A Proposal for a New Neutron Moderator Material for Accelerator-Based BNCT

Total Page：16

File Type：pdf, Size：1020Kb

A proposal for a new neutron moderator material for Accelerator-Based BNCT. Burlon Alejandro, Kreiner Andres, Minsky Daniel1 and Valda Alejandro2 1Escuela de Ciencia y Tecnolog´ıa-UNSAM. Argentina// CNEA-Argentina 2Escuela de Ciencia y Tecnolog´ıa-UNSAM - Argentina The use of nuclear reactions induced by accelerated particle beams, as neutron sources for Boron Neutron Capture Therapy (BNCT) appears as a competitive option to nuclear reactors. This technique involves the administration and selective uptake of the capture agent 10B (who has a thermal neutron capture cross section of 3840 b) in tumor cells and a suitable thermal neutron flux. The capture reaction 10B(n,α)7Li generates the heavy fragments α and 7Li that will damage the tumor cells. BNCT aims at treating pathologies that cannot be treated satisfactorily with conventional therapy (especially brain tumors). The neutron spectrum obtained from a nuclear reaction must be optimized using a moderator-reflector assembly before its application to a patient. It is important to get an epithermal neutron beam that termalizes inside the body in order to maximize the thermal flux at the tumor position. In the frame of a production-target optimization for accelerator-based BNCT, feasibility studies for a new moderator material have been performed. This material consists in successive stacks of Al, polytetrafluoroethylene (PTFE), commercially known as Teflon, and LiF. It is easy to build and its cost is relatively low. The interaction neutron cross section in Al and PTFE presents complementary resonances above 30 keV, which reduces the energy of the fast neutrons. On the other hand, the 6Li in the LiF captures the thermal neutrons in order to avoid high doses in the skin, where the boron uptake is greater than in the rest of the healthy tissue. An exhaustive Monte Carlo simulation study has been performed evaluating the doses delivered to a cylindrical water-filled head phantom by a neutron production Li-metal target based on the 7Li(p,n)7Be reaction. A 35.8 cm diameter lead reflector and a 7.5 cm diameter Al/PTFE/LiF moderator were considered. Several moderator thickness have been studied and the figures of merit show that it is possible to reach a 98% Tumor Control Probability for a 6.5 cm deep tumor with a 34 cm thick moderator and a 20 mA proton bean at a bombarding energy of 2.3 MeV in 50 minutes treatment time..

Copy Link

Recommended publications

Spent Nuclear Fuel Pools in the US
Spent Nuclear Fuel Pools in the U.S.: Reducing the Deadly Risks of Storage front cover WITH SUPPORT FROM: WITH SUPPORT FROM: By Robert Alvarez 1112 16th St. NW, Suite 600, Washington DC 20036 - www.ips-dc.org May 2011 About the Author Robert Alvarez, an Institute for Policy Studies senior scholar, served as a Senior Policy Advisor to the Secre- tary of Energy during the Clinton administration. Institute for Policy Studies (IPS-DC.org) is a community of public scholars and organizers linking peace, justice, and the environment in the U.S. and globally. We work with social movements to promote true democracy and challenge concentrated wealth, corporate influence, and military power. Project On Government Oversight (POGO.org) was founded in 1981 as an independent nonprofit that investigates and exposes corruption and other misconduct in order to achieve a more effective, accountable, open, and ethical federal government. Institute for Policy Studies 1112 16th St. NW, Suite 600 Washington, DC 20036 http://www.ips-dc.org © 2011 Institute for Policy Studies [email protected] For additional copies of this report, see www.ips-dc.org Table of Contents Summary ...............................................................................................................................1 Introduction ..........................................................................................................................4 Figure 1: Explosion Sequence at Reactor No. 3 ........................................................4 Figure 2: Reactor No. 3
[Show full text]
Neutron-Induced Fission Cross Section of 240,242Pu up to En = 3 Mev
