Czech Technical University in Prague

Transcription

1 Czech Technical University in Prague Dissertation Thesis 1

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3 Czech Technical University in Prague Faculty of Nuclear Sciences and Physical Engineering Department of Nuclear Reactors Ing. Ondřej Svoboda Experimental Study of Neutron Production and Transport for ADTT Dissertation Thesis Prague, 2011 III

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5 Declaration I declare that this dissertation thesis was done in the internal and combined form of postgradual study at the Department of Nuclear Reactors at the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague. All data, tables, figures and ideas stated in this work are results of my own work unless otherwise stated and referred. This work has not been submitted for any other qualification to this or any other university. Aspirant: Ing. Ondřej Svoboda Postgradual study program: Application of Natural Sciences Study field: Nuclear Engineering Supervisor: RNDr. Vladimír Wagner, CSc. Affiliation: Department of Nuclear Spectroscopy Nuclear Physics Institute Academy of Sciences of the Czech Republic public research institution Řež near Prague V

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7 Acknowledgements My thanks belong to: RNDr. Vladimír Wagner CSc., my great supervisor. He was always ready selflessly to help and guide me, with huge patience and with sense for showing the problems in consequences. the head of the department of Nuclear Spectroscopy at Nuclear Physics Institute Řež, RNDr. Andrej Kugler CSc., for his support, understanding and valuable suggestions and remarks. my colleague Mgr. Antonín Krása PhD., for the support and care that he gave me from my early beginnings in the Academy of Sciences up to now. my colleague Mitja Majerle PhD. for his dedication during introducing me to MCNPX simulations and helping me programming. my colleague Ing. Marek Fikrle, for preparing of iodine samples, the help with irradiations on LVR-15 reactor, and for valuable advice on the field of HPGe detectors and activation analysis. PhD students Ing. Jitka Vrzalová and Ing. Martin Suchopár and also to grammar school students Pavel Motal, Ondřej Novák, and Ondřej Sláma, same as to our foreign students Anne Larédo, Gail de Cargouet, Tazio Torrieri, and Havard Farder for their kind cooperation and a lot of interesting questions that forced me to think again about the physics I am dealing with and which lead me to better understanding. This work would not exist without the kind help of the people around the accelerators in Dubna, Uppsala, and Řež; namely M. I. Krivopustov, A. Prokofiev, P. Bém and theirs teams. I am also grateful to J. Frána, for enabling me to use his HPGe gamma-detector during measurements in Řež and the program DEIMOS32 for evaluating of measured gamma-spectra. Last but not least I would like to express thanks to my parents Zdena and František Svobodovi and to my girlfriend Kateřina Blažková for support, encouragement and love they gave me. This work was financially supported from the Grant Agency of the Czech Republic (grant No. 202/03/H043), Internal Grant Competition (grant number CTU ), Grant Agency of the Academy of Sciences of the Czech Republic (grant No. K ), F4E program of the Nuclear Reaction Department of the Nuclear Physics Institute (grant number F4E-2008-GRT-014), and from the EFNUDAT (European Facilities for Nuclear Data Measurements). VII

8 Abstract High energy neutron production in spallation reactions and their transport in the system of massive lead target and uranium blanket were studied within the international project Energy and Transmutation of Radioactive Waste. A setup called Energy plus Transmutation placed in Dubna, Russia, was irradiated with 1.6 GeV up to 4 GeV deuterons. Threshold reactions on activation detectors from Al, Au, Bi, Co, In, Ta, and Y were used for neutron measurements. Activated foils were measured on HPGe detectors. Spectroscopic corrections were applied during data analysis to find the yields of produced isotopes. The experimental results were compared with MCNPX calculations. These experiments are a continuation of previous research of the above mentioned setup with relativistic protons. No serious disagreement in neutron production to backward angles was observed for deuteron experiments on contrary to the proton ones. Cross-sections of used threshold reactions were measured on quasimonoenergetic neutron sources at Nuclear Physics Institute in Řež and at The Svedberg Laboratory in Uppsala, Sweden. In total eleven irradiations were done in the energy range MeV. Threshold reactions were measured up to (n,10n), the results were compared with the data from EXFOR, EAF, and with the calculated values from TALYS code with good agreement. Cross-sections for reactions over 40 MeV and (n,4n) are unique and were measured for the first time. A part of the data has already been published and presented at international conferences. Key words: spallation reaction, Energy plus Transmutation of Radioactive Waste, neutron activation analysis, HPGe gamma-ray detector, gamma-spectroscopy, MCNPX code, threshold reaction, cross-section. VIII

9 Abstrakt Produkce vysokoenergetických neutronů ve spalačních reakcích a jejich transport v systému masivního olověného terče a uranového blanketu byly studovány v rámci mezinárodního projektu Energy and Transmutation of Radioactive Waste. Sestava nazvaná Energy plus Transmutation umístěná v Dubně, Rusko, byla ozářena deuterony o energiích 1,6 GeV až 4 GeV. Pro měření neutronů byly použity prahové reakce na aktivačních detektorech z Al, Au, Bi, Co, In, Ta a Y. Záření gama aktivovaných fólií bylo měřeno pomocí polovodičových HPGe detektorů. Při analýze získaných dat byla aplikována řada spektroskopických korekcí za účelem nalezení výtěžku sledovaných isotopů. Experimentální data byla nakonec porovnána s výsledky simulací sestavy provedených pomocí programu MCNPX. Tyto experimenty navázaly na předchozí výzkum zmíněné sestavy pomocí relativistických protonů. Pro deuteronové experimenty nebyla na rozdíl od protonových pozorována žádná výraznější neshoda v produkci vysokoenergetických neutronů do zpětných úhlů. Účinné průřezy užitých prahových reakcí byly změřeny pomocí quasimonoenergetických neutronových zdrojů v Ústavu jaderné fyziky, Řež, a ve Svedbergově laboratoři, Uppsala, Švédsko. Bylo provedeno celkem 11 ozařování v energetickém rozsahu 17 až 94 MeV. Prahové reakce byly změřeny až do (n,10n), výsledky byly porovnány s daty z databází EXFOR, EAF a s hodnotami vypočtenými pomocí programu TALYS. Byla pozorována dobrá shoda. Účinné průřezy pro reakce nad 40 MeV a (n,4n) jsou unikátní a byly změřeny vůbec poprvé. Část naměřených dat již byla publikována a prezentována na mezinárodních konferencích. Klíčová slova: tříštivá reakce, Energy and Transmutation of Radioactive Waste, neutronová aktivační analýza, HPGe detektor záření gama, spektroskopie gama záření, program MCNPX, prahová reakce, účinný průřez. IX

10 List of abbreviations ABC Accelerator Based Conversion ADEP Accelerator Driven Energy Production ADS Accelerator Driven System ADTT Accelerator Driven Transmutation Technology AFCI Advanced Fuel Cycle Initiative AGS Alternating Gradient Synchrotron AMAVET Asociace pro mládež, vědu a techniku ATW Accelerator Transmutation of Waste ASCR Academy of Sciences of the Czech Republic barn unit of cross-section used in nuclear physics (1 barn = m 2 ) CEA Commissariat à l énergie atomique CERN Conseil Européen pour la Recherche Nucléaire (European Organization for Nuclear Research) CNGS CERN Neutrinos to Gran Sasso project CONFIRM Collaboration on Nitride Fuel Irradiation and Modeling CSMSR Cascade Subcritical Molten Salt Reactor CSNS China Spallation Neutron Source E+T Energy plus Transmutation setup E&T RAW Energy and Transmutations of Radioactive Waste project EAF European Activation File EFNUDAT European Facilities for NUclear DATa measurements ENDF Evaluated Nuclear Data File EPAC European Particle Accelerator Conference ESS European Spallation Source EUROTRANS EUROpean Research Programme for the TRANSmutation X

11 EUROPART EUROpean research programme for the PARTitioning of minor actinides ev electron volt EXFOR Experimental Nuclear Reaction Data FUS FUsion association GeV gigaelectron volt GW e GigaWatt electrical GWd/MTHM GigaWatt days per Metric Ton of Heavy Metal HINDAS High and Intermediate energy Nuclear Data for Accelerator-driven System HM Heavy Metal HPGe High Purity Germanium detector IAEA International Atomic Energy Agency IEEE Institute of Electrical and Electronics Engineers INDC International Nuclear Data Committee INR Institute for Nuclear Research of the Russian Academy of Sciences ISNS India Spallation Neutron Source JASNAPP nuclear spectroscopy on proton beam (from Russian) JINR Joint Institute for Nuclear Research, Dubna, Russia J-PARC Japan Proton Accelerator Research Complex kev kiloelectron volt kw kilowatt LAHET Los Alamos High-Energy Transport code LAMF Los Alamos Meson Physics Facility LANSCE Los Alamos Neutron Science Center LEDA Low Energy Demonstration Accelerator MCNP Monte-Carlo N-Particle MCNPX Monte-Carlo N-Particle extended XI

12 MEGAPIE MEGAwatt spallation target PIlot Experiment MeV Megaelectron Volt MLF Material and Life science Facility of the J-PARC MOX Mixed Oxide Fuel MTIHM Metric Tons of Initial Heavy Metal MW - megawatt MYRRHA Multi-purpose hybrid research reactor for high-tech applications NPI Nuclear Physics Institute of the Academy of Sciences of the Czech Republic NuMI Neutrinos at the Main Injector NWB Nuclear Waste Burner OMEGA Options Making Extra Gain from Actinides and fission products PEFP Proton Engineering Frontier Project PSI Paul Scherrer Institut PSR Proton Storage Ring of the LAMF SINQ - Schweizer Institut fur Nuklearforschung Quelle SNS Spallation Neutron Source in Oak Ridge, USA SSNTD Solid State Nuclear Track Detectors SÚRAO Správa úložišť radioaktivních odpadů (RAWRA - Radioactive Waste Repository Authority) TEF Transmutation Experimental Facility of the J-PARC TEF-P Transmutation Physics Experimental Facility in TEF of the J-PARC TEF-T ADS Target Test Facility in TEF of the J-PARC TIARA Takasaki Ion accelerators for Advanced Radiation Application TRIUMF Tri-University Meson Facility TSL The Svedberg Laboratory of the Uppsala University, Sweden UKAEA United Kingdom Atomic Energy Autohority XADS experimental Accelerator Driven System XII

13 Table of contents Introduction Accelerator Driven Systems Motivation for transmutation studies Transmutation Spallation reaction History of accelerator driven systems Modern spallation neutron sources Concepts of accelerator driven transmutation technologies Spallation neutron sources for ADTT research Experiments focused on nuclear data measurements Summary of ADS research goals Energy and Transmutation of Radioactive Waste project Introduction to the E&T RAW project Gamma E+T setup Gamma Kvinta setup EZHIK Placement of the E&T RAW targets Experimental background Activation detectors Correction on decay of the isotope between the end of irradiation and beginning of the measurement Correction on decay during irradiation Correction on the intensity of the I transition Correction on dead-time of the detector Correction on real - cascade coincidence Correction on changed detector efficiency due to sample dimensions Self-absorption correction Square-emitter correction (geometrical correction) Beam instability correction HPGe detectors DEIMOS32 program Yield evaluation Sources of uncertainties Background Beam diagnostics on Nuclotron accelerator Nuclotron accelerator Irradiation course Beam position and shape Beam intensity XIII

14 5. E+T results of deuteron irradiation Plain experimental results Ratios of yields for different thresholds Spectral indexes Comparisons between deuteron experiments Total neutron production MCNPX simulations of the Energy plus Transmutation setup MCNPX code Limitations of MCNPX code Simulation of the E+T setup Neutron fluxes in the E+T setup Calculation of the yields in used activation foils Normalized experiment/simulation ratios Yields for different beam particles of the same total energy Summary of the MCNPX simulations Cross-section measurements of the (n,xn) threshold reactions State-of-the-art of the neutron cross-section libraries Limitations on neutron source EFNUDAT project Quasi-monoenergetic neutron source at The Svedberg laboratory Cross-section estimation and planning of the experiment Neutron beams at TSL Quasi-monoenergetic neutron source at Nuclear Physics Institute Studied materials Evaluation procedure Background subtraction Uncertainty analysis Discussion of the cross-section results TALYS Introduction to TALYS Comparison among various models Comparison between TALYS 1.0 and TALYS Conclusion Appendix A - Threshold and non-threshold reactions on activation samples Appendix B - Placement of the foils during Energy plus Transmutation deuteron experiments Appendix C - List of spectra measured in E+T deuteron experiments Appendix D - Correction factor on beam instability Appendix E - Examples of correction factors on real coincidences Appendix F - Yields of isotopes produced on activation foils during 1.6 and 2.52 GeV deuteron experiments on Energy plus Transmutation setup Appendix G - Graphs with yields of isotopes produced on activation foils in E+T deuteron experiments G.1. Longitudinal yields at 3 cm over the target axis XIV

15 G.2. Radial yields in the first gap G.3. Spectral indexes G.4. Ratios of the yield in dependence on the threshold G.5. Comparison between experiments G.6. Ratios of the yields for various deuteron experiments Appendix H - Example of MCNPX input file Au in 4 GeV deuteron experiment Appendix I - Results of MCNPX simulations I.1. Deuteron and proton spectra I.2. Experiment/simulation ratios I.3. Normalized experiment/simulation ratios Appendix J - Cross-sections of threshold reactions from EXFOR and TALYS compared with my data Appendix K - Comparison between TALYS 1.0 and TALYS Appendix L - Measured cross-section values Appendix M - Equations of detector calibration for Excel Addin Bibliography List of tables List of figures XV

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17 Introduction Spallation reaction as a perspective source of neutrons has been studied with an increased interest in the last two decades. These studies are motivated by the need of high neutron fluxes for material research, transmutation of nuclear waste or production of nuclear fuel from thorium. New spallation sources are planned (European Spallation Source) or already commissioned (American Spallation Neutron Source) to fulfill scientist requirements. With advances in accelerator technology Accelerator Driven Systems, thanks to their high safety and unique properties, seem to be a perspective energy source for the future. This work is a part of the international research program Energy and Transmutation of Radioactive Waste. Within this project, groups from 15 countries study various aspects of spallation reaction, neutron production, transport and its usage for transmutation of nuclear waste. Six different setups of massive target surrounded with blanket and neutron moderator are used to measure differential as well as global data for ADS (chapter 2). Three of the setups are already acknowledged as IAEA benchmark targets. This thesis is experimentally oriented and discusses results of deuteron irradiations of the Energy plus Transmutation setup. This setup consists of a massive lead target surrounded with natural uranium blanket and polyethylene biological shielding. High energy neutrons from spallation reactions were measured using activation detectors from Al, Au, Bi, Co, In, Ta, and Y materials. A detailed description of used activation foils, reactions, HPGe detectors, and spectroscopic corrections is written in chapter 3. Aluminum and copper activation foils were also used to measure beam intensities, positions and shapes of all deuteron irradiations, details are stated in chapter 4. The (n,xn), (n, ), and (n,p) threshold reactions have been used to distinguish neutrons with different energies. Non-threshold (n, ) reactions with combination of polyethylene shielding have been used to assess total number of produced neutrons. Experimental results and comparisons are in chapter 5. This thesis carries on the work of my colleagues Antonín Krása and Mitja Majerle, who studied in their PhD theses properties of the Energy plus Transmutation setup irradiated with proton beams. The main aim of this work is to study the high energy neutrons in already well known setup irradiated with different (deuteron) beams. Various spectroscopic corrections are studied and routinely applied for the first time in order to produce more precise results. Experiments have been performed using the Nuclotron Accelerator at the Veksler and Baldin Laboratory of High Energy Physics of the Joint Institute for Nuclear Research (JINR) in Dubna, Russia. Energy plus Transmutation setup was irradiated with deuterons of 1.6, 2.52, and 4 GeV. Irradiated foils were measured using HPGe detectors at JINR and Nuclear Physics Institute (NPI) of the Academy of Sciences of the Czech Republic (ASCR). MCNPX simulations of the experiments were done and calculated data was compared with the experimental data in chapter 6. 1

18 After a long time of using threshold detectors at NPI there appeared a opportunity to measure their cross-sections in the energy regions where no data had existed so far. Up to now only calculated cross-sections were used for the reactions over 40 MeV or order of reaction higher than (n,4n). With the financial support from European Facilities for Nuclear Data Measurements grant organization (EFNUDAT) quasi-monoenergetic neutron source at The Svedberg Laboratory (TSL) at Uppsala, Sweden, was used. Three irradiations with energies 22, 47, and 94 MeV were performed in June They were supplemented with 17, 22, 30, and 35 MeV irradiations on similar neutron source at NPI Řež. A detail description of used neutron sources, evaluation procedure and neutron background subtraction as well as cross-section results can be found in chapter 7. The analysis of the deterministic code TALYS which was used for the neutron background subtraction at cross-section measurements is presented in Chapter 8. Two versions of TALYS (1.0 and 1.2) are compared, the same as different settings of the TALYS code. Their influence on the amount of subtracted background and thus on cross-sections is discussed. The summary of the main goals of PhD thesis is following: prepare, perform and evaluate 1.6 GeV and 2.52 GeV deuteron experiments on the E+T setup, study and apply spectroscopic corrections needed for the data evaluation, measure the beam intensities, positions and shapes during 1.6 GeV and 2.52 GeV deuteron experiments on the E+T setup, provide the results to the whole E&T RAW collaboration, compare experimental results within each experiment, between deuteron experiments and with previous proton experiments performed on the E+T setup, perform MCNPX simulation of deuteron experiments, make comparisons between experimental and simulated data, prepare, perform and evaluate cross-section measurements of (n,xn) threshold reactions used for high energy neutron measurements in the E+T setup, namely TSL experiments at 22, 47, and 94 MeV and NPI experiments at 17 and 22 MeV. Beside these PhD goals I have voluntarily worked on some topics of the 4 GeV deuteron experiment. I also show in my PhD thesis these data because they supplement deuteron systematics on the Energy plus Transmutation setup. This thesis was written with respect to its potential users from the Energy and Transmutation community as well as to other students from Nuclear Physics Institute of the ASCR, who are interested in this field of physics. In the work there are maybe more detailed descriptions and examples than would be necessary for the PhD work, but I tried to present a clear description of all the aspects of my work in order to enable easier continuation in these studies. With the constituency of the readers is connected also the choice of used language. 2

19 Chapter 1 Accelerator Driven Systems 1.1. Motivation for transmutation studies First nuclear reactor was started on December 2 nd, 1942, almost 70 years ago. Since that day, nuclear industry has undergone an amazing evolution. Nowadays, 437 energetic nuclear reactors produce 371 GW e (14% of electricity consumption) and 56 new reactors are under construction [1]. Rising demand on electricity and worldwide efforts on decrease of carbon dioxide emissions, as well as the oncoming insufficiency of crude oil will push nuclear industry forward in the next decades. After the gas crisis in the Central Europe in 2009, politically independent energy sources start to have a high importance in many countries. Nuclear energy can be completely independent at least in the period of several years. After the Chernobyl accident, a strong public opinion against the nuclear energy developed all over the world. Some countries even closed their nuclear power plants and become non-nuclear. Today, public meaning is slowly changing and nuclear energy is acceptable for most people under fulfillment of the following rules: - any serious accident with effects outside the power plant area must be reliably excluded, - proliferation of nuclear materials (enriched uranium in fresh fuel, plutonium in spent fuel, fission products etc.) must be out of question due to combination of technical and organizational rules, - time of nuclear power plant construction should by adequate, price of nuclear energy must be comparable to other energy sources, - question of spent fuel and high level radioactive waste generally must be reliably solved out. The last demand has not been fully solved out up to now. The total amount of spent fuel that has been discharged globally is approximately tones of heavy metal (HM). There are nowadays three possible ways how to handle spent fuel store it in geological repositories, reprocess it and store only currently unusable items or involve transmutation after the reprocessing. Geological repositories are one of the possibilities, which cannot be omitted in any scenario of spent fuel handling. Geological repository is a final storage place build deep under the earth surface in suitable rock formation. Special attention is paid to the stability and compactness of the rock massive, same as on the presence of underground water. Dense urban settlements nearby the location as well as the presence of valuable resources in the rock limit the choice of the repository site. Underground repository is based on the principle of multiple physical barriers that should stop potential leak of stored radioactive materials without future human 3

20 1. ACCELERATOR DRIVEN SYSTEMS assistance. Barriers should also ensure safety to future generations, they should embarrass the manipulation with such dangerous materials. Life-time of the deep underground repository is planned to be long enough to let most of the stored radionuclide to decay and to decrease the activity below the natural background level. Figure 1: Geological repository for nuclear waste [2]. Originally it was planned to store whole used fuel rods in the underground repository. This approach poses the easiest and safest way of spent fuel removal, but its massive usage is nowadays improbable because of its unthrift. Reactors can currently use only 3 4 percents of the total energy contained in the fuel. These 3 4 percents of the fuel represents ash after the nuclear burning, mainly high active fission products (most important fission products are summarized in the Table 1 bellow). These radioisotopes cannot be further used and must be separated from the biosphere for a long time (or transmuted). Vitrified fission products are nowadays the most probable content for underground repositories, when they will be opened in the second half of the 21 st century. 4

21 1.1. Motivation for transmutation studies Table 1: Annual production of the most important transuranides and fission fragments in light water reactor of thermal power 3000 MW [3]. Transuranides Production Half-life Fission Production Half-life kg/year [years] fragment kg/year [years] 238 Pu Se Pu Kr Pu Sr Pu Zr Pu Tc Np Pd Am Sn Am I Am Cs Cm Cm Cs Sm Another limitation for the final deposition of spent fuel in the geological repository is a residual heat production. Energy released in the decay of radioactive isotopes is finally converted into heat, which must be safely diverted. In the first years, spent fuel must be cooled in water nearby the reactor, otherwise it would melt. Later it can be stored under a gas atmosphere, but heat removal must be still ensured. In the geological repository, containers with spent fuel (or vitrified fission products) are planned to be buried in bentonite with rock around, so the heat production at that time must be smaller than the possible heat removal by conduction in used materials. Main heat sources in spent fuel are displayed in following Figure 2. Goal of the research is to eliminate components of the nuclear waste stream that account for the majority of the heat load and toxicity over the 300 to year time frame. Build-up and operation of geological repository is a long-distance run, it can take up a century until the repository site will be fully closed. Most of the states that use nuclear energy are in various stages of the repository build up. The Swedish Nuclear Fuel and Waste Management Company (SKB) selected locality Östhammar as the site for a final spent fuel geological repository, following a nearly 20 year process that narrowed the list of applicant sites to two in Site investigations for repositories at Olkiluoto in Finland and in the Bure region in France continued on the schedule with operation targeted for 2020 and 2025, respectively. In the USA, the Government decided to terminate its development of a permanent repository for high level waste at Yucca Mountain. It plans to establish a commission to evaluate alternatives. In the UK, a voluntary sitting process has been initiated, as well as in many other countries. Czech Republic stacks in the process of repository site selection. Initial study of six localities with similar geological underground as in Sweden or Finland was finished, 5

