Neutron Fission of Pu-239
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Compilation and Evaluation of Fission Yield Nuclear Data Iaea, Vienna, 2000 Iaea-Tecdoc-1168 Issn 1011–4289
IAEA-TECDOC-1168 Compilation and evaluation of fission yield nuclear data Final report of a co-ordinated research project 1991–1996 December 2000 The originating Section of this publication in the IAEA was: Nuclear Data Section International Atomic Energy Agency Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria COMPILATION AND EVALUATION OF FISSION YIELD NUCLEAR DATA IAEA, VIENNA, 2000 IAEA-TECDOC-1168 ISSN 1011–4289 © IAEA, 2000 Printed by the IAEA in Austria December 2000 FOREWORD Fission product yields are required at several stages of the nuclear fuel cycle and are therefore included in all large international data files for reactor calculations and related applications. Such files are maintained and disseminated by the Nuclear Data Section of the IAEA as a member of an international data centres network. Users of these data are from the fields of reactor design and operation, waste management and nuclear materials safeguards, all of which are essential parts of the IAEA programme. In the 1980s, the number of measured fission yields increased so drastically that the manpower available for evaluating them to meet specific user needs was insufficient. To cope with this task, it was concluded in several meetings on fission product nuclear data, some of them convened by the IAEA, that international co-operation was required, and an IAEA co-ordinated research project (CRP) was recommended. This recommendation was endorsed by the International Nuclear Data Committee, an advisory body for the nuclear data programme of the IAEA. As a consequence, the CRP on the Compilation and Evaluation of Fission Yield Nuclear Data was initiated in 1991, after its scope, objectives and tasks had been defined by a preparatory meeting. -
Wo 2009/108331 A2
(12) INTERNATIONAL APPLICATION PUBLISHED UNDER THE PATENT COOPERATION TREATY (PCT) (19) World Intellectual Property Organization International Bureau (10) International Publication Number (43) International Publication Date 3 September 2009 (03.09.2009) WO 2009/108331 A2 (51) International Patent Classification: AO, AT, AU, AZ, BA, BB, BG, BH, BR, BW, BY, BZ, A61K 38/22 (2006.01) CA, CH, CN, CO, CR, CU, CZ, DE, DK, DM, DO, DZ, EC, EE, EG, ES, FI, GB, GD, GE, GH, GM, GT, HN, (21) International Application Number: HR, HU, ID, IL, IN, IS, JP, KE, KG, KM, KN, KP, KR, PCT/US2009/001213 KZ, LA, LC, LK, LR, LS, LT, LU, LY, MA, MD, ME, (22) International Filing Date: MG, MK, MN, MW, MX, MY, MZ, NA, NG, NI, NO, 25 February 2009 (25.02.2009) NZ, OM, PG, PH, PL, PT, RO, RS, RU, SC, SD, SE, SG, SK, SL, SM, ST, SV, SY, TJ, TM, TN, TR, TT, TZ, UA, (25) Filing Language: English UG, US, UZ, VC, VN, ZA, ZM, ZW. (26) Publication Language: English (84) Designated States (unless otherwise indicated, for every (30) Priority Data: kind of regional protection available): ARIPO (BW, GH, 61/066,959 25 February 2008 (25.02.2008) US GM, KE, LS, MW, MZ, NA, SD, SL, SZ, TZ, UG, ZM, Not furnished 25 February 2009 (25.02.2009) US ZW), Eurasian (AM, AZ, BY, KG, KZ, MD, RU, TJ, TM), European (AT, BE, BG, CH, CY, CZ, DE, DK, EE, (71) Applicant and ES, FI, FR, GB, GR, HR, HU, IE, IS, IT, LT, LU, LV, (72) Inventor: FORSLEY, Lawrence, Parker, Galloway MC, MK, MT, NL, NO, PL, PT, RO, SE, SI, SK, TR), [US/US]; 70 Elmwood Avenue, Rochester, NY 1461 1 OAPI (BF, BJ, CF, CG, CI, CM, GA, GN, GQ, GW, ML, (US). -
Relative Fission Product Yield Determination in the Usgs
