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Distribution of fissionable material in thermal reactors of spherical geometry for uniform power generation

Description: Report presenting a solution for the two-group equations for reflected thermal reactors of spherical geometry for constant power generation over the reactor-core volume. Solution is obtained both analytically and by means of an electrical analogue simulator.
Date: June 1952
Creator: McCready, Robert R.; Spooner, Robert B. & Valerino, Michael F.
Partner: UNT Libraries Government Documents Department

Review of Current Experience on Intermediate Heat Exchanger (IHX) and A Recommended Code Approach

Description: The purpose of the ASME/DOE Gen IV Task 7 Part I is to review the current experience on various high temperature reactor intermediate heat exchanger (IHX) concepts. There are several different IHX concepts that could be envisioned for HTR/VHTR applications in a range of temperature from 850C to 950C. The concepts that will be primarily discussed herein are: (1) Tubular Helical Coil Heat Exchanger (THCHE); (2) Plate-Stamped Heat Exchanger (PSHE); (3) Plate-Fin Heat Exchanger (PFHE); and (4) Plate-Machined Heat Exchanger (PMHE). The primary coolant of the NGNP is potentially subject to radioactive contamination by the core as well as contamination from the secondary loop fluid. To isolate the radioactivity to minimize radiation doses to personnel, and protect the primary circuit from contamination, intermediate heat exchangers (IHXs) have been proposed as a means for separating the primary circuit of the NGNP (Next Generation Nuclear Plant) or other process heat application from the remainder of the plant. This task will first review the different concepts of IHX that could be envisioned for HTR/VHTR applications in a range of temperature from 850 to 950 C. This will cover shell-and-tube and compact designs (including the platefin concept). The review will then discuss the maturity of the concepts in terms of design, fabricability and component testing (or feedback from experience when applicable). Particular attention will be paid to the feasibility of developing the IHX concepts for the NGNP with operation expected in 2018-2021. This report will also discuss material candidates for IHX applications and will discuss specific issues that will have to be addressed in the context of the HTR design (thermal aging, corrosion, creep, creep-fatigue, etc). Particular attention will be paid to specific issues associated with operation at the upper end of the creep regime.
Date: February 2, 2010
Creator: Spencer, Duane & McCoy, Kevin
Partner: UNT Libraries Government Documents Department

Thorium: Uranium fuel cycle in safe reactors, the time is now

Description: The thorium-uranium fuel cycle has several advantages that make it attractive. Some of these beneficial properties are of particular interest now as they help alleviate current concerns. The Th-U cycle has neutronic advantages when utilized in thermal or epithermal reactors. Some of these reactors enjoy extraordinary safety qualities. The combination of these traits suggest that now is an appropriate time to deploy and begin exploiting the Th-U fuel cycle.
Date: December 31, 1995
Creator: Gat, Uri
Partner: UNT Libraries Government Documents Department

Update and Improve Subsection NH - Simplified Elastic and Inelastic Design Analysis Methods

Description: The objective of this subtask is to develop a template for the 'Ideal' high temperature design Code, in which individual topics can be identified and worked on separately in order to provide the detail necessary to comprise a comprehensive Code. Like all ideals, this one may not be attainable as a practical matter. The purpose is to set a goal for what is believed the 'Ideal' design Code should address, recognizing that some elements are not mutually exclusive and that the same objectives can be achieved in different way. Most, if not all existing Codes may therefore be found to be lacking in some respects, but this does not mean necessarily that they are not comprehensive. While this subtask does attempt to list the elements which individually or in combination are considered essential in such a Code, the authors do not presume to recommend how these elements should be implemented or even, that they should all be implemented at all. The scope of this subtask is limited to compiling the list of elements thought to be necessary or at minimum, useful in such an 'Ideal' Code; suggestions are provided as to their relationship to one another. Except for brief descriptions, where these are needed for clarification, neither this repot, nor Task 9 as a whole, attempts to address details of the contents of all these elements. Some, namely primary load limits (elastic, limit load, reference stress), and ratcheting (elastic, e-p, reference stress) are dealt with specifically in other subtasks of Task 9. All others are merely listed; the expectation is that they will either be the focus of attention of other active DOE-ASME GenIV Materials Tasks, e.g. creep-fatigue, or to be considered in future DOE-ASME GenIV Materials Tasks. Since the focus of this Task is specifically approximate methods, the authors ...
Date: June 27, 2009
Creator: Abou-Hanna, Jeries J.; Marriott, Douglas L. & McGreevy, Timothy E.
Partner: UNT Libraries Government Documents Department

What can Recycling in Thermal Reactors Accomplish?