Neutron-induced ﬁssion cross section of 240;242Pu Paula Salvador-Castineira~ Supervisors: Dr. Franz-Josef Hambsch Dr. Carme Pretel Doctoral dissertation PhD in Nuclear Engineering and Ionizing Radiation Barcelona, September 2014 Curs acadèmic: Acta de qualificació de tesi doctoral Nom i cognoms Programa de doctorat Unitat estructural responsable del programa Resolució del Tribunal Reunit el Tribunal designat a l'efecte, el doctorand / la doctoranda exposa el tema de la seva tesi doctoral titulada __________________________________________________________________________________________ _________________________________________________________________________________________. Acabada la lectura i després de donar resposta a les qüestions formulades pels membres titulars del tribunal, aquest atorga la qualificació: NO APTE APROVAT NOTABLE EXCEL·LENT (Nom, cognoms i signatura) (Nom, cognoms i signatura) President/a Secretari/ària (Nom, cognoms i signatura) (Nom, cognoms i signatura) (Nom, cognoms i signatura) Vocal Vocal Vocal ______________________, _______ d'/de __________________ de _______________ El resultat de l’escrutini dels vots emesos pels membres titulars del tribunal, efectuat per l’Escola de Doctorat, a instància de la Comissió de Doctorat de la UPC, atorga la MENCIÓ CUM LAUDE: SÍ NO (Nom, cognoms i signatura) (Nom, cognoms i signatura) President de la Comissió Permanent de l’Escola de Doctorat Secretària de la Comissió Permanent de l’Escola de Doctorat Barcelona, _______ d'/de ____________________ de _________ Universitat Politecnica`
[Show full text]
Progress in Fusion-Driven Transmutation Research in China
Progress in Fusion-Driven Transmutation Research in China Yican WU Institute of Plasma Physics, Chinese Academy of Sciences P.O. Box 1126, Hefei, Anhui, 230031, China E-mail: [email protected] Fax: +86 551 5591310 1. Introduction Although the recent experiments and associated theoretical studies of fusion energy development have proven the feasibility of fusion power, it's commonly realized that it needs hard work before pure fusion energy could commercially and economically utilized. On the other hand, the fission nuclear industry has been falling on hard times recently since so far there has been no conclusion about how to deal with the long-lived wastes produced from the nuclear spent fuel and about how to solve the shortage of natural uranium ore in addition to nuclear safety and proliferation. It's a natural way to develop fusion-fission hybrid reactors including fuel producing reactors and waste transmuting reactors as an alternate strategy to speed up the time for producing energy since a hybrid reactor as a subcritical system can operate with lower fusion energy gain ratios Q, therefore the design of the fusion core for a hybrid system is easier than for a pure fusion reactor. The fusion-fission hybrid concept dates back to the earliest days of the fusion project when it was recognized that using fusion neutrons to breed nuclear fuel would vastly increase the energy from fusion plant. It appears to receive almost no attention since the mid 80's in the world, except in China who has given very serious consideration and has strong hybrid reactor activities since 1986 in the framework of the National Hi’Tech Program supported by the State Science and Technology Commission (SSTC) of China.
[Show full text]
PGNAA Neutron Source Moderation Setup Optimization
Submitted to ‘Chinese Physics C PGNAA neutron source moderation setup optimization Zhang Jinzhao1（张金钊）Tuo Xianguo1（庹先国） (1.Chengdu University of Technology Applied Nuclear Techniques in Geoscience Key Laboratory of Sichuan Province，Chengdu 610059，China) Abstract: Monte Carlo simulations were carried out to design a prompt γ-ray neutron activation analysis (PGNAA) thermal neutron output setup using MCNP5 computer code. In these simulations the moderator materials, reflective materials and structure of the PGNAA 252Cf neutrons of thermal neutron output setup were optimized. Results of the calcuations revealed that the thin layer paraffin and the thick layer of heavy water moderated effect is best for 252Cf neutrons spectrum. The new design compared with the conventional neutron source design, the thermal neutron flux and rate were increased by 3.02 times and 3.27 times. Results indicate that the use of this design should increase the neutron flux of prompt gamma-ray neutron activation analysis significantly. Key word: PGNAA; neutron source; thermal neutron; moderation; reflection 1. Introduction study, Monte Carlo calculation was carried out for the Prompt gamma ray neutron activation analysis design of a 252Cf neutron source moderation setup for the (PGNAA) is a rapid, nondestructive, powerful analysis cement samples[7]. The model of Monte Carlo multielemental analysis technique, large samples of some simulation was verified by experiment[8, 9].We improve minor, trace light elements and is used in industrial the thermal neutron source yield rate of 252Cf neutron by control[1-5]. In a PGNAA analysis, the sample nuclear the PGNAA neutron source structure to the design. The composition is determined from prompt gamma rays calculation results for the new design were compared which produced through neutron inelastic scattering and with the previous, example: themal neutron flux rate, fast thermal neutron capture.
