DOCSLIB.ORG

Reactor Physics Design of Supercritical Co2-Cooled At

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Reactor Physics Design of Supercritical Co2-Cooled At REACTOR PHYSICS DESIGN OF SUPERCRITICAL CO2-COOLED FAST REACTORS by Michael A. Pope B.S. Nuclear Engineering Texas A&M University, 2002 SUBMITTEDTO THE DEPARTMENT OF NUCLEARENGINEERING IN PARTIAL FULFILLMENTOF THE REQUIREMENTS FOR THE DEGREE OF MASTER OF SCIENCE IN NUCLEAR ENGINEERING AT THE MASSACHUSETTS INSTITUTE OF TECHNOLOGY SEPTEMBER 2004 The author hereby grants MIT permission to reproduce and to distribute publicly paper and electronic copies of this report document in whole or in part MASSACHUSETTSNSTTTE Copyright © Massachusetts Institute of Technology (MIT) OF TECHNOLOGY All rights reserved OCT I 112005 LIBRARIES Signature of Author: f Department of Nuclear Engineering August 6, 2004 Certified by: AT Dr. Michael J. Driscoll Professor Emeritus of Nuclear Engineering Thesis Supervisor Certified by: "/ -Dr. Pavel Hejzlar ___zPrincipal Research Scientist in Nuclear Engineering /4 A'.· Thesis Reader Accepted by: -r'x 7 IT A I - LJr.Jeiirey A. Coaerre Chairman, Department Committee on Graduate Students ARCHIVES REACTOR PHYSICS DESIGN OF SUPERCRITICAL CO2-COOLED FAST REACTORS By Michael A. Pope Submitted to the Department of Nuclear Engineering on August 6, 2004 in partial fulfillment of the requirements for the degree of Master of Science in Nuclear Engineering Abstract Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO2 (S-CO 2) as a Brayton cycle working fluid in a direct cycle is evaluated. By using S- ° CO 2 at turbine inlet conditions of 20 MPa and 550 C - 700°C, efficiencies between 45% and 50% can be achieved with extremely compact components. Neutronic evaluation of candidate core materials was performed for potential use in block-type matrix fueled GFRs with particular concentration on lowering coolant void reactivity to less than $1. SiC cercer fuel was found to have relatively low coolant void worth (+22¢ upon complete depressurization of S-CO2 coolant) and tolerable reactivity- limited burnup at matrix volume fractions of 60% or less in a 600 MWth core having H/D of 0.85 and titanium reflectors. Pin-type cores were also evaluated and demonstrated higher kff versus burnup, and higher coolant void reactivity than the SiC cercer cores (+$2.00 in ODS MA956-clad case having H/D of 1). It was shown, however, that S-CO2 coolant void reactivity could be lowered significantly - to less than $1 - in pin cores by increasing neutron leakage (e.g. lowering the core H/D ratio to 0.625 in a pin core with ODS MA956 cladding), an effect not observed in cores using helium coolant at 8 MPa and 500°C. An innovative "block"-geometry tube-in-duct fuel consisting of canisters of vibrationally compacted (VIPAC) oxide fuel was introduced and some preliminary calculations were performed. A reference tube-in-duct core was shown to exhibit favorable neutron economy with a conversion ratio (CR) at beginning of life (BOL) of 1.37, but had a coolant void reactivity of +$ 1.4. The high CR should allow designers to lower coolant void worth by increasing leakage while preserving the ability of the core to reach high burnup. Titanium, vanadium and scandium were found to be excellent reflector materials from the standpoint of koff and coolant void reactivity due to their unique elastic scattering cross-section profiles: for example, the SiC cercer core having void reactivity of +$0.22 with titanium was shown to have +$0.57 with Zr3Si2. Overall, present work confirmed that the S-CO 2-cooled GFR concept has promising characteristics and a sufficiently broad opion space such that a safe and competitive design could be developed in future work with considerably less than $1 void reactivity and a controllable Ak due to burnup. Thesis Supervisor: Michael J. Driscoll Title: Professor Emeritus of Nuclear Engineering Thesis Reader: Pavel Hejzlar Title: Principal Research Scientist and Program Director of the Advanced Reactor Technology Program at the Center for Advanced Nuclear Energy Systems (CANES) 2 Acknowledgments I would first like to thank my research advisor, Professor Emeritus Michael Driscoll. This work would not have been possible without his patient guidance and continuous stream of ideas. I also want to thank my thesis reader, Dr. Pavel Hejzlar, for his scrutiny of the technical content of this thesis. His broad knowledge has been invaluable to this thesis. This financial support for this work was provided by Idaho National Engineering and Environmental Laboratory (INEEL) through their Lab Directed Research and Development (LDRD) program entitled "An Innovative Gas-Cooled Fast Reactor". Nathan Carstens assembled and maintains the Echelon Beowulf cluster for the Department of Nuclear Engineering at