Description: Thermal recycle provides several potential benefits when used as stop-gap, mixed, or backup recycling to recycling in fast reactors. These three roles involve a mixture of thermal and fast recycling; fast reactors are required to some degree at some time. Stop-gap uses thermal reactors only until fast reactors are adequately deployed and until any thermal-recycle-only facilities have met their economic lifetime. Mixed uses thermal and fast reactors symbiotically for an extended period of time. Backup uses thermal reactors only if problems later develop in the fast reactor portion of a recycling system. Thermal recycle can also provide benefits when used as pure thermal recycling, with no intention to use fast reactors. However, long term, the pure thermal recycling approach is inadequate to meet several objectives.
Date: September 1, 2007
Creator: Piet, Steven; Matthern, Gretchen E. & Jacobson, Jacob J.
Partner: UNT Libraries Government Documents Department

Current Comparison of Advanced Nuclear Fuel Cycles

Description: This paper compares potential nuclear fuel cycle strategies – once-through, recycling in thermal reactors, sustained recycle with a mix of thermal and fast reactors, and sustained recycle with fast reactors. Initiation of recycle starts the draw-down of weapons-usable material and starts accruing improvements for geologic repositories and energy sustainability. It reduces the motivation to search for potential second geologic repository sites. Recycle in thermal-spectru
Date: April 1, 2007
Creator: Piet, Steven; Bjornard, Trond; Dixon, Brent; Hill, Robert; Matthern, Gretchen & Shropshire, David
Partner: UNT Libraries Government Documents Department

Uncertainties in the analysis of plutonium fueled light water moderated assemblies

Description: A theoretical analysis of UO/sub 2/-- PuO/sub 2/ fueled, light-water- moderated lattice experiments has been performed to aid in establishing technical bases and design criteria for the utilization of plutonium bearing fuel in thermal power reactors. Results for UO/sub 2/ and Al-- Pu lattices are included in order to understand the effects due to uranium and plutonium separately. The problems involved in calculating high leakage critical experiments are discussed. Estimates of the effects of various approximations inherent in the theoretical methods and/or analysis procedures are included along with the consequence on the results of the correlation. Uncertainties in the measurements and the neutron crosssection data are related to uncertainties in the calculated values K/sub eff/ .The results of other studies which bear on evaluating the calculational methods are summarized. Areas which should be investigated in future analyses are also identified. (111 references) (auth)
Date: May 1, 1973
Creator: Liikala, R.C.; Uotinen, V.O. & Jenquin, U.P.
Partner: UNT Libraries Government Documents Department

HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

Description: This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication.
Date: November 1, 2010
Creator: Piet, Steven J.; Bays, Samuel E. & Soelberg, Nick R.
Partner: UNT Libraries Government Documents Department

A users guide for the REBUS-PC code, version 1.4.

Description: The Reduced Enrichment Research and Test Reactor (RERTR) Program uses the REBUS-PC computer code to provide reactor physics and core design information such as neutron flux distributions in space, energy, and time, and to track isotopic changes in fuel and neutron absorbers with burnup. REBUS-PC has evolved away from the original REBUS code, which was created starting in the 1960's to study large liquid metal cooled fast breeder reactors. REBUS and REBUS-PC both model the external cycle, and are very general codes with 1D, 2D, and 3D neutronics capabilities, and with complete fuel shuffling capabilities. REBUS-PC has evolved to its present status over the past decade. While it incorporates the same neutronics capabilities from DIF3D 9.0 as does REBUS 9.0 created by the RAE Division of ANL, REBUS-PC has numerous changes and enhancements directed toward the needs of the thermal reactor analyst using WINDOWS or linux-based PC's.
Date: January 30, 2002
Creator: Olson, A. P.
Partner: UNT Libraries Government Documents Department

Power Systems Development Facility. First quarterly report, 1997

Description: The objective of this project, herein referred to as the Power Systems Development Facility (PSDF), is to evaluate hot gas particle control technologies using coal derived gas streams. This project entails the design, construction, installation, and use of a flexible test facility which can operate under realistic gasification and combustion conditions. The major particulate control device (PCD) issues to be addressed include the integration of the PCDs into coal utilization systems, on-line cleaning techniques, chemical and thermal degradation of components, fatigue or structural failures, blinding, collection efficiency as a function of particle size, and scale-up of particulate control systems to commercial size.
Date: July 1, 1997
Partner: UNT Libraries Government Documents Department