22 1. ACCELERATOR DRIVEN SYSTEMS new locality is studied inside former military area Boletice. Start of the repository construction is planned beyond the year 2050 and operation after 2065 [4]. Figure 2: Dominant decay heat contributors in spent PWR fuel irradiated to 50 GWd/MTIHM [5]. The isotopes circled in red are the major contributors to the decay heat in 300 to year time frame. If these isotopes are removed then the solid blue line shows the decay heat of the remaining waste; the green dashed line shows the time at which the surface temperature of the waste container is below the boiling point of water; and the blue dashed line gives the time at which the waste radiotoxicity is below Class C nuclear waste 1. About 97 percent of the spent fuel contains uranium and plutonium, which can be reused after the reprocessing. Up to now, t of HM spent fuel were already reprocessed. Total global reprocessing capacity is about 5000 t of HM per year. Uranium gained from the reprocessing can be again enriched and fabricated to the fuel. Cumulated amount of plutonium possesses a safety risk, so there is a rising interest in the use of MOX fuel (mixed oxide fuel with uranium-235 partially replaced by plutonium-239). At the beginning of the year 2010 there was a 250 t HM MOX fuel fabrication capacity and 31 thermal reactors licensed for MOX fuel use in the world. 1 USA definition of radioactive waste classification, Class C is similar to Czech definition of low level waste 6

23 1.1. Motivation for transmutation studies Higher actinides contained in the spent fuel cannot be effectively burned in present types of reactors. Higher actinides can be most efficiently eliminated through nuclear transmutation using high intensive fields of fast neutrons Transmutation Transmutation is, generally said, every reaction, in which the composition of the atom nucleus is changed. Nuclei differ apart not only in the number of protons that defines the element, but they differ also in the number of neutrons. Neutrons impress besides other the stability of the nucleus. Adding or removing of a neutron can lead to a dramatic change in the nucleus, a new (stable) element can be produced in the following decay. Transmutation reactions are quite common in the nature. Production of 14 C and tritium production in the upper parts of atmosphere can be introduced as an example of cosmic rays induced transmutation reactions N n C H N n C H In 1951 Sir John D. Cockroft and Ernst T. S. Walton obtained the Nobel Prize for discovery of the transmutation of atom nucleus by accelerated particles. Fission products in the burned-up fuel are mostly -radioactive with a short halflife. Only a few of them are long-lived. To make these materials stable, multiple neutron absorption and consequent nucleus decay or fission is needed. Typical example can be the 99 Tc with the half-life years. 1 Figure 3: Transmutation of 99 Tc [6]. Plutonium and higher actinides which cannot be easily fissioned in thermal reactor can be also transmuted. A single neutron capture can change a non-fissile nuclide to a fissile one, which can be consequently fissioned in proper neutron spectrum. Basic physical requirement for successful transmutation of long lived waste is highly intensive field of neutrons. High transmutation rates can be achieved by combination of high neutron intensity, proper neutron energy and reaction cross-section. 7

24 1. ACCELERATOR DRIVEN SYSTEMS Main difficulty of the transmutation is thus in ensuring strong neutron field of proper energy. Spallation reaction is an ideal source of such neutrons Spallation reaction Spallation reaction is a process, in which a relativistic light ion (proton, deuteron or heavier nucleus) interacts with a massive heavy metal target, resulting in the breakup of the heavy nucleus and in production of wide range of new particles. Substantial parts of these particles are neutrons with relatively high energy. Number of these neutrons depends on the energy and mass of the interacting ion and on the target material. Spallation reaction can be divided into a few stages. Spallation starts with the accelerated proton (for example) interacting with the target nucleus of heavy element (e.g. Pb). The proton penetrates the target nucleus, and distributes its energy to a few nucleons of the nucleus. This stage is called intra-nuclear cascade. Target nucleus is afterwards in highly excited state and undergoes a pre-equilibrium emission of particles and photons. Particles are at this stage of process emitted unisotropicaly, most of them in the forward direction. After this emission, energy is in the nucleus uniformly distributed, but the nucleus is still highly excited. Such a nucleus can than disintegrate or massively evaporate particles and photons to lower its energy. Particle and photon production is isotropic in this phase. Neutrons occurring in the spallation reaction can have a wide range of energies (see e.g. Figure 63 in chapter 6 section 4). Highest energy of the neutrons can reach up to the energy of the particles in the incident beam. In the low energy part of the spectrum number of neutrons is decreasing significantly below the energy one MeV. In order to produce intense thermal neutron fluxes various moderators must be used. Figure 4: Principal schema of the spallation reaction [7]. 8

25 1.3. Spallation reaction With growing energy of the primary particle, course of the reaction substantially changes. In the energy interval GeV all interactions are only on the level of nucleons. Towards higher energies, other reaction channels are opening and new particles are produced in the nucleon-nucleon interactions. In the region hundreds of MeV first pions are produced, at 2 10 GeV heavier hadrons occur. Produced particles can further interact with other nuclei and a hadron showers are developed. First on the list of the accelerated particles used in the spallation sources are protons. Their accelerating is efficiently managed and commonly used process. Most effective energy of the protons used for spallation lies in the region MeV, where the neutron production per MeV per particle has its maximum. A little better situation is for deuterons; they have the same ionization losses (at the same energy per nucleon), but bring twice the amount of energy into the target. On the other hand, deuteron acceleration is more complicated process resulting in lower beam intensities. Very important for the spallation neutron sources is target material selection. Target material must fulfill a wide range of criteria, often contradictory. Suitable target material must have at first good spallation properties - high atom number and density. Moreover good thermal conductivity, low melting point for liquid targets or high melting point for solid targets, generally high boiling point, low activation and only short-lived activation products are required. Target materials for high power ADS can be sorted into three groups: a) liquid non-fissionable targets Targets in the form of molten material Hg, Bi, Pb or eutectics. Main advantage of this conception represents the cooling of such a target, liquid metal can circulate and outer cooling loops can be used. On the other hand, both heavy metals (Bi and Pb) produce long lived products when being irradiated ( 205 Pb 1.53(3) 10 7 years, 208 Bi 3.68(4) 10 5 years, 210 Bi 3,04(6) 10 6 years). b) solid non-fissionable targets Tantalum or wolfram metal formed to wafers. After the irradiation these materials show low radioactivity and residual heat. Wolfram has one of the biggest densities from considered materials 19.3 g/cm 3. c) solid natural uranium or thorium Targets from fissionable materials offer some fast neutron bonus through the fission. Among disadvantages can be named high cross-section for neutron absorption and production of long-lived radioisotopes (e.g. 236 U 2.342(3) 10 7 years). In some scenarios the transmuted material is placed directly into the target. 9

26 1. ACCELERATOR DRIVEN SYSTEMS Table 2: Overview of the properties of the most convenient materials for the spallation targets [8]. Isotope Z Relative atomic mass Density [g/cm 3 ] Melting point [ o C] Boiling point [ o C] Heat capacity [Jg -1 K -1 ] Heat conductivity [Jcm -1 s -1 K -1 ] Ta W Pb Bi Th U Np 93 (237) Pu 94 (244) Am 95 (243) History of accelerator driven systems Accelerator driven systems - ADS (including also accelerator driven transmutation technologies ADTT) come out from following four main research directions: a) ATW (Accelerator Transmutation of Waste) proposed by C. D. Bowman [9] and developed in Los Alamos, USA. Main aim of this project is to substantially shorten the half-life of the isotopes in the spent fuel by means of transmutation. b) ADEP (Accelerator Driven Energy Production) or Energy Amplifier (CERN project) [10] idea of C. Rubbia is based on the fission of 233 U. This isotope of uranium would be produced from thorium in the following reactions: n Th 22,3 m Th 26,967d Pa Th Pa e 233 U e Thorium is the fortieth most frequent element in the Earth crust. Few states headed by India and China have thorium resources, but lack of uranium or fossil fuels to meet their energy needs. Problem in usage of thorium is in the need of neutron source at the beginning of the 233 U production process, one needs something to start the breeding reaction. Spallation source as a representative of a strong neutron sources can be one of the solutions. c) APT (Accelerator Production of Tritium) [11] tritium was formerly used in fusion bombs. Nowadays, there starts to be a strong demand from the fusion community as the tritium is important fuel in fusion reactors. 10

27 1.4. History of accelerator driven systems d) ABC (Accelerator Based Conversion) [11] accelerator steered conversion of plutonium was proposed to liquidate huge plutonium resources from the reprocessing and nuclear weapon programs. Practical use of this research is nowadays less probable thanks to MOX fuels and development of new reactor types, but in the past it was one of the important branches of accelerator driven systems research Modern spallation neutron sources At the beginning of 2010, there were nine spallation neutron sources distributed in five countries. Another 50 synchrotron light sources of neutrons are located in over 20 countries [1]. Most of these neutron sources are used for material science and in related branches. Neutron scattering is one of the most effective ways to obtain information on both, the structure and the dynamics of condensed matter. A wide scope of problems, ranging from fundamental to solid state physics and chemistry, and from materials science to biology, medicine and environmental science, can be investigated with neutrons. Aside from the scattering techniques, non-diffractive methods like imaging techniques can also be applied with increasing relevance for industrial applications. In last decade, new international workplaces with intense spallation neutron sources are being built or planed, below is a list of the most important and strongest ones. European spallation source (ESS) European spallation source is a project, which involve partners from 16 countries. Now it is in the pre-construction phase, in 2012 should start the build-up phase. Spallation source should be commissioned in 2019 and fully operational in 2025 with total cost 1.48 billion Euro [12]. Spallation Neutron Source (SNS) American spallation neutron source located in Oak Ridge National Laboratory is nowadays the world s most advanced high flux neutron source for material science. It was launched in 2006 and offers 18 beam lines to 25 different experiments [13]. Linear accelerator provides 1 GeV H - beam of 1.4 MW to the mercury target (beam current is 1.4 ma, repetition rate 60 Hz). Facility holds Guinness World Record for the most powerful pulsed spallation neutron source. Japan Proton Accelerator Research Complex (J-PARC) In Japan Proton Accelerator Research Complex a mercury target irradiated by 3 GeV H - beam is used to produce neutrons. Current in pulsed proton beam can be up to 0.333mA, but total power deposited in current type of target is only 0.12 MW. Neutrons are guided to various experiments in Material & Life Science Experimental Facility (MLF) [14]. Spallation neutron source SINQ Spallation neutron source SINQ is situated in Paul Scherrer Institut (PSI), Switzerland. Cascade of three accelerators deliver protons with energy 0.59 GeV at a 11

28 1. ACCELERATOR DRIVEN SYSTEMS current up to 2.3 ma [15]. Target is an array of lead rods enclosed in zircaloy tubes and cooled by heavy water. SINQ is designed as a neutron source mainly for research with extracted beams of thermal and cold neutrons, but hosts also facilities for isotope production and neutron activation analysis. China Spallation Neutron Source (CSNS) China started to build their own spallation neutron source in May 2010 and plan to have first neutrons in 2015 [16]. It will be based on 1.6 GeV H - beam at 25 Hz repetition rate and MW power in the first stage (up to 0.5 MW in the third stage). Total cost of the facility is about 293 million US dollars. India Spallation Neutron Source (ISNS) India plans also to build its own spallation neutron source in near future. It will be based on the experiences gathered at existing high flux spallation neutron sources. Proposed parameters are 1 GeV proton beam on lead target, average beam current 0.1 ma at 25 Hz [17]. There is a long row of less powerful and older, but still excellent spallation neutron sources. For example, at Los Alamos National Laboratory - USA, meson physics facility (LAMPF) is working since At Rutherford Appleton Laboratory in Oxfordshire UK, ISIS pulsed neutron and muon source is used since 1985 [18]. Overview of spallation neutron sources from the beam power point of view is on the following Figure 5. Figure 5: Current powerful proton accelerators, SP - short pulsed, CW continuous wave, LP long pulsed [19]. Acronyms are described in the list of abbreviations. 12

29 1.6. Concepts of accelerator driven transmutation technologies 1.6. Concepts of accelerator driven transmutation technologies ADTT (Accelerator Driven Transmutation Technology) can be a future solution for the rising amount of high-level nuclear waste from the nuclear reactors, as well as a new source of energy. It is a combination of a subcritical reactor with an accelerator. The basic principle is in production of a large number of high energy neutrons in the spallation process (relativistic ions + heavy metal target), and their multiplication in sub-critical blanket. In dense field of high energy neutrons lot of actinides and/or fission products can be burned or effectively transmuted to short lived products. This approach can minimize demands on the geological repository. Dense neutron field can be also used to produce fuel from 232 Th. The main advantage of this technology is its safety; switch off of the accelerator means a switch off of the system (with proper design there can hardly be a criticality accident). Figure 6: Scheme of the typical ADS proposal [20] Spallation neutron sources for ADTT research In last decade, three main experiments with spallation neutron sources were started and they were focused on future transmutation use of the accelerator driven systems. Megapie experiment described below studied the behavior of a target under extreme thermal and radiation load. TEF experiment in J-PARC studies behavior of subcritical ADS under various beam conditions. Planned project MYRRHA will combine both directions. 13

30 1. ACCELERATOR DRIVEN SYSTEMS Megawatt Spallation Target Pilot Experiment (Megapie) Megawatt Spallation Target Pilot Experiment was the first project, where the target had the full power load as it is considered for the future ADS systems. Experiment was involved in the Fifth Framework program of the European Union. Megapie was an experiment aiming to demonstrate the safe operation of a liquid metal spallation target at a beam power level of 1 MW in the SINQ target station at the Paul Scherrer Institut (PSI). It was running successfully for four months and accumulated total charge 2.8 Ah [21]. Now the decommissioning of the target is ongoing. Transmutation Experimental Facility (TEF) in J-PARC Transmutation Experimental Facility (TEF) is situated in Japan Proton Accelerator Research Complex. It consists of two experiments: Transmutation Physics Experimental Facility (TEF-P) and ADS Target Test Facility (TEF-T) [22]. TEF-P is equipped with a critical assembly to investigate physical and dynamic properties of the accelerator-driven system by using low power (10W) proton beam. Uranium, plutonium and minor actinide fuels are planned to be loaded into the assembly. TEF-T is a facility to examine the existence of ADS (Accelerator-driven System) by engineering viewpoint. Liquid lead-bismuth spallation target is installed to the TEF-T and is irradiated by 600 MeV-0.2 MW proton beam. Multi-purpose hybrid research reactor for high-tech applications (MYRRHA) SCK CEN, the Belgian Nuclear Research Centre in Mol is the home institute of the MYRRHA project. It will be a multi-purpose hybrid research reactor for high-tech applications. It should replace ageing BR2 reactor, a multi-functional material testing reactor that is in operation since MYRRHA will be a flexible fast spectrum research reactor ( MW th ), it is conceived as an accelerator driven system (ADS), able to operate in sub-critical and critical modes. It contains a proton accelerator of 600 MeV, a spallation target and a multiplying core with MOX fuel, cooled by liquid lead-bismuth (Pb-Bi). Construction of the facility is foreseen in the period , full operation by 2023 [23]. There can be stated a long row of ADS experiments and facilities with focus on transmutation that were proposed and developed in the past, but never transformed into real scientific facility. With the development of new Generation IV reactors and mainly molten salt reactors there is an evident decrease in the interest in ADTT. This is connected also with the funding, so only a few projects have survived. Examples can be found in the Table 3. 14

31 1.7. Spallation neutron sources for ADTT research Table 3: Parameters of different ADS projects [24]. Acronyms are explained in the list of abbreviations. Project / Country ABC-ADTT-ATW -AFCI (USA) OMEGA (Japan 1997) JAERI-ADS (Japan 2004) HYPER (Korea) XADS Design A (Italy) XADS Design B (France) XADS Design C (Belgium) INR (Russia) NWB (Russia) CSMSR (Russia) Accelerator / blanket power [MW] 4.8 / 250 (800 MeV, 6 ma) 58 / 820 (1.5 GeV, 39 ma) 27 / 800 (1.5 GeV 18 ma) 15 / 1000 (1 GeV, ma) 3.6 / 80 (600 MeV, 3-6 ma) 3.6 / 80 (600 MeV, 3-6 ma) 1.75 / 50 (350 MeV, 5 ma) 0.15 / 5 (500 MeV, 10 ma) 3 /100 (380 MeV, 10 ma) 10 / 800 (1 GeV, 10 ma) k eff Flux / spectrum [n/cm 2 s] 0.95 Thermal Pb ThU Target Fuel References 0.9 4x10 15 Fast W Np/5Pu/30Zr 0.97 Fast Pb-Bi MA/Pu/ZrN DOE/RW(1999), Oct, OECD/IAEA (2005) Status Report, 5421 Nakamura et al. (1992), Takizuka et al. (1997) Ikegani et al.(2004), Kikuchi et al. (2004) 0.98 Fast Pb-Bi Ma/Pu Yoo ( Fast Pb-Bi U/Pu/MOX Abderrahim et al. (2004) Fast Steel U/Pu/MOX Abderrahim et al. (2004) x10 15 Fast Pb-Bi windowless U/Pu/MOX Abderrahim et al. (2004) Fast W MA/MOX Markov et al. (2003) Fast Pb-Bi x10 14 Intermediate Pb-Bi UO2/UN U/MA/Zr Np/Pu/MA molten salt Pavlopoulos et al. (2003) Degtyarev et al. (2005, 2006) From US projects mentioned in the Table 3 only the Spallation Neutron Source (SNS) was finished up to now. ATW project was postponed due to its inutility. Japan projects are further developed in the JAERI under the TEF facilities. HYPER project in South Korea was closed in Italian and French XADS (experimental Accelerator Driven System) stayed up to now only in the planning phase. Belgium XADS project developed into European project called MYRRHA with start of construction in All three Russian projects were stopped in the planning phase because of the lack of money Experiments focused on nuclear data measurements Cross-sections of various reactions are of fundamental importance for future ADS. Many construction materials, which are nowadays commonly used in nuclear reactors, will be exposed to extreme neutron fluxes of high energies. At this region of energies, only very few cross-sections are known. Precision of cross-section knowledge is even more important for materials of transmutation interest. Bad knowledge of crosssections and properties of nuclear reactions in general can lead in production of even longer-lived isotopes than is the transmuted one or at least to low transmutation rates. On the other hand, with good knowledge of cross-sections and thus proper choice of neutron energy and time of irradiation, negative effects can be minimized or eliminated. A few international initiatives were established to gain nuclear data for future ADS. Within the Fifth Framework Programme (FP5) of the European Union following researches were done [25]: 15

32 1. ACCELERATOR DRIVEN SYSTEMS Thorium Cycle project Thorium cycle project coordinated by Nuclear Research and Consultancy Group from Netherlands was focused on the measurement of key data for thorium fuel cycle in reactors and ADS systems. Various mixtures of Th/Pu fuel were studied under high burn-up in order to decrease its long-term radioactivity and thus demands on geological repositories. CONFIRM CONFIRM project was collaboration on Nitride Fuel Irradiation and Modeling. Research was oriented on the oxide and nitride ADS fuels without uranium. Special design of fuel pellets was developed in order to reach extremely high burn up. Special attention was paid to safety parameters of the fuel, which must be fulfilled through whole fuel irradiation. Coordinator of this project was Royal Institute of Technology from Stockholm, Sweden. HINDAS High and intermediate energy nuclear data for accelerator-driven system (HINDAS) was European project focused directly on nuclear data. Experimental data were measured on various accelerators throughout Europe. Nuclear models were improved according to experimental data. Energy scope of the HINDAS project was on energies from 20 to 2000 MeV. Libraries of nuclear data were extended up to 200 MeV (format ENDF was used). n-tof The neutron time-of-flight facility (n-tof) has been developed in the European Organization for Nuclear Research (CERN) since Time-of-flight method with fly path 200 meters is used to determine energy of the neutrons. The main goal of the project is to produce, evaluate and disseminate high precision cross sections for the majority of the isotopes relevant to the waste incineration and the ADS design. The Sixth Framework Programme [26] followed in the support of various research activities related to ADS and transmutation. Direct relation to the ADS has following four sub-programmes: EUROTRANS (EUROpean Research Programme for the TRANSmutation of High Level Nuclear Waste in a Accelerator Driven System). EUROPART (EUROpean Research Programme for the Partitioning of Minor Actinides). EFNUDAT (European Facilites for Nuclear Data Measurements). We used this programme to get access to the quasi-monoenergetic neutron source in The Svedberg laboratory in Uppsala, Sweden. More details about this programme are described in the chapter 7 Section 3. 16

33 1.8. Experiments focused on nuclear data measurements RED IMPACT (Impact of Partitioning, Transmutation and Waste Reduction Technologies on the Final Waste Disposal Project) Summary of ADS research goals Research of various ADS aspects continues nowadays both on simple setups and experiments, and on more complicated assemblies. Simple setups are used to measure the cross-sections of GeV down to MeV neutrons, and to study the spallation reaction and high energy neutron transport in more detail. More complex systems verify neutron multiplication, transmutation rates, heat production, long-term stability and overall suitable concepts for future XADS. Special attention starts to be paid to the engineering problems in construction of future ADS systems. There is also increasing motivation towards improving the precision of predictions of the codes used to simulate production and transport of high-energy spallation products in material. More realistic simulations will help to design more effective spallation neutron sources, subcritical blankets or better radiation shielding. Good codes can also spare budget in all stages of ADS life. But for codes development and improvements, a lot of real experimental data for comparisons and benchmark tests is needed. My research in the field of accelerator driven systems involves both the simple and complex experiments. The simple experiments are represented by the neutron crosssection measurements of the (n,xn) threshold reactions. Spallation experiments on the Energy plus Transmutation (E+T) setup belong to the complex experiments. Series of experiments of both types are described and compared with simulations in the following chapters. 17

