DOCSLIB.ORG

High-Temperature XAFS Measurement of Molten Salt Systems

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
High-Temperature XAFS Measurement of Molten Salt Systems GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1017 Study on High Performance MOX Fuel and Core Design in Full MOX ABWR(1) by GNF-J Sadayuki Izutsu1*, Daisuke Goto1, Jun Saeki1, Takehiro Kokubun1 and Jun Yokoya2 1Global Nuclear Fuel-Japan Co.,Ltd., Yokosuka, Kanagawa, 239-0836, Japan 2Electric Power Development Co.,Ltd., Tokyo, 104-8165, Japan The concepts of high-performance MOX fuel using 10x10 lattices suitable for full-MOX ABWR are shown in this paper, in which average discharge exposure is extended up to 45GWd/t with heavy-metal inventory increased over current MOX, reducing the number of refueling bundles, resulting in fuel cycle cost reduction and core performance satisfaction. Also, the increase of Pu inventory is taken into account from the viewpoint to extend the flexibility of MOX fuel utilization. KEYWORDS: High Performance MOX Fuel, Full MOX ABWR, 10x10 lattice, Discharge exposure I. Introduction for thermal margin, the power-exposure envelope can be set Discharge exposure is progressively extended for uranium lower than that of 9×9 uranium fuel by increase of the fuel used for BWR currently in operation, and thus fuel number of fuel rods, and therefore the active fuel length of cycle economy is being improved. In contrast, discharge MOX rods could be extended to 3.71m from 3.55m of exposure (33GWd/t) of current MOX fuel to be loaded in current MOX and Pu inventory can be increased. BWRs including ABWR as Pu-thermal utilization program Reprocessed material composition is assumed to be is less than that (45GWd/t) of uranium fuel by 25%. mixture of the reprocessed Pu isotope and U isotope. Besides, current MOX fuel bundle has slightly reduced heavy-metal inventory owing to the shortened active fuel Table 1 Basic specification of core and fuel design length compared to uranium fuel as the countermeasure to fuel rod internal pressure increase. Items Specification The concepts of high-performance MOX fuel suitable for Core full-MOX ABWR are shown in this paper, in which Type Advanced Boiling discharge exposure is extended up to that of current Water Reactor (ABWR) high-burnup uranium fuel with heavy-metal inventory Thermal power (MW) 3,926 increased over current MOX, reducing the number of Core flow (t/h) 52.2x103 refueling bundles, resulting in fuel cycle cost reduction and Core pressure (MPa) 7.17 core performance satisfaction. Also, the increase of Pu Number of fuel bundles 872 inventory is taken into account from the viewpoint to extend Number of control rods 205 the flexibility of MOX fuel utilization. Fuel Assembly Two types of lattice concepts are approached aiming at Lattice configuration 10x10 average discharge exposure of 45GWd/t on the basis of the Average discharge exposure 45 10x10 lattice configuration, that is, the high moderation (GWd/t) concept and the low moderation concept are possible. Peak pellet exposure (GWd/t) about 75 The channel, core, and regional stabilities are analyzed, Pellet material UO2-PuO2 (MOX rods) and satisfy design target. The SLMCPR is shown to be UO2-Gd2O3 (UO2 rods) almost equal to full MOX-fueled core loaded with current MOX matrix material Depleted U (0.22wt%) 8×8 MOX fuel. The ∆MCPR is estimated to be higher for Cladding outside diameter (mm) 10.3 the first type concept and lower for the second type concept Cladding wall thickness (mm) 0.66 compared to current MOX within the plant design margin. Cladding material Zirc-2 (with Zr liner) Operating Condition II. Design Conditions Cycle length (month) 13 (Equilibrium) High burnup MOX fuel designs of average discharge Reprocessed material exposure 45GWd/t have been studied for full-MOX ABWR composition 239 241 under the conditions shown in Table 1. Based on the Pu isotope composition Pu =61%,Pu =7% 235 10x10 lattices selected to improve performance, especially U isotope composition U =0.9% * Corresponding author, Tel. +81-468-33-9053, Fax. +81-468-33- 9248, E-mail: [email protected] III. Fuel and Core Design Concepts The current MOX design concept of discharge exposure High Burnup MOX 33GWd/t as the reference is shown in Fig.1, the (fuel economy enhancement) configuration of which is selected to be the same design as STEP-2 bundle (39.5GWd/t discharge exposure) with a large water rod at the center of 8x8 fuel rod configuration and Void Coefficient Gd Rod Fraction Fuel Rod Internal Increase Increase Pressure Increase with 12 Gd-contained UO2 rods. The MOX fuel rod is designed to make the fuel stack length shorter and the plenum volume larger than those