RELATIVE FISSION PRODUCT YIELD DETERMINATION IN THE USGS TRIGA MARK I REACTOR by Michael A. Koehl © Copyright by Michael A. Koehl, 2016 All Rights Reserved A thesis submitted to the Faculty and the Board of Trustees of the Colorado School of Mines in partial fulfillment of the requirements for the degree of Doctor of Philosophy (Nuclear Engineering). Golden, Colorado Date: ____________________ Signed: ________________________ Michael A. Koehl Signed: ________________________ Dr. Jenifer C. Braley Thesis Advisor Golden, Colorado Date: ____________________ Signed: ________________________ Dr. Mark P. Jensen Professor and Director Nuclear Science and Engineering Program ii ABSTRACT Fission product yield data sets are one of the most important and fundamental compilations of basic information in the nuclear industry. This data has a wide range of applications which include nuclear fuel burnup and nonproliferation safeguards. Relative fission yields constitute a major fraction of the reported yield data and reduce the number of required absolute measurements. Radiochemical separations of fission products reduce interferences, facilitate the measurement of low level radionuclides, and are instrumental in the analysis of low-yielding symmetrical fission products. It is especially useful in the measurement of the valley nuclides and those on the extreme wings of the mass yield curve, including lanthanides, where absolute yields have high errors. This overall project was conducted in three stages: characterization of the neutron flux in irradiation positions within the U.S. Geological Survey TRIGA Mark I Reactor (GSTR), determining the mass attenuation coefficients of precipitates used in radiochemical separations, and measuring the relative fission products in the GSTR. Using the Westcott convention, the Westcott flux, ; modified spectral index, ; neutron temperature, ; and gold-based cadmium ratiosφ were determined for various sampling√⁄ positions in the USGS TRIGA Mark I reactor. -
The Electronuclear Conversion of Fertile to Fissile Material
UCRL-52144 THE ELECTRONUCLEAR CONVERSION OF FERTILE TO FISSILE MATERIAL C. M. Van Atta J. D. Lee H. Heckrotto October 11, 1976 Prepared for U.S. Energy Research & Development Administration under contract No. W-7405-Eng-48 II\M LAWRENCE lUg LIVERMORE k^tf LABORATORY UnrmsilyotCatftxna/lJvofmofe s$ PC DISTRIBUTION OF THIS DOCUMENmmT IS UMUMWTED NOTICE Thii npoit WM prepared w u account of wot* •pomond by UM Uiilttd Stalwi GovcmiMM. Nittim Uw United Stain nor the United Statn Energy tUwardi it Development AdrnWrtrtUon, not «y of thei* employee!, nor any of their contricton, •ubcontrecton, or their employe*!, makti any warranty, expreai « Implied, or muMi any toga) liability oc reeponafctUty for the accuracy, completenni or uMfulntat of uy Information, apperatui, product or proem dlecloead, or repreienl. that tu UM would not *nfrkig» prrrtt*)}MWiwd r%hl». NOTICE Reference to a oompmy or product rum don not imply approval or nconm ndatjon of the product by the Untnrtity of California or the US. Energy Research A Devetopnient Administration to the uchvton of others that may be suitable. Printed In the United Stitei or America Available from National Technical Information Service U.S. Department of Commerce 528S Port Royal Road Springfield, VA 22161 Price: Printed Copy S : Microfiche $2.25 DMimtic P»t» Ranflt PHca *•#» Rtnft MM 001-025 $ 3.50 326-350 10.00 026-050 4.00 351-375 10.50 051-075 4.50 376-400 10.75 076-100 5.00 401-425 11.00 101-125 5.50 426-450 31.75 126-150 6.00 451-475 12.00 151-175 6.75 476-500 12.50 176-200 7.50 501-525 12.75 201-225 7.75 526-550 13.00 526-250 8.00 551-575 13.50 251-275 9.00 576-600 I3.7S 276-300 9.25 601 -up 301-325 9.75 *Ml J2.50 fot «ch iddltlOMl 100 pip tacmitMt from 601 ID 1,000 flfcK IM 54.50 for eicli iMUknal I0O plfe feenmMI om 1,000 p*o. -