34 1. ACCELERATOR DRIVEN SYSTEMS 18

35 Chapter 2 Energy and Transmutation of Radioactive Waste project 2.1. Introduction to the E&T RAW project There is a long tradition of spallation and high energy neutron studies in the Joint Institute for Nuclear Research (JINR Dubna, Russia). During the 1980s and 1990s, wide range of spallation targets was irradiated and the neutron production was studied with the respect to the target shape, dimensions, material and to the surrounding volumes. This aim culminated at the end of 1990s in the Energy plus Transmutation (E+T) project. The leader of this project was for almost last two decades M. I. Krivopustov, who established a big international team with interest in transmutation studies. Target systems Gamma-2, Energy plus Transmutation and Gamma-MD were developed and irradiated with protons and deuterons from the Nuclotron accelerator. Since 2009, M. Kadykov has been a new leader of the collaboration. The collaboration was renamed to Energy and Transmutation of Radioactive Waste (E&T RAW) and got a better position in the JINR structure, so a further development is foreseen. Collaboration is still growing and has nowadays approximately 85 members from 15 countries (Armenia, Australia, Bulgaria, Czech Republic, Poland, Germany, Russian federation, Belarus, Ukraine, Mongolia, Serbia, Kazakhstan, Greece, India, and Moldova). Two new target systems are developed, the first setup called Kvinta was already tested in experiment, the second setup called Ezhik is in the phase of technical design. Focus of our group from Řež is on high energy neutron measurement and beam diagnostics. We use so called reversed activation neutron detectors we put foil of a known isotope into unknown neutron field. Energy range of studied neutrons is from 5 up to about 80 MeV. Other groups from the collaboration use activation analysis on different isotopes, solid state nuclear track detectors (SSNTD), He-3 counters and nuclear emulsions to study other parts of the neutron spectrum. E&T RAW targets will be described shortly in the following sections. Main physical purpose of all targets is to study spallation reactions caused by GeV protons and deuterons, transport of high energy neutrons and transmutation. Use of various target and blanket materials, geometries and surrounding moderators enables to study their influence on neutron field. Systems have a big advantage in possibility of measuring integral data transmutation rates of actinides in real spallation field. GAMMA-2, E+T and GAMMA-3 setups were introduced into a Coordinated Research Project of IAEA and these targets are now acknowledged as IAEA benchmark targets Gamma-2 Gamma-2 setup consists of a lead target 8 cm in diameter and 20 cm long. Later the target was prolonged to 50 cm. It is surrounded with paraffin moderator of 6 cm thickness. Gamma-2 setup was irradiated with protons in the energy range

36 2. ENERGY AND TRANSMUTATION OF RADIOACTIVE WASTE PROJECT 4.15 GeV [27], respectively 1 2 GeV at the prolonged version [28]. Main experimental task of this setup was a study of spallation reactions and transport of high energy neutrons. First measurements with radioactive samples and their transmutation in the field of moderated neutrons were done. Scientific program on this target was more or less closed, but the target will be still ready for new irradiations if there is a need. Figure 7: Gamma-2 setup consisting of lead target (discs) and paraffin moderator E+T setup Further step in the transmutation studies was a more complex target system called Energy plus Transmutation setup (E+T setup). Setup was irradiated with 0.7, 1, 1.5, and 2 GeV protons, results of 1.6, 2.52, and partly 4 GeV deuteron irradiations are the main topic of this PhD thesis, that is why I will describe the target in more detail. The 0.7 GeV proton experiment was a subject of my diploma thesis [29], results of all proton experiments were the main subject of successfully defended PhD thesis of A. Krása [30]. Results concerning the proton experiments were also published as JINR Preprints [31], [32], [33], [34]; and presented on many conferences and workshops. The E+T setup consists of a cylindrical lead target (diameter 84 mm, total length 480 mm) and a surrounding subcritical uranium blanket (206.4 kg of natural uranium). Target and blanket are divided into four sections. Between the sections there are 8 mm gaps for user s samples, detectors and emulsions. Each section contains target cylinder 114 mm long and 30 identical natural uranium rods, which are secured in a hexagonal steel container with a wall thickness of 4 mm. The front and back of each section are covered with hexagonal aluminum plate 6 mm thick. The four target-blanket sections are mounted along the target axis on a wooden plate of 68 mm thickness, which is moreover covered with 4 mm thick steel sheet. Uranium rods are hermetically encapsulated in aluminum coverage of thickness 1 mm, respectively 2 mm at the bases. Each rod has an outside diameter of 36 mm, a length of 104 mm, and a weight of 1.72 kg. Density of the uranium is considered to be g cm -3. Around the blanket, there is a radiation shielding consisting of a wooden box, cadmium plates and polyethylene ((CH 2 ) n ) in the box walls. Cadmium plates have thickness of 1 mm and are mounted on the inner walls of the box. Polyethylene has a 20

37 2.3. E+T setup density of 0.8 g cm -3 and is granulated. On the floor inside the shielding box a 38 mm thick textolite 2 plate is placed. Shielding moderates and absorbs only a part of the high energy neutrons emerging from the setup, so there is a dosimetry limit on the beam flux. Figure 8: Cross-sectional side view (left) and front view (right) of the "Energy plus Transmutation" setup. All dimensions are in millimeters. Figure 9: Photo of the Energy plus Transmutation setup with the biological shielding (left). Detail of the natural uranium blanket (right). 2 Textolite (Latin textus a cloth, and Greek lithos stone) is a material consisting of several layers of fabric (filler); it is soaked by a synthetic resin. 21

38 2. ENERGY AND TRANSMUTATION OF RADIOACTIVE WASTE PROJECT More detailed information about the setup can be found for example in [31]. The detailed analysis of the influence of different setup parts and uncertainties in their geometrical and physical definitions on the neutron flux and possible sources of systematic uncertainties of obtained experimental data are analyzed using MCNPX simulation code in [31] Gamma-3 Gamma-3 setup, sometimes also called Gamma M-D 3, is a setup consisting of cylindrical lead target and big graphite moderator. Target has a diameter of 8 cm and length of 60 cm. Graphite moderator consists of blocks 25x25x60 cm 3 and 20x20x60 cm 3 big; total volume is 110x110x60 cm 3. In the moderator there are four cylinders, that can be pulled out and contain holes for sensors. Besides this, there are a few small plain holes through the moderator. Whole setup is placed on rails in F3 hall for easier manipulation. Figure 10: Photo of Gamma-3 setup in F3 experimental hall (left) and graphite cylinder with holes for samples. Up to now there was only one experiment on Gamma-3 setup with 2.33 GeV deuterons. Main experimental task was a study of radioactive sample transmutation; 129 I, 237 Np, 238 Pu, 239 Pu, and 241 Am were used. Next experiments are planned in the first half of the year Kvinta setup A new ready to use target has been available for the E&T RAW collaboration since the end of the year It is a setup of massive uranium target and lead shielding. Target has three sections of the same shape as E+T blanket a hexagon, but filled completely with uranium rods (weight 315 kg). Target is surrounded with massive 3 Minsk Dubna are names of the cities with the main institutes involved in the target construction 22

39 2.5. Kvinta setup lead shielding of total weight 1780 kg. Target is permanently placed in the shielding and the inner volume of the target is accessible only through four thin slots. Plastic holders are used to place samples inside the target [35]. There were two pilot measurements done during the 2009 winter run of the Nuclotron, setup was shortly irradiated with deuterons of 1 and 4 GeV energy. d Figure 11: Schema of Kvinta target. On the left there is a cut-view on the uranium target with supporting structures and plastics used for sample placement, on the right there is a view on the lead shielding enfolding the target [35] EZHIK Completely new target complex called EZHIK is nowadays projected and it should be ready to use by the end of Then it will be the main experimental device of the E&T RAW collaboration, although all previous targets will still exist and will be available for users. Name EZHIK means hedgehog in Russian, the parallel with hedgehog is because of the vertical channels sticking from the target. The target complex EZHIK is a quasi-infinite target from metallic uranium with wide range of measurement channels and positions. Basic scheme can be seen in Figure 12. The original technical solution of asymmetric beam input into a quasi-infinite target (first applied in [36]) is implemented in somewhat modified form. It provides results equivalent to those that could be obtained with 8 t uranium target in the case of conventional axial beam input into a cylindrical symmetric target, but with just about 3 t of target material from natural uranium [37]. Scientific program of the target EZHIK will be developed in three main fields. First direction will be focused on gathering of integral data, mainly in the direction of fission rates and transmutation cross-sections of actinide fission fragments. For this, wide range of support data will be measured particle fluxes, energy and heat distribution, isotopes equilibrium, neutron multiplication and dosimetry quantities. Second direction will be devoted to simulations. It is expected that all differences between the models and experiments, which were observed in the past, will be more 23

40 2. ENERGY AND TRANSMUTATION OF RADIOACTIVE WASTE PROJECT pronounced at quasi-infinite target and thus it will be easier to find the reason and correct it. Third direction will be focused on structural and fuel materials irradiated with large doses of relativistic beams and high energy neutrons. Radiation damage and gas production will be studied. Measurement channels Lead Uranium Graphite Figure 12: Scheme of the new target EZHIK [35]. Besides the basic version with uranium marked EZHIK-U will be developed also a version EZHIK-Pb, which will be geometrically identical, but whole inner volume will be filled by natural lead. EZHIK-Pb will be used for verification and adjustment of basic measurement systems and methods as well as background measurements with proton and deuteron beams in the projected energy range before main experiments with uranium target EZHIK-U will be made Placement of the E&T RAW targets For the E&T RAW collaboration is now allocated whole F3 experimental hall at the Nuclotron accelerator, see Figure 13. Targets stand in the hall on rails, so they can be quickly moved in/out of the beam. There is a crane in the hall to manipulate with heavier parts of the targets and equipment. 24

41 2.7. Placement of the E&T RAW targets Experimental setups A «Energy+Тransmutation»; B «Еzhik»; C «Gamma-3»; D «Кvinta». Нейтронная защита 4 6 Room for E&T RAO personal D Hall diagnostic system: 1 Ionization chamber 2 Activation foils 3 - Profile meter 4 Scintillator telescope 5 Pneumatic transport system 6 - B F 3 detector 7 Neutron spectrometer 8 Stilben detector; 9 - Detector «Isомеr»( 3 He); 10 - Detector LaBr 3 (Ce) B Beam 4 A C Scale: 0 1m 5m 10m Figure 13: Placement of E&T RAW targets inside the F3 experimental hall [35]. 25

42 2. ENERGY AND TRANSMUTATION OF RADIOACTIVE WASTE PROJECT 26

43 Chapter 3 Experimental background My work is focused on the studies of high energy neutrons produced in spallation reactions and their transport in the setup. High energy neutrons are in this case neutrons with energies from approximately 5 MeV up to 100 MeV. These neutrons can be measured by multiple methods (time of flight, nuclear emulsions, proton recoil detectors etc.) but specific conditions in the E+T setup makes these methods hard to use or unsuitable. Main limitations of high energy neutron measurements are the following: - lack of space a need to measure the neutrons inside the setup, - neutron field is changing on centimeter scale, - presence of thermal, epithermal and resonance neutrons, - presence of protons, deuterons and heavier charged particles, - huge gamma background, - specific conditions in JINR Dubna - problems with transport of electronics and with its operation due to highly intensive short bunches from the accelerator. Method of neutron activation detectors solves most of these problems. Samples can be small, thin, are insensitive to gamma and they do not need any power or maintenance during irradiation. Unirradiated samples can be easily transported, are simple to handle and relatively cheap (compared to electronic equipment). Last but not least there is a long tradition in using neutron activation detectors for high energy neutrons measurement at NPI. Following chapter will discuss the equipment, methods, and corrections used in experiments. My PhD work is focused mainly on the experimental part of the E+T experiment, so this description will go into detail on some places. I was the first one in our group who routinely applied some of the corrections into the experimental data and studied their effect. Results of these studies are also presented Activation detectors Neutron activation analysis method is mostly used for detecting the small amount of some isotope in compound. It is a very sensible method with sensitivity level up to gram per gram [38]. It can measure qualitative as well as quantitative content of tens of isotopes in one measurement. It uses known fields of neutrons or a system of standards (reference materials with known content of studied isotopes). We used it reversed - we placed a known amount of some elements into unknown high energy neutron field in order to measure the neutron spectrum and flux. Activation samples were made from pure aluminum, gold, tantalum, indium, cobalt and bismuth, see Figure 14. In the evaluation of the experiments are shown also 27

44 3. EXPERIMENTAL BACKGROUND results for yttrium, which was used by polish group, but was measured on our detectors and evaluated independently by us. Chemical purity of the materials was better than % 4. Figure 14: Activation materials used in the E+T for the study of high energy neutron field. Above mentioned elements were chosen, because they are mostly naturally mono-isotopic or one of the isotopes is dominant. They are also cheap, nontoxic and have good physical properties (melting point, ductile, no long-live isotopes). Further dominant criteria for choosing these elements were the decay times of the isotopes, that were produced in observed (n,xn) threshold reactions. Isotopes with half lives shorter than ~30 minutes or longer than ~year are not acceptable for us (we are not able to measure them with current equipment). For more details of used reactions see Table 4 or Appendix A. Table 4: Threshold reactions on aluminum activation samples. Reaction Threshold energy [MeV] Half-life 5 Used -line [kev] Intensity of used -line [%] 27 Al (n,p) 27 Mg min Al (n, ) 24 Na h Al (n, +2n) 22 Na y Materials were bought mostly from Goodfellow with the support from various grants, I personally had a grant from Czech Technical University from Internal grant competition (CTU ) and bought of it bismuth and gold foils. Bought foils had to be cut into smaller pieces suitable for us (we bought price convenient but bigger pieces). 5 Half-life of isotopes and gamma line energies were taken from [39]. Threshold energies were taken from [40]. Isotopes without listed gamma-line and intensity have not been detected. 28

45 Threshold energy [MeV] 3.1. Activation detectors Y(n,xn) 115In(n,xn) 181Ta(n,xn) 197Au(n,xn) 209Bi(n,xn) n,2n n,3n n,4n n,5n n,6n n,7n n,8n n,9n n,10n Order of threshold reaction [-] Figure 15: The threshold energies of (n,xn) reactions in Au, Bi, In, Ta, and Y detectors. Activation detectors were placed in the setup in two main directions longitudinal and radial, see Figure 16. List of all detectors is in Table 5 or in Appendix B. The foils had dimensions mostly 20x20x mm 3 and were twice wrapped up in the paper. Outer paper layer stopped most of the radioisotopes coming from the setup and was removed before the measurement. Inner paper layer stopped radioisotopes coming out from the foil and was present during all measurements [41]. Activation foils with the paper package were sticked on a plastic plane with holders and put into the slots in the setup (totally 5 planes, ~ 100 detectors/one experiment). After the irradiation and one to two hours cooling time (for decrease of the setup radioactivity) the foils could be removed. Figure 16: Placement of the gold and aluminum activation foils. Others were placed in the same way, only in another direction (e.g. bismuth in the right-down direction from the target axis). 29

46 1. plane 3. EXPERIMENTAL BACKGROUND Figure 17: Plastic plane with sticked samples (left) and the plane holders (right). Figure 18: Energy plus Transmutation setup with inserted plane holders, top view left, and side view right. Table 5: Placement of the activation samples in 1.6 GeV deuteron experiment. Distance from the target axis [cm] Foil labels in 1.6 GeV deuteron experiment 6 0 Y_5 3 Al1 Au1 Ta01 Bi1 In1 Y_8 6 Al2 Au2 Ta02 Y_ Al3 Au3 Ta03 Y_ Y_ Al4 Au4 Ta Y_9 up Y_19 down Y_21 left Y_38 right Y_20 6 Samples printed in normal letters were placed in the upward direction from the target axis (on the vertical axis). Samples printed in bold letters were placed in the right-down direction 30 from the horizontal axis. Samples printed in cursive were placed in the up-left direction 30 from the vertical axis. 30

47 5. plane 4. plane 3. plane 2. plane 3.1. Activation detectors 0 Y_10 3 Al5 Au5 Ta05 Bi2 In2 Y_1 6 Al6 Au6 Ta06 Bi3 In3 Y_6 8.5 Al7 Au7 Ta07 Bi4 In4 Y_ Y_ Al8 Au8 Ta Bi5 In Y_2 0 Y_4 3 Al9 Au9 Ta09 Bi6 In6 Y_35 6 Al10 Au10 Ta10 Y_ Al11 Au11 Ta11 Y_ Y_ Al12 Au12 Ta Y_27 0 Y_41 3 Al13 Au13 Ta13 Bi7 In7 Y_25 6 Al14 Au14 Ta14 Y_ Al15 Au15 Ta15 Y_ Y_ Al16 Au16 Ta Y_16 0 Y_17 3 Al17 Au17 Ta17 Bi8 In8 Y_11 6 Al18 Au18 Ta18 Y_ Al19 Au19 Ta19 Y_ Y_ Al20 Au20 Ta Y_12 List of all spectra measured on the samples is shown in Appendix C. Following paragraphs will contain description of various spectroscopic corrections that I have used and applied to evaluate right yield of the isotopes. These equations of corrections are the same for detector calibration, beam intensity and position measurement as well as for experimental data analysis Correction on decay of the isotope between the end of irradiation and beginning of the measurement Decay of all radioactive materials obeys the decay law: N t ( t) N0 e (3.1) In this equation decay constant) is the most important quantity, which says us how quickly is the nuclide decaying. Decay constant can be expressed by using half-live of the nuclide as follows: 31

48 3. EXPERIMENTAL BACKGROUND ln 2 (3.2) T 1 2 When we define the time t 0 as the time between the end of irradiation and the beginning of measurement, and the measurement lasted for a period t real, then the number of nuclei at the end of irradiation can be expressed like a product of the peak area and a factor e 1 e t 0 t real (3.3) This relation can be derived using following arithmetic process. If N(t) is a number of nuclei in time t, than the decay law has the form N t ( t) N0 e (3.4) In our case is N 0 the number of nuclei of studied isotope at the end of irradiation. Number of registered decays during the measurement can be marked like N. N is equal to the difference between the number of nuclei at the beginning and at the end of measurement N N t ) N( t t ) (3.5) ( 0 0 real When we introduce into this equation from the decay law than will be N equal to N N 0 e t N 0 e ( t t 0 0 real ) (3.6) From this we can express the ratio between the number of nuclei at the end of irradiation and number of registered nuclei during the measurement time t real : 32 t0 N0 e t N 1 e real (3.7) 3.3. Correction on decay during irradiation Studied radioactive isotopes decay already during irradiation. Let us assume that t irr is a time of irradiation and that at the beginning of the irradiation there are no nuclei of studied isotope in the sample. At the end of irradiation there is N o of nuclei in the foil. Next presumption is the rate of production studied nuclei are produced in the foil with stable rate P per unit of time. Number of radioactive nuclei N of studied material in irradiated sample follows differential equation: dn P N dt (3.8) This equation can be solved by the method of separation of variables:

49 3.3. Correction on decay during irradiation tirr 0 dt N0 0 dn P N (3.9) With the substitution x P N the equation becomes: t irr P N 0 P dx x (3.10) When we integrate the equation within the limits, we will get the term: p N ln P (0) t irr (3.11) From this we can derive the rate of production of radioactive isotope, the quantity P: N P 1 e (0) t irr (3.12) This equation can be transformed into the form, from which can be easily seen how many times more nuclei of studied material were produced during the whole time of irradiation t irr than it has remained in the sample at the end of its irradiation. P tirr tirr t N 1 e irr ( 0) (3.13) The right side of this equation is the searched correction on decay of studied isotope during irradiation Correction on the intensity of the I transition Gamma decay of the excited state of the daughter s nuclei can pass over various energy levels. Intensity of the gamma transition I is defined as the probability, that a gamma photon of energy E will be emitted during the decay of the nucleus (it is usually given in percentage and its value is from almost zero to 100 %) Correction on dead-time of the detector Dead time of the detector (and attached electronics) is the time, in which the detector collects and process previous impulse and during this time the detector is not able to handle next impulse. If a new impulse comes during this (dead) time, it is not recorded. Theoretically, there can be three main dead time types: cumulative, uncumulative and zero [42]. 33

50 3. EXPERIMENTAL BACKGROUND If the dead time is increasing with rising number of incoming gamma photons, it is called a cumulative dead time 7. At uncumulative dead time the detector has a fixed maximal signal rate, which it can handle. From some intensity of the gamma source the detector registers and processes one gamma photon and all others are ignored during this time, but the dead time is not prolonged. When the signal is processed, the detector opens for new gamma photons. The new photon is immediately registered and the detector is again closed and works on processing of newly registered photon. Output from the detector will display a constant activity of the source, although the real source activity can rise further. Some detectors can have a zero dead time, this is valid for example for gas-filled detectors working in current regime. Our HPGe detectors have a cumulative dead time. Measurement runs over the time t real, but the detector was able to accept new impulses/gamma photons during the treal time t live. Correction on dead time follows the equation Cdead. In the Dubna t measurements, there was a limit on dead time given by the electronics and wiring of the detectors to common ADC, more details are in chapter 3 section Correction on real - cascade coincidence Most of the nuclei have complicated decay schemes and various energy levels are fed. This leads to a complicated set of gamma and x-ray photons, which are emitted during deexcitation of daughter s nucleus. In such cases, a correction on real - cascade coincidence has to be used [43]. As an example for demonstration of the - cascade coincidence effect I will assume general decay scheme, where the studied isotope has basically two possibilities how to get from the excited state to the ground state. Process of emission of the photon (A) competes with the process of photon emission (B), whose emission brings the nucleus in other excited state with lower energy. From this excited state the nucleus can emit a photon (C) and comes to its ground state. live Figure 19: General decay scheme. 7 gamma quantum incoming to the detector during processing of signal /dead time/ caused by previous gamma-photon leads to prolongation of this dead time 34

51 3.6. Correction on real - cascade coincidence In the first approximation we can say, that the emission directions of both photons (B) and (C) coming from the same decay are independent. Then, with a certain probability, both two photons will interact inside the detector. This probability is growing with the decreasing distance between sample and the detector. Except some cases the life time of the energetic level between transitions (B) and (C) is negligible compared to the time that is needed for the signal collection after the absorption of the photon (B). That is why the detector registers both photons at the same time and summarizes the signals from photon (B) and (C). Energy of the summarized signal is naturally the same as of the photon (A), and the peak respective to the photon (A) is falsely increased. This effect is called - cascade coincidence. Ratio between the area of the summing peak (B)+(C) and area of the peak accordant to the -transition (A), which represents the coefficient of enlargement of the area of the peak (A), is determined by the following relation: I ( B) p ( C) S( A B C) a c (3.14) C C I ( A) ( A) p where I is absolute intensity of the gamma line, a is the branching ratio, c /(1 ), 1 t where t is total conversion coefficient and p is the peak efficiency of the detector. Coefficient representing a decrease of the area of the peak B is equal to L( B C) a c ( C), where a C is the probability that the transition (B) will be c C t followed by the transition (C), c C is the probability that a photon (C) will be emitted and t (C) is a probability that photon (C) will interact in the detector leaving there at least some part of its energy C, which will be bigger than the energy resolution of the detector. Than the signal of energy (B) + C is registered outside the full absorption peak of the transition B. In the same way can be derived also the coefficient, which relates to the reduction of the peak area connected with transition C, which follows after the transition B: I ( B) L( B C ) accc t ( B) (3.15) I ( C) These equations can be derived analogically in the case of multiple cascades. When we make corrections corresponding to the coincidence summation S(A) and to coincidence losses L(A), it is possible to formulate the number of measured impulses N det (A) in the peak of full absorption of the gamma transition A like: N det ( A) N( A) N( A) S( A) L( A) N( A) L( A) S( A) N( A) (3.16) With the coincidence factor defined as COI ( 1 L( A))(1 S( A)) the right number of pulses N(A) coming from the transition (A) can be the expressed by the equation: 35