Neutron cross sections for uranium-235 (ENDF/B-VI release 3)

Description: The resonance parameters in ENDF6 (Release 2) U235 were adjusted to make the average capture and fission cross sections below 900 eV agree with selected differential capture and fission measurements. The measurements chosen were the higher of the credible capture measurements and the lower of the fission results, yielding a higher epithermal alpha. In addition, the 2,200 m/s cross sections were adjusted to obtain agreement with the integral value of K1. As a result, criticality calculations for thermal benchmarks, and agreement with a variety of integral parameters, are improved.
Date: May 1, 1996
Creator: Lubitz, C.
Partner: UNT Libraries Government Documents Department

Accuracy of the Quasistatic Method for Two-Dimensional Thermal Reactor Transients with Feedback

Description: An important aspect in the design and safe operation of a nuclear reactor is the behavior of a reactor in a transient, or nonsteady state, condition. This study shows that the quasistatic method is capable of producing highly accurate results, relative to the direct finite-difference method, for two-dimensional thermal reactor transients with feedback.
Date: October 23, 2001
Creator: Dodds, H.L. Jr.
Partner: UNT Libraries Government Documents Department

Measurement and accounting of the minor actinides produced in nuclear power reactors

Description: Because of their value as nuclear fuels and their impact on long-term storage of high-level radioactive waste, measurement and accounting for minor actinides (MAs) produced in nuclear power reactors are becoming significant issues. This report attempts to put the issues in perspective by reviewing the commercial nuclear fuel cycle with emphasis on reprocessing plants and key measurement points therein. Radiation signatures and characteristics are compared and contrasted for special nuclear materials (SNMs) and MAs. Also, inventories and relative amounts of SNMs and MAs are generally described for irradiated nuclear fuel and reprocessing plants. The bulk of the report describes appropriate measurement technologies, capabilities, and development needs to satisfy material accounting requirements for MAs, with emphasis on adaptation of current technologies. Recommendations for future systems studies and development of measurement methods are also included. 38 refs., 3 figs., 12 tabs.
Date: January 1, 1996
Creator: Stewart, J.E.; Walton, R.B.; Phillips, J.R.; Hsue, S.T.; Eccleston, G.W.; Menlove, H.O. et al.
Partner: UNT Libraries Government Documents Department

Isolation of Metals from Liquid Wastes: Reactive Scavenging in Turbulent Thermal Reactors

Description: The Overall project demonstrated that toxic metals (cesium Cs and strontium Sr) in aqueous and organic wastes can be isolated from the environment through reaction with kaolinite based sorbent substrates in high temperature reactor environments. In addition, a state-of-the art laser diagnostic tool to measure droplet characteristic in practical 'dirty' laboratory environments was developed, and was featured on the cover of a recent edition of the scientific journal ''applied Spectroscopy''. Furthermore, great strides have been made in developing a theoretical model that has the potential to allow prediction of the position and life history of every particle of waste in a high temperature, turbulent flow field, a very challenging problem involving as it does, the fundamentals of two phase turbulence and of particle drag physics.
Date: August 6, 2003
Creator: Wendt, Jost O. L.; Kerstein, Alan R.; Scheeline, Alexander; Pearlstein, Arne & Linak, William
Partner: UNT Libraries Government Documents Department

Resonance region neutronics of unit cells in fast and thermal reactors

Description: A method has been developed for generating resonance-self-shielded cross sections based upon an improved equivalence theorem, which appears to allow extension of the self-shielding-factor (Bondarenko f-factor) method, now mainly applied to fast reactors, to thermal reactors as well. The method is based on the use of simple prescriptions for the ratio of coolant-to-fuel region-averaged fluxes, in the equations defining cell averaged cross sections. Linearization of the dependence of these functions on absorber optical thickness is found to be a necessary and sufficient condition for the existence of an equivalence theorem. Results are given for cylindrical, spherical and slab geometries. The functional form of the flux ratio relations is developed from theoretical considerations, but some of the parameters are adjusted to force-fit numerical results. Good agreement over the entire range of fuel and coolant optical thicknesses is demonstrated with numerical results calculated using the ANISN program in the S/sub 8/P/sub 1/ option.
Date: May 1, 1977
Creator: Salehi, A.A.; Driscoll, M.J. & Deutsch, O.L.
Partner: UNT Libraries Government Documents Department

