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Description: Maximum HRT dome wall temperatures for reactor powers up to 10 Mw were calculated for both an insulated amd uninsulated dome. Temperatures exceeding 315 deg C in the fuel solution flowing through the HRT dome section can cause phase separation which results in excessive corrosion rates. The results, based on very conservative assumptions, inticated wall temperatures exceeding 315 deg C could occur in the insulated dome when operating the reactor at 10 Mw. Using the same conservative assumptions, the analysis of the uninsulated dome indicated wall and bellows temperatures below 300 deg C. (auth)
Date: April 15, 1958
Creator: Claiborne, H.C.
Partner: UNT Libraries Government Documents Department


Description: The postulate that the average number of lattice displacements is directly proportional to the available energy is carried one step further; it is assumed that damage to steel (particularly in regard to brittle fracture) is proportional to the number of lattice vacancies that occur. The model, although crude, permits estimation of the relative damage resulting from differences in neutron spectra. The results can be used as a rough method of correcting damage data for the effect of the neutron-energy spectrum. Radiation damage calculations for steel, relative to those for a fission spectrum, were made for neutron spectra that result from fission neutrons penetrating water or graphite. The results were plotted as a function of effective distance from the fission source. From this plot it is possible to make a conservative estimate of the correction factor to apply to damage data obtained with different neutron spectra. (auth)
Date: July 27, 1962
Creator: Claiborne, H.C.
Partner: UNT Libraries Government Documents Department

Trip report: workshop on risk analysis and geologic modeling in relation to the disposal of radioactive wastes into geological formations

Description: The Workshop was co-sponsored by the Commission of European Communities (CEC) and the Office of Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA), with primary object being to promote international cooperation in developing and using risk assessment techniques for the long-term safety assessment of waste disposal. The attendance was restricted to specialists in the field and a few observers; 43 people were in attendance representing 14 different countries. Nothing particularly new or novel was presented nor any formal cooperation agreed upon. However, there was a feeling that continued informal cooperation was helpful and should be continued. Greater or lesser degrees of formality could be decided later. The U.S. program was definitely more advanced and larger in scope than the others that were discussed. Countries that seemed to have significant programs include the Federal Republic of Germany, France, Canada, Sweden, and the CEC. Abstracts of papers are presented together with consensus reports on containment failure modes and geosphere transport modeling.
Date: January 1, 1977
Creator: Claiborne, H.C.
Partner: UNT Libraries Government Documents Department

Efficacy of backfilling and other engineered barriers in a radioactive waste repository in salt

Description: In the United States, investigation of potential host geologic formations was expanded in 1975 to include hard rocks. Potential groundwater intrusion is leading to very conservative and expensive waste package designs. Recent studies have concluded that incentives for engineered barriers and 1000-year canisters probably do not exist for reasonable breach scenarios. The assumption that multibarriers will significantly increase the safety margin is also questioned. Use of a bentonite backfill for surrounding a canister of exotic materials was developed in Sweden and is being considered in the US. The expectation that bentonite will remain essentially unchanged for hundreds of years for US repository designs may be unrealistic. In addition, thick bentonite backfills will increase the canister surface temperature and add much more water around the canister. The use of desiccant materials, such as CaO or MgO, for backfilling seems to be a better method of protecting the canister. An argument can also be made for not using backfill material in salt repositories since the 30-cm-thick space will provide for hole closure for many years and will promote heat transfer via natural convection. It is concluded that expensive safety systems are being considered for repository designs that do not necessarily increase the safety margin. It is recommended that the safety systems for waste repositories in different geologic media be addressed individually and that cost-benefit analyses be performed.
Date: September 1, 1982
Creator: Claiborne, H.C.
Partner: UNT Libraries Government Documents Department