52 3. EXPERIMENTAL BACKGROUND N N ( A) A) (3.17) COI ( det Except the real cascade coincidences, a stochastic coincidence can also occur. During stochastic coincidence, two photons from two different decays hit the detector at the same time and they are summarized. Because of low activities of our samples and small efficiency of used HPGe detectors this effect is negligibly small in our case. On the beginning of my PhD work I calculated the correction on real coincidences with hand-made equations made by my colleague A. Krása [44]. Equations were made up according to the k-0 standardization method [43]. With the rising number of studied isotopes, it was not further possible to set up the equations. Set up of the COI equations is sometimes extremely time-consuming due to high number of emitted gamma lines. With the number of gamma lines also the possibility of errors in the equations is quickly growing. My second colleague M. Majerle developed a program in Excel (Excel Addin in Visual Basic, see example in Appendix M) for automatic calculation of real coincidences. The Addin needs as an input a table of gamma lines, energy levels of the nucleus, gamma intensities and branching ratios; all this converted to a special table. In the code, it is necessary to change the number of valid rows in the table. With another Addin for the detector efficiencies, the COI can be calculated. Using different detectors and geometries, COI must be calculated for all these combinations used in each experiment. I have compared both approaches of COI calculation with satisfactory results. Table with an example of the coincidence correction for various isotopes is in the Appendix E. For the calculation of real coincidence correction both peak and total detector efficiencies are needed. Peak efficiency of the detector appears also in the final equation for calculation of the yield, so in the following two paragraphs I will make a short description of peak ( p ) and total ( t ) efficiencies. Practical approach for their measurement is stated in the section dealing with the detector calibration. Peak efficiency of the detector Peak efficiency of the detector in dependence on energy of registered photon is defined as: S (3.18) p N 0 Peak efficiency is a ratio between the number of gamma photons from a level transition in a calibration source registered into the peak of full absorption (per unit of time) divided by the activity of the calibration source, recalculated to the day of measurement. Peak efficiency depends on the photon energy, on the distance between emitter and detector (solid angle) and on the detector type and quality. 36

53 Correction factor [-] 3.6. Correction on real - cascade coincidence Total detector efficiency t Total detector efficiency t is a summary of efficiencies of all partial processes, which leads to any deposition of the energy of emitted gamma-photon in the detector resulting in an output signal. Main three processes are Compton scattering, photoeffect and production of electron-positron pairs. Total efficiency t depends again on photon energy, distance between emitter and detector and on type and quality of detector. Knowledge about the total detector efficiency is necessary for calculation of the correction on real coincidences. Total efficiency is determined during the calibration of each detector, results for ORTEC(new2) detector can be seen in Figure Correction on changed detector efficiency due to sample dimensions Correction on changed detector efficiency was used at thick foils (typically beam monitor foil 3 mm thick). During the measurement on the detector, calibration sources same as studied foils were mounted on the same plane. Most of the foils have thickness smaller than 1 mm ( m), so the center of mass is approximately at the same place like the calibration sample was. In the case of foils thicker than 1 mm this is not true, so I had to recalculate the efficiency of the detector taking into account that center of the foil is closer to the detector kev 2754 kev Distance from sample to detector [cm] Figure 20: Correction on the change in detector efficiency in the case of 3 mm thick Al foil measured on Ortec(new2) detector. I used efficiencies of the detector measured for a certain energy at different distances and fitted them with a second order polynomial. Then I used this fit to 37

54 Self-absorption correrction factor [ - ] 3. EXPERIMENTAL BACKGROUND calculate the efficiency for the closer position. Ratio between the recalculated and old efficiency is the searched correction. In the Figure 20 the specimen values are used in the 4 GeV deuteron run for 3 mm thick Al foil measured on Ortec(new2) detector. Uncertainty of this correction is negligibly small compared to other uncertainties Self-absorption correction I have calculated the self absorption correction for each isotope and used foil thickness. I used a formula (3.19), which can be derived as a ratio between gamma fluxes from the foil with and without self-absorption. Quantity cm -1 used in the equation is the total mass attenuation coefficient T [cm 2 /g] divided by density [g/cm 3 ]. Values of comes from Handbook of Nuclear Data for Neutron Activation Analysis [45] and were verified in web database Mass attenuation coefficients [46]. Quantity I 0 is unattenuated intensity of the gamma photons produced in the foil and D is foil thickness. I used a smooth curve between tabular values and calculated correction values for all used gamma energies and materials. In the following Figure 21 there is an example of self-absorption correction factor for 1 mm thick Bi foil (at this foil the correction was most important because of the big thickness of the foil, dense and heavy material, and low energies of used gamma lines). C abs D 0 D 0 I 0 dx D I 0 e D x dx D 1 e D (3.19) Tabular value Bi - 1 mm thickness Used gamma-energies Photon energy [MeV] Figure 21: Self-absorption correction factors for 1 mm thick Bi foil. 38

55 3.8. Self-absorption correction Only gamma-rays going parallel with the detector axis were taken into account in the self-absorption analysis stated above. Effective thickness of the sample can be bigger than the real one for close sample to detector distances (gamma-ray going sideways in the sample can be also detected), but this effect was assessed to be negligibly small in our case Square-emitter correction (geometrical correction) I had to calibrate all detectors that I used. I exploited standard laboratory pointlike calibration sources, see chapter 3 section 11. The biggest activation samples had dimensions 2.5x2.5 cm 2 with more or less equally distributed activity. It was clear that the detector efficiency was in close distances not the same for point and non-point like emitter (so the efficiency calibration is not precise enough). To assess somehow this effect, M. Majerle made a MCNPX calculation and studied the response of the detector for point-like and nonpoint-like emitter (correction is defined according to the relation (3.20)). Main problem of these simulations was missing knowledge about the detectors geometry size and shape of the crystal, thickness of dead layer and aluminum coating. More details about the MCNPX simulations can be found in M. Majerle s PhD thesis [41]. p ( foil ) cg (3.20) ( point) p To verify M. Majerle s MCNPX calculations a long row of experiments was done. These experiments were done originally by M. Majerle, later by me and also by our young student from Czech grammar school (O. Sláma, he was involved in the project Open science) and foreign student S. Peeterman. Gold foils were irradiated in LVR-15 reactor and cyclotron in Řež for these experiments. We used standard 2x2 cm 2 foil and small approximately 1x1 mm 2 - piece of Al with admixture of gold. Yields of 198 Au isotope were normalized to the measurement done in the biggest detector to sample distance, where the difference between point and nonpoint-like emitter can be neglected. Within the uncertainty bars, results of all measurements are comparable among themselves and also with M. Majerle s MCNPX calculations, example can be seen in following Figure 22. One can object why not to buy a square-emitter calibration source. Unfortunately, this was not a usable solution for us. The economic and administrative obstacles were one of the main reasons. With this is connected also the number of needed calibration sources at least five various dimensions from 1.5x1.5 cm 2 to 3x3 cm 2 would be necessary to measure for covering the main dimensions of our samples. Another question is the package type of these square-emitter samples, if they would be closed emitters in the term of law. We would need these calibration sources at three different places Řež, Uppsala in Sweden and Dubna in Russia. Cross-border transport of radioactive materials is a difficult task, which cost a lot of money and manpower (and both we do not have in sufficient amount). 39

56 Square-emitter correction [-] Square-emitter correction [-] 3. EXPERIMENTAL BACKGROUND MCNPX calculation Sláma - measurement 1 Sláma - measurement 2 Peeterman measurement Distance from sample to detector [cm] Figure 22: Comparison between measured and simulated square-emitter correction for 2x2 cm 2 foil and detector in Řež. Lines are only to guide reader s eyes. Sláma s data come from [47], Peeterman s data are from [48] x1.25 cm 2x2 cm 2.5x2.5 cm 3x3 cm iodine 3cm round Distance from sample to detector [cm] Figure 23: Square-emitter correction for the detector in Řež calculated in MCNPX for all sizes of measured samples. Distances correspond with used geometries (15 mm, 23 mm, 33 mm, 53 mm, 70 mm, 93 mm, and 178 mm). 40

57 3.9. Square-emitter correction (geometrical correction) I calculated square-emitter correction in MCNPX and brought it to routine use. I used the input file developed by M. Majerle and modified it to calculate the correction for all sizes of foils, detector types and geometries used in Energy plus Transmutation experiments and for cross-section measurements. Example of calculated correction can be seen in Figure 23. Up to now I have considered thin foils, where I can neglect the thickness of the foil. In the case of Al beam monitor foils, this is not true. Aluminum foil for beam intensity measurement is 10x10 cm 2 big, as described in Chapter 4, section 3. For gamma measurement, the foil was bended to half, and again and again, so it finally had dimensions of 2.5x2.5 cm 2 and thickness 3 mm. Due to the small beam spot and bending procedure, the activity is not distributed in the foil homogenously. Thus, there can be some uncertainty in measurement of such a foil. To assess the effect of various placement of activity in the thick foil, I modified the MCNPX simulation and used volume source instead of surface one. Schematic plot of the geometrical situation is in following Figure 24, it is a visualization of the MCNPX input file made in VISED code [49]. I put 80 % of activity in 20 % of the foil volume and opposite. The distance between the foil and detector was changed according to experimentally used geometries. Figure 24: Detector with inhomogeneous volume source representing Al foil that is used for beam intensity measurements. 41

58 Efficiency [-] 3. EXPERIMENTAL BACKGROUND Ep - activity up Ep - activity down Et - activity up Et - activity down Distance from sample to detector [cm] Figure 25: Peak and total efficiencies of the ORTEC(new1) detector calculated for inhomogeneous 25x25x3 mm 3 volume source hypothetic case of activity distribution in Al monitor foil. Efficiencies are calculated for positions p3, p4, and p5 used in the experiment. Uncertainties of the calculation are below 0.1% (high statistics). I calculated peak and total efficiency of the detector for the case when the main activity is on top or opposite. The results for the 4 GeV deuteron experiment are in Figure 25. From the figure we can see, that the differences are small (mean value of the p -up/ p -down ratios is 0.972) in comparison with the uncertainties related with the spectra evaluation in DEIMOS32, detector calibration etc Beam instability correction Irradiations on the Nuclotron accelerator were unfortunately not too much stabile, see Figure 35 - Figure 37. To correct the beam instabilities I used an easy program developed in Dubna, which counts production and decay of each isotope for each bunch. The input to the program consists of two files describing the beam structure and half-lives of the isotopes, for which the correction should be calculated. The program works according to the equation (3.21), [50]. Isotope production and decay are calculated for each beam bunch. Less accurate correction factor can be obtained by a manual calculation according to the equation (3.21) when the irradiation process is divided into sections with the same beam intensity. By this manual procedure I checked the function of the program getting almost the same correction factor values (differing at the third decimal place). More details can be found in my Diploma thesis [29]. 42

59 3.10. Beam instability correction B a t irr N i t 1 ( p 1 e W ( i) e i) t irr t ( i) t ( i) e (1 e p ) (3.21) where: t irr total irradiation time t e (i) time from the end of the irradiation interval till the end of whole irradiation t p (i) time of calculated irradiation interval W (i) ratio between the number of protons in the interval and in the whole irradiation N total number of intervals decay constant Beam correction factor is in most cases very close to the one, but it strongly depends on the decay time of the isotope and on the irradiation structure. For small decay times or more complicated beam it would naturally differ more from the one, but as we cannot measure isotopes with decay time shorter than one hour, the biggest beam correction is in my case equal to the factor of 0.8. Complete list of beam instability correction factors for all observed isotopes in deuteron experiments is in the Appendix D HPGe detectors For the measurement of all activated samples (beam monitors as well as threshold detectors) we used High Purity Germanium (HPGe) detectors. Detectors are placed in the JASNAPP laboratory in JINR, Dubna. In this laboratory there are four HPGe detectors for gamma-measurements and one planar detector for X-ray measurements [51]. We used detectors marked like Ortec new1 and Ortec new2. Their parameters are in following Table 6. These detectors were connected together with the planar detector to common ADC, so they had a collective dead time (high dead time caused by one detector was added to the dead times of the others all three detectors had the same dead time value). Both used detectors were equipped with small Dewar flask, so I had to refill the nitrogen every two days (to be sure there is always enough liquid nitrogen there are no scales). Detectors are during the year operated by J. Adam and his group from RChL (Radiochemical Laboratory department of JINR). Detectors were placed in lead shielding with the back and front wall partially opened, see Figure 26. Shielding was built up of various types of lead bricks, minimal thickness was 5 cm Pb, maximal 8 cm. This shielding suppressed the background approximately ten times; moreover it shielded personnel from measured radioactive samples (in the same room 3 detectors were operated at the same time and many people were all the time present). During the upgrade of the shielding in summer 2009 I helped to bring one tone of lead bricks from a very old beam dump. These bricks seemed to be without any radioactivity, but after the building of the shielding gamma lines of 207 Bi occurred (half-live years). Intensity of this isotope was approximately five times 43

60 3. EXPERIMENTAL BACKGROUND stronger than that of 40 K and the spectra from 4 GeV deuteron experiment are contaminated. There was built new shielding made of unactivated lead recently. Table 6: Parameters of used HPGe detectors, partly overtaken from [51]. Dubna detectors Řež detector Manufacturer/Name ORTEC(new1) ORTEC(new2) Intertechnique /IAA Type GMX GMX-30 EGNC20 Resolution [kev] (E =1332 kev) Relative efficiency [%] (E =1332 kev ) 1.80 kev 1.80 kev 1.80 kev Coating [mm] Be Al Al Dead Ge layer [ m] Detector bias supply ORTEC 659 ORTEC 660 ND360 Spectroscopy preamplifier Canberra 2024 Canberra 2026 Canberra 9615 ADC Multichannel buffer - ORTEC 919 Canberra 9635 Bias voltage [V] Shaping time [ s] P8 P7 P6 P5 P4 P3 P2 P1 Figure 26: HPGe detector Ortec(new) with lead shielding (left) and the bank with sample holder (right). Inside the shielding there was a bank for exact placement of plastic holders with samples. From one side it had a special hole, which fits firmly together with detector. The bank was situated on a plate with adjustable height. It was manufactured from acrylic glass, so it was light and easy to handle. Inside the bank there were eight 44

61 3.11. HPGe detectors positions for holders at distances of 12 mm, 24 mm, 41 mm, 65 mm, 99 mm, 147 mm, 216 mm, and 311 mm from the front of the detector 8. I calibrated detectors with a standard set of calibration etalons 9 : 54 Mn, 22 Na, 57 Co, 60 Co, 65 Zn, 88 Y, 109 Cd, 113 Sn, 133 Ba, 137 Cs, 139 Ce, 152 Eu, 228 Th, 226 Ra, and 241 Am (all isotopes were not available during all calibration measurements). After the end of measurements of samples from experiment I checked the calibration once more to control the calibration stability (temperature and voltage changes during days and nights were demonstrably changing energy calibration). I had to repeat the calibration before each experiment because of changes in geometry (manipulation with the detector and shielding during the year), changes in electronic settings and generally long time between experiments. Calibration was performed for the positions P2 up to P5. Measured activities of the calibration etalons were corrected on decay between manufacture and current date of measurement; experimental efficiencies for each used gamma line were calculated. Total number of used gamma lines was between 30 and 50 depending on the geometry and available time for measurement. Peak efficiency of the detector was calculated according to the equation (3.22), where S peak is the area of the peak related with the calibration isotope and A 0 [Bq] is the activity of the calibration etalon at the date of manufacture. S e t0 peak treal p treal A I C (1 COI e ) (3.22) 0 tlive C COI is a correction on real coincidences, it was necessary for following isotopes: 133 Ba, 60 Co, 152 Eu, 228 Th, and 88 Y. For these isotopes an iteration loop was necessary change in measured activities invoked change in efficiencies and consequently in the fit of the points, which invoked change in fitted efficiency and in real coincidences correction, which back invoked change in measured activities of calibration samples. Iteration was repeated so far the efficiencies were not changing in reasonable number of digits (mostly 8 decimal places). Values of the experimental efficiency were taken into logarithm and fitted with one or two curves so that the differences between the experimental values and the fit would be minimal (3.23). I have omitted several experimental points that were too far from the curve, probably because of the coincidence with natural background or because of the complicated evaluation in DEIMOS ( a ln( E) b ln( E) c ln( E) d ) p e respectively 2 ( a ln( E) b ln( E) c) p e (3. 23) 8 For the newest detector ORTEC(new2) I bought the material from MK Plexi s.r.o. I paid it from my grant CTU and I transported it after the assemblage in NPI to JINR Dubna. 9 Half-lifes of the isotopes were taken from the etalon certificates. Intensities of used gamma-lines were taken from [39]. 45

62 Efficiency [-] 3. EXPERIMENTAL BACKGROUND Total uncertainty of the peak efficiency calibration is assessed to be at least 1%. It comes from the uncertainty of the calibrations etalons (1% - 2% uncertainty in the knowledge of activity) and from the fit of the experimental points. In the case of total efficiency, only a few isotopes, which have one or two gamma lines, could be used ( 57 Co, 139 Ce, 113 Sn, 137 Cs, 54 Mn, 60 Co up to energy 1253 kev). In the DEIMOS32 code [52] I have summarized number of counts from the beginning of the spectra up to the end of the peak (S total ), activity of the etalon at the date of manufacture is A 0. S e t t0 total real t treal A I (1 e ) (3.24) 0 tlive I have interlayed logarithmic values of experimental points with third order polynomial, see equation (3.25). 3 2 ( a ln( E) b ln( E) c ln( E) d ) t e (3.25) Finally, I put calibration curves into the Excel as an Addin, so they can be easily used as a function (function ep for peak efficiency and function et for total efficiency), see Appendix M. Example of peak and total efficiencies for the ORTEC(new2) detector in the 4 GeV experiment are in following Figure 27. In Figure 27 three geometries (p3, p4, and p5 are shown, distances between sample and detector are according to the Figure 26. Exp means experimental points from the calibration sample measurements, fit indicates mathematical fit of experimental points Ep-p3-exp Ep-p4-exp Ep-p5-exp Ep-p3-fit Ep-p4-fit Ep-p5-fit Et-p3-fit Et-p4-fit Et-p5-fit Et-p3-exp Et-p4-exp Et-p5-exp Energy [MeV] Figure 27: Example of peak and total efficiencies for the ORTEC(new2) detector in the 4 GeV experiment.

63 Displaced to centre ratio [-] HPGe detectors I have also studied homogeneity of the detector in X and Y axis, because foils in our experiments are not irradiated homogenously 10. There can be a difference in the activity of the foil on its sides. I have measured the difference in the response of the detector on the same gamma source placed on different sides from the centre of the crystal to assess a possible uncertainty coming from the detector in-homogeneity in combination with un-homogenously irradiated foils. I used a point-like laboratory etalon and displaced it in X and Y axis in a grid of 3 mm or more. Detector in JINR Dubna is homogenous within 2 percent, what is the uncertainty of this type of measurement (uncertainty comes from the peak evaluation in DEIMOS32). HPGe detector in Řež is not homogenous because it was accidentally irradiated by neutrons in the past. After this irradiation, the crystal had to be newly surfaced. Results of one of these measurements are in following Figure 28, it is a comparison of the detector response to point-like 137 Cs source placed in the centre of the detector and displaced to the left and to the right (in left right direction is the biggest difference). Conclusion is that we have to be careful when measuring big foils in close distances, but under normal conditions no special precautions have to be done (there is only a difference of 2.6 % between left and right side 27 mm from the centre, measured at 15 mm distance from the detector cap) right side left side Distance from the centre of detector [cm] Figure 28: Homogeneity of Řež HPGe detector in X-axis. On Y-axis of the graph there is a ratio between the responses of the detector to 137 Cs point-like source placed in the centre and displaced to left and right. Homogeneity measured at the distance 15 mm from the detector cap. 10 Change of the neutron field in the E+T setup is not negligible on the distance comparable with the dimension of the foils [31], [53]. 47

64 3. EXPERIMENTAL BACKGROUND DEIMOS32 program I used DEIMOS32 code to evaluate measured -spectra. This program was developed at the Nuclear spectroscopy department by J. Frána [52]. Simplified principle of the gamma spectra analysis is a fit of selected gamma peaks with Gauss curve. DEIMOS32 code has a lot of functions, settings and possible ways of use; it was one of the main tools I have used, so more detailed description of the code follows. Basic function of the DEIMOS32 is a spectrum display. One can see whole spectrum (with already evaluated regions if this function is used) or one can work directly with a chosen part of the spectrum. Three modes of the Y-axis depiction can be used: linear, square root, and logarithmic. Position of the mouse pointer is displayed above as a function of energy (number of channel) and number of counts at this energy. Evaluation of the spectra is possible in automatic or manual mode; both regimes can be switched anytime. I have used only manual evaluation, because it enables the best control over the peak fit procedure. Peak fit procedure is based on the non-linear least squares method. In each evaluated part of the spectrum positions, heights, common widths of the peaks, and parabolic or linear background are fitted. Widths of the peaks can be pre-calibrated and held fixed, or maintained in preset range. Regions for evaluation can be chosen manually or found by automatic searching procedure. Peak distance can be also fixed. I used a manual process of energy calibration, by which I evaluated a spectrum with known peaks/energies and then modified the calibration file. Two-point energy calibration was used. Program can work with the following list of spectra: DAT (AccuSpec, Silena),.MCA (S100),.CHN a.spc (Ortec),.SPE (Sampo),.CNF (Genie). Reading of ASCII files is also possible. At some types of these spectra only some parts of the file header are red. This caused to me a lot of problems, because I processed a lot of CNF files, where the headers could not been displayed, so I had to store the times and dates of the measurements separately. DEIMOS32 solves also the staircase increase of the background on the lowenergy part of the peak. This jump in the background is small for energies over 300 kev and can be considered stable for all energies (in our case of detectors with relative efficiency ~ 20 % it is approximately 1 % of the peak height). In the region of energies lower than 300 kev I have used preset values already involved in the code, that were successfully tested on the same type of the detectors like I have used. 48

65 3.12. DEIMOS32 program Figure 29: Graphical interface of the DEIMOS32 code [52]. Peak area can be established by two different procedures. A simple one is a summary of the number of counts in each channel involved in the region that is evaluated. This I have used during measurement of the total detector efficiency. More sophisticated is the fit procedure with non-linear least squares fit. The code enables to run the procedure step by step and change the parameters of the fit (background description, region of fit, number of peaks in the fit, fixed FWHM etc.) between each step. I have widely used this feature and I have evaluated thanks to it also very complicated spectra. The disadvantage is an extreme time consumption of this procedure. It can be chosen many types of output files and data written in it, I used only a simple one with the extension PRN. PRN file is a text file, which contains basic information about the spectra (red from the file header if possible), peak position in channels and energy, peak area and its uncertainty and several statistical parameters related to the Gauss-fit. Most of the peaks that I evaluated in DEIMOS32 were small ones on a background of much bigger peaks from the isotopes produced in non-threshold (n, ) reactions (intensities of studied peaks are comparable with the peaks from unshielded background). Precision of the DEIMOS32 evaluation was crucial for the experiment, so I have tested differences between results of evaluation of the same spectra in DEIMOS32 and Genie, as a representative of automatic commercial software for spectra analysis [54]. 49

