430 Matching Results

Search Results

Advanced search parameters have been applied.


Description: A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.
Date: June 1, 2007
Creator: CHENG,L.Y. & LUDEWIG, H.
Partner: UNT Libraries Government Documents Department

Development Of An Experiment For Measuring Flow Phenomena Occurring In A Lower Plenum For VHTR CFD Assessment

Description: The objective of the present report is to document the design of our first experiment to measure generic flow phenomena expected to occur in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In the process, fabrication sketches are provided for the use of CFD (computational fluid dynamics) analysts wishing to employ the data for assessment of their proposed codes. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. One aspect of the complex flow in a prismatic VHTR is being addressed: flow and thermal mixing in the lower plenum ("hot streaking" issue). Current prismatic VHTR concepts were examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses were applied to determine key non-dimensional parameters and their magnitudes over this operating range. The flow in the lower plenum can locally be considered to be a situation of multiple jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentum-dominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other.
Date: September 1, 2005
Creator: McEligot, D. M.; Condie, K.G.; Creery, G. E. Mc & Ilroy, H. M. Mc
Partner: UNT Libraries Government Documents Department


Description: Three physically distinguishable, poorly crystalline forms of Ca(OH){sub 2} which are made by reactions of water vapor at 25 C with CaO powders evolve heat and develop sharper X-ray diffraction patterns when heated to 300 C. Measurements of X-ray diffraction peak line breadths, surface areas, porosities, and pore size distributions were made before and after heat treatment. The exothermic process of one of these forms of Ca(OH){sub 2} which is nonporous is recrystallization of the highly strained solid to essentially unstrained crystals. The exothermic process in two porous forms of Ca(OH){sub 2}, not only reduces their internal strains, but also reduces their surface areas and porosities.
Date: February 1, 1980
Creator: Beruto, Dario; Searcy, Alan W.; Barco, Luigi & Belleri, Gabriele
Partner: UNT Libraries Government Documents Department

Radionuclide mass inventory, activity, decay heat, and dose rate parametric data for TRIGA spent nuclear fuels

Description: Parametric burnup calculations are performed to estimate radionuclide isotopic mass and activity concentrations for four different Training, Research, and Isotope General Atomics (TRIGA) nuclear reactor fuel element types: (1) Aluminum-clad standard, (2) Stainless Steel-clad standard, (3) High-enrichment Fuel Life Improvement Program (FLIP), and (4) Low-enrichment Fuel Life Improvement Program (FLIP-LEU-1). Parametric activity data are tabulated for 145 important radionuclides that can be used to generate gamma-ray emission source terms or provide mass quantity estimates as a function of decay time. Fuel element decay heats and dose rates are also presented parametrically as a function of burnup and decay time. Dose rates are given at the fuel element midplane for contact, 3.0-feet, and 3.0-meter detector locations in air. The data herein are estimates based on specially derived Beginning-of-Life (BOL) neutron cross sections using geometrically-explicit TRIGA reactor core models. The calculated parametric data should represent good estimates relative to actual values, although no experimental data were available for direct comparison and validation. However, because the cross sections were not updated as a function of burnup, the actinide concentrations may deviate from the actual values at the higher burnups.
Date: March 1, 1997
Creator: Sterbentz, J.W.
Partner: UNT Libraries Government Documents Department

Thermal Performance of Radioactive Material Packages in Transport Configuration

Description: Drum type packages are routinely used to transport radioactive material (RAM) in the U.S. Department of Energy (DOE) complex. These packages are designed to meet the federal regulations described in 10 CFR Part 71. The packages are transported in specially designed vehicles like Safe Secure Transport (SST) for safety and security. In the transport vehicles, the packages are placed close to each other to maximize the number of units in the vehicle. Since the RAM contents in the packagings produce decay heat, it is important that they are spaced sufficiently apart to prevent overheating of the containment vessel (CV) seals and the impact limiter to ensure the structural integrity of the package. This paper presents a simple methodology to assess thermal performance of a typical 9975 packaging in a transport configuration.
Date: March 4, 2010
Creator: Gupta, N.
Partner: UNT Libraries Government Documents Department