of the UO fuel rod so as to 2 Peripheral Gd 10x10 suppress the rod internal pressure. W/R W/R addition minimization Rod Disposition Configuration The high burnup to increase the generation of energy per (Concept(I)) (Concept(II)) heavy metal weight is effective to enhance fuel economy due to the reduction of fabricated fuel bundles and spent fuel Fig. 2 High-burnup MOX design concept bundles. On the other hand, the high burnup to increase the Pu content has the tendency toward the absolute value increase of void coefficient and Gd worth reduction, causing Table 2 Main features of 10x10 MOX concepts severer transient result and Gd-contained UO rod fraction 2 Current Concept (I) Concept (II) increase. Also, it may make bundle heavy metal weight MOX (10x10 MOX) (10x10 MOX) reduce through the rod plenum volume increase against rod (8x8) internal pressure. Therefore, on the basis of the 10x10 bundle configuration Number of 1 (4) 2 (4) 1 (1) selected to reduce linear heat generation rate mitigating the water rod types 2 (1) rod internal pressure, two types of lattice concepts are (*) approached: the concept (I) where the water rods are added Number of 48 / 12 78 / 12 99 / 6 (High moderation concept) and the concept (II) where the MOX rods / water rods are minimized (Low moderation concept), in Gd rods Number of Pu 4 2 3 addition to the optimization of Gd-contained UO2 rod locations. Fig.2 shows the relation on the influence of high content rod burnup and the selected concepts. The main features of the types concepts are shown in Table 2. Fuel active 3.55 3.71 3.71 length (m) (same as UO2 ) (same as UO2 ) Heavy metal base + 5% + 16% inventory MOX Bundle Pu weight base x 1.6 x 2.2 24 * : Number of fuel rods corresponding to each W/R area 18 1.Concept (I) The configuration of concept (I) is shown in Fig.3. In order to avoid the absolute value increase of void coefficient, two large water rods with each area corresponding to four 12 fuel rods and two water rods with the same diameter as fuel rod are disposed in the central region of 10x10 bundle Axial Hight configuration, resulting in the moderator to heavy metal volume ratio of 3.02 similar to 3.07 of STEP-3(9x9) UO2 6 UO -Gd O fuel rod 2 2 3 fuel. Water rod The fundamental idea of nuclear design is that most MOX fuel rod Gd-contained UO2 rods are placed in the bundle peripheral region and the water rod neighboring region so as to reduce the Gd-contained rods and to increase Pu inventory. Fig.4 shows the relation between Gd rod worth and Gd-contained Fig. 1 Current MOX design concept rod location. The Gd rod worth placed in the bundle (33GWd/t discharge exposure ) peripheral region is 35% improved than that in the bundle internal region. It is possible to extend burnup exposure of MOX fuel without the decrease of MOX fuel rod fraction with Gd rod worth increase effect and power flattening effect by the Gd-contained rods placed in the bundle showing that the absolute value of void coefficient is smaller peripheral region. Two rod types of Pu content, the in the case of larger H/HM fuel. To the contrary, the fabrication cost reduction per unit Pu weight due to Pu moderator density coefficient for MOX fuel has a maximum inventory increase and uranium enrichment decrease are point and decreasing tendency in the both sides of H/HM, realized. Twelve partial fuel rods (PLR) are utilized to showing that the absolute value of void coefficient is smaller keep the same pressure loss as 8x8 or 9x9 bundles. The in the case of larger Pu content. This concept (II) utilizes average Puf content of the bundle is 4.2wt% and the the decreasing feature of moderator density coefficient in the maximum PuO2 content of the MOX rod is 10wt%, resulting low H/HM region, and the increasing feature of Pu and in the increase of 5% in heavy metal weight and x1.6 in Pu heavy metal inventory. weight compared to the current 8x8 MOX design. The fundamental idea of nuclear design is that Gd-contained UO2 rods are placed in the bundle corner G G G region so as to reduce the Gd-contained rods and to increase Low Puf content Pu inventory. Because the UO2 rods disposed in the bundle corner region contain depleted uranium without Pu, the part High Puf content G G of which contains Gd, the low enrichment of high neutron G PLR(Low Puf content) importance region contributes the reduction of void G coefficient and hot operation to cold reactivity swing. G G G Nat-U+Gd conc. Three types of Puf content, the fabrication cost reduction per unit Pu weight due to Pu inventory increase and uranium Water rod enrichment decrease are realized. Twelve partial fuel rods G G G (PLR) are utilized to keep the same pressure loss as 8x8 or 9x9 bundle.