66 3. EXPERIMENTAL BACKGROUND I used spectra with a lot of peaks coming from threshold reactions, namely Al, Au, and Bi. I chose the most and the less active sample of each isotope. I came to following conclusions. In the case of huge peaks on clear background (Al), the differences between the programs are smaller than 0.1 %. When the background is more complicated and a lot of peaks is nearby the main peak (e.g. Al with small activity and long time of measurement = bigger background), differences between the fits are 3% on average. When the size of the studied peak is comparable with the peaks from background, studied peak lies on a complicated background or is nearby a strong peak, differences between the two programs are up to tens of percent in some cases, on average 7%. There is no clear trend in the differences, values of the differences between the programs go equally to plus and minus. Differences between these two programs are in any case smaller than the uncertainty of the peak fit in each program. DEIMOS32 code is for my purpose of use much more convenient than the automatic codes. It is much more complicated and time consuming to use, but I have full control over the whole fitting process. I can focus on studied peak and do much more for its analysis than any automatic code can ever do Yield evaluation Finally, yield of observed isotopes (products of the (n,xn) reactions) was calculated with respect to the various spectroscopic corrections according to the equation (3.26). Weight normalization is involved in the yield. Weight-normalized yields from various foils (of the same material) can be compared within one experiment. To be able to compare the yields among all E+T experiments, I have finally divided the yields also by the total number of beam particles N d (is discussed in following chapter). Peak area Self-absorption correction Beam correction Dead time correction Decay during cooling and measurement N yield I S ( ) 0 p Cabs E Ba treal 1 e ( E) C Coi C t m 1 e P g area live foil ( t ) ( t ) t 1 e irr ( real t irr ) line intensity per decay Detector efficiency Correction for efficiency change Correction for coincidences Square-emitter correction Weight normalization Decay during irradiation (3.26) Following notation is used in the equation: t real measurement time on the detector t live live time of the detector t irr irradiation time 50

67 3.13. Yield evaluation t 0 time between end of irradiation and beginning of measurement decay constant Weighted average (X) according to the equation (3.27) was used for the isotopes with more detected lines or in the case of multiple measurements. It is a standard weighted least-square average over n values (x i ) and their uncertainties (Δx i ). I have used the equations according to the publication Review of Particle Physics 2000 Edition [55]. n x X i 1 n i 1 xi 1 i x i 2 2 (3.27) I have determined the uncertainty of the weighted average (ΔX i ) using the equation (3.28): X i n i x i 2 (3.28) I have also calculated χ 2 and compared it with (n-1), which is the expected value of χ 2 if the measurements are from a Gaussian distribution. 2 If / n n 1 n i 1 xi xi n 1 2 X 2 (3.29) is less than or equal to one, I accepted the weighted average. If 2 / n was slightly bigger than one, I increased the uncertainty of weighted average 2 by multiplying it with / n 1. 2 My reasoning for this is following: value of / n 1 bigger than one means that at least one of the partial values is too far from the Gaussian distribution and has too small uncertainty (to be connected with the rest of the data). Not knowing which uncertainty of which value is underestimated, I assumed they are all underestimated by 2 the same factor / n 1. If I scaled up all the input uncertainties by this factor, the χ 2 become equal to (n-1) and also the output uncertainty of the weighted average scales 2 up by the same factor / n 1 2 If the / n 1. was very large, I searched in the data for the values that were far from the weighted average. I have reanalyzed these values in order to find possible 51

68 3. EXPERIMENTAL BACKGROUND source of discrepancy 11. When the reanalysis did not helped I have excluded these values from further analysis. This approach was not usable in all cases, because the spread of all values was big at some isotopes and it was not possible to decide which value should be excluded. Yields of the most important isotopes produced in various reactions and samples during 1.6 GeV and 2.52 GeV experiment are in the Appendix F Sources of uncertainties In the following Figure 30 there are shown basic sources of uncertainties (measured quantities, corrections, given data), their partial steps and partial uncertainties, that come out in each step (display idea comes from [56]). Dashed lines represent most important relations between the quantities from the uncertainty point of view. Red marked uncertainties I have already involved into the analysis, blue marked uncertainties are exactly known, but are negligibly small or have practically no influence on the result. Rest of uncertainties (marked black) is not exactly known but they are negligible compared to the red ones. Beam intensity determination was always the biggest source of uncertainties in the E+T experiments. Besides the uncertainty coming from the spectroscopic evaluation of the foils that are used for intensity measurement there is an uncertainty from the cross-section and uncertainty from the shift of the cross-section to the used beam energy (hard to assess). Anyway, uncertainty of the beam intensity is the same for all samples and reactions in one experiment, so it has no influence on the shape of the relative yields of the isotopes in radial or longitudinal direction. This is also the reason for stating this uncertainty separately. Other uncertainties in experiment come from the peak fit in DEIMOS32, these vary from units of percents up to tens of percents. Detector calibration is known with at least 1 % uncertainty, another 1 % uncertainty is contained in spectroscopic corrections. Uncertainty coming from the foil placement can be up to 20 % at 5 mm foil displacement [41]. Within the Energy and Transmutation RAW collaboration there has not been a clear statement how to handle all these uncertainties (if involve them into data or put them separately) up to now. Some of the uncertainties have changed in the time deuteron experiments were the first ones where I have involved some corrections (e.g. on self-absorption) or modified older ones (COI correction). That is why I state my experimental data only with the DEIMOS32 uncertainty, other uncertainties can be modified and involved by the users when needed. 11 This was done in various ways by repeated DEIMOS32 evaluation, searching in the background for possible sources of interference, comparisons among the yields on all foils of the same type, comparison between multiple measurements, looking for decay products of the isotope, comparison with the MCNPX simulation etc. 52

69 3.14. Sources of uncertainties Detector efficiency Position uncertainty Activity uncertainty Calibration samples DEIMOS uncertainty Gamma line identification Spectra evaluation Fit uncertainty Fit of the points Intensity Gamma-line intensity correction uncertainty Beam Gamma line intensity Real coincidences Beam intensity Self absorption Foil material Weight of the foils COI correction uncertainty Detector efficiency uncertainty COI correction Position uncertainty Cross-section uncertainty Detector efficiency uncertainty DEIMOS uncertainty Monitor placement Gamma line identification Cross - section Monitor measurement Spectra evaluation Attenuation coefficient uncertainty Self absorption correction uncertainty uncertainty Purity uncertainty uncertainty Thickness Size Weight measurement uncertainty Density Weight measurement uncertainty uncertainty uncertainty uncertainty uncertainty Start of measurement Live time Real time End of irradiation Start of irradiation Dead time uncertainty Detector dead time correction Beam correction uncertainty correction Non-point like emitter correction uncertainty Non-point like emitter correction Time measurement Detector dead time Beam stability Non-point like emitter Yield Figure 30: Schema of the uncertainties. 53

70 3. EXPERIMENTAL BACKGROUND Background Natural background played a significant role in all my spectroscopic measurements. Due to small cross-sections as well as low high energy neutron intensities I worked with small activities, so some detector shielding was necessary. HPGe detectors that I used in JINR Dubna had lead shielding described more in the chapter 3 section 11, see e.g. Figure 26. I measured the background before each experiment at JINR Dubna and also afterwards I measured last sample. I supposed the background was not changing during the sample measurements, although it had to be not necessarily true (laboratory with detectors is in the building neighbouring with the Phasotron accelerator. During the Phasotron operation the background was up to one order of magnitude higher than usually. Nevertheless, after the fire of the power supply of beam line guiding magnets, Phasotron is working rarely.). Spectroscopic laboratory at JINR Dubna is placed in the first floor of the building, so radon and its daughter products are not a problem. There are two more floors with massive concrete ceilings over it, so the cosmics is also partially suppressed. After the long time of the laboratory usage there can be seen artificial isotopes in the background spectra ( 152 Eu and 137 Cs). I had to subtract the background at some isotopes where the energy of studied gamma line was close or even the same like the energy of some gamma-line in background (typically 207 Bi all gamma lines or 192 Au kev gamma line). 54

71 Chapter 4 Beam diagnostics on Nuclotron accelerator 4.1. Nuclotron accelerator Irradiation of the E+T setup was carried out in the Laboratory of High Energies by 1.6 GeV, 2.52 GeV and 4 GeV deuteron beam extracted from the Nuclotron accelerator. These deuteron irradiations were a continuance of previous protons experiments, in which the Energy plus Transmutation setup was irradiated by 0.7, 1, 1.5 and 2 GeV protons. Table 7: Irradiation parameters of three deuteron experiments on the E+T setup. Deuteron beam energy [GeV] Beam start :55:33 7:01:11 23:59:20 Beam end :42:18 15:00:48 17:47:37 Time of irradiation [h] Beam intensity measured by operators [10 13 ] The Nuclotron ring is installed in the tunnel around the synchrophasotron, see Figure 31 and 33. This tunnel was originally built for cable communications and the equipment of the synchrophasotron vacuum system. The Nuclotron median plane is at 3.7 m below the synchrophasotron one. The Nuclotron lattice is typical for strong-focusing synchrotrons with separated functions. It contains 8 super periods and 8 straight sections, one of which is "warm". General view of the Nuclotron dipole and quadrupole magnet is presented on Figure 32. The magnets are fastened to a vacuum shell of the cryostat 540 mm by 8 suspension parts of stainless steel. A nitrogen shield 490 mm covered with 20 layers of super insulation is placed in the vacuum space between the magnet and the vacuum shell. The dipole magnet has a window-frame type iron yoke with the sizes of window of 110x55 mm 2. The quadrupole lens has the iron yoke with hyperbolic poles. The SCcable was manufactured of a 5 mm in diameter copper-nickel tube with a wall thickness of 0.5 mm and 31 in parallel connected multifilament strands of 0.5 mm in diameter covering an outer surface of the tube. The strand consist of 1045 NbTi filaments 10 m in diameter stabilized by copper [57]. The design parameters of the dipoles are: B=2.2 T and db/dt = 2 4 T/s. Nominal current amplitudes are: up to 6.3 ka and 6 ka for the dipoles and quadrupoles respectively. There are 96 dipoles, 64 quadrupole, and 32 correcting SC-magnets in the 55

72 4. BEAM DIAGNOSTICS ON NUCLOTRON ACCELERATOR Nuclotron ring with circumference of m. Averaged specific weight of the magnetic system is only 0.32 t/m. Figure 31: Nuclotron site scheme. Energy plus Transmutation setup is placed in Hall B [57]. Figure 32: One section of the Nuclotron accelerator (inside beam tube surrounded with superconducting magnet, He and LN2-cooling pipes, isolation and steel shell) [57]. Figure 33: Nuclotron accelerator ring in the Synchrophasotron cable tunnel (own photo). Underlying is the beam outlet. 56

73 4.1. Nuclotron accelerator Table 8: Selected parameters of Nuclotron accelerator compared to the older Synchrophasotron accelerator [58]. Parameter Synchrophasotron Nuclotron Maximal kinetic energy - protons [GeV] Maximal kinetic energy - Z/A=1/2 [GeV/A] 4 6 Repetition time (p.p.s.) Extraction time [s] Vacuum [torr] Maximal magnetic field in magnets [T] Circumference [m] Accelerator consumption [MW] 8.5 note Figure 34: General scheme of the Nuclotron cryogenics. 1 - vacuum shell; 2 - heat shield; 3 - supply header; 4 - return header; 5 - dipole magnet; 6 - quadrupole magnet; 7 - subcooler; 8 - separator; 9 - helium flow from the refrigerator; 10 - return helium flow to the refrigerator, [59]. General scheme of the Nuclotron cryogenics is presented in Figure 34. All the magnets are connected in parallel with supply and return helium headers. The internal diameters of the headers are 36 mm and 52 mm respectively. The cooling of the magnets is performed by two-phase helium flow. The liquid-vapor content varies from MW is the consumption of the accelerator. Total consumption of the accelerator complex is ~8 MW (cooling, vacuum etc.). Consumption of the beam guide to user area is another 8 13 MW according to the place, energy and guided particles [60]. 57

74 Beam intensity [deuteron/bunch] 4. BEAM DIAGNOSTICS ON NUCLOTRON ACCELERATOR at the inlet of the magnet to 0.9 at its outlet. The temperature sensors are placed at the helium inlet and outlet of the winding and also at the helium outlet of the iron yoke of each magnet. Totally, the temperature measuring system includes about 600 points. The Nuclotron operational temperature is K. The cryogenic supply system is based on three industrial helium refrigerator/liquefiers with a total capacity of 4.8 kw at 4.5 K. Helium cooling of the accelerator is nowadays the biggest limitation for our experiments. High price of helium in coincidence with leaking compressors enable to accelerate particles only a few weeks in the year. For every accelerator run, operators receive much more beam requests than they can accommodate. Moreover, cooling system is a source of frequent failures, which canceled or postponed our irradiation in the past. In the year 2011, a new system of cryogenic cooling should be installed and thus annual year load should increase significantly Irradiation course The runs of the accelerator were unfortunately not too much stable during our experiments, below are the figures from the irradiation process :00 1:00 2:00 3:00 4:00 5:00 6:00 7:00 Time Figure 35: Beam intensity during 1.6 GeV deuteron irradiation of the E+T setup. 58

75 Beam intensity [deuteron/bunch] Beam intensity [deuteron/bunch] 4.2. Irradiation course :00 7:00 8:00 9:00 10:00 11:00 12:00 13:00 14:00 15:00 16:00 Time Figure 36: Beam intensity during 2.52 GeV deuteron irradiation of the E+T setup :00 2:00 4:00 6:00 8:00 10:00 12:00 14:00 16:00 Time Figure 37: Beam intensity during 4 GeV deuteron irradiation of the E+T setup. Irradiation instabilities had to be corrected; so-called beam instability correction is described in chapter 3, section

76 4. BEAM DIAGNOSTICS ON NUCLOTRON ACCELERATOR 4.3. Beam position and shape Knowledge of the beam position and shape is crucial for the experiment evaluation. Significant influence of the beam position on experimental results was observed in the activation detectors placed close to the target axis, MCNPX simulations were done to assess this effect [31]. In MCNPX simulations, experimentally measured beam position and shape parameters are used in order to calculate comparable data. In the following paragraphs a description of the beam measurement procedure and examples of its application on deuteron experiments will be shown. The geometrical adjustment of the experimental setup with respect to the deuteron beam was tuned before the irradiation by the means of sensitive Polaroid films. Only one low intensity bunch was necessary to get a visible trace on the Polariod film. Films were placed directly in front of the target to see the position and profile at the point, where the beam entered the target. Another Polaroid film was placed behind the target to check the direction of the beam in the target. When the beam was not in the target centre, whole setup was moved on the rails and lifted by screws and the Polaroid procedure was repeated, until the target axis was reached. Examples of the irradiated Polaroid films are on the Figure 38. A new method of beam alignment has to be developed for future, because the Polaroid films are not produced anymore and the rest of the reserves has already expired. Figure 38: Polaroid films for pre-irradiation beam alignment (2.52 GeV deuteron experiment). Beam parameters during the irradiation were determined independently from solid state nuclear track detectors (Belarus group) and from a set of copper activation foils. Gained results were compared and a common beam-report was always prepared, e.g. [61]. I will further discuss my results and compare them with the results of other E&T groups. I used a set of copper foils, which I placed directly in front of the target and behind it. The copper was chosen, because in interaction with deuteron a lot of radioactive isotopes are produced, but none of them are produced by neutrons 13. On the 13 Aluminum cannot be used at this position because of a significant production of the same isotope 24 Na by the neutrons. 60

77 4.3. Beam position and shape other hand, no experimental cross-sections are known for interaction of relativistic deuteron and copper. I could make only relative comparison between the foils. For measurement of the beam position in front of the target, 60x60 mm 2 copper foil was used in 1.6 and 2.52 GeV experiment. To cover bigger area of the beam monitoring, I have enlarged the size of monitoring foil to 80x80 mm 2 in the case of 4 GeV experiment, see Figure 39. Thickness of the foil was 100 m, respectively 70 m in the case of 4 GeV deuteron experiment. I cut the foil after the irradiation into 20x20 mm 2 pieces (totally 9 pieces, respectively 16 pieces in the case of 4 GeV experiment), and I measured these pieces separately. Following isotopes were observed: 24 Na, 43 K, 47 Sc, 48 Sc, 44m Sc, 44 Sc, 48 V, 48 Cr, 52 Mn, 58 Co, 56 Co, 55 Co, 57 Ni, and 61 Cu. Totally 19 lines were used for the final evaluation. I have observed above mentioned isotopes only in the most active foils, in other foils they were not detected or were on the level of detection limit (this represents relative production between 1 % and 6 % in non-hit foils). None of these isotopes was visible in all foils and with similar activities; this lead me to the presumption that all the isotopes I used were produced by the deuterons from the beam and not by back-scattered neutrons from the target. Yields of each isotope were normalized to the most active foil and a weighted average over all reactions and used gamma lines was made. Figure 39: Photo of the copper foil used for front beam monitor (left) and its paper envelope (right). I will discuss 4 GeV experiment as a representative of the beam position measurement (this beam analysis was done most detailed because of the large beam shift). Relevant weighted averages are in the Table 9 and Figure

78 Beam profile in front of the target - big monitors 4. BEAM DIAGNOSTICS ON NUCLOTRON ACCELERATOR Target left centre right top centre down Figure 40: Weighted average over relative yields of 19 different gamma-lines in the forward Cu beam monitor during 4 GeV experiment (left). Schema of the foil-cut and target projection (right). Even during gamma spectra measurement I could see (according to the activities in the foils), where the beam passed through the uncut foil. In the case of 4 GeV experiment I saw that the highest activity was in the right-up foils on the edge (and at that moment I did not know if some or how much deuterons went out of the foil). Beam profile from small monitors I decided to cut the most active foils 3, 4, 7, and 8 onto smaller pieces 1x1x0.007 cm 3 and I measured each of them once again. I did the same procedure as in the previous case and got results presented in Figure 41. I did these measurements two days later and some of the isotopes were already decayed, but final weighted average is still over 11 gamma-lines. Values with their uncertainties are summarized in the Table 9. left centre right top centre down Target Figure 41: Weighted average over relative yields of 11 different gamma-lines in the double cut Cu beam monitor irradiated in 4 GeV deuteron experiment (left). Schema of the foil-cut and target projection is on the right. 62

79 4.3. Beam position and shape Table 9: Weighted average over relative yields in forward Cu monitor during 4 GeV deuteron experiment. Number of foil Relative yield (uncertainty) Number of foil Relative yield (uncertainty) (11) (5) (24) (5) (4) (5) (6) (5) (22) (5) (18) (9) (5) (13) (7) (7) (12) (6) (21) (8) (19) (11) (18) (6) (8) (22) (5) (5) (14) (4) (18) (27) A Gaussian beam profile in X and Y axis was assumed. Measured data were fitted in the PAW program [62]. The final shift of the beam was during deuteron irradiations in order of millimeters up to units of centimeters and is summarized in following Table 10. In 1.6 GeV experiment I also used two copper foils placed in front of the target to measure exactly how many deuterons went out of the target. I used circles 84 mm in diameter (same as target) and 120 mm in diameter. For the gamma measurement I bended the foils to a smaller pieces approximately 25x25x3 mm 3 big. Further analysis procedure was the same as at forward beam monitor. Results can be seen on Figure 42. In 2.52 GeV and 4 GeV experiment it was unfortunately not possible to place these foils in the setup, so I can assess number of the out-of-target deuterons only from the fit. 63

80 4. BEAM DIAGNOSTICS ON NUCLOTRON ACCELERATOR Table 10: Beam position, shape and intensity during deuteron experiments, comparison of data from various groups. Deuteron beam energy [GeV] Me I. Zhuk Me I. Zhuk Me I. Zhuk X-axis shift [cm] Y-axis shift [cm] FWHM in x [cm] FWHM in y [cm] Measured beam intensity 2.4± ± ± [10 13 ] - Me Measured beam intensity [10 13 ] - W. Westmeier 2.45 ± ± ± 0.15 Final beam intensity [10 13 ] 2.45 ± ±0.013 Deuterons out of the target - Me Deuterons out of the target - W. Westmeier Deuterons out of the target - I. Zhuk Final number of deuterons out of the target up to now unknown 6 % - < 8 % 21.3 % 1 % 43.7 % 0.3 % 3 % 8.2 % 0.3 % 3 % up to now unknown 12 cm 8.4 cm ± Figure 42: Relative number of deuterons that did not hit the target during 2.52 GeV deuteron experiment. I placed copper foil also behind the target to check the beam direction in the target (if the beam goes parallel with the target axis). I used foil with dimensions 90x90x0.12 mm 3 and made of the same copper as the front foil. After the irradiation I 14 Value determined only from 22 Na and 7 Be reactions using method proposed by J. Blocki [63]. 64

81 4.3. Beam position and shape cut it onto 9 pieces 30x30x0.12 mm 3 big and I measured each part separately. I detected the same isotopes as in the forward foil and I was able to do the weighted average over 19 lines. From the results (Figure 43) we can see that the beam was during the 4 GeV deuteron run more or less parallel with the target axis (the same I observed also in other deuteron experiments). Behind target left centre right 1.00 Target 0.49 Z-1 Z-2 Z top centre Z-4 Z-5 Z down Z-7 Z-8 Z-9 Figure 43: Weighted average over relative yields of 19 different gamma-lines in the Cu beam monitor placed behind the target during 4 GeV deuteron experiment (left). Schema of the foil-cut and target dimension (right) Beam intensity Beam intensity was measured using aluminum foils. W. Westmeier used concentric rings placed few meters in front of the setup, I used a square foil 100x100x0.2 mm 3 placed close to W. Westmeier s ones. Number of neutrons coming from the target is negligible at this distance; it was proven by M. Majerle using MCNPX [64]. Cross-section of the 27 Al(d,3p2n) 24 Na reaction is the only one known crosssection at the region of GeV energies of deuterons with suitable half-life and energies of gammas. It was measured by J. Banaigs [65] at deuteron energy 2330 MeV (15.25 ± 1.5 mbarn), see Figure 44. I made a fit of the data in order to assess the change of the cross-section value within our energy region. The most suitable seemed to be a function y a x b, which described well the data both in low and high energy region. Finally, I used the same value as W. Westmeier in his analysis in order to get comparable data. His fitted values are close to my fit, changes come from the selection of the beginning of fitted curve. Cross-section uncertainties are not involved in the uncertainty of the beam intensity. 65