On the Calculation of the Fast Fission Factor
AE-27 On the Calculation of the w < Fast Fission Factor B. Almgren AKTIEBOLAGET ATOMENERGI STOCKHOLM • SWEDEN • I960 AE-27 ON THE CALCULATION OF THE FAST FISSION FACTOR E. Almgren Summary: Definitions of the fast fission factor e ars discussed. Different methods of calculation of e are compared. Group constants for one -, two- and three- group calculations have been evaluated using the best obtainable basic data. The effects of back-scattering, coupling and (n, 2n)-reactions are discussed. Completion of manuscript in June 1960 Printed and distributed in November 1960 LIST OF CONTENTS Page 1. Definitions 3 2. Formulas for the fast fission factor e and the fast fission ratio R 3 3. Calculation of group constants 7 4. Collision probabilities 10 5. Numerical results 12 6. Discussion 12 7. Acknowledgements 13 Preferences 14 On the calculation of the fast fission factor. 1. Definitions The fast fission factor e may be defined in different ways. Carlvik and Pershagen (3) have defined e as "the number of neutrons which are either slowed down below 0, 1 MeV in the fuel or leave the fuel, per primary neu- tron from thermal fission". This definition has been used in Sweden, since it was proposed (1956). Another commonly used definition is due to Spinrad (2) "the number of neutrons making first collision with moderator per neutron arising from thermal fission". The choice of definition must be consistent with the definition of the resonance escape probability. The above-mentioned definitions include in e some of the capture below the fast fission threshold. However, they do not include the effects of (n, 2n)-reactions or capture of high energy neutrons in the moderator. -
The Transmutation of Fission Products (Cs-137, Sr-90) in a Liquid Fuelled Fast Fission Reactor with Thermal Column
EIR-Bericht Nr . 270 Eidg . Institut fur Reaktorforschung Wurenlingen Schweiz The transmutation of fission products (Cs-137, Sr-90) in a liquid fuelled fast fission reactor with thermal column M . Taube, J . Ligou, K. H . Bucher Wurenlingen, Februar 1975 EIR-Bericht Nr . 270 The transmutation of fission products (Cs-137, Sr-90) in a liquid fuelled fast fission reactor with thermal column M. Taube, J . Ligou, K .H . Bucher February 1975 Summary `,he possibilities for the transmutation of caesium-137 and strontium-90 in high-flux fast reactor with molten plutonium chlorides and with thermal column is discussed . The effecitve half life of Cs-137 could be decreased from 30 years to 4-5 years, and for Sr-90 to 2-3 years . Introduction The problem of management of highly radioactive fission product waste has been intensely and extensively discussed in the recent paper WASH - 1297 and especially in BNWL 1900 (Fig . 1) . Here only the transmutation of Fission Products (F .P .) without the recycling of the actinides is discussed, making use of neutron irradiation by means of a fission reactor (Fig . 2) . A short outline of this paper is as follows 1) Why, contrary to many assertions, is neutron transmutation in a fusion reactor not feasible . 2) Why recent discussions concerning transmutation in fission reactors are rather pessimistic . 3) Could the transmutation in a fission reactor be possible taking into account the neutron balance in a breeding system? 4) Which are the F .P . candidates for irradiation in a fission reactor? 5) Is the rate of transmutation sufficiently high in a fission reactor? 6) In what type of fission reactor is the transmutation physically possible? 7) What are the limiting parameters for transmutation in a solid fuelled fission reactor? 8) Is a very high flux fission reactor possible if the fuel is in the liquid state instead of the solid state? 9) How could such a high flux fast reactor with circulating liquid fuel and a thermal column operate as a 'burner' for some F .P . -