Description: Dye hold-up time studies were completed for three 1/16-scale lucite models of possible HRE-3 configurations. The models studied had a spherical (#MF- 2) and a cylindrical (#MF-1)(core with flat heads on the pressure vessels. Model MF-2 was also run with a ''dished'' head on the pressure vessel (#MF-2D). Results for these models indicate a change from laminar to turbulent flow when the equivalent Reynolds number was within a range of 900 to 1,500. Better mixing was found in model MF-1 than in MF-2, but the difference was small beyond the transition point. Addition of a dished head to model MF-2 caused the dye hold-up times to decrease at Reynolds numbers less than 2,500. (auth)
Date: June 1, 1958
Creator: Mixon, W.R. & Claiborne, H.C.
Partner: UNT Libraries Government Documents Department

Brine migration in salt and its implications in the geologic disposal of nuclear waste

Description: This report respresents a comprehensive review and analysis of available information relating to brine migration in salt surrounding radioactive waste in a salt repository. The topics covered relate to (1) the characteristics of salt formations and waste packages pertinent to considerations of rates, amounts, and effects of brine migration, (2) experimental and theoretical information on brine migration, and (3) means of designing to minimize any adverse effects of brine migration. Flooding, brine pockets, and other topics were not considered, since these features will presumably be eliminated by appropriate site selection and repository design. 115 references.
Date: December 1, 1981
Creator: Jenks, G.H. & Claiborne, H.C.
Partner: UNT Libraries Government Documents Department

Expected near-field thermal environments in a sequentially loaded spent-fuel or high-level waste repository in salt

Description: This report describes the effect of realistic waste emplacement schedules on repository thermal environments. Virtually all estimates to date have been based on instantaneous loading of wastes having uniform properties throughout the repository. However, more realistic scenarios involving sequential emplacement of wastes reflect the gradual filling of the repository over its lifetime. These cases provide temperatures that can be less extreme than with the simple approximation. At isolated locations in the repository, the temperatures approach the instantaneous-loading limit. However, for most of the repository, temperature rises in the near-field are 10 to 40 years behind the conservative estimates depending on the waste type and the location in the repository. Results are presented for both spent-fuel and high-level reprocessing waste repositories in salt, for a regional repository concept, and for a single national repository concept. The national repository is filled sooner and therefore more closely approximates the instantaneously loaded repository. However, temperatures in the near-field are still 20/sup 0/C or more below the values in the simple model for 40 years after startup of repository emplacement operations. The results suggest that current repository design concepts based on the instantaneous-loading predictions are very conservative. Therefore, experiments to monitor temperatures in a test and evaluation facility, for example, will need to take into account the reduced temperatures in order to provide data used in predicting repository performance.
Date: January 1, 1982
Creator: Rickertsen, L.D.; Arbital, J.G. & Claiborne, H.C.
Partner: UNT Libraries Government Documents Department

Comparison of the results of several heat transfer computer codes when applied to a hypothetical nuclear waste repository

Description: A direct comparison of transient thermal calculations was made with the heat transfer codes HEATING5, THAC-SIP-3D, ADINAT, SINDA, TRUMP, and TRANCO for a hypothetical nuclear waste repository. With the exception of TRUMP and SINDA (actually closer to the earlier CINDA3G version), the other codes agreed to within +-5% for the temperature rises as a function of time. The TRUMP results agreed within +-5% up to about 50 years, where the maximum temperature occurs, and then began an oscillary behavior with up to 25% deviations at longer times. This could have resulted from time steps that were too large or from some unknown system problems. The available version of the SINDA code was not compatible with the IBM compiler without using an alternative method for handling a variable thermal conductivity. The results were about 40% low, but a reasonable agreement was obtained by assuming a uniform thermal conductivity; however, a programming error was later discovered in the alternative method. Some work is required on the IBM version to make it compatible with the system and still use the recommended method of handling variable thermal conductivity. TRANCO can only be run as a 2-D model, and TRUMP and CINDA apparently required longer running times and did not agree in the 2-D case; therefore, only HEATING5, THAC-SIP-3D, and ADINAT were used for the 3-D model calculations. The codes agreed within +-5%; at distances of about 1 ft from the waste canister edge, temperature rises were also close to that predicted by the 3-D model.
Date: December 1, 1979
Creator: Claiborne, H.C.; Wagner, R.S. & Just, R.A.
Partner: UNT Libraries Government Documents Department