82 Cross-section [barn] 4. BEAM DIAGNOSTICS ON NUCLOTRON ACCELERATOR For the gamma measurement on the detector I middled the foil few times to get a dimension approximately 25x25x3 mm 3. I measured this packed foil on the detector a several times in various geometries (and also on different detectors if possible) to suppress the uncertainty coming from detector calibration. The foil was also rather thick, so a correction on detector efficiency was necessary, see chapter 3 section 7. Same corrections as described in the section related to the sample evaluation were used. Example of the summary of the intensity measurement is in the Table 11 (in this case it is concretely 4 GeV deuteron experiment) Al(d,3p2n) 24 Na measured value GeV mbarn 4 GeV mbarn GeV mbarn 2.52 GeV mbarn Deuteron energy [MeV] Figure 44: Cross-section 15 of the 27 Al(d,3p2n) 24 Na reaction from EXFOR [66] and fit between the values for used deuteron energies. Beam intensity N d was calculated according to the following equation (4.1), where N yield (of 24 Na) is calculated using the (3.26) equation. where: S area of the foil A molar weight N A Avogadro number - 27 Al(d,3p2n) 24 Na reaction cross-section N d N Yield N S A A (4.1) 15 1 barn = m 2 66

83 2754 kev 1368 kev Efficiency of detector 2754 kev 1368 kev Yields of 24 Na from separate gamma-lines 17 uncert 2754 kev uncert 1368 kev walbigp51 walbigp54 walbigp46 walbigp5ort3 walbigp4ort2 walbigp3ort Beam intensity Table 11: Summary of the beam intensity measurements done in 4 GeV deuteron experiment. Spectrum E E E E E E E E E E E E E E E E E E E E E E E E Changed efficiency correction 18 COI Non-point like emitter correction Spectra label is constructed in following way, e.g. walbigp51 w is from Wagner (our spectrum), Al material, big label for beam monitor foil, p5 position on the detector, 1 number of measurement. 17 Grey marked numbers were omitted from the further evaluation because of their deviousness. 18 Changed efficiency correction means change of the detector peak efficiency due to the thickness of the sample (3 mm) in comparison to standard one centre of the thick sample is closer to the detector. 67

84 Weighted average for 2754 kev 1368 kev Selfabsorption correction 2754 kev 1368 kev 4. BEAM DIAGNOSTICS ON NUCLOTRON ACCELERATOR Beam correction E+08 uncert. 2.44E+06 χ E+08 uncert. 2.23E+06 χ Total weighted average (except 3.10E+08 value) Number of deuterons 3.43E+08 uncert. 1.76E+06 χ E+13 uncert. 1.91E+11 In each experiment there was a difference between the beam intensity value measured by me and by W. Westmeier, but till the 4 GeV deuteron experiment, differences were always within the uncertainty bars. In 4 GeV deuteron experiment, a new additional electronic system of beam monitoring was installed. Unfortunately, beam intensity measured by this device is by 25 percent higher than my value, which is of 30 percent higher than the value of W. Westmeier. Up to now, no clear reason for such differences was found 19. When searching for the source of uncertainties I tried to calculate the beam intensity using also other reactions in the Al monitor and in the copper monitor of beam shape and profile. There are no experimental data for cross-section of deuterons and copper (or Al except those leading to 24 Na), but a lot of data exist for protons on copper and aluminum. It is possible to recalculate the cross-sections from proton to deuteron using following method (proposed by J. Blocki [63], used also by W. Westmeier to analyze the yields of 22 Na and 7 Be in his Al beam intensity monitor). Cross-section recalculation is based on the presumption that there is a fixed ratio between the inelastic cross-section for proton and deuteron (at relativistic energies two nucleons in 2 H behave as two separate items). Cross-section for protons and deuterons seems to change slowly and run parallel at GeV energies. I have started from already mentioned reaction 27 Al(d,3p2n) 24 Na, where I know the cross-section for deuterons at 2330 MeV (J. Banaigs, [65]). I found cross-section for protons leading also to 24 Na (reaction 27 Al(p,3pn) 24 Na) at similar energy 1200 MeV: Dittrich B. 12 mbarn [67], 19 Data analysis of this experiment was not closed up to now and will be a subject of M. Suchopár s PhD work. 68

85 4.4. Beam intensity Michel R mbarn [68], and Titarenko 12.9 mbarn [69]. Mean cross-section value is 11.9 mbarn. Ratio between the deuteron and proton cross-section is thus (this should be the same for all reactions on Al). With this ratio I multiplied cross-sections of proton induced reactions 27 Al(p,3p3n) 22 Na and 27 Al(p,10p11n) 7 Be and calculated beam intensity from the 22 Na and 7 Be yields produced by deuteron beams. In the best case the differences from the directly evaluated intensity were smaller than 4 % at the 22 Na and 2 % at 7 Be (in the case of 1.6 GeV deuteron experiment). During 2.52 GeV deuteron experiment no long-time measurements of Al beam monitors were done (because of lack of time), so it was not possible to test this procedure. In the 4 GeV experiment, the differences were much higher: beam intensity was 15 % higher in the case of 22 Na, respectively 50 % higher in the case of 7 Be (statistical uncertainty from DEIMOS32 is six percent at 22 Na and 20 percent at 7 Be). The reason for the disagreement can be partially in bad statistics; long term measurement for 22 Na and 7 Be was done in the case of 4 GeV experiment half a year after the irradiation and took place for a few weeks. The natural background played a significant role in this case and had to be subtracted at 7 Be. Another source of disagreement can be in chosen cross-section of protons with Al or in the efficiency of the detector. Finally, I tried to calculate deuteron beam intensity from the copper foils. No experimental cross-sections for suitable nat Cu(d,x) reactions are known at used energy region, so I had to calculate my own cross-sections. Above mentioned procedure was not usable because of missing cross-sections, so I assumed the beam intensity in 2.52 GeV deuteron experiment is determined properly. With this beam intensity I have calculated cross-sections of various reactions observed on copper during 2.52 GeV deuteron experiment. These cross-sections I have to shift to 1.6 GeV and 4 GeV energy. I have done the shift according three various reactions for protons, for which I have found experimental cross-section at 1.6 GeV, 2.52 GeV and 4 GeV. I have determined average ratio between the cross-sections (1.6 GeV / 2.52 GeV, respectively 4 GeV / 2.52 GeV). With this ratio I have shifted the cross-sections and calculated deuteron beam intensity for 1.6 GeV and 4 GeV E+T experiment. For some of the reactions the beam intensity values were close to the intensity value determined by 24 Na, but some of them were one order of magnitude higher or lower (no serious reason for the discrepancy was found). The final result (average over 10 reactions) is in the case of 1.6 GeV experiment (2.24 ± 0.08) deuterons in the beam (value determined from the 24 Na is (2.45 ± 0.04) 10 13, so this procedure gives rather good results, but is less reliable. In the case of 4 GeV experiment this analysis has not been finished yet, but the preliminary value of the beam intensity determined by using the data from copper foil is (2.5 ± 0.7) (intensity determined from 24 Na is (1.985 ± 0.019) ). 69

86 4. BEAM DIAGNOSTICS ON NUCLOTRON ACCELERATOR 70

87 Yield [1/g*deuteron] Chapter 5 E+T results of deuteron irradiation 5.1. Plain experimental results After the gamma-spectra evaluation and application of necessary spectroscopic corrections, I have determined the yields of produced isotopes. These yields are proportional to the neutron field in the place of the foil 20. The uncertainty bars in the graphs bellow are only from the Gauss fit in the DEIMOS32 and are hardly visible in the logarithmic scale (are only a few %). Lines in the graphs are only to guide reader s eyes and have no real physical meaning Au 196Au 194Au 192Au 24Na Position along the target [cm] Figure 45: Yields of the isotopes produced in Au and Al activation detectors in longitudinal direction, 3 cm over the target axis, 1.6 GeV deuteron experiment. Products of the threshold reactions have their maxima near to the first gap (~12 cm from the target beginning). This value does not differ for higher beam energies very much, although the deuteron range in the lead is rising. The reason is in the probability of the first collision (spallation), which takes place for most of the deuterons in first ~20 cm. During the spallation reaction high energy neutrons are produced mostly to the forward angles (intranuclear cascade), neutrons from high energy fission and evaporation are produced isotropicaly. These isotropicaly emitted neutrons cause most of the threshold reactions in the foils placed in front of the lead target. 20 The higher the yield of some (n,xn) reaction is, the more neutrons with the energy higher than the relevant threshold had to be in that place. 71

88 Yield [1/g*deuteron] 5. ENERGY + TRANSMUTATION RESULTS OF DEUTERON IRRADIATION Au 196Au 194Au 192Au 24Na Position along the target[cm] Figure 46: Yields of the isotopes produced in Au and Al activation detectors in longitudinal direction, 10.7 cm over the target axis, 1.6 GeV deuteron experiment. Non-threshold 197 Au(n, ) 198 Au reaction is caused by the epithermal and resonance neutrons coming from the biological shielding. High energy neutrons escaping from the target and blanket are moderated in the polyethylene inside the shielding and some of them are backscattered into the inner volume of the biological shielding. Cadmium layer on the inner walls of the shielding absorbs only neutrons with energies under the cadmium edge (0.5 ev). Neutrons with higher energies create inside the biological shielding almost constant field, which is not so strong like the field of high energy neutrons, see Figure 63. But, the yields of 198 Au or 182 Ta are by one to two orders of magnitude higher than the yields of threshold reactions due to the high crosssection values of the non-threshold reactions, see Figure 45 or Figure 46. Field of epithermal and resonance neutrons inside the biological shielding is disturbed only in the beginning and at the end of the setup due to the holes in the shielding (used for manipulation and beam entrance). This can be documented in the Figure 46, where the outer points of 198 Au yield are lower than the average between. In radial direction the yields of threshold reactions are quickly (almost exponentially) decreasing. This can be read out from the lines in Figure 47 and Figure 48. From the product of non-threshold 198 Au can be seen, that the epithermal and resonance neutron field is really homogenous in radial direction. It is slightly disturbed close to the target by difference in neutron absorption in uranium and lead (there are a lot of resonances of neutron capture in 238 U, see Figure 64). Outside the blanket (10.7 cm position) there is visible a small influence of the moderator/reflector. 72

89 Yield [1/g*deuteron] Yield [1/g*deuteron] 5.1. Plain experimental results Au 196Au 194Au 192Au 24Na Radial distance from the target axis [cm] Figure 47: Yields of the isotopes produced in Au and Al activation detectors in radial direction, first gap of the E+T setup 12.2 cm from the target beginning, 1.6 GeV deuteron experiment Au 196Au 194Au 192Au 24Na Radial distance from the target axis [cm] Figure 48: Yields of the isotopes produced in Au and Al activation detectors in radial direction, behind the E+T setup 48.8 cm from the target beginning, 1.6 GeV deuteron experiment. 73

90 Yield [1/g*deuteron] Yield [1/g*deuteron] 5. ENERGY + TRANSMUTATION RESULTS OF DEUTERON IRRADIATION More figures for reactions on Bi, In, Ta and Y from 1.6 GeV experiment as well as from 2.52 GeV and 4 GeV deuteron experiment can be found in Appendix G Au 3 cm cm 8.5 cm cm Distance along the target [cm] Figure 49: Yields of the 196 Au isotope produced in Au activation detectors in longitudinal direction, various distance from the target axis, 1.6 GeV deuteron experiment Au in front of 1st gap 2nd gap 3rd gap behind target Radial distance from the target axis [cm] Figure 50: Yields of the 196 Au isotope produced in Au activation detectors in radial direction, various distance from the target beginning, 1.6 GeV deuteron experiment. 74

91 5.1. Plain experimental results Example of the comparison of the yields of threshold reaction 197 Au(n,2n) 196 Au (threshold 8.1 MeV) is shown on the Figure 49 and Figure 50. Normal scale is used for the Y-axis (not a logarithmic one), thus the statistic uncertainties from Gauss fit in DEIMOS32 are better visible. In the longitudinal direction it can be seen that the highest production was close to the target centre (3 cm from the target axis). In the radial direction the highest flux of the neutrons with E > 8 MeV was in the first gap (12.2 cm from the beginning of the target), than in the second gap and the lowest flux was behind the target. Most of the data were already published on national conferences (6. Mikulášské setkání sekce mladých při České nukleární společnosti [70]) and international conferences (Baldin conference [71], ND2007 [72], NEMEA-4 [73]). General overview can be found in article in revised journal Nuclear Instruments and Methods in Physics Research [74] Ratios of yields for different thresholds In Figure 51 are plotted ratios of the yields of various threshold reactions in front of and behind the target in dependence on their threshold (longitudinal direction 3 cm over the target axis). Neither in 1.6 GeV deuteron experiment nor in 2.52 GeV deuteron experiment (see Appendix G, section 4) a clear dependence is visible like it was during proton experiments (see e.g. [30] or [32]). There is some trend that shows a decrease of the ratio with rising threshold energy, that means the difference in neutron flux in front of and behind the target is smaller for neutron energies higher than ~20 MeV. The difference comes from the probability of the first interaction, respectively spallation reaction. Neutron field inside the setup is a complicated mixture of spallation, fission, moderated and back-scattered neutrons. Neutron field has its maximum around 12 cm from the target beginning, see e.g. Figure 45. Neutrons with higher energies come from the intranuclear phase of the spallation reaction and are emitted more forward, in contradiction to neutrons below 20 MeV, which come from evaporation and fission phase of the spallation reaction and are emitted isotropicaly. Epithermal and resonance neutrons come from the biological shielding. Combination of the spallation probability and various sources of neutrons in spallation reaction causes observed difference in front/end yield ratio for threshold energy approximately 20 MeV. In radial direction the ratios are calculated from the yields at 3 cm and 10.7 cm from the target axis. Ratios are made of the foils placed in the first gap of the setup (place with maximal neutron flux). The ratios oscillate around the value 6.5 up to the neutron energy 35 MeV in the case of 1.6 GeV experiment, see Figure 52. Above 35 MeV there is a steep increase. The situation is very similar in the case of 2.52 GeV deuteron experiment, see Figure 126 in the Appendix G. This difference originates from the course of spallation reaction neutrons with higher energies are produced mainly in intranuclear cascade and move to forward angels, so they can hardly get far from the target in radial direction. 75

92 Ratio 3/10.5 cm [-] Ratio in front of / behind target [-] 5. ENERGY + TRANSMUTATION RESULTS OF DEUTERON IRRADIATION 5 Al Au Y Bi In Threshold energy [MeV] Figure 51: Ratio in front of and behind the target for various threshold reactions, 1.6 GeV deuteron experiment Al Au Y Bi In Threshold energy [MeV] Figure 52: Ratio in 3cm and 10.7 cm (11.5 cm) in the first gap of the target for various threshold reactions, 1.6 GeV deuteron experiment. Yields of various threshold reactions can be theoretically used for deconvolution and acquiring of the neutron spectrum in certain point. Problem can be in the knowledge of cross-section data, there are except Bi no experimental cross-section data for reactions higher than (n,4n) or energies over 40 MeV. Spectrum calculated with the use 76

93 5.2. Ratios of yields for different thresholds of simulated/calculated cross-sections would be no more experimental. This is one of the reasons for my cross-section measurements described in chapter 7. Polish E+T group tried neutron spectra deconvolution from their yttrium samples [75]. They used simulated cross-sections from the TALYS code [76], where the experimental ones were missing. They divided the spectrum to three parts according to various thresholds of three used (n,xn) reactions. They got three Fredholm equations for the yields of different threshold reactions. Then they assumed that the yield is a continuous function of threshold energy. Using further mathematical presumptions they converted Fredholm equations to the Volterra s equation of the first kind and solved them. Calculated neutron spectrum is in MeV energy region close to the simulated one, but its maximum shifted to higher energies. Usage of this method for the deconvolution of the yields of threshold reactions has to be further studied; authors admit problems caused by used presumptions (final function describing the neutron spectra is concave instead of convex function observed in MCNPX simulations). Authors also use presumption about the shape of neutron spectrum below the threshold of the first reaction, although they do not have any experimental sign for it Spectral indexes When I compared yields of reactions with different threshold (e.g. 196 Au and 192 Au) I have observed a spectrum hardening at the end of the target (see Figure 53 or Figure 54). Figure 53 or Figure 54 are in principal similar to previous Figure 51 and Figure 52. Threshold energy is here hidden in the ratio of two reactions with different threshold. Observed spectrum hardening is specific for the spallation reaction; high energy neutrons are produced more into the forward direction. In comparison between experiments it can be seen that the differences in spectral indexes in front of and behind the target are decreasing with rising beam energy. More spectral indexes can be found in Appendix G.3. 77

94 Spectral index 192 Au/ 196 Au [-] Spectral index 192 Au/ 196 Au [-] 5. ENERGY + TRANSMUTATION RESULTS OF DEUTERON IRRADIATION cm 6 cm 8.5 cm 10.7 cm Figure 53: Neutron spectra hardening along the target in 1.6 GeV deuteron experiment (ratio between 192 Au and 196 Au) cm cm Figure 54: Neutron spectra hardening along the target in 4 GeV deuteron experiment (ratio between 192 Au and 196 Au). 78

95 Yield [ - ] 5.4. Comparison between deuteron experiments 5.4. Comparisons between deuteron experiments It is well-known from the experiments 21 that the most effective energy of the proton beam for spallation neutron production is around 800 MeV - 1 GeV. In this interval is the biggest neutron production per MeV per proton on heavy target. The usage of deuterons brings another bonus in neutron production, but complicates accelerating. This leads to lower intensity of the beam. Experimental results (non-threshold and threshold yields per 1 gram of foil material and one beam particle) are in following Figure 55 and Figure 56. Data are normalized to the second foil to see the difference in the shape. More figures with comparison of experiments (normalized and unnormalized) are in Appendix G.5. Except the 4 GeV deuteron experiment there can be seen an increase in the neutron flux behind the first gap (maximum) with rising beam energy. Differences in the shape of the yields of 4 GeV deuteron experiment can be caused by the beam placement beam was displaced to the right and up, almost to the corner of the target close to the Au samples (see Table 10). Some spallation in uranium was thus possible and it is probably the reason of much higher neutron yield due to additional fission in irradiated uranium. Yield normalization led to difference in the shape of the yields of 4 GeV deuteron experiment GeV d 2.52 GeV d 1.6 GeV d 0.7 GeV p Au Position along the target [cm] Figure 55: Comparison of non-threshold 198 Au yields in longitudinal direction at 3 cm from the target axis, deuterons and 0.7 GeV proton experiment on E+T setup. Values are normalized to the second foil. Results of the 4 GeV experiment are preliminary. 21 Example of multiplicity experiments can be found for example in the summary article of A.V. Dementyev [77]. 79

96 Yield [ - ] 5. ENERGY + TRANSMUTATION RESULTS OF DEUTERON IRRADIATION Comparison of unnormalized yields from various experiments shows problems with the beam position, results of threshold 197 Au(n,2n) 196 Au reaction are close together or even disordered in front of the target and in the first gap (=maximum), where the beam position influence is the most significant, see Figure 130 and Figure 132 in Appendix G.5. The yields of 4 GeV deuteron experiment are much higher than the rest, probably because of the uranium spallation and additional fission 22. I have calculated a ratio of the 198 Au, 196 Au and 194 Au yields for 2.52 GeV / 1.6 GeV and 4 GeV / 1.6 GeV deuteron experiments. There is a clearly visible trend for 198 Au, with rising beam energy the ratio is increasing in the radial direction (groups of foils 1-4, 5-8 etc). That means the number of epithermal and resonance neutrons is rising more rapidly with rising beam energy. This is valid up to approximately the half of the setup, then the trend is changing and behind the target the epithermal and resonance neutron flux is decreasing more rapidly when moving out from the target axis (valid for increase in beam energy). For more details see following Figure 57 and Figure 133 in Appendix G.5. In the case of 196 Au and 194 Au threshold reactions there is no visible trend in the yield ratios for 2.52 GeV / 1.6 GeV and 4 GeV / 1.6 GeV deuteron experiments. For more details see Figure 134 and Figure 135 in the Appendix G GeV d 2.52 GeV d 1.6 GeV d 0.7 GeV p Au Position along the target [cm] Figure 56: Comparison of threshold 196 Au yields in longitudinal direction, deuterons and 0.7 GeV proton experiment on E+T setup. Values are normalized to the second foil. Results of the 4 GeV experiment are preliminary. 22 Higher neutron fluxes due to probable spallation of uranium were observed also by other E&T RAW groups, but their results are still preliminary. 80

97 Ratio 2.52GeV / 1.6 GeV [-] 5.4. Comparison between deuteron experiments Au Number of foil [-] Figure 57: Ratio of the 198 Au yields for 2.52 GeV and 1.6 GeV deuteron experiments in all twenty Au foils, which were used Total neutron production The so-called water-bath/activation foil method [78] is often used for the determination of the integral numbers of neutrons produced in thick targets. The conventional variant of this method uses two basic premises: neutrons from the source are predominantly contained within the moderator volume; and it is possible to integrate the measured thermal flux distribution over the water volume with adequate precision. As the latter requires the usage of a large-scale grid of activation foils, I have used a new form of this method [79], which replaces the flux integration by relating a small-scale set of foil activities to the integral quantity the integral number of neutrons sim produced per one beam particle (so-called neutron multiplicity) n total obtained by simulation. Polyethylene in the biological shielding of the E+T setup worked as a water bath it moderated outgoing neutrons. I neglected front and back openings of the biological shielding. I did multiplicity simulations in MCNPX 2.7.a using INCL4 + ABLA models 23. For calculation of the neutron multiplicity, I determined the ratios between experimental and simulated yields of 198 Au in all gold samples. I tried to use also tantalum samples for the first time, because tantalum has similar cross-section for (n, ) reaction like the gold has (see Figure 58), and tantalum samples were placed close (or even at the same place) like the gold samples. I calculated weighted average over these 23 neutron multiplicity does not depend significantly on the combination of the models available in MCNPX in this energy region, proven by A. Krása in his PhD work [30] 81

98 5. ENERGY + TRANSMUTATION RESULTS OF DEUTERON IRRADIATION ratios and I multiplied it with the simulated neutron multiplicity see following equation 5.1: multiplicity sim N n exp total ntotal (5.1) N exp Yield sim Yield The advantage of this procedure is that the experimental value of neutron exp n total is highly insensitive to the simulated value sim n total and its uncertainty. Assuming that the MCNPX describes well the spatial distribution of the neutrons as well as the shape of low energy part of neutron spectrum and its approximate magnitude; the product of the two terms in equation (5.1) effectively cancels out the dependence on n sim total. Neutron multiplicity results for deuteron experiments are summarized in Table 12, Figure 59 and Figure 60. Results from gold and tantalum samples are comparable within the uncertainties. Multiplicity determined by tantalum seems to be closer to the simulated multiplicity of the E+T setup. Figure 58: Cross-section of the (n, ) reaction on Au and Ta, overtaken from ENDF/B- VII. [85]. 82

