terms as coming from multidimensional consistency of the Hirota system itself.

[en] The types and the forms of radioactive source term are getting a variety due to expansion of the radiation industry such as medical radiation treatments and research. of materials. For neutron sources, it is necessary to account for the neutron energy distribution to evaluate exactly the dose equivalent, because the energy range of the neutron from thermal neutrons (0.025 eV) to a few GeV. The neutron dose equivalent depends heavily on the energy of the neutrons. The neutron field spectra from a DT neutron generator at Korea Atomic Energy research Institute (KAERI) and the proton accelerators at Korea Institute of Radiological and Medical Science (KIRAMS) were measured by using the Bonner sphere measurement system (BS system) and several neutron detectors. The spectrum weighted response was obtained by using the response function of the neutron detectors as given in the IAEA Technical Report Series 403 (TRS)

[en] In this article experimental investigation is described into the spread of fission products within a nuclear power plant, which after an accident involving melting of the nucleus, will be possible in spite of prohibiting constructions for the case of severe unbalancing of generated and carried-off energy. 6 refs.; 4 figs

[en] This technical manual presents updated generic source terms (i.e., unitized amounts and radionuclide compositions) which have been developed for use in the Integrated Data Base (IDB) Program of the U.S. Department of Energy (DOE). These source terms were used in the IDB annual report, Integrated Data Base for 1992: Spent Fuel and Radioactive Waste Inventories, Projections, and Characteristics, DOE/RW-0006, Rev. 8, October 1992. They are useful as a basis for projecting future amounts (volume and radioactivity) of low-level radioactive waste (LLW) shipped for disposal at commercial burial grounds or sent for storage at DOE solid-waste sites. Commercial fuel cycle LLW categories include boiling-water reactor, pressurized-water reactor, fuel fabrication, and uranium hexafluoride (UF₆) conversion. Commercial nonfuel cycle LLW includes institutional/industrial (I/I) waste. The LLW from DOE operations is category as uranium/thorium fission product, induced activity, tritium, alpha, and open-quotes otherclose quotes. Fuel cycle commercial LLW source terms are normalized on the basis of net electrical output [MW(e)-year], except for UF₆ conversion, which is normalized on the basis of heavy metal requirement [metric tons of initial heavy metal ]. The nonfuel cycle commercial LLW source term is normalized on the basis of volume (cubic meters) and radioactivity (curies) for each subclass within the I/I category. The DOE LLW is normalized in a manner similar to that for commercial I/I waste. The revised source terms are based on the best available historical data through 1992

[en] We introduce a new treatment of the difference formulation[1] for photon radiation transport without scattering in 1d slab geometry that is closely analogous to that of Fleck and Cummings[2] for the traditional formulation. The resulting form is free of implicit source terms and has the familiar effective scattering of the field of transport

[en] To calculate the source term behaviour of nuclear power plants close at the border the Austrian Research Centre Seibersdorf uses the 'Source Term Code Package'. This package was installed on the CYBER 180-860 of the Technical University in Vienna, which has been changed to a CONVEX C3220 at the beginning of 1992. Therefore it was necessary to implement the program package on the new machine. This report is a compilation of the program changes. Finally the experience of this work is discussed. (authors)

[en] The purpose of this conceptual model analysis is to assist Performance Assessment Operations (PA) in modeling the volume of in-package water available for contact with the wasteform. The conceptual model developed in this analysis model report (AMR) represents an enhancement of the in-package water volume model that was abstracted into TSPA-VA performance assessment modeling. This enhanced conceptual model will allow PA to provide a more detailed and complete representation of the volume of in-package water available for contact with the wasteform as a function of time. The conceptual model is presented as a methodology appropriate for use in the total system performance model

[en] Most long-lived radionuclides associated with an underground nuclear test are incorporated into a melt glass and are released by glass dissolution to become part of the hydrologic source term (HST) (Pawloski et al., 2001). Although the rates of rhyolite glass dissolution are well known under conditions where the fluid is far from saturation with respect to glass, the rates are not well known under conditions where the fluid approaches saturation. These rates are commonly much lower than the far-fromsaturation rates, often by a factor greater than 100. In recent HST simulations (Pawloski et al., 2001; Pawloski et al., 2000; Tompson et al., 1999), we conservatively estimated steady-state release rates based on a far-from-saturation fluid conditions. In recent CHESHIRE near-field simulations (Pawloski et al., 2001), it was predicted that ∼30% of the nuclear melt glass dissolved over 1000 years. Although the ''far-from-saturation rate'' approach provides a conservative estimate of glass dissolution, it may greatly overestimate the rates of melt glass dissolution. At CHESHIRE, less conservative estimates suggest that only ∼1% of the nuclear melt glass will dissolve in 1000 years. Lower glass dissolution rates result in lower radionuclide release rates from nuclear melt glass. The following report documents glass dissolution experiments performed to measure glass dissolution rates close to saturation

[en] The 232-Z waste incinerator at the Hanford plutonium finishing facility was used to incinerate plutonium-bearing combustible materials generated during normal plant operations. Nondestructive analysis performed after the incinerator ceased operations indicated high plutonium loading in exhaust ductwork near the incinerator glove box, while the incinerator was found to have only low quantities. Measurements following a campaign to remove some of the ductwork resulted in a markedly higher assay valve for the incinerator glove box itself. Subsequent assays confirmed the most recent results and pointed to a potential further underestimation of the holdup, in part due to the attenuation due to fire brick which could not be seen and which had been thought to be present. Resolution of the raised concerns entailed forming a task team to perform further assay based on gamma and neutron NDA methods. This paper is a discussion of the unreviewed safety question regarding the source term in this area