A Phased Development of Breed-And-Burn Reactors for Enhanced Nuclear Energy Sustainability
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Doe Nuclear Physics Reactor Theory Handbook
DOE-HDBK-1019/2-93 JANUARY 1993 DOE FUNDAMENTALS HANDBOOK NUCLEAR PHYSICS AND REACTOR THEORY Volume 2 of 2 U.S. Department of Energy FSC-6910 Washington, D.C. 20585 Distribution Statement A. Approved for public release; distribution is unlimited. This document has been reproduced directly from the best available copy. Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831. Available to the public from the National Technical Information Service, U.S. Department of Commerce, 5285 Port Royal., Springfield, VA 22161. Order No. DE93012223 DOE-HDBK-1019/1-93 NUCLEAR PHYSICS AND REACTOR THEORY ABSTRACT The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance. Key Words: Training Material, Atomic Physics, The Chart of the Nuclides, Radioactivity, Radioactive Decay, Neutron Interaction, Fission, Reactor Theory, Neutron Characteristics, Neutron Life Cycle, Reactor Kinetics Rev. 0 NP DOE-HDBK-1019/1-93 NUCLEAR PHYSICS AND REACTOR THEORY FOREWORD The Department of Energy (DOE) Fundamentals Handbooks consist of ten academic subjects, which include Mathematics; Classical Physics; Thermodynamics, Heat Transfer, and Fluid Flow; Instrumentation and Control; Electrical Science; Material Science; Mechanical Science; Chemistry; Engineering Symbology, Prints, and Drawings; and Nuclear Physics and Reactor Theory. -
Vver and Rbmk Cross Section Libraries for Origen-Arp
VVER AND RBMK CROSS SECTION LIBRARIES FOR ORIGEN-ARP Germina Ilas, Brian D. Murphy, and Ian C. Gauld, Oak Ridge National Laboratory, USA Introduction An accurate treatment of neutron transport and depletion in modern fuel assemblies characterized by heterogeneous, complex designs, such as the VVER or RBMK assembly configurations, requires the use of advanced computational tools capable of simulating multi-dimensional geometries. The depletion module TRITON [1], which is part of the SCALE code system [2] that was developed and is maintained at the Oak Ridge National Laboratory (ORNL), allows the depletion simulation of two- or three-dimensional assembly configurations and the generation of burnup-dependent cross section libraries. These libraries can be saved for subsequent use with the ORIGEN-ARP module in SCALE. This later module is a faster alternative to TRITON for fuel depletion, decay, and source term analyses at an accuracy level comparable to that of a direct TRITON simulation. This paper summarizes the methodology used to generate cross section libraries for VVER and RBMK assembly configurations that can be employed in ORIGEN-ARP depletion and decay simulations. It briefly describes the computational tools and provides details of the steps involved. Results of validation studies for some of the libraries, which were performed using isotopic assay measurement data for spent fuel, are provided and discussed. Cross section libraries for ORIGEN-ARP Methodology The TRITON capability to perform depletion simulations for two-dimensional (2-D) configurations was implemented by coupling of the 2-D transport code NEWT with the point depletion and decay code ORIGEN-S. NEWT solves the transport equation on a 2-D arbitrary geometry grid by using an SN approach, with a treatment of the spatial variable that is based on an extended step characteristic method [3]. -
The Effect of Burnup and Separation Efficiency on Uranium Utilization and Radiotoxicity
INL/CON-10-20154 PREPRINT The Effect of Burnup and Separation Efficiency on Uranium Utilization and Radiotoxicity Eleventh Information Exchange Meeting on Partitioning and Transmutation Samuel Bays Steven Piet November 2010 This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party’s use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the United States Government or the sponsoring agency. THE EFFECT OF BURNUP AND SEPARATION EFFICIENCY ON URANIUM UTILIZATION AND RADIOTOXICITY Samuel Bays, Steven Piet Idaho National Laboratory, United States Abstract This paper addresses two fundamental cradle-to-grave issues of fuel cycle sustainability. The two primary issues of interest are efficient use of the natural uranium resource (cradle), and management of nuclear waste radiotoxicity (grave). Both uranium utilization and radiotoxicity are directly influenced by the burnup achieved during irradiation (transmutation related) and where applicable the separation efficiency (partitioning related). Burnup influences the in-growth of transuranics by breeding them into the fuel cycle. -
Nuclear Waste Transmutation in Subcritical Reactors Driven by Target-Distributed Accelerators