99 Neutrons per beam particle [-] 5.5. Total neutron production Table 12: Experimental neutron multiplicities for deuteron experiments 24. Beam energy [GeV] N exp Yield N sim Yield exp n total sim n total exp n to ta l sim n total per GeV per GeV 198 Au ± ± ± ± ± ± ± ± ± Ta ± ± ± ± ± ± ± ± ± protons -exp deuterons - exp - Au deuterons - exp - Ta sim-p sim-d Beam energy [GeV] Figure 59: Neutron multiplicities for E+T setup (proton experimental points overtaken from the PhD thesis of A. Krása, [30]). 24 Data for the 182 Ta in 4 GeV deuteron experiment were evaluated by our grammar school student Ondřej Novák. 83

100 Neutrons per beam particle per GeV [-] 5. ENERGY + TRANSMUTATION RESULTS OF DEUTERON IRRADIATION protons - exp deuterons - exp - Au deuterons - exp - Ta protons - sim deuterons - sim Beam energy [GeV] Figure 60: Neutron multiplicities for E+T setup normalized per GeV (proton experimental points overtaken from the PhD thesis of A. Krása, [30]). 84

101 Chapter 6 MCNPX simulations of the Energy plus Transmutation setup 6.1. MCNPX code MCNPX is a general purpose Monte Carlo radiation transport code designed to track many particle types over broad ranges of energies [80]. MCNPX means Monte Carlo N-Particle transport code extended. It is the next generation in the series of Monte Carlo transport codes that began at Los Alamos National Laboratory nearly sixty years ago. The MCNPX program began in 1994 as an extension of MCNP4B and LAHET 2.8 in support of the Accelerator Production of Tritium project (APT) [81]. The work envisioned a formal extension of MCNP to 34 particle types and up to teraelectronvolt energy range; improvement of physics simulation models; extension of neutron, proton, and photonuclear libraries to 150 MeV; and the formulation of new variance-reduction and data-analysis techniques. The APT project also included crosssection measurements, benchmark experiments, deterministic code development, and improvements in transmutation code and library tools. Our group from Řež is a member of the MCNPX beta tester team, so for a long time we have had access to the newest versions of MCNPX. In this PhD thesis, all simulations were done in the version 2.7.a, which is the newest version we are allowed to use. Beta tester mailing list is another helpful tool we use. Through this we can get help and advice from the people from MCNPX community almost immediately Limitations of MCNPX code MCNPX code has specific restrictions coming out from its Monte-Carlo approach. Correctness and accuracy of the MCNPX calculation is limited by used crosssection data libraries, physical models and intrinsic imprecision. Spectra of different particles are weakly dependant on the choice of used library (proven in M. Majerle s PhD thesis, [41]). In my calculations I used LA150 library for proton and neutron transport (it seems to give the best results). Energy range of the library is limited to 150 MeV, cross-section models are used for higher energies. For (n,xn) reactions, crosssection combined from TALYS [82] and MCNPX - CEM were used to get the most reliable results. Correctness of the MCNPX is also determined by the description of physical processes. In the case of MCNPX descriptions are based both on empirical fits of experimental data and on mathematical idealizations of predicted theories. In my case a description of the spallation reaction and transport of high energy particles is the most crucial point. MCNPX offers a following set of intranuclear cascade models (BERTINI, ISABEL, INCL4) and evaporation models (DRESNER, ABLA), which can be combined, or all in one model (CEM03). Results of the MCNPX calculations with different combinations were one of the topics studied by A. Krása. In his PhD thesis he 85

102 6. MCNPX SIMULATIONS OF THE E+T SETUP used all available combinations of models for protons and studied the changes in the yields of threshold and non-threshold reactions in our samples. Up to the MCNPX version 2.6.e only the combination of INCL/ABLA was able to handle the deuterons with energy higher than 2 GeV. From MCNPX 2.6.e all models can be used also for higher deuteron beam energies. I used INCL4/ABLA combination because it was available in all previous versions and thus it has been better tested. Another reason is that INCL4/ABLA was used by A. Krása and M. Majerle in their proton calculations prepared for comparison with deuterons (e.g. multiplicity). Unfortunately, INCL4 model is very slow and the simulation takes ~10 times more computer time than the other models. PC cluster of 64 processors was installed at Nuclear Spectroscopy Department of the NPI in Řež and it is used for calculations, so one E+T calculation with 10 7 source particles takes approximately one day. Under intrinsic imprecision are meant rounding errors, interpretation of numbers, inaccuracies in numerical solving of equations etc. These imprecision cannot be usually influenced and changed by the user. Another limitation of the code is due to the random nature of generated events. Large number of single events is necessary to collect enough statistics in our small volumes representing activation foils. Typical example can be in the case of 196 Au and 198 Au isotopes. In the spallation reaction mostly high energy neutrons are produced, so enough neutrons above the reaction threshold can be collected in the foil for e.g source particles. 198 Au is produced mostly by the resonance neutrons, whose number is at 10 6 source particles much smaller and hence determined with worse accuracy. Number of source particles (statistics) had to be thus enlarged in some cases. High statistics is connected with the problem of computer time, as mentioned above. Precision of the calculation depends inversely on the square-root of the number of processed events, so there is not too much place for a radical decrease of the statistical uncertainties in simulations Simulation of the E+T setup For every E+T experiment we made a set of MCNPX simulations, e.g. [32]. After a few years of improving the setup description in the code, we have a detailed model of the experimental setup now (see Appendix H). Two-dimensional visualization can be seen in Figure 61-right, three dimensional visualization of some setup parts can be seen on Figure 61-left and Figure 62. Development and improvements of the MCNPX input file and calculation procedure was one of the main tasks of my colleague M. Majerle and it is described in his PhD thesis [41]. M. Majerle also studied various details of the E+T setup using MCNPX. He tested the influence of the proton beam shape and position, foil placement and thickness, setup composition etc. Results of these studies are also described in his PhD work and in publications (e.g. [31] or [53]). 86

103 6.3. Simulation of the E+T setup Figure 61: Visualization of the Energy plus Transmutation setup as defined in MCNPX input file. On the left is SABRINA [83] plot provided by Jaroslav Šolc. Figure 62: Model of the parts of E+T setup in MCNPX, rendered in Povray 25 code [84], author M. Majerle Neutron fluxes in the E+T setup Advantage of the MCNPX simulation is a possibility of easy calculation of practically immeasurable things. In the calculation I can change material composition, density and add or except parts of the setup like shielding, structural materials, uranium etc. I repeated calculations of the neutron, proton and deuteron fluxes in the four target cylinders that I formerly performed for 0.7 GeV proton experiment [29]. This time I used deuterons with energy 2.52 GeV. In the Figure 63 we can see dependence between the presence of various parts of the setup and produced neutron fluxes. 25 The Persistence of Vision Raytracer (POVRAY) is a high-quality, totally free tool for creating stunning three-dimensional graphics [84]. It is available in official versions for Windows, Mac OS/Mac OS X and i86 Linux. 87

104 Neutron flux E [deuteron -1.cm -2.MeV -1 ] 6. MCNPX SIMULATIONS OF THE E+T SETUP 10 0 Pb Pb+const Pb+U+const without Cd whole E+T Neutron energy [MeV] Figure 63: Neutron flux (multiplied by energy because of binning) in the first target cylinder irradiated with 2.52 GeV deuterons, log-log scale, various parts of the setup are omitted. Uncertainties are on the level of 1 percent. Figure 64: Cross-section of the (n, ) reaction on 238 U in ENDF database [85]. 88

105 6.4. Neutron fluxes in the E+T setup Difference between bare Pb target and target with all constructions (Al and Fe support structures, U-rod cover from Al etc.) is almost negligible, support structures add some more high energy neutrons due to the spallation induced on them by scattered neutrons. Addition of natural uranium causes more neutrons in the region between 1 kev and 1 MeV due to the high energy fission. Biological shielding adds further neutrons to the low energy region bellow 10 kev and also a second maximum of the neutron spectrum around ev. Addition of the cadmium layer on the inner walls of the biological shielding suppresses this thermal energy peak. In all cases, a small peak can be seen close to the highest neutron energies. These neutrons come from the deuteron disintegration. Absorptions on the resonances in 238 U are also visible in the Figure 63. First depression on the low energy part of the neutron spectrum corresponds with the first important resonance in 238 U(n, ) 239 U reaction at 6.67 ev (is visible directly from the Figure 63 and Figure 64 comparison). Several conclusions can be drawn from Figure 63. First of all, most high energy neutrons are produced in the lead target and they are not notably moderated by the target/blanket setup. Support structures have no influence on this part of the spectrum, what is positive for the spallation spectrum studies. Addition of natural uranium causes addition of neutrons with energies lower than 10 MeV, but this addition is not a fundamental one. Biological shielding is fully responsible for the thermal, epithermal and resonance neutrons, but it is not changing number and spectrum of neutrons with energy higher than 8 MeV (differences are smaller than calculation uncertainty, which is below one percent at this energy region). Cadmium layer is an effective absorber of neutrons below 0.5 ev. I have made the same flux calculations also for protons and deuterons to see the production of these particles inside the target. Results can be seen in the Figure 136 and Figure 137 in Appendix I Section 1. Deuterons are in the target mainly slowed down and dismissed due to spallation (high energy peak), small amount of low energy deuterons is produced in the spallation reaction (low energy part of the spectrum in Figure 136 with four orders of magnitude lower intensity). Protons come from the deuteron disintegration (high energy peak in Figure 137) and from spallation reaction (low energy peak with three times lower intensity). No differences were observed in the proton and deuteron spectra within the simulation uncertainties, when various parts of the E+T setup were removed Calculation of the yields in used activation foils Due to the bad knowledge of experimental cross-sections of used reactions, our experimental evaluation ended always at the yields of isotopes. To get the same value from the simulation can be more complicated than the calculation of neutron spectra but is still possible with a good accuracy. Non-threshold reactions can be calculated directly using f4+fm tally. For threshold reactions the situation is more complicated because of the missing crosssections. Products of some (n,xn) reactions can be also calculated with f4+fm tally, but 89

106 6. MCNPX SIMULATIONS OF THE E+T SETUP the MCNPX handles with cross-sections not ideally. It uses libraries up to their highest energy, than when it has no model, it takes the last value in library and use it for the convolution with the rest of the neutron spectrum. In reality, (n,xn) cross-sections decrease slightly after their peak (see figures in Appendix J), so this approach is not suitable. We solved this problem in the following way. We add small volumes to the E+T model correspondent to the specific detector positions during each irradiation and we calculate the neutron, proton, deuteron and charged pion fluxes in these volumes using MCNPX. We calculate cross-sections of the (n,xn), (p,pxn), (d,dxn), and (, xn) reactions in TALYS and MCNPX and we connect them together. We make manual folding of the fluxes and cross-sections in Excel, according to the following equation (6.1): N Yield 1 A m r u E beam 0 ( E) ( E) ( E) ( E) ( E) ( E) ( E) ( E) de n n p p (6.1) where A r is the specific atomic mass of a chemical element from which the foil was made and m u is the unified atomic mass unit. Final outputs from the simulation part are directly the yields of isotopes. Contributions of various particles to the total isotope production in gold during 2.52 GeV deuteron experiment are displayed in the following Table 13 (result of MCNPX spectra simulation and manual folding). Most important is the contribution of neutrons, protons can create also a substantial part of the yield. Contribution of deuterons and pions is under the level of neutron spectra uncertainty and could be thus negligible. Nevertheless deuterons and charged pions are always included. Table 13: Contribution of various particles to the total yield, result of MCNPX simulation and manual folding. Positions in the first gap and behind the target, radial distance 3 cm and 10.7 cm, 2.52 GeV deuteron experiment. pi pi d d 196 Au first gap of the setup 194 Au 192 Au 3 cm 10.7 cm 3 cm 10.7 cm 3 cm 10.7 cm neutrons 98.8% 99.7% 94.6% 98.7% 92.8% 98.2% protons 1.1% 0.26% 5.06% 1.14% 6.63% 1.53% deuterons 0.03% 0.05% 0.07% 0.08% 0.11% 0.07% charged pions 0.08% 0.02% 0.24% 0.08% 0.45% 0.16% behind the setup Au 194 Au 192 Au 3 cm 10.7 cm 3 cm 10.7 cm 3 cm 10.7 cm neutrons 97.9% 99.2% 92.7% 97.3% 91.3% 96.6% protons 1.96% 0.70% 6.99% 2.51% 8.14% 3.05% deuterons 0.06% 0.02% 0.15% 0.06% 0.28% 0.12% charged pions 0.08% 0.05% 0.18% 0.12% 0.31% 0.21%

107 Exp. yield / sim. yield [-] 6.5. Calculation of the yields in used activation foils Manual folding of simulated spectra and cross-sections is a time consuming procedure, but it gives better results and a contribution of various types of particles to the total yield can be easily controlled. Dependence of the yield on neutron spectrum changes (or cross-section changes) can be also directly studied. Examples of the experiment to simulation ratios for 2.52 GeV deuteron experiment are in following Figure 65 and Figure 66. More experiment/simulation ratios are in the Appendix I. Uncertainty bars contain only statistical uncertainty from the DEIMOS32 and MCNPX, because the main purpose of this comparison is to see the relative differences between various isotopes and different measurement points (some uncertainties are the same for all points e.g. beam intensity uncertainty, and their involvement would be misleading in this case). If there would be an interest to compare absolute values of the exp/sim ratios to the one, other uncertainties must be also involved. Beside the statistical uncertainty from the DEIMOS32, three percent uncertainty from the HPGe detector calibration and spectroscopic corrections must be included in the experimental yield uncertainty, the same way as the additional uncertainty (at least ten percent) from the beam intensity determination. Uncertainties are believed to be independent and thus they should be summarized according to the equation X X y y y..., where X is the final experimental yield and y a is partial relative uncertainty (can be calculated as x1 y1 ). x Au 196Au 194Au 192Au 24Na Distance along the target [cm] Figure 65: Ratio between experiment and simulation in longitudinal direction for 2.52 GeV deuteron experiment, Au and Al samples at 3 cm from the target axis. 91

108 Exp. yield / sim. yield [-] 6. MCNPX SIMULATIONS OF THE E+T SETUP Au 196Au 194Au 192Au 24Na Radial distance from the target axis [cm] Figure 66: Ratio between experiment and simulation in radial direction for 2.52 GeV deuteron experiment, Au and Al samples in the first gap. Absolute values of the experiment/simulation ratio are strongly dependant on the beam intensity determination. From the multiplicity calculation (Figure 60) can be seen, that the beam intensity determination is probably not too correct in the case of deuteron experiments. However, also in proton experiments the experimental neutron multiplicity was slightly higher than the simulated one. Beam overestimation confirms also average value of the exp/sim ratios, which is higher than one, especially in the case of 1.6 GeV experiment (Figure 138 and Figure 139 in Appendix I, Section 2). There is a clearly visible trend in the 198 Au exp/sim ratios in all three deuteron experiments. In Figure 67 there are exp/sim ratios for the 198 Au yields on foils placed in the first gap at distance 3, 6, 8.5, and 10.7 cm (1-4), the same in the second gap (5-8) etc. I have observed the same behavior at 182 Ta, product of non-threshold (n, ) reaction in 181 Ta. The MCNPX simulation over-predicts number of epithermal and resonance neutrons inside the setup, with rising distance from the axis the ratio is decreasing. I have performed a few simulations with different uranium enrichment and density in order to clear up this effect, but without a satisfactory result. No such a behavior was observed at threshold reactions. 92

109 Exp. yield / sim. yield [-] 6.5. Calculation of the yields in used activation foils Au 1.6 GeV 198Au 2.52 GeV 198Au 4 GeV d Number of the foil [-] Figure 67: Ratio between experiment and simulation for all three deuteron experiments and 198 Au isotope Normalized experiment/simulation ratios To see clearly the shape of the exp/sim ratio, I normalized the values to the first foil in radial distance, respectively to the second foil in longitudinal distance (foils with maximal yields). This normalization cancels out also the influence of the absolute value of the cross-sections and of some correction uncertainties that are the same for each isotope (possible differences in the shape of the cross-section in MCNPX and reality are preserved). Only DEIMOS32 uncertainties are involved in these comparisons. Most of the normalized exp/sim ratios are close to the one (see Figure 68 and Figure 69, more figures can be found in Appendix I, Section 3). Discrepancies can have multiple sources; starting from beam position and shape during whole irradiation (uncertainty is hard to assess), ongoing with the cross-section shapes (can lead up to ten percent uncertainty) and closing with discrepancies coming from the foil placement in the experiment (imprecision of 5 mm can change the yield up to 20 percent proven by M. Majerle, see his PhD [41]). Exp/sim ratio is after including all of these uncertainties equal to one (within the uncertainty bars). No serious disagreements in the exp/sim ratios were found, so the INCL4/ABLA models seem to be generally precise in the case of deuteron beams. This can be confirmed also by the comparison between the figures with experimental yields, where a maximum both in longitudinal and radial direction can be seen, and in the figures of exp/sim ratios, where these maxima are missing (the simulation describes well the shape of yield curves). 93

110 Exp. yield / sim. yield [-] Exp. yield / sim. yield [-] 6. MCNPX SIMULATIONS OF THE E+T SETUP Au 196Au 194Au 192Au 24Na Distance along the target [cm] Figure 68: Ratio between experiment and simulation in longitudinal direction for 2.52 GeV deuteron experiment, Au and Al samples at 3 cm from the target axis. Ratios are normalized to the second foil Au 196Au 194Au 192Au 24Na Radial distance from the target axis [cm] Figure 69: Ratio between experiment and simulation in radial direction for 2.52 GeV deuteron experiment, Au and Al samples in the first gap. Ratios are normalized to the first foil. 94

111 Simulated yield [1/g*beam particle] 6.7. Yields for different beam particles of the same total energy 6.7. Yields for different beam particles of the same total energy I tried to compare yields of Au isotopes for different beam particles of the same total energy. I used deuteron, proton, and proton plus neutron (50:50, 1.26 GeV) beam of 2.52 GeV. I calculated neutron, proton, charged pion and deuteron spectra in the foil volume and made a folding with calculated cross-sections. Yields were calculated at 3 cm over the target axis in longitudinal direction and in the first gap of the setup in radial direction (places with highest neutron flux and thus also with highest yields). Examples of the results of beam particle comparison are in the following Figure 70 and Figure 71. Deuteron beam is the most efficient for neutron production. At same total energy it has lower ionization loses in the target than the proton beam, and so more energy can be used for spallation. In comparison with hypothetic mixed beam of protons and neutrons of the same total energy, deuterons generate slightly more neutrons (but still within calculation uncertainty). Difference between the deuteron and mixed beam is most probably caused by the difference in the MCNPX calculation of protons, neutrons and deuterons (description of their behavior in the model) p+n 1260 MeV d 2520 MeV p 2520 MeV Au Distance along the target [cm] Figure 70: Comparison of simulated longitudinal 196 Au yields for various beams of the same total energy, samples placed at 3 cm from the target axis. 95

112 Simulated yield [1/g*beam particle] 6. MCNPX SIMULATIONS OF THE E+T SETUP p+n 1260 MeV d 2520 MeV p 2520 MeV Au Radial distance from the target axis [cm] Figure 71: Comparison of simulated radial 196 Au yields for various beams of the same total energy, samples placed in the first gap of the setup Summary of the MCNPX simulations By summarizing MCNPX results from previous proton experiments, we observed an increasing difference in the radial direction between experiment and simulation for proton energies higher than 1.5 GeV, see Figure 72. For deuteron experiments there is a good agreement for all three measured energies (from 1.6 GeV up to 4 GeV), see Figure 73. This result prefers the hypothesis that in proton experiments the problem is rather in the experimental part than in the simulations. 1.5 GeV and 2 GeV proton experiments were first experiments on the E+T setup. At that time the influence of a lot of aspects was unknown (importance of proper beam position, foil placement etc.). When the proton beams will be again available on the Nuclotron, we will propose to perform an experiment at the beam energy equal or higher than 1.5 GeV to confirm this conclusion. 96

113 Exp. yield / sim. yield [-] exp. yield / sim. yield [-] 6.8. Summary of the MCNPX simulations GeV 1.5 GeV 1.0 GeV 0.7 GeV Au Radial distance from the target axis [cm] Figure 72: Ratio between experiment and simulation for different proton beam energies and 194 Au (overtaken from A. Krása [44]). Samples were placed in radial direction in the first gap of the setup GeV 2.52 GeV 4 GeV Au Radial distance from the target axis [cm] Figure 73: Ratio between experiment and simulation for different deuteron beam energies and 194 Au. Samples were placed in radial direction in the first gap of the setup. 97

114 6. MCNPX SIMULATIONS OF THE E+T SETUP 98

115 Cross-section [barn] Cross-section [barn] Cross-section [barn] Cross-section [barn] Chapter 7 Cross-section measurements of the (n,xn) threshold reactions My motivation for the cross-section measurements comes from the "Energy plus Transmutation" project discussed in the previous chapters. Au, Al, Bi, Co, In, Ta, and Y foils were used as activation neutron detectors, but unfortunately almost no experimental cross-section data for most of the observed threshold (n,xn), (n,p), and (n, ) reactions are available for higher neutron energies State-of-the-art of the neutron cross-section libraries The present status of knowledge of cross-sections for the (n,xn) reactions is poor. Figure 74 shows measured (from EXFOR [66]) and evaluated (from ENDF [85]) cross-sections for (n,xn) reactions in Au and Bi Au(n,2n) 196 Au EXFOR ENDF Neutron energy [MeV] Au(n,4n) 194 Au EXFOR ENDF Neutron energy [MeV] Bi(n,xn) (n,4n)206bi (n,5n)205bi Bi(n,xn) Neutron energy [MeV] Neutron energy [MeV] Figure 74: Neutron cross-sections for the Au and Bi (n,xn) threshold reactions. Data are taken from the EXFOR [66] and ENDF [85] (n,6n)204bi (n,8n)202bi (n,10n)200bi (n,7n)203bi (n,9n)201bi (n,11n)199bi In the case of gold, only (n,2n) reaction was measured in detail and by more authors (Figure 74 left up), but only up to less than 40 MeV. (n,4n) reaction on natural Au was measured also only up to 40 MeV (Figure 74 - right up). Higher (n,xn) reactions on Au have not been studied yet. 99