Neutron-Physical Simulation of Fast Nuclear Reactor Cores
Neutron-Physical Simulation of Fast Nuclear Reactor Cores Investigation of New and Emerging Nuclear Reactor Systems Zur Erlangung des Grades eines Doktors der Naturwissenschaften (Dr. rer. nat.) genehmigte Dissertation von Friederike Renate Frieß, M.Sc., aus Cuxhaven Tag der Einreichung: 13.6.2017, Tag der Prüfung: 12.7.2017 Juli 2017 — Darmstadt — D 17 1. Gutachten: Prof. Dr. Barbara Drossel 2. Gutachten: Prof. Dr. Wolfgang Liebert Institut für Festkörperphysik Fachbereich Physik Neutron-Physical Simulation of Fast Nuclear Reactor Cores Investigation of New and Emerging Nuclear Reactor Systems Genehmigte Dissertation von Friederike Renate Frieß, M.Sc., aus Cuxhaven 1. Gutachten: Prof. Dr. Barbara Drossel 2. Gutachten: Prof. Dr. Wolfgang Liebert Tag der Einreichung: 13.6.2017 Tag der Prüfung: 12.7.2017 Juli 2017 — Darmstadt — D 17 Contents 1. Introduction 1 1.1. Basics of (Fast) Nuclear Reactors . .2 1.2. Generation IV Nuclear Power Plants . .4 1.3. Special Applications for Fast Nuclear Reactors . .5 1.3.1. Disposition of Weapon-grade Plutonium . .5 1.3.2. Powering Remote Energy Grids . .6 1.3.3. Partitioning and Transmutation . .8 1.4. Research Synopsis . 12 2. Material Assessment & Non-Proliferation 15 2.1. Fissile Materials in the Nuclear Fuel Cycle . 15 2.2. The Existing Non-Proliferation Regime . 16 2.3. General Considerations on Assessing Proliferation Risk . 17 2.4. Spent Fuel Standard . 18 2.5. Figures of Merit . 19 3. Theoretical Basis 23 3.1. Neutron Flux Distribution . 23 3.2. Source Efficiency . 26 3.3. Material Depletion . 27 3.4. Spallation Reaction . 28 3.5. Interaction of Radiation and Tissue . -
Some Effects of Pu-240, Pu-241, and Pu-242 on the Pu-239 Critical Mass of a Fast Reactor Frank Peter Mertes Jr
Iowa State University Capstones, Theses and Retrospective Theses and Dissertations Dissertations 1962 Some effects of Pu-240, Pu-241, and Pu-242 on the Pu-239 critical mass of a fast reactor Frank Peter Mertes Jr. Iowa State University Follow this and additional works at: https://lib.dr.iastate.edu/rtd Part of the Nuclear Commons, and the Oil, Gas, and Energy Commons Recommended Citation Mertes, Frank Peter Jr., "Some effects of Pu-240, Pu-241, and Pu-242 on the Pu-239 critical mass of a fast reactor " (1962). Retrospective Theses and Dissertations. 2313. https://lib.dr.iastate.edu/rtd/2313 This Dissertation is brought to you for free and open access by the Iowa State University Capstones, Theses and Dissertations at Iowa State University Digital Repository. It has been accepted for inclusion in Retrospective Theses and Dissertations by an authorized administrator of Iowa State University Digital Repository. For more information, please contact [email protected]. This dissertation has been 63—2991 microfilmed exactly as received MERTES, Jr., Frank Peter. 1935— SOME EFFECTS OF PU-240, PU-241, AND PU-242 ON THE PU-239 CRITICAL MASS OF A FAST REACTOR. Iowa State University of Science and Technology Ph.D., 1962 Physics, nuclear University Microfilms, Inc., Ann Arbor, Michigan SOME EFFECTS OF PU-2ltO, PU-2liL, AND PU-2lj2 ON THE PU-239 CRITICAL ÎS.SS OF A FAST REA.CTGR Frank Peter Martes, Jr. A Dissertation Submitted to the Graduate Faculty in Partial Fulfillment of The Requirements for the Degree of DOCTOR OF PHILOSOPHY Major Subject: Nuclear Engineering Approved: Signature was redacted for privacy. -