Nuclear Waste Transmutation in Subcritical Reactors Driven by Target-Distributed Accelerators Anatoly Blanovsky Teacher Technology Center, 7850 Melrose Ave., Los Angeles, CA 90046, USA E-mail [email protected] Abstract-A radioactive waste transmutation system based extensively on existing nuclear power technology is presented. By replacing the control rods with neutron generators, we could maintain good power distribution and perform long-lived waste burning in high flux subcritical reactors (HFSR). To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed. The design is based on a small pressurized water reactor (PWR), fission electric cell (FEC), target-distributed accelerator (TDA) and power monitoring system with in-core gamma-ray detectors, now under development in several countries. The FEC is essentially a high-voltage power source that directly converts the kinetic energy of the fission fragments into electrical potential of about 2MV. The TDA, in which an FEC electric field compensates for lost beam energy in the target, offers a new approach to obtain large neutron fluxes. INTRODUCTION I. HFSR CONCEPT In the conventional reactor, nuclear energy With self-amplifying due to feedback, (electrical in nature) is converted to thermal, then to blanket and booster multiplication factors of k=0.95 mechanical and then to electrical energy. To achieve and 0.98, respectively, an external neutron source the high degree of the burn-up, only fresh low rate of at least 10.sup.15 n/s is needed to drive the enriched uranium fuel is used. -
Nureg/Cr-6760 Ornl/Tm-2000-321
NUREG/CR- 6760 ORNL/TM-2000-321 Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit Oak Ridge National Laboratory U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, DC 20555-0001 AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS NRC Reference Material Non-NRC Reference Material technical As of November 1999, you may electronically access Documents available from public and special NUREG-series publications and other NRC records at libraries include all open literature items, such as NRC's Public Electronic Reading Room at books, journal articles, and transactions, Federal www.nrc.gov/NRC/ADAMS/index.html. Register notices, Federal and State legislation, and Publicly released records include, to name a few, congressional reports. Such documents as theses, NUREG-series publications; FederalRegister notices; dissertations, foreign reports and translations, and applicant, licensee, and vendor documents and non-NRC conference proceedings may be purchased organization. correspondence; NRC correspondence and internal from their sponsoring memoranda; bulletins and information notices; inspection and investigative reports; licensee event Copies of industry codes and standards used in a reports, and Commission papers and their substantive manner in the NRC regulatory process are attachments. maintained at The NRC Technical Library NRC publications in the NUREG series, NRC Two White Flint North regulations, and Title 10, Energy, in the Code of 11545 Rockville Pike Federal Regulations may also be purchased from one Rockville, MD 20852-2738 of these two sources. 1. The Superintendent of Documents These standards are available in the library for are U.S. Government Printing Office reference use by the public. -
Load Following with a Passive Reactor Core Using the SPARC Design
UPTEC ES 16 023 Examensarbete 30 hp 13 Juni 2016 Load following with a passive reactor core using the SPARC design Sebastian Leo Eile Svanström Abstract Load following with a passive reactor core using the SPARC design Sebastian Leo Eile Svanström Teknisk- naturvetenskaplig fakultet UTH-enheten This thesis is a follow up on "SPARC fast reactor design: Design of two passively metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Besöksadress: Reactivity Control" by Tobias Lindström (2015). In this thesis the two reactors Ångströmlaboratoriet Lägerhyddsvägen 1 designed by Lindström in said thesis were evaluated. The goal was to determine the Hus 4, Plan 0 reactors ability to load follow as well as the burnup of the neutron absorber used in the passive control system. Postadress: Box 536 751 21 Uppsala To be able to determine the dynamic behaviour of the reactors the reactivity feedbacks of the cores were modelled using Serpent, a Monte Carlo simulation Telefon: software for 3D neutron transport calculations. These feedbacks were then 018 – 471 30 03 implemented into a dynamic simulation of the core, primary and secondary circulation Telefax: and steam generator. The secondary circulation and feedwater flow were used to 018 – 471 30 00 regulate steam temperature and turbine power. The core was left at constant coolant flow and no control rods were used. Hemsida: http://www.teknat.uu.se/student The simulations showed that the reactor was able to load follow between 100 % and 40 % of rated power at a speed of 6 % per minute. It was also shown that the reactor could safely adjust its power between 100 % and 10 % of rated power suggesting that load following is possible below 40 % of rated power but at a lower speed. -