Load more

Recommended publications
  • Institute for Clinical and Economic Review
    INSTITUTE FOR CLINICAL AND ECONOMIC REVIEW FINAL APPRAISAL DOCUMENT BRACHYTHERAPY & PROTON BEAM THERAPY FOR TREATMENT OF CLINICALLY-LOCALIZED, LOW-RISK PROSTATE CANCER December 22, 2008 Senior Staff Daniel A. Ollendorf, MPH, ARM Chief Review Officer Julia Hayes, MD Lead Decision Scientist Pamela McMahon, PhD Sr. Decision Scientist Steven D. Pearson, MD, MSc President, ICER Associate Staff Michelle Kuba, MPH Sr. Technology Analyst Angela Tramontano, MPH Research Assistant © ICER, 2008 1 CONTENTS About ICER .................................................................................................................................. 3 Acknowledgments ...................................................................................................................... 4 Executive Summary .................................................................................................................... 5 Evidence Review Group Deliberation.................................................................................. 15 ICER Integrated Evidence Rating.......................................................................................... 21 Evidence Review Group Members........................................................................................ 24 Appraisal Overview.................................................................................................................. 28 Background ...............................................................................................................................
    [Show full text]
  • 小型飛翔体／海外 ［Format 2］ Technical Catalog Category
    小型飛翔体／海外 ［Format 2］ Technical Catalog Category Airborne contamination sensor Title Depth Evaluation of Entrained Products (DEEP) Proposed by Create Technologies Ltd & Costain Group PLC １．DEEP is a sensor analysis software for analysing contamination. DEEP can distinguish between surface contamination and internal / absorbed contamination. The software measures contamination depth by analysing distortions in the gamma spectrum. The method can be applied to data gathered using any spectrometer. Because DEEP provides a means of discriminating surface contamination from other radiation sources, DEEP can be used to provide an estimate of surface contamination without physical sampling. DEEP is a real-time method which enables the user to generate a large number of rapid contamination assessments- this data is complementary to physical samples, providing a sound basis for extrapolation from point samples. It also helps identify anomalies enabling targeted sampling startegies. DEEP is compatible with small airborne spectrometer/ processor combinations, such as that proposed by the ARM-U project – please refer to the ARM-U proposal for more details of the air vehicle. Figure 1: DEEP system core components are small, light, low power and can be integrated via USB, serial or Ethernet interfaces. 小型飛翔体／海外 Figure 2: DEEP prototype software ２．Past experience (plants in Japan, overseas plant, applications in other industries, etc) Create technologies is a specialist R&D firm with a focus on imaging and sensing in the nuclear industry. Createc has developed and delivered several novel nuclear technologies, including the N-Visage gamma camera system. Costainis a leading UK construction and civil engineering firm with almost 150 years of history.
    [Show full text]
  • A 50-100 Kwe Gas-Cooled Reactor for Use on Mars
    SANDIA REPORT SAND2006-2189 Unlimited Release Printed April 2006 A 50-100 kWe Gas-cooled Reactor For Use On Mars Curtis D. Peters Prepared by Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550 Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under Contract DE-AC04-94AL85000. Approved for public release; further dissemination unlimited. Issued by Sandia National Laboratories, operated for the United States Department of Energy by Sandia Corporation. NOTICE: 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 any of their employees, nor any of their contractors, subcontractors, or their employees, make any warranty, express or implied, or assume any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represent 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, any agency thereof, or any of their contractors or subcontractors. The views and opinions expressed herein do not necessarily state or reflect those of the United States Government, any agency thereof, or any of their contractors. Printed in the United States of America. This report has been reproduced directly from the best available copy. Available to DOE and DOE contractors from U.S.