MIT. This machine allowed the many thousands of processor-hours represented by this thesis to be performed in a manageable amount of time. Peter Yarsky provided valuable feedback and comments on earlier drafts of this thesis. Jonathan Plaue gave me vials containing titanium and silicon carbide so that I could see what they look like. He also reminded me from time to time that I cannot simply make up chemicals to satisfy neutronic requirements. Jan Outcalt created some of the drawings used as figures in this thesis. I would finally like to thank my parents, Alex and Barbara Pope, for their financial and moral support throughout my education. You are wonderful parents and will soon be wonderful grandparents. Thank you and I love you both very much. 3 Table of Contents Abstract........... 2.................. Acknowledgments ............................................................................................................... 3 Table of Contents .................................... ................... .................. ..............................4 Table of Figures ..................................................................................................................6 List of T ables ......................................................................................................................9 Dedication ......................................................................................................................... 10 1 Introduction and Background ......................................... 11 1.1 Gas-Cooled Fast Reactors .......................................................... ...................... I 1.2 Supercritical CO2 Brayton Cycle ......................................... 11 1.3 Matrix Fuel........................................................................................................12 1.4 Coolant Void Reactivity ......................................... 12 1.5 Objectives and Organization of Thesis .............................................................15 2 Reactor Modeling Methodology ......................................... 17 2.1 Introduction .......................................................................................................17 2.2 MCNP Model ....................................................................................................17 2.3 Burnup Calculations..........................................................................................22 2.4 Summary ......................................... 24 3 Comparison of the Reactivity Effects of Reflector, Matrix and Shielding Materials25 3.1 Introduction .......................................................................................................25 3.2 Multiplication Factor and Coolant Void Reactivity. ............................. 25 3.2.1 Required TRU Enrichment ................................................................... 27 3.2.2 Results of Comparison ......................................... 27 3.2.3 Helium versus CO 2 Coolant ......................................................................28 3.2.4 Sensitivity to Cross-Section Sets ......................................... 29 3.3 Neutronic Advantage of Titanium, Vanadium and Scandium Reflectors ........ 30 3.4 Burnup Comparison ......................................... 35 3.5 Neutron Fluence in Reactor Vessel ...................................................................37 3.6 Summary ......................................... 38 4 Parametric Studies ....................................................................................................40 4.1 Introduction .......................................................................................................40 4.2 Effects of Matrix Volume Fraction...................................................................40 4.3 Effects of Variation of Core Size......................................................................42 4.4 Prospect of Pin-Type Cores ..............................................................................43 4.4.1 Validity of Single-Enrichment Calculations ........................ 47 4.4.2 Determination of Axial Reflector Height .. ..................... 48 4.4.3 Coolant Void Reactivity versus Core H/D .. ................. 49 4.4.4 Effect of Pressure on Coolant Void Reactivity .. ............... 54 4.4.5 Comparison of Cladding Material Effect on Burnup ...................... 56 4.4.6 Neutron Fluence in Cladding Material .. ................... 57 4.4.7 Burnup Comparison of S-CO2 and Helium Coolants ....................... 58 4.4.8 Use of Oxide Fuel in Pin Cores ......................................... 59 4.4.9 Power Shaping in Pin Cores ......................................... 61 4.5 Tube-In-Duct Fuel Assembly ........................................