[Show full text]
Nuclear Weapons Technology 101 for Policy Wonks Bruce T
NUCLEAR WEAPONS TECHNOLOGY FOR POLICY WONKS NUCLEAR WEAPONS TECHNOLOGY 101 FOR POLICY WONKS BRUCE T. GOODWIN BRUCE T. GOODWIN BRUCE T. Center for Global Security Research Lawrence Livermore National Laboratory August 2021 NUCLEAR WEAPONS TECHNOLOGY 101 FOR POLICY WONKS BRUCE T. GOODWIN Center for Global Security Research Lawrence Livermore National Laboratory August 2021 NUCLEAR WEAPONS TECHNOLOGY 101 FOR POLICY WONKS | 1 This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory in part under Contract W-7405-Eng-48 and in part under Contract DE-AC52-07NA27344. The views and opinions of the author expressed herein do not necessarily state or reflect those of the United States government or Lawrence Livermore National Security, LLC. ISBN-978-1-952565-11-3 LCCN-2021907474 LLNL-MI-823628 TID-61681 2 | BRUCE T. GOODWIN Table of Contents About the Author. 2 Introduction . .3 The Revolution in Physics That Led to the Bomb . 4 The Nuclear Arms Race Begins. 6 Fission and Fusion are "Natural" Processes . 7 The Basics of the Operation of Nuclear Explosives. 8 The Atom . .9 Isotopes . .9 Half-life . 10 Fission . 10 Chain Reaction . 11 Critical Mass . 11 Fusion . 14 Types of Nuclear Weapons . 16 Finally, How Nuclear Weapons Work . 19 Fission Explosives . 19 Fusion Explosives . 22 Staged Thermonuclear Explosives: the H-bomb . 23 The Modern, Miniature Hydrogen Bomb . 25 Intrinsically Safe Nuclear Weapons . 32 Underground Testing . 35 The End of Nuclear Testing and the Advent of Science-Based Stockpile Stewardship . 39 Stockpile Stewardship Today . 41 Appendix 1: The Nuclear Weapons Complex .
[Show full text]
Neutron Capture Cross Sections of Cadmium Isotopes
Neutron Capture Cross Sections of Cadmium Isotopes By Allison Gicking A thesis submitted to Oregon State University In partial fulfillment of the requirements for the degree of Bachelor of Science Presented June 8, 2011 Commencement June 17, 2012 Abstract The neutron capture cross sections of 106Cd, 108Cd, 110Cd, 112Cd, 114Cd and 116Cd were determined in the present project. Four different OSU TRIGA reactor facilities were used to produce redundancy in the results and to measure the thermal cross section and resonance integral separately. When the present values were compared with previously measured values, the differences were mostly due to the kind of detector used or whether or not the samples were natural cadmium. Some of the isotopes did not have any previously measured values, and in that case, new information about the cross sections of those cadmium isotopes has been provided. Table of Contents I. Introduction………………………………………………………………….…….…1 II. Theory………………………………………………………………………...…...…3 1. Neutron Capture…………………………………………………….….……3 2. Resonance Integral vs. Effective Thermal Cross Section…………...………5 3. Derivation of the Activity Equations…………………………………....…..8 III. Methods………………………………………………………….................…...…...12 1. Irradiation of the Samples………………………………………….….....…12 2. Sample Preparation and Parameters………………..………...………..……16 3. Efficiency Calibration of Detectors…………………………..………....…..18 4. Data Analysis…………………………………...…….………………...…..19 5. Absorption by 113Cd……………………………………...……...….………20 IV. Results………………………………………………….……………..……….…….22
[Show full text]
EVALUATION of NEUTRON CROSS SECTIONS for Cm-242, -243, -244
XA0100537 -67- UDC 539.172 EVALUATION OF NEUTRON CROSS SECTIONS FOR Cm-242, -243, -244 A.I. Blokhin, A.S. Badikov, A.V. Ignatyuk, V.P. Lunev, V.N. Manokhin, G. Ya. Tertychny, K.I. Zolotarev Institute of Physics and Power Engineering, Obninsk, Russia EVALUATION OF NEUTRON CROSS SECTIONS FOR Cm-242, -243, -244. The work is devoted to the analysis of available experimental and evaluated data on the neutron cross sections for Cm-242, Cm-243, Cm-244. A comparison of experimental data with the results of theoretical calculations and the evaluations of the most important cross sections were performed. As a result the new version of complete files of evaluated neutron cross sections for Cm-242, Cm-243, Cm- 244 were performed. These files were included into the BROND-3 library the formations of which is under development in Russian Nuclear Data Center. For curium isotopes there are considerably less experimental data than for americium isotopes. The measurements of the total and radiative capture cross sections are restricted to the neutron resonance region and only the fission cross sections were measured in the fast neutron energy region. That is why all the neutron cross section evaluations are mainly based on the analysis of the available measured fission cross sections and on the statistical calculations of other cross sections. For Cm-245 and Cm-246 the new experimental fission cross sections, obtained in the framework of the Project [18] agree good enough with the available experimental data and the new cross section evaluations made by Minsk's group in 1996 [1, 2].