    [Show full text]
  • Advances in Small Modular Reactor Technology Developments
    Advances in Small Modular Reactor Technology Developments Advances in Small Modular Reactor Technology Developments Technology in Small Modular Reactor Advances A Supplement to: IAEA Advanced Reactors Information System (ARIS) 2018 Edition For further information: Nuclear Power Technology Development Section (NPTDS) Division of Nuclear Power IAEA Department of Nuclear Energy International Atomic Energy Agency Vienna International Centre PO Box 100 1400 Vienna, Austria Telephone: +43 1 2600-0 Fax: +43 1 2600-7 Email: [email protected] Internet: http://www.iaea.org Printed by IAEA in Austria September 2018 18-02989E ADVANCES IN SMALL MODULAR REACTOR TECHNOLOGY DEVELOPMENTS 2018 Edition A Supplement to: IAEA Advanced Reactors Information System (ARIS) http://aris.iaea.org DISCLAIMER This is not an official IAEA publication. The material has not undergone an official review by the IAEA. The views expressed do not necessarily reflect those of the International Atomic Energy Agency or its Member States and remain the responsibility of the contributors. Although great care has been taken to maintain the accuracy of information contained in this publication, neither the IAEA nor its Member States assume any responsibility for consequences which may arise from its use. The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA.
    [Show full text]
  • Self-Sustaining Thorium Boiling Water Reactors
    Project No. 11-3023 Self-Sustaining Thorium Boiling Water Reactors Reactor Concepts Dr. Ehud Greenspan University of California-Berkeley In collaboraon with: Massachuses Instute of Technology University of Michigan Brookhaven Naonal Laboratory Brian Robinson, Federal POC Temitope Taiwo, Technical POC NEUP Project # 11-3023 September 2011 to December 2014 Self-sustaining thorium boiling water reactors Summary Report Ehud Greenspan, Phillip M. Gorman, Sandra Bogetic, Jeffrey E. Seifried, Guanheng Zhang, Christopher R. Varela, Massimiliano Fratoni and Jasmina L. Vujic UC Berkeley, [email protected] Thomas Downar, Andrew Hall, Andrew Ward, Michael Jarrett, Aaron Wysocki and Yunlin Xu University of Michigan, [email protected] Mujid Kazimi, Koroush Shirvan and Alexander Mieloszyk MIT, [email protected] Michael Todosow, Nicholas Brown and Lap Cheng BNL, [email protected] 28 February 2015 1 Table of Contents Overview Page Executive summary 3 1. Introduction 7 2. The RBWR cores 7 2.1 Hitachi RBWR cores 7 2.2 Independent evaluation of the Hitachi designs 14 2.3 The incentive for thorium-based RBWR 15 2.4 The thorium-based RBWR core designs 17 3. Development of computational methods for RBWR type cores 19 3.1 3-D core simulator 20 3.2 Monte-Carlo based coupled neutronics – TH – depletion analysis capability 21 3.3 Stability and safety analysis capability 21 3.4 Fuel performance analysis capability 22 3.5 Thermal hydraulic correlations for tight lattice high void cores 24 4. Feasibility of fuel-self-sustaining RBWR-Th cores 24 4.1 Study strategy 24 4.2 RBWR-Th core designs 26 4.3 RBWR-Th safety and stability 30 4.4 RBWR-Th fuel performance 31 5.
