DOCSLIB.ORG

Measured and Calculated Effective Delayed Neutron Fraction of the Ipr-R1 Triga Reactor

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Measured and Calculated Effective Delayed Neutron Fraction of the Ipr-R1 Triga Reactor 2011 International Nuclear Atlantic Conference - INAC 2011 Belo Horizonte,MG, Brazil, October 24-28, 2011 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-04-5 MEASURED AND CALCULATED EFFECTIVE DELAYED NEUTRON FRACTION OF THE IPR-R1 TRIGA REACTOR Rose Mary G. P. Souza, Hugo M. Dalle, Daniel A. M. Campolina Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) Comissão Nacional de Energia Nuclear (CNEN) Campus da UFMG - Pampulha – Caixa Postal 941 30123-970, Belo Horizonte, MG, Brazil. [email protected], [email protected]; [email protected] ABSTRACT The effective delayed neutron fraction, beff , one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U235. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the beff value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current beff parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 ± 38) pcm and (744 ± 20) pcm, respectively. The beff was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. 1. INTRODUCTION The nominal thermal power of the IPR-R1 TRIGA reactor, at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, is 100 kW. The work presented in this paper shows the effective fraction of delayed neutrons (beff) calculated by Monte Carlo transport code MCNP5, and estimated experimentally by two methodologies: the inhour equation and the rod drop. One of the most important aspects of the fission process from the point of view of controlling the reactor is the presence of delayed neutrons. In general, delayed neutrons are more effective to control the reactor than prompt neutrons because they are born at lower energy compared to prompt (fission) neutrons. Thus, they have a better chance to survive leakage and resonance absorption. A delayed neutron is a neutron emitted by an excited fission product nucleus during beta disintegration some appreciable time after the fission. There are six decay chains which are of significance in the emission of delayed neutrons. Correspondingly, delayed neutrons are commonly discussed as being in six groups. Each of these groups, (i), are characterized by a fractional yield bi and a decay constant li. The amount bi is defined as the fraction of the total number of fission neutrons that are emitted by the decay of fission products in the group i. The effective total delayed neutron fraction is designated beff. The value of this parameter for a given fuel will vary over the lifetime of the reactor core, and also with the average energy of the neutrons producing fission. βeff depends on the core size and presents higher values for smaller cores. The effective fraction of delayed neutrons recommended by the General Atomic Technologies inc. (manufacturer of TRIGA) is 790 pcm [1, 2]. 2. BRIEF DESCRIPTION OF THE IPR-R1 REACTOR The IPR-R1 TRIGA reactor is a 100 kW facility. The reactor reached its first criticality on November 1960, and since then its core configuration and the number of fuel elements have been changed several times. As calculated in [3] about 4% of the U-235 in the core was burned during its 50 years. The fuel is a homogeneous mixture of uranium-zirconium hydride containing about 8.5% and 8% by weight of uranium, for stainless-steel and aluminum clad elements, respectively, enriched to 20% in 235U [1, 2]. The reactor core has 91 positions of which 63 are filled with fuel elements (59 Al-clad and 4 SS-clad elements) arranged in five concentric rings, inside a water tank. In the F ring there are 23 graphite dummy elements. The power level of the reactor is controlled by three control rods: Regulating, Shim, and Safety. The reactor is equipped with three irradiation facilities: rotary specimen rack, pneumatic transfer tube, and central thimble. A complete description of the IPR-R1 reactor can be found in [4]. Figure 1 shows the current core configuration which was used in the experiments. Figure 1. IPR-R1 TRIGA reactor core. INAC 2011, Belo Horizonte, MG, Brazil. 3. MONTE CARLO CALCULATION METHOD AND MODEL The general purpose, continuous energy, generalized-geometry Monte Carlo transport code MCNP5 was used in the calculations. In MCNP it is possible to model the reactor in 3-D geometry in details. A very detailed description of the MCNP reactor core model can be found in [3]. In this model the fuel burnup since the first criticality of the IPR-R1 was simulated taking into account the power history as recorded in the reactor log books. ENDF/B-VII pointwise cross sections data libraries were used in the simulations. When such data were not available then ENDF/B-VI data was used, instead. Three core configurations were simulated. The first two are the Beginning Of Life (BOL) core, as it was in 1960, considering one configuration with graphite elements filling the positions into the F ring and the other with water filling the positions in F ring. The third core is the current core, as shown in Fig. 1. The effective delayed neutron fraction, βeff, was calculated using the following equation [5, 6]: kP b eff = 1 - (1) keff where kp is the effective neutron multiplication factor only for prompt neutrons and keff is the effective neutron multiplication factor for both, prompt and delayed neutrons. The MCNP calculations were run with 4.0x105 histories per cycle and 2500 active cycles of neutrons, leading to a standard deviation within 0.00008 (8 pcm). The results of calculated βeff for all the three core configurations are presented in the Table 1. One can notice that the calculation result of the BOL core without any dummy elements surrounding the core agrees very well with the nominal value of 0,00790 for βeff informed by the reactor manufacturer, GA. This may be an indirect indication in favor of the adopted calculation model. Furthermore, it can be observed that when graphite dummy elements are filled into the F ring, then the βeff value reduces, as expected because the graphite elements work as a reflector, diminishing the leakage of the high energy prompt neutrons and thus improving the contribution of these neutrons in the balance. The current core is larger than the BOL, thus, the additional reduction in the βeff is also expected as the effective delayed neutron fraction depends on the core size and presents higher values for smaller cores. Such results agree with other authors [5, 6] as the delayed neutrons are born with lower energies and are more effective in inducing fission in system with larger leakage. INAC 2011, Belo Horizonte, MG, Brazil. Table 1. Reactor core configurations and calculated βeff Number of Date of the Element in Error β Fuel Factor Mean eff Configuration the F Ring (pcm) (pcm) Elements keff 0.99576 8 06/11/1960 56 H2O 789 kp 0.98790 8 keff 1.02254 8 06/11/1960 56 Graphite 750 kp 1.01487 8 keff 1.02183 8 06/11/2004 63 Graphite 747 kp 1.01420 8 4. EXPERIMENTAL DETERMINATION OF DELAYED NEUTRON FRACTION A brief discussion of the methodologies that were adopted to obtain beff and the results of the experiments performed for such purposes are given in the following items. 4.1. Measurement Results Based on Inhour Equation Theoretically, the reactivity of a control rod can be given approximately by [7, 8]: é Z 1 æ 2p Z öù r (Z ) = r (H ) ê - sen ç ÷ú (2) ë H 2p è H øû where r (Z) and r (H) refer to the control rod reactivity worth inserted the distance Z and fully inserted, respectively. Knowing the control rod worth r (H), the value of the reactivity at different Z positions can be calculated. The integral curve of the Regulating control rod is shown in Fig. 2, and of the Shim and Safety rods in Fig. 3. These curves were obtained by the Eq. (2). The total reactivity worth of the Regulating, Safety and Shim rods are 379 pcm, 2244 pcm and 2536 pcm, respectively [9]. INAC 2011, Belo Horizonte, MG, Brazil. 400 350 300 250 200 150 Reactivity (pcm) 100 50 0 0 200 400 600 800 1000 Regulating Rod Position Figure 2. Integral curve of the Regulating control rod. 3000 2500 2000 Safety 1500 Shim 1000 Reactivity (pcm) 500 0 0 200 400 600 800 1000 Rod Position Figure 3. Integral curves of the Shim and Safety control rods. The relationship between a step change in reactivity (r) and the resulting reactor period (T) is given by the inhour equation: w m b w r = l + å i (3) k 1 w + li In which l is the prompt neutron generation time; m is the number of delayed neutron groups; w are the (m+1) solutions, or roots, of the inhour equation; and k is the effective neutron multiplication factor. The stable period of the reactor is defined by the Eq.
Load more