Load more

Recommended publications
  • KINETIC ANALYSIS of SUB-PROMPT-CRITICAL REACTOR ASSEMBLIES Hy S Das Theoretical Physics Division
    BARC/1992/E/011 O 1 "o KINETIC ANALYSIS OF SUB-PROMPT-CRITICAL REACTOR ASSEMBLIES hy S Das Theoretical Physics Division 1992 BARC/1992/E/011 o UJ GOVERNMENT OF INDIA - AATOMIT C ENERGY COMMISSION U KINETIC ANALYSIS OF SUB-PROMPT-CRITICAL REACTOR ASSEMBLIES by S. Das Theoretical Physics Division BHABHA ATOMIC RESEARCH CENTRE BOMBAY, INDIA 1992 BARC/1992/E/011 BIBLIOBRAPHIC DESCRIPTION SHEET FOR TECHNICAL REPORT (as par IS i 94BNB - 1980) 01 Security classification i Unclassified 02 Distribution t External 03 Report status i New 04 Series i BARC External 05 Report type i Technical Report 06 Report No. i BARC/1992/E/011 07 Part No. or Volume No. i 08 Contract No. i 10 Title and subtitle i Kinetic analysis of sub-prompt- critical reactor assemblies 11 Collation i 14 p.f 4 figs., 1 tab. IS Project No. i 20 Personal author<s> i S. Das 21 Affiliation of author(s) i Theoretical Physics Division, Bhabha Atomic Research Centre, Bombay 22 Corporate author(s) i Bhabha Atomic Research Centre, Bombay - 400 085 23 Originating unit i Theoretical Physics Division, BARC, Bombay 24 Sponsor(s) Name i Department of Atomic Energy Type i Government 30 Date of submission i May 1992 31 Publication/Issue date i June 1992 Contd... (ii) 1 40 Publ i»h»r/Distributor t Head, Library and Information Division, Bhabha Atomic Rasaarch Contra, Bombay 42 Form of distribution t Hard Copy 50 Languaga of taxt i English 51 Languaga o-f summary i English 92 No. of references i % rafs. 53 Givas data on i 60 Abstract x Nautronic analysis of safaty-ralatad kinetics problams in experimental neutron multiplying assemblies has bean carried out using a sub-prompt-critical reactor model.
    [Show full text]
  • Regulatory Guide 3.35, Revision 1
    Revision 1 U.S. NUCLEAR REGULATORY COMMISSION July 1979 :*REGULATORY GUIDE OFFICE OF STANDARDS DEVELOPMENT REGULATORY GUIDE 335 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN A PLUTONIUM PROCESSING AND FUEL FABRICATION PLANT A. INTRODUCTION will review the proposal and approve its use, if found acceptable. Section 70.22, "Contents of Applications," of 10 CFR Part 70, "Domestic Licensing of Special B. DISCUSSION Nuclear Materials," requires, that each appli- cation for a license to possess and use special In the process of reviewing applications for nuclear material in a plutonium processing and permits and licenses authorizing the construc- fuel fabrication plant contain a description and tion or operation of plutonium processing and safety assessment of the design bases of the fuel fabrication plants, the NRC staff has principal structures, systems, and components developed a number of appropriately conser- of the plant. Section 70.23(a)(3) states that vative assumptions that are used by the staff applications will be approved if the Commission to evaluate an estimate of the radiological determines that, among other factors, the consequences of various postulated accidents. applicant's proposed equipment and facilities These assumptions are based on previous are adequate to protect health and minimize accident experience, engineering judgment, danger to life and property, and Sec- and on the analysis of applicable experimental tion 70.23(b) states that the Commission will results from safety research programs. This approve construction of the principal struc- guide lists assumptions used by the staff to tures, systems, and components of the plant evaluate the magnitude and radiological conse- when the Commission has determined that the quences of a criticality accident in a plutonium design bases of the principal structures, sys- processing and fuel fabrication plant.