Fission and Corrosion Product Behaviour in Liquid Metal Fast Breeder Reactors (Lmfbrs)
IAEA-TECDOC-687 Fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs) INTERNATIONAL ATOMIC ENERGY AGENCY The IAEA does not normally maintain stocks of reports in this series. However, microfiche copie f thesso e reportobtainee b n sca d from INIS Clearinghouse International Atomic Energy Agency Wagramerstrasse5 P.O. Box 100 A-1400 Vienna, Austria Orders should be accompanied by prepayment of Austrian Schillings 100,- fore for e chequa th f m th IAEf m o n i o n i r eAo microfiche service coupons orderee whicb y hdma separately fro INIe mth S Clearinghouse. FISSIO CORROSIOD NAN N PRODUCT BEHAVIOUR IN LIQUID METAL FAST BREEDER REACTORS (LMFBRs) IAEA, VIENNA, 1993 IAEA-TECDOC-687 ISSN 1011-4289 Printed by the IAEA in Austria February 1993 FOREWORD In 1969 e Internationath , l Atomic Energy Agency decide o includt d e th e topic of fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs} in its programme of specialists meetings. It was apparent at the time that with the advent of experimental fast reactor systems in the USAe USSRth , ,e Unite Francth d d an eKingdom e expansioth , f supportino n g research programme e fragmenteth d an s d natur f informatioo e n arising from these various activities, it would be beneficial to co-ordinate certain topics at an international level. Suppor thir tfo s approac provides hwa Membey db r States and, against this background e IAEs helth ,Aha d three international specialists meetingse on , Germanyn i o purpose meetinge tw Th provido USSe th .d t i th n f an R s o e wa se a better understanding of the way various radionuclides behave in operating systemo providt d an se guidance safth e r developmenfo e f prototypo t d an e eventually commercial systems o meeT .t these objective e specialistth s s discussed experiment results, reviewed experimental programme identified san d area further fo s r work. -
Fast Breeder Reactor Programs: History and Status
Fast Breeder Reactor Programs: 8 History and Status Thomas B. Cochran, Harold A. Feiveson, Walt Patterson, Gennadi Pshakin, M.V. Ramana, Mycle Schneider, Tatsujiro Suzuki, Frank von Hippel A research report of the International Panel on Fissile Materials February 2010 Research Report 8 International Panel on Fissile Materials Fast Breeder Reactor Programs: History and Status Thomas B. Cochran, Harold A. Feiveson, Walt Patterson, Gennadi Pshakin, M.V. Ramana, Mycle Schneider, Tatsujiro Suzuki, Frank von Hippel www.fissilematerials.org February 2010 © 2010 International Panel on Fissile Materials ISBN 978-0-9819275-6-5 This work is licensed under the Creative Commons Attribution-Noncommercial License. To view a copy of this license, visit www.creativecommons.org/licenses/by-nc/3.0 Table of Contents About the IPFM i 1 Overview: The Rise and Fall of Plutonium Breeder Reactors Frank von Hippel 1 2 Fast Breeder Reactors in France Mycle Schneider 17 3 India and Fast Breeder Reactors M. V. Ramana 37 4 Japan’s Plutonium Breeder Reactor and its Fuel Cycle Tatsujiro Suzuki 53 5 The USSR-Russia Fast-Neutron Reactor Program Gennadi Pshakin 63 6 Fast Breeder Reactors in the United Kingdom Walt Patterson 73 7 Fast Reactor Development in the United States Thomas B. Cochran, Harold A. Feiveson, and Frank von Hippel 89 Contributors 113 Fast Breeder Reactor Programs: History and Status Figures Overview: The Rise and Fall of Plutonium Breeder Reactors Figure 1.1 Plutonium breeding. 3 Figure 1.2 History of the price of uranium. 6 Figure 1.3 Funding in OECD countries. 7 Figure 1.4 Dose rate of separated transuranics. -
