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Description: Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an ''upper-limit'' (i.e., instantaneous degradation) model for use in the TSPA-LA model. ''Best-estimate'' models for the degradation of the fuels in each of ...
Date: November 19, 2004
Creator: CUNNANE, J.
Partner: UNT Libraries Government Documents Department

In-Situ Method for Treating Residual Sodium

Description: A unique process for deactivating residual sodium in Liquid Metal Fast Breeder Reactor (LMFBR) systems which uses humidified (but not saturated) carbon dioxide at ambient temperature and pressure to convert residual sodium into solid sodium bicarbonate.
Date: July 19, 2005
Creator: Sherman, Steven R. & Henslee, S. Paul
Partner: UNT Libraries Government Documents Department

Reactivity feedback from irradiated pin failure in unprotected slow TOP accidents in LMFBR's

Description: The present work is an outgrowth of studies made in support of CRBR licensing, but the conclusions drawn should be generally applicable to oxide-fueled LMFBR's. The accident under consideration is a 10 cents/s unprotected TOP (transient overpower), for which a series of PLUTO2/SAS4A calculations has been performed using a higher power CRBR EOC3 fuel pin which had 275 days irradiation. The assumption was made in the licensing work that a short pin failure will occur at the axial midplane, maximizing the positive fuel motion reactivity effect, as it was felt that a less conservative assumption could not be conclusively justified. This assumption is also made in the present case.
Date: September 19, 1984
Creator: Hummel, H.H. & Pizzica, P.A.
Partner: UNT Libraries Government Documents Department

Seismic design and development for fast reactors: a design-application perspective with directions for improvement

Description: Applications of seismic design criteria and qualification methods to the US breeder reactor projects have developed new findings, improvements in design methods, and identified areas for further development. Discussions are presented regarding site free field motion, soil-structure interaction, equipment response spectra, piping, snubbers and support design analyses, dynamic decoupling, seismic qualification testing, and protection of Seismic Category I components from Non-Category I equipment failures.
Date: July 19, 1979
Creator: Severud, L. K.
Partner: UNT Libraries Government Documents Department

Sodium pool fire model for CONACS code. [LMFBR]

Description: The modeling of sodium pool fires constitutes an important ingredient in conducting LMFBR accident analysis. Such modeling capability has recently come under scrutiny at Westinghouse Hanford Company (WHC) within the context of developing CONACS, the Containment Analysis Code System. One of the efforts in the CONACS program is to model various combustion processes anticipated to occur during postulated accident paths. This effort includes the selection or modification of an existing model and development of a new model if it clearly contributes to the program purpose. As part of this effort, a new sodium pool fire model has been developed that is directed at removing some of the deficiencies in the existing models, such as SOFIRE-II and FEUNA.
Date: October 19, 1982
Creator: Yung, S.C.
Partner: UNT Libraries Government Documents Department

Fuel Coolant Thermal Interaction Project. Quarterly progress report No. 1, July 1, 1975--September 30, 1975. [LMFBR]

Description: The objective of the work reported is to experimentally and analytically study the dominant mechanism in fuel--coolant thermal interactions which could lead to vapor explosions. The exploration of mechanisms is focused in two areas: (a) mechanisms responsible for fragmentation in molten metal droplet experiments (including assessment of the validity of the proposed spontaneous nucleation mechanism), and (b) thermal stress initiated fracture as a fragmentation mechanism. Work being performed in these areas is briefly described.
Date: November 19, 1975
Creator: Todreas, N. E.
Partner: UNT Libraries Government Documents Department

Comparisons of PRD (power-reactivity-decrements) components for various EBR-II configurations

Description: Comparison of detailed calculations of contributions by region and component of the power-reactivity-decrements (PRD) for four differing loading configurations of the Experimental Breeder Reactor-II (EBR-II) are given. The linear components and Doppler components are calculated. The non-linear (primarily subassembly bowing) components are deduced by differences relative to measured total PRD values. Variations in linear components range from about 10% to as much as about 100% depending upon the component. The deduced non-linear components differ both in magnitude and sign as functions of reactor power. Effects of differing assumptions of the nature of the fuel-to-clad interactions upon the PRD components are also calculated.
Date: September 19, 1986
Creator: Meneghetti, D. & Kucera, D. A.
Partner: UNT Libraries Government Documents Department