Downward heat transfer in a miscible melting system

Description: The integrity of an ex-vessel core-retention system in the event of core meltdown is of concern in PAHR safety assessment. Several ex-vessel core retention concepts incorporate sacrificial beds. The integrity of the ex-vessel core-retention system is dependent on the directional growth of the molten pool into soluble boundaries of the sacrificial bed. Mutual dissolution of the molten pool of core-debris and the sacrificial material is expected to change the thermal characteristics of the pool and thus affect the heat transfer to the boundaries. The two-dimensional simulation study of the penetration of a dense, hot liquid pool into the boundaries of a meltable, soluble solid revealed the dependency of the directional pool growth on the density ratio, rho*, of the liquid pool to the meltable solid. In the one-dimensional study of the downward penetration of the hot pool into a soluble boundary four different hydrodynamic flow regimes were identified that occurred at different ranges of rho*. The downward heat transfer enhanced beyond rho* approx. = 1.1. The present study investigates the effect of test cell geometry and material properties on the downward heat transfer in a horizontal melting system.
Date: January 1, 1981
Creator: Farhadieh, R.
Partner: UNT Libraries Government Documents Department

Fractional heat generation rates in Hanford reactors after shutdown

Description: The knowledge of the fraction of decay-heat which is absorbed in fuel elements after a reactor is shutdown is important for many reasons. For example, allowable, reduced flow-rates after shutdown are very sensitive to the manner in which the decay-heat is distributed. Also, the temperature-rise in a discharged, uncooled fuel element is dependent on the total heat generated in the slug. Apart from any heat consideration, the escape of rays from a discharged fuel element is also of importance in certain applications. A significant refinement in the knowledge pertaining to decay-heat in irradiated uranium in recent years warrants a complete review of the fractional heat generation in the Hanford reactors. Earlier work was based on very qualitative aspects of fission product decay rates and energy spectra following reactor shutdown. The results of these early calculations on the original, solid, Hanford slug indicated that about 20 per cent of the energy generated in a fuel element from both fission product decay and delayed fission escaped from slug. The results reported here show a much smaller escape-fraction. Also, the time-dependence of the escape-fraction is considered for times greater than 100 seconds after reactor shutdown. Fractional heat-generation rates are calculated for various Hanford fuel types. The results, condensed to the maximum fraction of energy which escapes from the fuel elements over the period 10{sup 2}--10{sup 7} seconds after shutdown, are given in table below. Time-dependent Curves from 100 seconds to 10{sup 7} seconds after shutdown are given in the body of the report.
Date: May 17, 1961
Creator: Nilson, R. & Meichle, R. H.
Partner: UNT Libraries Government Documents Department

Thermal issues with the US high-level waste repository and the potential benefits of waste transmutation

Description: This paper provides a qualitative update of the thermal issues arising from the decay heat in the proposed U.S. high level waste repository at Yucca Mountain. Significant questions about the ability to license this site are envisioned due to the difficulties in predicting perturbations to the site that arise from the decay heat. The hydrology of Yucca Mountain would be affected. It is suggested that waste transmutation (fuel reprocessing and use of Pu and other transuranic elements as fuel) may provide significant benefits to the repository by removing the long-term heat source posed by actinides.
Date: December 31, 1995
Creator: Michaels, G.E.
Partner: UNT Libraries Government Documents Department

Calorimetric measurement of afterheat in target materials for the accelerator production of tritium

Description: The estimate of afterheat in a spallation target of lead (Pb) or tungsten (W), by calorimetry, is the purpose of this experiment in support of the Accelerator Production of Tritium (APT). Such measurements are needed to confirm code calculations, these being the only practical way of gaining this type of information in a form suitable to aid the design of the APT machine. Knowledge of the magnitude and duration of afterheat resulting from decay of activation products produced by proton bombardment of the target is necessary to quantify APT safety assumptions, to design target cooling and safety systems, and to reduce technical risk. Direct calorimetric measurement of the afterheat for the appropriate incident proton energies is more reliable than the available alternative, which is indirect, based on data from gamma-ray spectroscopy measurements. The basic concept, a direct measurement of decay afterheat which bypasses the laborious classical way of determining this quantity, has been demonstrated to work. The gamma-ray energy given off by the decay products produced in the activation of lead or tungsten with high-energy protons apparently does represent a significant fraction of the total decay energy. A calorimeter designed for measurement of isotopes decaying by alpha emission must be modified to reduce energy lost with escaping gamma rays. Replacement of the aluminum liner with a tungsten liner in the SSC measurement chamber resulted in a 270% increase in measured heat, proving that the energy loss in the earlier (1992) measurements was significant. Gamma-ray measurements are needed to confirm the gamma-ray absorption calculations for the calorimeter to determine the correction for loss of heat due to transmission of high-energy gamma rays through the calorimeter walls. The experiments at BLIP have shown that calorimetry can be a useful tool in measuring the afterheat in APT target materials.
Date: June 1, 1994
Creator: Perry, R.B. & Zucker, M.S.
Partner: UNT Libraries Government Documents Department