    [Show full text]
  • Fundamentals of Power Reactors Module One Science & Engineering Fundamentals
    Training Centre / Centre de formation Fundamentals of Power Reactors Module One Science & Engineering Fundamentals Copyright Notice ©HER MAJESTY THE QUEEN IN RIGHT OF CANADA (1993) as represented by the Atomic Energy Control Board All rights reserved. No part of this publication may be reproduced, stored in a retrieval system or transmitted in any form or by any means, electronic, electrostatic, magnetic tape, mechanical photocopying, recording or otherwise, without permission from the Atomic Energy Control Board of Canada. Training Centre / Centre de formation Training Centre / Centre de formation Basis of Nuclear Structure and Fission Training Objectives The participant will be able to describe or understand: 1 the atomic and nuclear structure, 2 the basic vocabulary of nuclear energy: typical length scales, mass and energy units, etc., 3 the atomic and nuclear phenomena, 4 the fission and energy release processes, 5 the types of radiation, 6 the concept of neutron spectrum, 7 the conecpt of cross section 8 the concept of irradiation Nuc Struc & Fission Training Centre / Centre de formation Nuc Struc & Fission Training Centre / Centre de formation Basis of Nuclear Structure and Fission Table of Contents 1 Introduction ...............................................................................................3 2 Basics of Nuclear Physics ........................................................................4 2.1 Structure of the atom and nucleus ................................................4 2.1.1 Nuclear structure ..........................................................................4
    [Show full text]
  • Introductory Nuclear Physics Problem Set 05
    UNIVERSITY OF CAPE TOWN EEE4106Z: Introductory Nuclear Physics Problem Set 05 Due 1400 Tuesday 2 June 2015 This problem set should prove valuable in preparation for the final examination. The questions allow you to possibly gain a fuller understanding of important concepts that were not familiar to you at the time of preparing your answers to the preceding problem set. At Koeberg hand in to Z Isaacs. At UCT place in Aschman mailslot in mailboxes opposite room 508, top floor RW James Building UCT. Late answers will incur a penalty. Students may work together on the problems, and discuss the results together, but a handed-in script must be each student's own work. Copied work gets zero. Discuss the problems before tackling them! 1) Explain how the fuel temperature coefficient of reactivity is largely determined by the resonances in the cross-section for the 238U(n; γ) capture of neutrons. Provide a clear explanation of the Doppler broadening of the resonance peaks. SOLUTION: At higher temperatures there is greater thermal motion of the fuel nuclei, in particular U-238 nuclei. This increases the spread in the velocities, and hence the energies, with with the neutrons approach the U-238 nuclei. This causes Doppler broadening of the neutron capture resonances in the 238U(n; γ) cross-section in the epi-thermal energy region, so there is a greater chance of the a neutron being captured in the fuel as the neutron energy gets stepped down, from fast to thermal, by the moderator. This reduces the resonance escape factor p, and hence k.