[Show full text]
Fundamentals of Nuclear Power
Fundamentals of Nuclear Power Juan S. Giraldo Douglas J. Gotham David G. Nderitu Paul V. Preckel Darla J. Mize State Utility Forecasting Group December 2012 Table of Contents List of Figures .................................................................................................................................. iii List of Tables ................................................................................................................................... iv Acronyms and Abbreviations ........................................................................................................... v Glossary ........................................................................................................................................... vi Foreword ........................................................................................................................................ vii 1. Overview ............................................................................................................................. 1 1.1 Current state of nuclear power generation in the U.S. ......................................... 1 1.2 Nuclear power around the world ........................................................................... 4 2. Nuclear Energy .................................................................................................................... 9 2.1 How nuclear power plants generate electricity ..................................................... 9 2.2 Radioactive decay .................................................................................................
[Show full text]
Testing of Evaluated Transactinium Isotope Neutron Data and Remaining Data Requirements
Testing of Evaluated Transactinium Isotope Neutron Data and Remaining Data Requirements H. Kiisters Nuclear Research Center Karlsruhe Institute of Neutron Physics and Nuclear Engineering 1. Introduction The first international meeting on transactinium isotope nuclear data, held in Karlsruhe in 1975 /l/, mainly dealt with neutron cross-sections and other nuclear properties as e.g. decay constants of minor actinide isotopes, abbreviated in this paper to MINACs. For the plutonium and the thorium fuel cycle, the MINAC isotopes are 230Th up to 244C, with the exception of the main nuclei 232Th, 233U, 235U, 23aU, 23gPu, 240Pu and 241Pu (sometimes the higher Pu isotopes are incorrectly counted as MINACs; for special interests the curium isotopes beyond mass number 244 are also included in the MINAC list). Minor actinide isotopes are understood to be of "minor importance" in thermal and fast reactor design. This minor importance is regarded with respect to the influence of these nuclei on the reactivity balance or the neutron spectrum of a nuclear plant. However, the MINACs can be of prime importance in the out-of-pile stages of the nuclear fuel cycle, especially in determini% the decay heat and the neutron as well as alpha, beta and gamma radiation of spent nuclear fuel after discharge from the reactor. These aspects are essential for designing safe transport casks and inter- mediate storage facilities, reprocessing units and waste disposal sites. The intensive discussion on the nuclear data aspects of MINACs originated from the predicted expansion of nuclear energy after the first oil crisis in 1973. Reprocessing units of sizes up to 1500 to/yr throughput were dis- cussed, recycling of Pu in thermal reactors was, and in some countries is considered as a serious strategy besides the introduction of Pu into fast ~breeder reactors; also U-recycling came into the discussion.