Calculation Methodology of Fast Fission Factor in a Thermal Reactor
MATEC Web of Conferences 19, 01 021 ( 2014) / / DOI: 10.1051 matecconf 201419 01 021 C Owned by the authors, published by EDP Sciences, 2014 Calculation methodology of fast fission factor in a thermal reactor Denis V. Grishko, Arian V. Kuzmin, Dmitry Y. Malishev National Research Tomsk Polytechnic University, 634050 Tomsk, Russia Abstract. This article describes the coefficient of the fast fission, which is part of the formula of «four factors». Considered, to exist at the moment, two methods for calculation of the coefficient of the fast fission in uranium-water tight lattices. Also presents the results of calculations and comparative analysis of the data obtained by two techniques. 1. Introduction Cross-section of fission fuel (235U, 239Pu, 233Th) is a continuous spectrum with energy distinct from zero. It is mean, that in all reactors there are processes of fission on a heat, resonance and fast neutrons. Such elements as uranium and thorium are fissionable at specific energy (higher than boundary energy). For example, fast neutrons with energy 1.1 Mev fission 238U. There is a chance that fast neutrons will cause fission. It appearance causes other fission neutrons. The coefficient of the fast fission accounts this process. Fig. 1 characteristic curve micro-cross-fission 238U of neutron energy. 2. Problem statement The Monte Carlo method is an accurate method in calculating the coefficient of the fast fission. But to use it requires writing a program and use of computers. Therefore, the main problem in this situation is the ability to quickly and accurately calculate this ratio without the use of complex calculations. -
Fission Yield Measurements from Deuterium-Tritium Fusion Produced Neutrons Using Cyclic Neutron Activation Analysis and Γ-Γ Coincidence Counting
Fission yield measurements from deuterium-tritium fusion produced neutrons using cyclic neutron activation analysis and γ-γ coincidence counting by Bruce D. Pierson A dissertation submitted in partial fulfillment of the requirements for the degree of Doctor of Philosophy (Nuclear Engineering & Radiological Sciences) in The University of Michigan 2016 Doctoral Committee: Professor Sara A. Pozzi, co-chair Assistant Professor Marek Flaska, Penn. State University, co-chair Professor, John E. Foster Larry R. Greenwood, Pacific Northwest National Laboratory Assistant Professor Physics Thomas Schwarz c Bruce D. Pierson 2016 All Rights Reserved This dissertation is dedicated to my family for their unyielding patience and support throughout my graduate career. ii ACKNOWLEDGEMENTS I would like to thank Drs. Marek Flaska, Larry Greenwood, Amanda Prinke, Sara Pozzi, and Sean Stave for their assistance, guidance, mentorship, and revisions to written works; their support and input drastically improved the quality of the final analysis and results (between the five them, I was getting at least one form of support from each of them). I'd also like to thank Drs. Ovidiu Toader and Joe Miklos for their assistance and support in maintaining and managing the Neutron Science Laboratory. Dr. Miklos was instrumental in amending the Nuclear Science Laboratory Nuclear Regulatory Commission license that allowed me to even do the work outlined in this document. He is a good friend and cheered me on to the finish at every opportunity. Dr. Toader was an invaluable resource for tools and ideas, and even emotional support when confronted with complex problems and the, what seemed to be, near endless graduate career.