    [Show full text]
  • Latest Fusion R&D Activities at INEST
    1 Latest Fusion R&D Activities at INEST Presented by Yican Wu (Zhibin Chen on behalf) Contributed by FDS Team Institute of Nuclear Energy Safety Technology (INEST) Chinese Academy of Sciences www.fds.org.cn FDS: Frontier Development of Science 2 Institute of Nuclear Energy Safety Technology Chinese Academy of Sciences ∙ FDS Team Main Research Areas Fundamental research: • Neutron Physics Applied research: 1. Advanced Fission Energy LFR etc. 2. Fusion Energy FNT & Materials etc. Headquarter in Hefei, Anhui, China 3. Extended Nuclear Technology Applications Key Programs: . Strategic Priority Research Program of CAS . ITER Related International and Domestic Program . Nuclear Energy & Safety Technology Innovation Program 400 Staff + 100 Students Three Integrated Test Reactors A professional institute is devoted to design and R&D of advanced nuclear energy and safety technology 3 Outline I. Highlights of Fusion R&D Activities • Fusion Neutron Sources • Neutronics Methodology and Simulation • Fusion Safety • Materials and TBM Technologies II. International Communication and Collaboration 4 1. Fusion Neutron Sources 5 Development of Fusion Neutron Sources at INEST ≥ 1018 n/ s 15 16 10 -10 n/s HINEG-IIIa 1012-1013 n/s Fusion Volumetric Neutron HINEG-IIa Source ( e.g. GDT-FVNS ) Mixed-beam HINEG-I Fusion Neutron Source (in Operation) . Fusion materials (≥20dpa/FPY) . Materials Irradiation . Multi-Physics Coupling Test . Nuclear Design and Performance . Nuclear Data . System Engineering Integration Test of Component Test . Methods and Software
    [Show full text]
  • Control Rod Safety Management in Nuclear Power Plant
    Journal of Advanced Review on Scientific Research 32, Issue 1 (2017) 14-27 Journal of Advanced Review on Scientific Penerbit Akademia Baru Research Journal homepage: www.akademiabaru.com/arsr.html ISSN: 2289-7887 Control rod safety m anage ment in nuclear power plant: A Open review Access Muhammad Adil Khattak 1, ∗, Filzah Hazirah Jaffar 1,Najaa Fadhilah Mohd Nasir 1, Nurfarhah 1 1 1 Ridzuan , Nurlaila Syamsul Bahri ,Wan Nur Iffah Haziqah Md Zamri 1 Department of Nuclear Engineering, Faculty of Chemical & Energy Engineering Universiti Teknologi Malaysia, 81310 Skudai Johor , Malaysia ARTICLE INFO ABSTRACT Article history: The primary objective of control rod management is to ensure the safe, reliable and Received 12 April 2017 optimum use of the nuclear fuel in the reactor, while remain within the limits imposed Received in revised form 24 May 2017 by the design of the fuel assembly and reactor w.r.t the safety analysis. In numerous Accepted 25 May 2017 reactors, the control rods perform the function of reactivity control, both globally and Available online 26 May 2017 locally, latter also control the power distribution of the core. Most control rods are completely withdrawn from the core during operations and fully inserted during shutdown of the reactor. This are the prove of an investigative study into optimization of the heterogeneous control rods, the safety management of an additional safety rod, and control rod drop hydrodynamic analysis which will be able to provide an efficient and maximum safety procedure for emergency shut down system in the reactor. It is also will be very important safety features of the reactor.
    [Show full text]
  • NRC Collection of Abbreviations
    I Nuclear Regulatory Commission c ElLc LI El LIL El, EEELIILE El ClV. El El, El1 ....... I -4 PI AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources: 1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2. The Superintendent of Documents, U.S. Government Printing Office, P. 0. Box 37082, Washington, DC 20402-9328 3. The National Technical Information Service, Springfield, VA 22161-0002 Although the listing that follows represents the majority of documents cited in NRC publica- tions, it is not intended to be exhaustive. Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu- ments and correspondence. The following documents in the NUREG series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro- ceedings, international agreement reports, grantee reports, and NRC booklets and bro- chures. Also available are regulatory guides, NRC regulations in the Code of Federal Regula- tions, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Information Service Include NUREG-series reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. Documents available from public and special technical libraries include all open literature items, such as books, journal articles, and transactions.
    [Show full text]
  • Irradiation Performance of Fast Reactor Fuels
    Oktober 1967 KFK 662 EUR 3698 e Institut für Angewandte Reaktorphysik Institut für Material- und Festkörperforschung Irradiation Performance of Fast Reactor Fuels D. Geithoff, G. Karsten, K. Kummerer GESELLSCHAFT FUR KERNF( KARLSRUHE Als Manuskript vervielfältigt Für diesen Bericht behalten wir uns alle Rechte vor Gesellschaft für 'Kernforschung m.b.H. Karlsruhe KERNFORSCHUNGSZENTRUM KARLSRUHE Oktober I967 KFK­662 EUR 3698 e Institut für Angewandte Reaktorphysik Institut für Material­ und Festkörperforschung IRRADIATION PERFORMANCE OF FAST REACTOR FUELS *' D. Geithoff, G. Karsten and Κ. Kümmerer Paper presented at the Symposium on Plutonium Fuels Technology, Nuclear Metallurgy Committee IMD­AIME, Oct.4­6,1967, Phoenix, Ariz., USA. Gesellschaft für Kernforschung mbH., Karlsruhe *) Work performed within the association in the field of fast reactors between the European Atomic Energy Community and Gesellschaft für Kernforschung mbH., Karlsruhe ABSTRACT The requirements for uranium-plutonium dioxide fuels are the basis for a program of irradiation tests which is outlined in the introduction. Some hypothetical considerations supply a rational background for the design of specimens and the chosen operation conditions. The experimental facilities and irradiation devices are shortly described and discussed with respect to the irradiation characteristics. The available hot cell equipment provides for the standard examination technique including X-raying by betatron and fission gas chromatography. One group of experiments refers to the short term behaviour of the oxide type fuel. In another group the long time behaviour of oxide fuel pins is studied. Different pins have been irradiated in capsules and in a helium loop. The present results indicate, that the basic assumptions for the pin design are principally con­ firmed, the specimen failures only arising from malfunctions of the irradiation devices.