[Show full text]
Behaviour of Neutron Moderator Materials at High Temperatures in CASTOR®-Casks: Qualification and Assessment
14th International Symposium on the Packaging and Paper # 235 Transportation of Radioactive Materials (PATRAM 2004), Berlin, Germany, September 20-24, 2004 Behaviour of Neutron Moderator Materials at High Temperatures in CASTOR®-Casks: Qualification and Assessment Torsten Krietsch1), Dietmar Wolff1), Ulrich Knopp2) and Hans-Dieter Brocke3) 1) Federal Institute for Materials Research and Testing (BAM), Berlin, Germany 2) Gesellschaft für Nuklear-Behälter mbH (GNB), Essen, Germany 3) TÜV Rheinland Group, Berlin, Germany 1. Introduction The Federal Institute for Materials Research and Testing (BAM) is the responsible German authority for the assessment of mechanical and thermal designs of transport and storage casks for radioactive materials. BAM checks up the proofs of the applicants in their safety reports and assesses the conformity to the Regulations for the Safe Transport of Radioactive Material [1, 2]. One applicant is the Gesellschaft für Nuklear-Behälter mbH (GNB) with a new generation of transport and storage casks of CASTOR®-design. GNB typically uses ultra high molecular weight Polyethylene (UHMW-PE) for the moderation of free neutrons. Rods made of UHMW-PE are positioned in axial bore holes in the wall of the cask and plates of UHMW-PE are in free spaces between primary and secondary lid and between the bottom of the cask and an outer plate (Figure 1). Because of the heat generated by the radioactive inventory and because of a strained spring at the bottom of every bore hole, UHMW- PE is subjected to permanent thermal and mechanical loads as well as loads from gamma and neutron radiation. UHMW-PE has been used under routine- and normal conditions of transport for maximum temperatures up to 130 °C.
[Show full text]
Integrated Waste Assay System (Iwas) and Analysis Enhancements
WM’05 Conference, February 27 - March 3, 2005, Tucson, AZ INTEGRATED WASTE ASSAY SYSTEM (IWAS) AND ANALYSIS ENHANCEMENTS R.D. McElroy, Jr. and S. Croft Canberra Industries, Inc. ABSTRACT The Integrated Waste Assay System (IWAS) is a hybrid waste assay system developed to provide high throughput and improved data quality for characterization of transuranic wastes contained within 200 liter drums and 320 liter overpacks. The IWAS combines high efficiency passive neutron coincidence counting, active neutron interrogation and high resolution gamma-ray spectroscopy into a single assay cavity. The complementary nature of the multiple assay modes yields a higher success rate than any single assay mode alone allowing the system to accommodate a wide range of source and matrix types. Performance of the individual assay modes and performance of the combined analyses is discussed. Modifications and enhancements of the traditional analysis algorithms for the passive and active neutron analyses were implemented in the IWAS. Reductions of the passive neutron detection levels of up to 50% have been achieved without the need to increase the neutron detection efficiency. Utilization of the Add-A- Source matrix correction measurement for determination of the active interrogation moderator index provided a means to implement a moderator correction for low Pu mass samples and uranium only samples. The modified algorithms and performance improvements are discussed. The IWAS analysis makes use of data obtained from the three assay modalities to identify and flag problem waste drums. The analysis makes use of the Total Measurement Uncertainty (TMU) analysis for each assay modality to identify the mode providing the result that most likely represents the actual drum contents.
[Show full text]
System Studies of Fission-Fusion Hybrid Molten Salt Reactors
University of Tennessee, Knoxville TRACE: Tennessee Research and Creative Exchange Doctoral Dissertations Graduate School 12-2013 SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS Robert D. Woolley University of Tennessee - Knoxville, [email protected] Follow this and additional works at: https://trace.tennessee.edu/utk_graddiss Part of the Nuclear Engineering Commons Recommended Citation Woolley, Robert D., "SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS. " PhD diss., University of Tennessee, 2013. https://trace.tennessee.edu/utk_graddiss/2628 This Dissertation is brought to you for free and open access by the Graduate School at TRACE: Tennessee Research and Creative Exchange. It has been accepted for inclusion in Doctoral Dissertations by an authorized administrator of TRACE: Tennessee Research and Creative Exchange. For more information, please contact [email protected]. To the Graduate Council: I am submitting herewith a dissertation written by Robert D. Woolley entitled "SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS." I have examined the final electronic copy of this dissertation for form and content and recommend that it be accepted in partial fulfillment of the equirr ements for the degree of Doctor of Philosophy, with a major in Nuclear Engineering. Laurence F. Miller, Major Professor We have read this dissertation and recommend its acceptance: Ronald E. Pevey, Arthur E. Ruggles, Robert M. Counce Accepted for the Council: Carolyn R. Hodges Vice Provost and Dean of the Graduate School (Original signatures are on file with official studentecor r ds.) SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS A Dissertation Presented for the Doctor of Philosophy Degree The University of Tennessee, Knoxville Robert D.
[Show full text]