    [Show full text]
  • Reactor Physics
    REACTOR PHYSICS COURSE INTRODUCTION This Training Manual assumes prior knowledge of Nuclear Theory. It extends this information into a discussion of Reactor Physics, particularly as it relates to CANDU reactors. The course begins with the general principles of reactor configuration required to maintain a self- sustaining chain reaction. It continues with reactor dynamics (in both the critical and subcritical core), reactivity feedback effects (temperature effects, fission product poisoning, and fuel burnup), and ends with operational considerations (at low and high power). The material covers four main areas, subdivided into eight major sections as follows: • The Critical Reactor at Steady Power Output (Section 1) • The Dynamic Reactor (Sections 2 and 3) • Reactivity Feedback Effects (Sections 4, 5, and 6) • Reactor Operations (Sections 7 and 8) CNSC i Science and Reactor Fundamentals – Reactor Physics Technical Training Group TABLE OF CONTENTS Objectives 1 The Critical Reactor at Steady Power Output 9 1.0 INTRODUCTION 9 1.1 Fission 9 1.2 Harnessing Fission 15 1.3 Movement of Neutrons Through the CANDU lattice Lattice 18 1.4 The Finite Reactor 26 Response of The Critical Reactor to a Reactivity Change 31 2.0 INTRODUCTION 31 2.1 Exponential Power Rise 31 2.2 Corrections to Exponential Reactor Response 33 2.3 The Effect of Delayed Neutrons 36 2.4 Prompt Criticality 40 2.5 Power Rundown: The Prompt Drop 42 Responsiveness of The Subcritical Reactor 45 3.0 INTRODUCTION 45 3.1 Neutron Flux in a “Shut Down” Reactor 45 3.2 Dynamics in the Subcritical
    [Show full text]
  • Accelerator-Driven Systems: Safety and Kinetics
    Accelerator-driven Systems: Safety and Kinetics Marcus Eriksson Doctoral Thesis Department of Nuclear and Reactor Physics Royal Institute of Technology Stockholm 2005 Akademisk avhandling som med tillstånd av Kungliga Tekniska Högskolan framlägges till offentlig granskning för avläggande av teknologie doktorsexamen fredagen den 18 mars 2004 kl. 10.00 i sal FA32, Albanova universitetscentrum, Roslagstullsbacken 21, Stockholm. ISBN 91-7283-988-0 TRITA-FYS 2005:13 ISSN 0280-316X ISRN KTH/FYS/--05:13—SE Copyright Marcus Eriksson Tryckeri: Universitetsservice US-AB, Stockholm 2005 2 Abstract The accelerator-driven system (ADS) is recognized as a promising system for the purpose of nuclear waste transmutation and minimization of spent fuel radiotoxicity. The primary cause for this derives from its accelerator-driven, sub-critical operating state, which introduces beneficial safety-related features allowing for application of cores employing fuel systems containing pure transuranics or minor actinides, thereby offering increased incineration rate of waste products and minimal deployment of advanced (and expensive) partitioning and transmutation technologies. The main theme of the thesis is safety and kinetics performance of accelerator-driven nuclear reactors. The studies are confined to the examination of ADS design proposals employing fast neutron spectrum, uranium-free lattice fuels, and liquid- metal cooling, with emphasis on lead-bismuth coolant. The thesis consists of computational studies under normal operation and hypothetical accidents, and of evaluation and identification of safety design features. By itself, subcritical operation provides a distinct safety advantage over critical reactor operation, distinguished by high operational stability and additional margins for positive reactivity insertion. For a uranium-free minor actinide based fuel important safety parameters deteriorate.
    [Show full text]
  • Regulatory Guide 3.34, Revision 1, Assumptions Used for Evaluating
    Revision I U.S. NUCLEAR REGULATORY COMMISSION July 1979 'PA' S t REGULATORY GUIDE "** OFFICE OF STANDARDS DEVELOPMENT REGULATORY GUIDE 334 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN A URANIUM FUEL FABRICATION PLANT A. INTRODUCTION B. DISCUSSION Section 70.23, "Requirements for the ap In the process of reviewing applications for proval of applications," of 10 CFR Part 70, licenses to operate uranium fuel fabrication "Domestic Licensing of Special Nuclear Mate plants, the NRC staff has developed appro rials," requires, among other things, that the priately conservative assumptions that are used applicant's proposed equipment and facilities be by the staff to evaluate an estimate of the adequate to protect health and minimize danger radiological consequences of various postulated to life or property. In order to demonstrate the accidents. These assumptions are based on adequacy of the facility, the applicant must previous accident experience, engineering provide an analysis and evaluation of the judgment, and the analysis of applicable design and performance* of structures, sys,' experimental results from safety research tems, and components of the facility. The programs. This guide lists assumptions used to objective of this analysis and evaluation is to evaluate the magnitude and radiological con assess the risk to public health and safety: sequences of a criticality accident in a uranium resulting from operation of the facility, includ fuel fabrication plant. ing determination of the adequacy of struc tures, systems, and components provided for A criticality accident is an accident resulting the prevention of accidents and the mitigation in the uncontrolled release of energy from an of the consequences of accidents.