    [Show full text]
  • Understanding Non-Hodgkin Lymphoma
    Understanding Non-Hodgkin Lymphoma A Guide for Patients, Survivors, and Loved Ones July 2018 Lymphoma Research Foundation (LRF) Helpline and Clinical Trials Information Service CONTACT THE LRF HELPLINE Trained staff are available to answer questions and provide support to patients, caregivers and healthcare professionals in any language. Our support services include: • Information on lymphoma, treatment options, side effect management and current research fi ndings • Financial assistance for eligible patients and referrals for additional fi nancial, legal and insurance help • Clinical trial searches based on patient’s diagnosis and treatment history • Support through LRF’s Lymphoma Support Network, a national one-to one volunteer patient peer program Monday through Friday, Toll-Free (800) 500-9976 or email [email protected] Understanding Non-Hodgkin Lymphoma A Guide For Patients, Survivors, and Loved Ones July 2018 This guide is an educational resource compiled by the Lymphoma Research Foundation to provide general information on adult non- Hodgkin lymphoma. Publication of this information is not intended to replace individualized medical care or the advice of a patient’s doctor. Patients are strongly encouraged to talk to their doctors for complete information on how their disease should be diagnosed, treated, and followed. Before starting treatment, patients should discuss the potential benefits and side effects of cancer therapies with their physician. Contact the Lymphoma Research Foundation Helpline: (800) 500-9976 [email protected] Website: lymphoma.org This patient guide is supported through unrestricted educational grants from: © 2018 Lymphoma Research Foundation. Information contained herein is the property of the Lymphoma Research Foundation (LRF). Any portion may be reprinted or reproduced provided that LRF is acknowledged to be the source and the Foundation’s website (lymphoma.org) is included in the citation.
    [Show full text]
  • Fundamentals of Boiling Water Reactor (Bwr)
    FUNDAMENTALS OF BOILING In this light a general description of BWR is presented, with emphasis on the reactor physics aspects (first lecture), and a survey of methods applied in fuel WATER REACTOR (BWR) and core design and operation is presented in order to indicate the main features of the calculation^ tools (second lecture). The third and fourth lectures are devoted to a review of BWR design bases, S. BOZZOLA reactivity requirements, reactivity and power control, fuel loading patternc. Moreover, operating 'innts are reviewed, as the actual limits during power AMN Ansaldo Impianti, operation and constraints for reactor physics analyses (design and operation). Genoa, The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality Italy and low power operation are illustrated in the fifth lecture. The banked position withdrawal sequence mode of operation is indeed an example of safety and design/operation combined approach. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact Abstract between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the These lectures on fundamentals of BWR reactor physics are a synthesis startup of a commercial BWR. A control .od calibration is also illustrated. of known and established concepts. These lecture are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation.
    [Show full text]
  • A Graphite Isotope Ratio Method Primer—A Method for Estimating Plutonium Production in Graphite Moderated Reactors
    PNNL-14568 A Graphite Isotope Ratio Method Primer—A Method for Estimating Plutonium Production in Graphite Moderated Reactors C. J. Gesh February 2004 Prepared for the U.S. Department of Energy under Contract DE-AC06-76RL01830 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 BATTELLE for the UNITED STATES DEPARTMENT OF ENERGY under Contract DE-AC06-76RL01830 This document was printed on recycled paper. PNNL-14568 A Graphite Isotope Ratio Method Primer—A Method for Estimating Plutonium Production in Graphite Moderated Reactors C. J. Gesh February 2004 Prepared for the U.S. Department of Energy under Contract DE-AC06-76RL01830 Pacific Northwest National Laboratory Richland, Washington 99352 Summary The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor.
    [Show full text]