Benchmark data testing for thermal-reactor applications

Description: The purpose of the thermal reactor data testing effort is to determine the adequacy of basic cross section data in thermal reactor applications, and particularly for beginning of life criticality calculations. This data testing effort has: pointed up inadequacies in Evaluated Nuclear Data (ENDF/B) sets, and thus has stimulated new differential data measurements and improved evaluations; provided standard calculated results against which approximate or newly developed accurate methods can be tested; identified inadequacies in Integral Experiments; and identified problems in specific calculational methods.
Date: January 1, 1981
Creator: Rose, P.F. & Pearlstein, S.
Partner: UNT Libraries Government Documents Department

Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors - methods

Description: Simple diffusion theory cannot be used to evaluate control rod worths in thermal neutron reactors because of the strongly absorbing character of the control material. However, reliable control rod worths can be obtained within the framework of diffusion theory if the control material is characterized by a set of mesh-dependent effective diffusion parameters. For thin slab absorbers the effective diffusion parameters can be expressed as functions of a suitably-defined pair of blackness coefficients. Methods for calculating these blackness coefficients in the P/sub 1/, P/sub 3/, and P/sub 5/ approximations, with and without scattering, are presented. For control elements whose geometry does not permit a thin slab treatment, other methods are needed for determining the effective diffusion parameters. One such method, based on reaction rate ratios, is discussed.
Date: January 1, 1984
Creator: Bretscher, M.M.
Partner: UNT Libraries Government Documents Department

Projections of ENDF/B version V performance for fast and thermal reactors using sensitivity coefficients

Description: Proposed reductions to /sup 235/U(anti ..nu..) and /sup 235/U(n,f) in the fast energy range have significant impact for uranium fueled fast critical assemblies. The long-standing LMFBR /sup 28/c//sup 49/f calculated overprediction is not resolved by proposed Version 5 cross section modifications for /sup 238/U(n,..gamma..) and /sup 239/Pu(n,f). The upward evaluation for the /sup 239/Pu(n,f)//sup 235/U(n,f) ratio improves criticality predictions for Pu fueled fast assemblies. For thermal reactors, changes to the /sup 238/U resonance parameters significantly reduce the long-standing /sup 28/rho discrepancy. Reduced resonance capture in the 1 eV /sup 240/Pu resonsnce has significant implications for LWR fuel cycle studies.
Date: January 1, 1978
Creator: Weisbin, C.R.; Marable, J.H.; Hardy, J. Jr. & McKnight, R.D.
Partner: UNT Libraries Government Documents Department

FY2012 summary of tasks completed on PROTEUS-thermal work.

Description: PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targeted reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference ...
Date: June 6, 2012
Creator: Lee, C.H. & Smith, M.A. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

Description: This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.
Date: August 1, 2010
Creator: Piet, Steven J.; Bays, Samuel E. & Soelberg, Nick
Partner: UNT Libraries Government Documents Department

Update and Improve Subsection NH –– Alternative Simplified Creep-Fatigue Design Methods

Description: This report described the results of investigation on Task 10 of DOE/ASME Materials NGNP/Generation IV Project based on a contract between ASME Standards Technology, LLC (ASME ST-LLC) and Japan Atomic Energy Agency (JAEA). Task 10 is to Update and Improve Subsection NH -- Alternative Simplified Creep-Fatigue Design Methods. Five newly proposed promising creep-fatigue evaluation methods were investigated. Those are (1) modified ductility exhaustion method, (2) strain range separation method, (3) approach for pressure vessel application, (4) hybrid method of time fraction and ductility exhaustion, and (5) simplified model test approach. The outlines of those methods are presented first, and predictability of experimental results of these methods is demonstrated using the creep-fatigue data collected in previous Tasks 3 and 5. All the methods (except the simplified model test approach which is not ready for application) predicted experimental results fairly accurately. On the other hand, predicted creep-fatigue life in long-term regions showed considerable differences among the methodologies. These differences come from the concepts each method is based on. All the new methods investigated in this report have advantages over the currently employed time fraction rule and offer technical insights that should be thought much of in the improvement of creep-fatigue evaluation procedures. The main points of the modified ductility exhaustion method, the strain range separation method, the approach for pressure vessel application and the hybrid method can be reflected in the improvement of the current time fraction rule. The simplified mode test approach would offer a whole new advantage including robustness and simplicity which are definitely attractive but this approach is yet to be validated for implementation at this point. Therefore, this report recommends the following two steps as a course of improvement of NH based on newly proposed creep-fatigue evaluation methodologies. The first step is to modify the current approach by ...
Date: October 26, 2009
Creator: Asayama, Tai
Partner: UNT Libraries Government Documents Department