    [Show full text]
  • Predicted Radionuclide Release from Marine Reactors Dumped in the Kara Sea
    IAEA-TECDOC-938 Predicted radionuclide release from marine reactors dumped in the Kara Sea ReportSourcethe of Term Working Groupthe of International Arctic Seas Assessment Project (IASAP) INTERNATIONAL ATOMIC ENERGY AGENCY The IAEA does not normally maintain stocks of reports in this series. However, microfiche copies of these reports can be obtained from ClearinghousS I N I e International Atomic Energy Agency Wagramerstrasse 5 P.O. Box 100 A-1400 Vienna, Austria Orders shoul accompaniee db prepaymeny db f Austriao t n Schillings 100, in the form of a cheque or in the form of IAEA microfiche service coupons which may be ordered separately from the INIS Clearinghouse. The originating Section of this publication in the IAEA was: Waste Safety Section International Atomic Energy Agency Wagramer Strasse5 0 10 x P.OBo . A-1400 Vienna, Austria PREDICTED RADIONUCLIDE RELEASE FROM MARINE REACTORS DUMPE KARE A TH ASE DN I IAEA, VIENNA, 1997 IAEA-TECDOC-938 ISSN 1011-4289 © IAEA, 1997 Printed by the IAEA in Austria April 1997 FOREWORD In 1992 the news that the former Soviet Union had, for over three decades, dumped high level radioactive wasteshalloe th n i s w waterKare a causeth Se af o s d widespread concern, especialln yi countries with Arctic coastlines. The IAEA responded by launching an international study, the International Arctic Seas Assessment Project (IASAP) orden i , asseso t r e potentiath s l healtd han environmental implications of the dumping and to examine the feasibility of remedial actions related to the dumped wastes. 199y Russiae Ith 3nMa n Federation provided informatio IAEe th Ao nt w aboulo hige d th t han level radioactive waste dumped in the Arctic Seas ("White Book-93").
    [Show full text]
  • Using ORGEN to Help Probe Gamma-Ray Burst Mysteries
    Using ORIGEN to Help Probe Gamma-Ray Burst Mysteries Robert B. Hayes, PhD, CHP, PE Associate Professor of Nuclear Engineering North Carolina State University SCALE workshop Introduction • The model presented here has the potential to explain some of the highly varied nature of a portion of measured gamma ray bursts (GRBs). – These include the random nature of the events in terms of their parametric values and ranges. – Specifically many measured values of energy and time distributions, pulsing, and afterglows appear to fit those attainable with the proposed criticality model. Oklo • The Oklo uranium mine in South Africa is a place where the natural abundance of U235 was found to be consistently less than the 0.72 % found everywhere else on the planet • Because the half life of U235 is almost an order of magnitude smaller than that of U238, the ratio of U235 to U238 increases in reverse time. Oklo burn up • By going back far enough in time, the abundance of the U nuclides would become comparable allowing rainwater to provide the moderation needed for criticality – The critical state would then drive out the water until more rainwater would allow the system to go critical again. – This probably continued many times over. – Kenneth Krane; Introductory Nuclear Physics; John Wiley and Sons Oklo in space • Just after the Super Novae (SN) (or other source) creation of the uranium and transuranic isotopes, most of the already radioactive elements created will quickly decay into stable atoms leaving these fissile and fissionable isotopes close to their original abundance. • A fast reactor is possible • Plentiful H makes a thermal system likely TRU isotopes from SN (or other) • Even if a large fraction of the Pu239 (which is fissile with a moderately long half life of just over 24 ka) were to decay away, there would still be almost the initial amount of U233 .