Expected environments in high-level nuclear waste and spent fuel repositories in salt

Description: The purpose of this report is to describe the expected environments associated with high-level waste (HLW) and spent fuel (SF) repositories in salt formations. These environments include the thermal, fluid, pressure, brine chemistry, and radiation fields predicted for the repository conceptual designs. In this study, it is assumed that the repository will be a room and pillar mine in a rock-salt formation, with the disposal horizon located approx. 2000 ft (610 m) below the surface of the earth. Canistered waste packages containing HLW in a solid matrix or SF elements are emplaced in vertical holes in the floor of the rooms. The emplacement holes are backfilled with crushed salt or other material and sealed at some later time. Sensitivity studies are presented to show the effect of changing the areal heat load, the canister heat load, the barrier material and thickness, ventilation of the storage room, and adding a second row to the emplacement configuration. The calculated thermal environment is used as input for brine migration calculations. The vapor and gas pressure will gradually attain the lithostatic pressure in a sealed repository. In the unlikely event that an emplacement hole will become sealed in relatively early years, the vapor space pressure was calculated for three scenarios (i.e., no hole closure - no backfill, no hole closure - backfill, and hole closure - no backfill). It was assumed that the gas in the system consisted of air and water vapor in equilibrium with brine. A computer code (REPRESS) was developed assuming that these changes occur slowly (equilibrium conditions). The brine chemical environment is outlined in terms of brine chemistry, corrosion, and compositions. The nuclear radiation environment emphasized in this report is the stored energy that can be released as a result of radiation damage or crystal dislocations within crystal lattices.
Date: August 1, 1980
Creator: Claiborne, H.C. & Rickertsen, L.D., Graham, R.F.
Partner: UNT Libraries Government Documents Department

Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

Description: This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs.
Date: September 1, 1987
Creator: Claiborne, H.C.; Croff, A.G.; Griess, J.C. & Smith, F.J.
Partner: UNT Libraries Government Documents Department

Physical and decay characteristics of commercial LWR spent fuel

Description: Information was collected from the literature and from major manufacturers that will be useful in the design and construction of a mined geologic repository for the disposal of light-water-reactor spent fuel. Pertinent data are included on mechanical design characteristics and materials of construction for fuel assemblies and fuel rods and computed values for heat generation rates, radioactivity, and photon and neutron emission rates as a function of time for four reference cases. Calculations were made with the ORIGEN2 computer code for burnups of 27,500 and 40,000 MWd for a typical boiling-water reactor and 33,000 and 60,000 MWd for a typical pressurized-water reactor. The results are presented in figures depicting the individual contributions per metric ton of initial heavy metal for the activation products, fission products, and actinides and their daughters to the radioactivity and thermal power as a function of time. Tables are also presented that list the contribution of each major nuclide to the radioactivity, thermal power, and photons and neutrons emitted for disposal emitted for disposal periods from 1 to 100,000 years.
Date: January 1, 1986
Creator: Roddy, J.W.; Claiborne, H.C.; Ashline, R.C.; Johnson, P.J. & Rhyne, B.T.
Partner: UNT Libraries Government Documents Department

Development of reference conditions for geologic repositories for nuclear waste in the USA

Description: Activities to determine interim reference conditions for temperatures, pressure, fluid, chemical, and radiation environments that are expected to exist in commercial and defense high-level nuclear waste and spent fuel repositories in salt, basalt, tuff, granite, and shale are summarized. These interim conditions are being generated by the Reference Repository Conditions Interface Working Groups (RRC-IWG), an ad hoc IWG established by the National Waste Terminal Storage Program's (NWTS) Isolation Interface Control Board (I-ICB).
Date: October 1, 1980
Creator: Raines, G.E.; Rickertsen, L.D.; Claiborne, H.C.; McElroy, J.L. & Lynch, R.W.
Partner: UNT Libraries Government Documents Department