116 7. CROSS-SECTION MEASUREMENTS OF THE (N,XN) REACTIONS In the case of bismuth, reactions from (n,4n) up to (n,12n) were already measured (Figure 74 - down), highest neutron energy is 150 MeV. Unfortunately, all these values are from one experiment only [86], so these values should be independently verified. There are also huge distances between separate energies, so the cross-section peak is not well described. There are no evaluated data available so far. The situation for Al, Co, In, Ta, and Y is similar to Au. Low threshold reactions (n,2n and 3n) were studied in detail up to 30 MeV, but no experimental data exist for higher energies or (n,xn) reactions. More graphs with the data from EXFOR for various reactions are further in the text in the section with cross-section results and in Appendix J. One can there easily make his opinion on the situation in cross-section libraries of (n,xn) reactions. It is really necessary to perform new cross-section measurements to fill in the gaps in the libraries and estimate possible systematic errors in already measured values. Not only for the high energy neutron measurements by the means of activation foils, but also for the authors of the codes handling with crosssections and last but not least also for the designers of new spallation devices like ESS [87] etc. For the (n,xn) cross-section measurements we decided to use the same method as we are familiar with from the Dubna activation measurements neutron activation and gamma spectra measurement. Usage of activation analysis brought us a lot of difficulties, which we had to solve and fight with (e.g. background subtraction). On the other hand, we had a good knowledge on working procedure and various spectroscopic corrections. It has to be mentioned, that the cross-sections of the threshold reactions can be measured also by other methods, e.g. by the time of flight method for the neutron beam and by on-line and X-Ray measurements of the samples. But, these methods require more complicated equipment and longer time for preparation at the irradiation place, which were both unavailable for us. One of this type of cross-section measurement is described in the first chapter Limitations on neutron source First difficulty was a selection of proper neutron source. Spallation neutron sources in Dubna mentioned in Chapter 2 are white neutron sources with unknown neutron spectrum (neutron spectrum was up to now only calculated). These sources cannot be used for this type of cross-section measurements, see Figure 75. In the world there is a lot of quasi-monoenergetic neutron sources with more or less known spectrum ([88], [89], [90]), but only a few of them have sufficiently high beam intensity for activation measurements ((n,xn) cross-section are in order of ~1 barn or lower). Most of the neutron sources are also limited with the maximal neutron energy they can deliver, usually MeV (this is the reason for quite a good cross-section data in libraries for lower energies, but no data for higher energies). Other limitations were presence of spectroscopic laboratory we needed at least one detector with good resolution and efficiency for two weeks of continuous measurement. Irradiation place had to be easily accessible for quick manipulation with 100

117 Neutron flux density [1/(cm 2.s)] 7.2. Limitations on neutron source the samples, a low radiation around the irradiation place was also important (shortest half-lives are in order of few minutes, so immediate access to the irradiation place was needed). Last but not least was the question of experiment funding; we needed to find money on transport, accommodation, diets for three people, beam-time and additional experimental costs (liquid nitrogen for cooling the detector etc.) spallation source at JINR quasi-monoenergetic source at TSL Neutron energy [MeV] Figure 75: Comparison of the spallation neutron spectrum in Dubna and quasimonoenergetic neutron spectrum in TSL EFNUDAT project We decided to use the EFNUDAT project (European Facilities for Nuclear Data Measurements) to get access to one of the supported facilities The Svedberg laboratory of the Uppsala University, Sweden. The EFNUDAT project is an Integrated Infrastructure Initiative (I3) funded under the 6th framework program (FP6) of the European Commission. The main objective of EFNUDAT is to promote the coherent use and integration of infrastructure related services via networking, transnational access to the participating facilities for nuclear data measurements and joint research activities [91]. Figure 76: Logo of the EFNUDAT project [91]. 101

118 7. CROSS-SECTION MEASUREMENTS OF THE (N,XN) REACTIONS Figure 77: Countries and institutes involved in EFNUDAT [91]. In 2007 we started discussions with The Svedberg laboratory about the possibilities of the (n,xn) cross-section measurements on their quasi-monoenergetic neutron source. I have prepared with my colleague A. Krása a proposal for the experiment, which was accepted in October In June 2008 we have performed three irradiations with neutron energies 22, 47, and 94 MeV, details are discussed in following text. With the preliminary, but successful results we decided to exploit the EFNUDAT once again to fill in the gap between 47 and 94 MeV. I have prepared new proposal for further six energies, which was accepted in September 2009 and the irradiation took place in February Quasi-monoenergetic neutron source at The Svedberg laboratory Main experimental equipment of The Svedberg laboratory (TSL) is a Gustav Werner cyclotron, which can deliver beams with the energies up to 180 MeV for protons or 45 MeV per nucleon for heavier ions up to Xe. The main activity of TSL is based on an agreement between Uppsala Academic Hospital and Uppsala University on proton therapy. Tens of patients are routinely irradiated every week. Beamtime not used for proton therapy is devoted to commercial neutron and proton irradiation projects, mainly tests of radiation endurance of the electronics. The beam can be for this purposes collimated to a very small shielded pencil, which can irradiate separate microchips on the printed circuits. But, there is still some time for basic (academic) research, but one must apply for long time in advance. 102

119 7.4. Quasi-monoenergetic neutron source at The Svedberg Laboratory Figure 78: Photo of the Gustav Werner cyclotron (author s photo). In this laboratory quasi-monoenergetic MeV neutron source based on the 7 Li(p,n) 7 Be reaction was developed [92]. High energy protons from the cyclotron at TSL are directed to a thin, lithium target, neutron flux density can be up to cm -2 s -1 at standard user position (373 cm from the target). Neutron flux density is limited only with available heat removal from the target. The half of intensity is in the peak with FWHM = 1 MeV (corresponds to the ground state and first excited state at 0.43 MeV in 7 Be) and half of intensity is in a continuum in lower energies (corresponds to higher excited states, multiple-particle emission etc.). Proton energy loss in the target amounts to 2-6 MeV depending on the incident beam energy and target thickness. Downstream the target, the proton beam is deflected by a magnet and guided into a graphite beam dump. The whole proton beam line, target, bending magnet and most active devices are hidden in concrete corridor, so the hall is accessible immediately after the shutdown of the beam. The neutron beam is formed by an iron collimator (50 cm in diameter and 100 cm long) with a hole of variable size and shape. Behind the collimator there is a large cave (so called Blue hall) with the neutron beam dump at the end, more than 15 meters of free space are ready for the users. Multiple system of laser surveying can be used for exact sample allocation. Beam can be handled directly by the users via a User control interface. After the operators set up the beam and make intensity and calibration checks, they give the beam control to the user. User can then remotely switch off or on the beam without any contact to the accelerator operators. One can also open the Blue hall to restricted or free access mode for manipulations. This appeared to be a very useful procedure, which we used for short interruptions during the irradiation and taking out some of the samples. 103

120 7. CROSS-SECTION MEASUREMENTS OF THE (N,XN) REACTIONS Figure 79: Blue hall with the quasi-monoenergetic target and shielding [92]. Figure 80: User control interface for beam handling in TSL. 104

121 7.5. Cross-section estimation and planning of the experiment 7.5. Cross-section estimation and planning of the experiment During planning of the irradiation it was necessary to have at least some knowledge about the possible cross-section course and values. I have used the knowledge about the reaction thresholds calculated for the E+T experiment see Chapter 3. For calculation of cross-section courses I have also used the TALYS 1.0 code [82]. For most of the isotopes it was possible to make a convolution of calculated cross-sections and neutron spectra at irradiation points. I have roughly calculated yields of the isotopes and with the knowledge about the detector efficiency I have planned the weights and dimensions of the foils in order to get enough activated nuclei. Then I could also calculate when and how long we had to measure the sample in order to catch enough counts in the detector of the observed isotope (the activity of the foils was very low, a few hours of measurements were necessary for each foil). Table 14: Example of the number of predicted counts in the strongest line of gold isotopes. I calculated the numbers for 89 MeV neutron beam in Uppsala, detector position p2, weight of the foil 1.2 g. Measurement time after the beam end Number of counts in the strongest line of the isotope [-] from [min] to[min] 196 Au 194 Au 193 Au 192 Au 191 Au 190 Au 189 Au 188 Au 187 Au Neutron beams at TSL For every irradiation we received from the TSL staff report on the irradiation, where all necessary data were summarized. To the report belonged also the file with the course of irradiation (it contained the beam intensity during each burst). With this I could calculate the correction on irradiation, see further. I prepared samples and was present at all irradiations and gamma measurements. I have completely analysed results from the first three irradiations. Data from the second campaign in TSL from February 2010 will be the main subject of the PhD work of J. Vrzalová, but I have already analysed a few parts of it and I act as a consultant of Vrzalová s work. 105

122 7. CROSS-SECTION MEASUREMENTS OF THE (N,XN) REACTIONS Table 15: Neutron beam parameters at TSL Uppsala for used energies. Proton beam energy [MeV] ± 0.04 June 2008 February ± 0.2 ± 0.3 ± 0.2 ± 0.2 ± ± 0.3 Li-target thickness 26 [mm] Proton beam current [ A] Average energy of peak neutrons [MeV] Fraction of neutrons in the peak [%] Peak neutron flux density [10 5 cm -2 s -1 ] Total peak neutron flux [10 9 cm -2 ] Quasi-monoenergetic neutron source at Nuclear Physics Institute Second neutron source that we used is in the Nuclear Physics Institute (NPI) of the Academy of Sciences of the Czech Republic in Řež. Protons from the isochronous cyclotron U120-M are directed to the 7 Li target and quasi-monoenergetic neutrons in the range MeV can be produced [93]. Figure 81: Isochronous cyclotron U-120M in NPI Řež (left - own photo, right photo from [94]). 26 target thickness is connected with the FWHM of the high energy neutron peak. For used thicknesses of the Li target, ~ 98% of the proton beam passed without producing a neutron, protons only lost energy [95]. Thermal charge on the target is the main limiting factor for the neutron intensity (at TSL it is solved by defocusation of the proton beam in front of the target and thus by hitting large area). 106

123 7.7. Quasi-monoenergetic neutron source at Nuclear Physics Institute Cyclotron U120-M was designed and completed in JINR Dubna in In the following years the cyclotron was gradually modernized, after the devastating floods in 2002 new systems of cooling, vacuum and power supplies were build. This quasi-monoenergetic neutron source (placed at the point NG-2) is based on the same reaction 7 Li(p,n) 7 Be like the TSL one, but the construction layout of the target is completely different. Behind the foil with enriched lithium there is no bending magnet, but a graphite beam dump, which stops the rest of the beam. Behind this is a holder with the samples, so no collimators or shielding are applied. Whole setup of target and graphite stopper is cooled by alcohol to 5 degrees of Celsius, 600 Watts of heat are reliably dispatched. Target and its cooling are not fixed, but are movable, because they possess the same beam-line like the targets for the production of radiofarmacs 27. Table 16: Neutron beam parameters at NPI Řež for used energies. Proton beam energy [MeV] Start of irradiation :22 15:26 13:34 14:47 End of irradiation :17 8:02 10:00 11:00 Time of irradiation 19h 55min 16h 36min 20h 26min 20h 13min Average energy of peak neutrons [MeV] Total peak neutron flux [10 12 cm -2 ] I prepared and was present at all four irradiations at NPI Řež. I have analyzed first two experiments completely, the second two experiments were a subject of the Diploma thesis of J. Vrzalová [96]. I was a consultant of this work and I helped J. Vrzalová to understand all experimental details of these cross-section measurements. This Diploma thesis was successfully defended in June I have also tested possible attenuation of the neutron beam in the foils. In NPI we have used spare positions behind the samples of P. Bém, so there were some thin foils in front of us (usually 50 m thick). Neutron beam was partially collimated by the holders to approximately 40 mm, so I could not place all our foils side-by-side. I used 4 to 6 holders and sticked few foils upon itself. The result was that there were multiple foils in common beam. At studied neutron energies (17-34 MeV) cross-sections are around one barn or smaller, so the attenuation should be negligibly small. 27 This quasi-monoenergetic source could be operated only during weekends or special occasions up to now, because the beam-line is occupied most of the time with the radiofarmacs. 107

124 7. CROSS-SECTION MEASUREMENTS OF THE (N,XN) REACTIONS Graphite stopper Beam line Samples Li-target Figure 82: Quasi-monoenergetic source in NPI Řež based on the design of Uwamino [97]; scheme (left) [93] and a real outlook (right). To assess this I made a simple MCNPX simulation. I put all foils in one row, set maximal used thickness (the worst case that never happened in reality) and added double paper wrap, for illustration see Figure 83. I put a pencil beam from one side of this packet and calculated number of neutrons in the cells between the foils. Cells had the same dimensions like the foils, but bigger thickness (1 cm). I also used 30 energy bins for each cell to see, how the neutrons are decelerated. Figure 83: The sequence of the foils in the MCNPX simulation of neutron beam attenuation. First two Au-Cu sets are samples of P. Bém, rest are ours. Attenuation of the neutron beam in simulation was for most of the foils negligible, difference between the total number of neutrons in front of the first foil and behind the last one was 4 %. More serious problem was the energy spread of the originally monoenergetic beam, number of the neutrons in the energetic group MeV decreased to 85 % of the original value. The simulation was done for the worst case, in real measurement some foils were side by side, had smaller thickness or were left out, so the total amount of mass in the beam was much lower. I have not involved these results into the cross-section data from Řež, as I think this problem needs to be further studied. It will be one of the topics of PhD thesis of J. Vrzalová. 108

125 7.8. Studied materials 7.8. Studied materials In all irradiations I studied Au, Al, Bi, In, and Ta materials, the same we use in the Energy plus Transmutation experiments for high energy neutron measurements. Other group from the E&T RAW collaboration studies transmutation of radioactive and stable iodine in the field of spallation neutrons, so we involved tablets of natural iodine to our cross-section studies. In the second irradiation campaign at TSL Uppsala we measured samples of Y for the Polish E&T group. Neutron source at NPI is due to closer distance between target and samples (10-15 cm) much more intensive (2 orders of magnitude), so the samples were after the irradiation more active. Higher activity shortened time of measurement on the detector and we could study more materials. We decided to test beside the above mentioned also the foils from Zn, Mg, Fe, Cu, and Ni, practically all suitable materials for (n,xn) measurements of high energy neutrons. Materials were except the iodine in form of foils with dimensions of 20x20x mm 3, weights of the foils varied from 0.2 up to 7 grams depending on the foil type and beam energy. Foils were wrapped in two layers of paper; outer coating was removed before gamma measurements. Iodine samples were in the form of solid KIO 4 tablet. These tablets were manufactured on a pressing machine in NPI and packed hermetically in plastic coating, its weight was between one and three grams and dimension of the pills were 15x3 mm Evaluation procedure Typical irradiation time was 8 hours at TSL Uppsala, respectively 15 hours at NPI Řež. Transport from the irradiation hall to the spectrometer took approximately two minutes in Uppsala, ten minutes in Řež. Principles of measurement of irradiated foils, - spectra processing in DEIMOS32 and evaluation of the yields was the same as for the Energy plus Transmutation experiments. I have calculated all necessary spectroscopic corrections and I have included the important ones in the data (eg. beam instability correction was negligible for TSL measurements due to high stability of the beam). Theoretical background of the corrections was the same as described in E+T evaluation section, only the numbers differed. Final yield of studied isotopes was calculated according to the equation (3.26). I have used following equation (7.1) for determination of reaction cross-sections. where: N yield yield of studied isotope S area of the foil A molar weight N n number of neutrons in the peak N A Avogadro number N yield S A (7.1) N N n A 109

126 Neutron flux [1/MeV (peak area=1)] 7. CROSS-SECTION MEASUREMENTS OF THE (N,XN) REACTIONS Background subtraction Every source of high energy neutrons is more or less quasi-monoenergetic and neutron spectrum contains beside the main neutron peak also lower continuum stretching up to the thermal energies. This spectrum is different at every irradiating facility because of different construction of the target and surrounding equipment, and also because of the method of spectrum determination (experiment / experiment+ calculation / calculation). Neutron spectra for TSL are in Figure 84, for NPI in Figure MeV p-beam, 2 mm Li-target 49.5 MeV p-beam, 4 mm Li-target 61.8 MeV p-beam, 4 mm Li-target 69.1 MeV p-beam, 4 mm Li-target 75.4 MeV p-beam, 4 mm Li-target 92.5 MeV p-beam, 8.5 mm Li-target 97.6 MeV p-beam, 8 mm Li-target Neutron energy [MeV] Figure 84: Quasi-monoenergetic neutron spectrum from 7 Li(p,n) 7 Be at the TSL. Data have been overtaken from the facility staff A. Prokofiev. Proper measurement of the high energy neutron spectrum (e.g. by the means of the time-of-flight method) is not possible at all accelerators because of the beam structure, small space, electromagnetic disturbance, etc. In TSL, the conditions for neutron spectrum measurement were good and the neutron spectrum is nowadays known with the uncertainty below 10% (from personal discussion with A. Prokofiev, TSL). In NPI, neutron spectrum from the quasi-monoenergetic source was never exactly measured because of long beam pulse and insufficient space in the cyclotron crypt. The source was manufactured according to the source designed and operated in Japan by Y. Uwamino, so the neutron spectrum is believed to be the same or very similar. More about the neutron spectra can be found in [97]. Neutron spectrum at NPI is more complicated compared to TSL because of the carbon beam stopper. Reaction nat C(p,xn)X is not negligible in low energy region, see Figure

127 Neutron flux [n/mev/sr/c] Number of neutrons [1/sr.MeV.C] Background subtraction MeV 25 MeV 30 MeV 35 MeV 40 MeV Neutron energy [MeV] Figure 85: Quasi-monoenergetic neutron spectrum from 7 Li(p,n) 7 Be at cyclotron Řež data overtaken from the facility staff M. Honusek. Neutron energy [MeV] Figure 86: Neutron spectrum produced in reaction with 7 Li target and nat C beam stopper in the case of NPI target station, overtaken from M. Honusek [98]. Because of the large amount of background neutrons, production of the isotope by these neutrons was not negligible for most of the isotopes. Only reactions with the threshold few MeV lower than the neutron peak could be used to direct cross-section evaluation (Figure 87 left). Number of these isotopes is not high, usually one or none per one beam energy and material. Originally it was planned not to evaluate other 111

128 Arbitrary units [-] 7. CROSS-SECTION MEASUREMENTS OF THE (N,XN) REACTIONS isotopes than these backgroundless. I have developed a procedure how to subtract the background in the case like in Figure 87 right. neutron spectrum simulated crosssection neutron spectrum simulated crosssection Neutron energy [MeV] Neutron energy [MeV] Figure 87: Example of folding of the quasi-monoenergetic neutron spectrum and simulated cross-section. I used TALYS to calculate cross-section of every reaction in the same energy bins like is in the neutron spectrum in TSL and NPI. Then, I made a folding of the neutron spectra and TALYS cross-section. I calculated isotope production in neutron peak and by the whole neutron spectrum. Finally I made a ratio of these two values; this ratio stands for the relative production of the isotope in the peak. I multiplied the experimental yield of the isotope by this number, what resulted in the subtraction of the amount of isotope produced by background neutrons. Peak to whole spectrum ratio varies from 10 to 100 percent. This background subtraction procedure is a potential source of unknown uncertainty. It is insensitive to the absolute value of the cross-section (I use the same cross-section both in numerator and denominator), but a modification in the crosssection shape or in the neutron spectrum shape can change it. Calculation of the crosssection in TALYS is also a weak point of this procedure. TALYS enables for example five basic settings of nuclear level densities. Cross-sections have slightly different shape for each of it (see Figure 95), so there is a space for variations and changes in background subtraction procedure. At the end of the year 2009 a new version of TALYS appeared, concretely version 1.2. I calculated the cross-sections of Au in this new version and got different results compared to those from TALYS 1.0, see e.g. Figure 99 - Figure 100. So, there is again a place for changes in background subtraction. However, most of the changes remain within 10 %. Direct uncertainty assessment of calculated cross-section is not involved in the TALYS up to now, but there are signs it will be possible in a new version. At ND2010 conference, one of the TALYS authors S. Hillaire showed his current work repeated calculations with automatically varied inner TALYS parameters. He got some region, where most of the calculated crosssection lies. This region can be connected with the uncertainty of calculated crosssection. More details about the TALYS calculations of cross-sections are discussed in following Chapter

129 7.10. Background subtraction I studied also other possibilities for background subtraction. M. Honusek from NPI uses routinely step-by-step method [99], [100]. He plans the irradiations so, that he has every 2 3 MeV one measurement. He starts with the cross-section value close to the threshold, which is not affected by the neutron background. With this cross-section value he moves step-by-step to higher energies and subtracts the background. Compared to our procedure this approach is safer in using real cross-sections, not calculated ones. We started cross-section measurements with three energies - 22, 47, and 94 MeV so this was no usable for us. But in the future, when we will have better coverage of the energy interval, it is planned to try also this process of neutron background subtraction. At some quasi-monoenergetic neutron sources the neutron background is independent within some angle, but the neutron peak disappears when moving from the beam axis. Then it is possible to irradiate the same samples in the direct beam and under certain angle from the beam axis and then subtract the yields [101]. In the case of TSL this is not possible because of the 1m thick iron collimator. In NPI I tried to place Au samples under the angle 30 and 60 from the beam axis during 32.5 MeV irradiation, see Figure 88. Comparison of the neutron spectra under selected angles are in Figure 89 (overtaken from Y. Uwamino [97]). From this figure it can be seen that in the case of NPI the neutron peak does not disappear completely and also that the background changes a bit. Figure 88: Placement of the Au and Al samples under the 30 and 60 from the beam axis. The results of this experiment (relative ratios of the production by background neutrons) were too far from values I got from TALYS/spectra convolution and most probably also from the reality. Results completely confirmed my presumption that this background subtraction procedure is not usable for us. Reasons can be found already in the neutron spectra shown in Figure 89 - they do not agree completely with the statements presented in the work of S. Sekimoto [101]. Second reason can be in the construction of NPI neutron source. Neutron spectra are overtaken from Y. Uwamino, who measured them on a neutron source similar to the NPI one only within the beam axis. Under non-zero angle there is much more material around the target in NPI than it was in Y. Uwamino s case, see Figure 90. This leads to probably higher differences 113

130 7. CROSS-SECTION MEASUREMENTS OF THE (N,XN) REACTIONS between the real neutron spectra in non-zero angles at NPI and those presented in Y. Uwamino s work. Figure 89: Neutron spectra under 0 and 60 angle from the beam axis, overtaken from Y. Uwamino [97]. 60 degrees 30 degrees Figure 90: Comparison between the neutron source construction of Y. Uwamino [97] and at NPI Řež [93]. Used angles are drawn in the right part of the figure Uncertainty analysis Main uncertainties in my cross-section measurements come from the neutron spectrum knowledge (10%), beam intensity (10%), gamma-detector calibration (3%), and from the Gauss-fit of the gamma peaks in DEIMOS32 (at least 2%). At this moment I consider that all these uncertainties are independent, so the final cross-section 114