    [Show full text]
  • NUCLEAR CRITICALITY SAFETY:^ R'
    LA-12387-M Manual UC-714 Issued: November 1992 Revised: June 1993 NUCLEAR CRITICALITY SAFETY:^ r' 3-DAY TRAINING COURSE Offered at The Los Alamos National Laboratory Edited and Compiled by John A. Schlesser MASTER Los Alamos NATIONAL LABORATORY Los Alamos, New Mexico 87545 DISTRIBUTION OF THIS DOCUMENT IS UNLIMITED An Affirmative Action/Equal Opportunity Employer The Los Alamos National Laboratory is operated by the University of California for the United States Department of Energy under contract W-7405-ENG-36. The illustration on the cover was designed by John Schlesser and Gary Webb and drawn by AnnMarie Dyson. It depicts one ofthe course's two critical experiments. DISCLAIMER This manuscript was prepared as an account of work sponsored by an agency of the United States Government. Neither the 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. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. ABSTRACT This compilation of notes is presented as a source reference for the criticality safety course. It represents the contribution? of many people, particularly Tom McLaughlin, the course's primary instructor. INTRODUCTION The 3-Day Training Course in an intensive course in criticality safety consisting of lectures and laboratory sessions, including active student participation in actual critical experiments, a visit to a plutonium processing facility, and in-depth discussions on safety philosophy.
    [Show full text]
  • Dynamics of Ramp-Lnitiated Reactor Transients
    KfK 5352 Juni 1994 Dynamics of Ramp-lnitiated Reactor Transients G. Kußmaul Institut für Neutronenphysik und Reaktorphysik Projekt Nukleare Sicherheitsforschung Kernforschungszentrum Karlsruhe Kernforschungszentrum Karlsruhe Institut für Neutronenphysik und Reaktortechnik Projekt Nukleare Sicherheitsforschung KfK 5352 DYNAMICS OF RAMP-INITIATED REACTOR TRANSIENTS Günter Kußmaul Kernforschungszentrum Karlsruhe GmbH, Karlsruhe Als Manuskript gedruckt Für diesen Bericht behalten wir uns alle Rechte vor Kernforschungszentrum Karlsruhe GmbH Postfach 3640, 76021 Karlsruhe ISSN 0303-4003 Abstract Transients in nuclear systems induced by reactivity ramps develop different types of power traces - exponential, single-pulse, oscillating - depending an the ramp rates and the neutron lifetime of the system. Starting from the kinetics equations of the point reactor and using a model of energy-dependent feedback reactivity, based an the prompt and inherent Doppler effect, simple relations for the de­ pendence of energy release and power shape on ramp rate, energy coefficient and neutron lifetime are derived. Depending on neutron lifetime of the nuclear system the ranges of ramp rate are determined, which define the type of power trace. ln the regime of superprompt-critical transients, the models of step and ramp insertions are especially investigated. The ranges of applicability have been checked by comparison with calculations using the dynamics code system DYANA2 which solves numerically the set of kinetics and thermal-hydraulics equations describing
    [Show full text]
  • Point Kinetic Model of the Early Phase of a Spherically Symmetric Nuclear Explosion
    Point kinetic model of the early phase of a spherically symmetric nuclear explosion Andreas Walter Aste Department of Physics, University of Basel, 4056 Basel, Switzerland June 6, 2016 Abstract A concise point kinetic model of the explosion of a prompt supercritical sphere driven by a nu- clear fission chain reaction is presented. The findings are in good agreement with the data available for Trinity, the first detonation of a nuclear weapon conducted by the United States Army as part of the Manhattan project. Results are presented for an implosion device containing pure plutonium- 239, although the model can be easily applied to, e.g., uranium-235. The fizzle probability and cor- responding yield of a fission bomb containing plutonium recovered from reactor fuel and therefore containing significant amounts of spontaneously fissioning plutonium-240 which can induce a pre- detonation of the device is illustrated by adding a corresponding source term in the presented model. Related questions whether a bomb could be made by developing countries or terrorist organizations can be tackled this way. Although the information needed to answer such questions is in the public domain, it is difficult to extract a consistent picture of the subject for members of organizations who are concerned about the proliferation of nuclear explosives. Physics and Astronomy Classification Scheme (2010). 28.20.-v Neutron physics; 25.85.Ec Neutron-induced fission; 28.70.+y Nuclear explosions. 1 Introduction Despite the terrifying fact that numerous operational fission or even thermonuclear bombs exist on our planet, there is a great interest in the basic principles and physics underlying the concept of nuclear weapons.
    [Show full text]