Concept of an Accelerator-Driven Advanced Nuclear Energy System
Article Concept of an Accelerator-Driven Advanced Nuclear Energy System Xuesong Yan, Lei Yang *, Xunchao Zhang and Wenlong Zhan Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000, China; [email protected] (X.Y.); [email protected] (X.Z.); [email protected] (W.Z.) * Correspondence: [email protected]; Tel.: +86-931-4969-187 Academic Editor: Hiroshi Sekimoto Received: 24 March 2017; Accepted: 10 May 2017; Published: 7 July 2017 Abstract: The utilization of clean energy is a matter of primary importance for sustainable development as well as a vital approach for solving worldwide energy-related issues. If the low utilization rate of nuclear fuel, nuclear proliferation, and insufficient nuclear safety can be solved, nuclear fission energy could be used as a sustainable and low-carbon clean energy form for thousands of years, providing steady and base-load electrical resources. To address these challenges, we propose an accelerator-driven advanced nuclear energy system (ADANES), consisting of a burner system and a fuel recycle system. In ADANES, the ideal utilization rate of nuclear fuel will be >95%, and the final disposal of nuclear waste will be minimized. The design of a high-temperature ceramic reactor makes the burner system safer. Part of fission products (FPs) are removed during the simple reprocessing in the fuel recycle system, significantly reducing the risks of nuclear proliferation of nuclear technology and materials. The ADANES concept integrates nuclear waste transmutation, nuclear fuel breeding, and safety power production, with an ideal closed loop operation of nuclear fission energy, constituting a major innovation of great potential interest for future energy applications. -
Nuclear Fuel Cycles Technology Assessment
Clean Power Quadrennial Technology Review 2015 Chapter 4: Advancing Clean Electric Power Technologies Technology Assessments Advanced Plant Technologies Biopower Carbon Dioxide Capture and Storage Clean Power Value-Added Options Carbon Dioxide Capture for Natural Gas and Industrial Applications Carbon Dioxide Capture Technologies Carbon Dioxide Storage Technologies Crosscutting Technologies in Carbon Dioxide Capture and Storage Fast-spectrum Reactors Geothermal Power High Temperature Reactors Hybrid Nuclear-Renewable Energy Systems Hydropower Light Water Reactors Marine and Hydrokinetic Power Nuclear Fuel Cycles Solar Power Stationary Fuel Cells Supercritical Carbon Dioxide Brayton Cycle U.S. DEPARTMENT OF Wind Power ENERGY Clean Power Quadrennial Technology Review 2015 Nuclear Fuel Cycles Chapter 4: Technology Assessments Introduction and Background The Nuclear Fuel Cycle (NFC) is defined as the total set of operations required to produce fission energy and manage the associated nuclear materials. It can have different attributes, including the extension of natural resources, or the minimization of waste disposal requirements. The NFC, as depicted in Figure 4.O.1, is comprised of a set of operations that include the extraction of uranium (U) resources from the earth (and possibly from seawater), uranium enrichment and fuel fabrication, use of the fuel in reactors, interim storage of used nuclear fuel, the optional recycle of the used fuel, and the final disposition of used fuel and waste forms from the recycling processes. Thorium (Th) fuel cycles have been proposed also, but have not been commercially implemented). Figure 4.O.1 Schematic of the uranium based Nuclear Fuel Cycle 1 Quadrennial Technology Review 2015 Clean Power TA 4.O: Nuclear Fuel Cycles The nuclear fuel cycle is often grouped into three classical components (front-end, reactor, and back-end): Front End: The focus of the front end of the nuclear fuel cycle is to deliver fabricated fuel to the reactor. -
Fuel Burnup Analysis of the TRIGA Mark II Reactor at the University of Pavia
Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia Davide Chiesaa,b, Massimiliano Clemenzaa,b, Stefano Pozzia,b, Ezio Previtalia,b, Monica Sistia,b, Daniele Allonic,d, Giovanni Magrottic,d, Sergio Manerac,d, Michele Pratac,d, Andrea Salvinic,d, Antonio Cammie,b, Matteo Zanettie, Alberto Sartorif aPhysics Department \G. Occhialini" of Milano-Bicocca University, piazza della Scienza 3, 20126 Italy bINFN section of Milano-Bicocca, piazza della Scienza 3, 20126 Italy cLaboratorio Energia Nucleare Applicata (L.E.N.A.) of the University of Pavia, via Aselli 41, Pavia 27100, Italy dINFN section of Pavia, via A. Bassi 6, Pavia 27100, Italy ePolitecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano, Italy fSISSA mathLab, International School for Advanced Studies, Via Bonomea 265, 34136 Trieste, Italy Abstract A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this pre- diction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to anal- yse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specif- ically to keep into account some features that differentiate experimental re- actors from power ones, such as the daily on/off cycle and the long fuel lifetime. -
A Fission-Fusion Hybrid Reactor in Steady-State L-Mode Tokamak Configuration with Natural Uranium
PSFC/RR-11-1 A Fission-Fusion Hybrid Reactor in Steady-State L-Mode Tokamak Configuration with Natural Uranium Reed, M., Parker, R., Forget, B.* *MIT Department of Nuclear Science & Engineering January 2011 Plasma Science and Fusion Center Massachusetts Institute of Technology Cambridge MA 02139 USA A Fission-Fusion Hybrid Reactor in Steady-State L-mode Tokamak Configuration with Natural Uranium Mark Reed, Ronald R. Parker, Benoit Forget Massachusetts Institute of Technology January 2011 Abstract The most prevalent criticism of fission-fusion hybrids is simply that they are too exotic - that they would exacerbate the challenges of both fission and fusion. This is not really true. Intriguingly, hybrids could actually be more viable than stand-alone fusion reactors while mitigating many challenges of fission. This work develops a conceptual design for a fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium- lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. Subcritical operation could obviate the most challenging fuel cycle aspects of fission. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of 7.7. -
The Burnup Dependence of Light Water Reactor Spent Fuel Oxidation
GNA bNLI-I MOL.19980810.0383 PNNL-11 929 Nationail Labolratory Opeate Batelb fornimthe"• The Burnup Dependence of Light Water Reactor Spent Fuel Oxidation B. D. Hanson July 1998 z"D z co Prepared for the U.S. Department of Energy under Contract DE-AC06-76RLO 1830 ' DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor Battelle Memorial Institute, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof, or Battelle Memorial Institute. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. PACIFIC NORTHWEST NATIONAL LABORATORY operated by BATTEELLE for the UNITED STATES DEPARTMENT OF ENERGY under ContractDE-ACO6-76RLO 1830 Printed in the United States of America Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; prices available from (615) 576-8401. Available to the public from the National Technical Information Service, U.S. Department of Commerce, 5285 Port Royal Rd., Springfield, VA 22161 This document was printed on recycled paper. -
Part Twenty-Nine Reactor Operation with Feedback Effects
22.05 Reactor Physics - Part Twenty-Nine Reactor Operation with Feedback Effects 1. Reference Material: See pp. 368 – 372 in “Light Water Reactor Control Systems,” in Wiley Encyclopedia of Electrical and Electronics Engineering. 2. Point of Adding Heat – Hot Operating Crucial concepts are: Types of power-dependent feedback — Coolant Temperature — Fuel Temperature (Doppler Effect) — Void Coefficient — Xenon/Samarium Time scales of each feedback mechanism Estimated Critical Position 3. Coolant Temperature As the temperature of the primary coolant increases, it becomes less dense. This causes: — Less neutron moderation — Increased neutron leakage Hence, as coolant temperature rises, negative reactivity is generated. This makes reactors self-regulating. Time scale is primary loop circuit time ~ 30 s. Negative temperature coefficients are required for all U.S., European, and Asian reactors. They provide a safety feature. But, under certain accidents they make the situation worse: — Steam line break — Control rod drop In both of these cases, the reactor cools off and the negative coefficient causes a positive reactivity insertion. 1 4. Void Coefficient The primary coolant in both PWRs and BWRs performs several functions. These are: — Removal of heat. — Moderation of neutrons so as to continue the neutron chain reaction. — Shielding during maintenance. If the coolant becomes less dense, it becomes less efficient as a moderator. One way to decrease coolant density is to decrease its temperature. Another is to create voids in it. This occurs in BWRs when steam bubbles form. BWRs can be controlled by adjusting the recirculation flow which in turn controls the rate at which voids are swept out of the core.