Fission-product afterheat: a review of experiments pertinent to the thermal-neutron fission of $sup 235$U

Description: Experimental data relating to beta and gamma energy release rates following thermal-neutron fission of /sup 235/U are reviewed. An afterheat function based on these experiments, for after-shutdown times from 1 s to 5 x 10/ sup 5/ s, is proposed and compared with the proposed ANS Standard ANS-5.1; and the uncertainty in the function proposed here is estimated. This proposed function varies between 0.98 and 1.08 times the proposed standard function; the standard deviation is estimated to be between 10% and 15%. (31 figures, 17 tables, 48 references) (auth)
Date: October 1, 1973
Creator: Perry, A. M.; Maienschein, F. C. & Vondy, D. R.
Partner: UNT Libraries Government Documents Department

Afterheat calculations for the HTGR

Description: Afterheat rates were calculated for the HTGR utilizing new input information. The new afterheat results are a few percent higher than the HTGR afterheat results previously obtained in the 100-to 10,000-sec time range and are 11 to 15% higher than the previous results in the one-day to ten-days time range. The uncertainties were studied in the time region from 100 sec to ten days. The new results were also compared with calculations at other laboratories. The gamma-ray spectra were computed at several shutdown times for future applications in shielding studies. Previous afterheat calculations for the HTGR are completely replaced and updated. (auth)
Date: November 1, 1973
Creator: Sund, R.E.
Partner: UNT Libraries Government Documents Department


Description: Decay heat removal at depressurized shutdown conditions has been regarded as one of the key areas where significant improvement in passive response was targeted for the GEN IV GFR over the GCFR designs of thirty years ago. It has been recognized that the poor heat transfer characteristics of gas coolant at lower pressures needed to be accommodated in the GEN IV design. The design envelope has therefore been extended to include a station blackout sequence simultaneous with a small break/leak. After an exploratory phase of scoping analysis in this project, together with CEA of France, it was decided that natural convection would be selected as the passive decay heat removal approach of preference. Furthermore, a double vessel/containment option, similar to the double vessel/guard vessel approach of the SFR, was selected as the means of design implementation to reduce the PRA risks of the depressurization accident. However additional calculations in conjunction with CEA showed that there was an economic penalty in terms of decay heat removal system heat exchanger size, elevation heights for thermal centers, and most of all in guard containment back pressure for complete reliance on natural convection only. The back pressure ranges complicated the design requirements for the guard containment. Recognizing that the definition of a loss-of-coolant-accident in the GFR is a misnomer, since gas coolant will always be present, and the availability of some driven blower would reduce fuel temperature transients significantly; it was decided instead to aim for a hybrid active/passive combination approach to the selected BDBA. Complete natural convection only would still be relied on for decay heat removal but only after the first twenty four hours after the initiation of the accident. During the first twenty four hour period an actively powered blower would be relied on to provide the emergency decay power removal. ...
Date: June 1, 2007
Creator: CHENG,L.Y. & LUDEWIG, H.
Partner: UNT Libraries Government Documents Department

Fission product decay heat studies of December 15, 1975

Description: The purpose of the project described is to study fission product decay heating rates, with emphasis on short decay times. Isothermal calorimetry is used to perform benchmark experiments for decay times between 20 seconds and 2000 seconds with an absolute accuracy of better than 5 percent. Experiments are being done with $sup 235$U and will be done eventually with $sup 239$Pu. Thermal neutron spectra are used for the irradiations. The project was initiated in July 1974. Final results for $sup 235$U are expected by 6-30-76, and for $sup 239$Pu by 1-1-77. Final reports for each task will follow the final data by 3 months. The work done to date is described together with the status of the final experimental configuration.
Date: February 26, 1976
Creator: Yarnell, J.L. & Bendt, P.J.
Partner: UNT Libraries Government Documents Department

Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

Description: A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of ...
Date: December 15, 2005
Creator: Ingersoll, DT
Partner: UNT Libraries Government Documents Department

Optimized, Competitive Supercritical-CO2 Cycle GFR for Gen IV Service

Description: An overall plant design was developed for a gas-cooled fast reactor employing a direct supercritical Brayton power conversion system. The most important findings were that (1) the concept could be capital-cost competitive, but startup fuel cycle costs are penalized by the low core power density, specified in large part to satisfy the goal of significatn post-accident passive natural convection cooling; (2) active decay heat removal is preferable as the first line of defense, with passive performance in a backup role; (3) an innovative tube-in-duct fuel assembly, vented to the primpary coolant, appears to be practicable; and (4) use of the S-Co2 GFR to support hydrogen production is a synergistic application, since sufficient energy can be recuperated from the product H2 and 02 to allow the electrolysis cell to run 250 C hotter than the reactor coolant, and the water boilers can be used for reactor decay heat removal. Increasing core poer density is identified as the top priority for future work on GFRs of this type.
Date: September 8, 2008
Creator: Driscoll, M.J.; Hejzlar, P. & Apostolakis, G.
Partner: UNT Libraries Government Documents Department

Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

Description: The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.
Date: January 1, 2003
Creator: Shaber, Eric; Baccaglini, G.; Ball, S.; Burchell, T.; Corwin, B.; Fewell, T. et al.
Partner: UNT Libraries Government Documents Department

HEat Decay Data Repository Footprint for Thermal-Hydrologic and Conduction-Only Models for TSPA-SR

Description: The repository heat decay data contained within this calculation is specified for both mountain-scale and drift-scale thermal-hydrologic (TH), thermal-hydrologic-mechanical (THM), and thermal-hydrologic-chemical (THC) simulations used in total systems performance assessments (TSPA). Repository thermal output data, and how it decays in time, is required by the models that compute changes to the geologic system as a result of a heat addition. The mountain-scale problem requires a repository-wide waste stream including the total heat output of each fuel type to be emplaced in the repository. These models apply a smeared heat source over a predefined repository footprint area specified in the model. The drift-scale problem requires the heat output of a number of representative (specific) waste package types. These models apply specific waste package heat outputs resolved at the scale of the waste package itself. The results of this calculation will supply details of the repository heat load for each model type. It also provides a schematic of the repository footprint outlines for the License Application Design Selection (LADS), the total repository footprint for TSPA site recommendation (SR) including the contingency area, and the actual loaded repository footprint. This calculation is performed under procedure AP-3.12Q, Rev. 0/ICN 0, Calculations. It is directed by the development plan TDP-MGR-HS-000001 (CRWMS M&O 1999f) which was developed under procedure AP-2.13Q, Rev. 0/ICN 1, Technical Product Development Plans for use in Performance Assessment activities.
Date: April 24, 2000
Creator: Francis, N.D.
Partner: UNT Libraries Government Documents Department

Prototype Testing for a Copper Rotatable Collimator for the LHC Collimation Upgrade

Description: The Phase II upgrade to the LHC collimation system calls for complementing the robust Phase I graphite collimators with high Z Phase II collimators. The design for the collimation upgrade has not been finalized. One option is to use metallic rotatable collimators and testing of this design will be discussed here. The Phase II collimators must be robust in various operating conditions and accident scenarios. A prototype collimator jaw referred to as RC0 has been tested for both mechanical and thermal compliance with the design goals. Thermal expansion bench-top tests are compared to ANSYS simulation results. The prototype has also been tested in vacuum bake-out to confirm compliance with the LHC vacuum spec. CMM equipment has been used to verify the flatness of the jaw surface after heat tests and bake-out.
Date: January 20, 2009
Creator: Smith, Jeffrey Claiborne; Anzalone, Gene; Doyle, Eric; Keller, Lewis; Lundgren, Steven; Markiewicz, Thomas Walter et al.
Partner: UNT Libraries Government Documents Department

Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

Description: This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.
Date: May 1, 2011
Creator: Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma (Computational Engineering Analysis, Albuquerque, NM); Al Rashdan, Ahmad (Texas A&M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A&M University, College Station, TX); Vernon, Milton E. et al.
Partner: UNT Libraries Government Documents Department

Plasma-sprayed ceramic coatings for molten metal environments.

Description: Coating porosity is an important parameter to optimize for plasma-sprayed ceramics which are intended for service in molten metal environments. Too much porosity and the coatings may be infiltrated by the molten metal causing corrosive attack of the substrate or destruction of the coating upon solidification of the metal. Too little porosity and the coating may fail due to its inability to absorb thermal strains. This study describes the testing and analysis of tungsten rods coated with aluminum oxide, yttria-stabilized zirconia, yttrium oxide, and erbium oxide deposited by atmospheric plasma spraying. The samples were immersed in molten aluminum and analyzed after immersion. One of the ceramic materials used, yttrium oxide, was heat treated at 1000 C and 2000 C and analyzed by X-ray diffractography and mercury intrusion porosimetry. Slight changes in crysl nl structure and significant changes in porosity were observed after heat treatments.
Date: January 1, 2002
Creator: Hollis, K. J. (Kendall J.); Peters, M. I. (Maria I.) & Bartram, B. D. (Brian D.)
Partner: UNT Libraries Government Documents Department