Recommended publications
  • KINETIC ANALYSIS of SUB-PROMPT-CRITICAL REACTOR ASSEMBLIES Hy S Das Theoretical Physics Division
    BARC/1992/E/011 O 1 "o KINETIC ANALYSIS OF SUB-PROMPT-CRITICAL REACTOR ASSEMBLIES hy S Das Theoretical Physics Division 1992 BARC/1992/E/011 o UJ GOVERNMENT OF INDIA - AATOMIT C ENERGY COMMISSION U KINETIC ANALYSIS OF SUB-PROMPT-CRITICAL REACTOR ASSEMBLIES by S. Das Theoretical Physics Division BHABHA ATOMIC RESEARCH CENTRE BOMBAY, INDIA 1992 BARC/1992/E/011 BIBLIOBRAPHIC DESCRIPTION SHEET FOR TECHNICAL REPORT (as par IS i 94BNB - 1980) 01 Security classification i Unclassified 02 Distribution t External 03 Report status i New 04 Series i BARC External 05 Report type i Technical Report 06 Report No. i BARC/1992/E/011 07 Part No. or Volume No. i 08 Contract No. i 10 Title and subtitle i Kinetic analysis of sub-prompt- critical reactor assemblies 11 Collation i 14 p.f 4 figs., 1 tab. IS Project No. i 20 Personal author<s> i S. Das 21 Affiliation of author(s) i Theoretical Physics Division, Bhabha Atomic Research Centre, Bombay 22 Corporate author(s) i Bhabha Atomic Research Centre, Bombay - 400 085 23 Originating unit i Theoretical Physics Division, BARC, Bombay 24 Sponsor(s) Name i Department of Atomic Energy Type i Government 30 Date of submission i May 1992 31 Publication/Issue date i June 1992 Contd... (ii) 1 40 Publ i»h»r/Distributor t Head, Library and Information Division, Bhabha Atomic Rasaarch Contra, Bombay 42 Form of distribution t Hard Copy 50 Languaga of taxt i English 51 Languaga o-f summary i English 92 No. of references i % rafs. 53 Givas data on i 60 Abstract x Nautronic analysis of safaty-ralatad kinetics problams in experimental neutron multiplying assemblies has bean carried out using a sub-prompt-critical reactor model.
    [Show full text]
  • Total Prompt Energy Release in the Neutron-Induced Fission Of
    Nuclear Physics A 772 (2006) 113–137 Total prompt energy release in the neutron-induced fission of 235U, 238U, and 239Pu D.G. Madland Theoretical Division, Los Alamos National Laboratory, Los Alamos, NM 87545, USA Received 12 January 2006; received in revised form 28 March 2006; accepted 29 March 2006 Available online 2 May 2006 Abstract This study addresses, for the first time, the total prompt energy release and its components for the fission of 235U, 238U, and 239Pu as a function of the kinetic energy of the neutron inducing the fission. The com- ponents are extracted from experimental measurements, where they exist, together with model-dependent calculation, interpolation, and extrapolation. While the components display clear dependencies upon the incident neutron energy, their sums display only weak, yet definite, energy dependencies. Also addressed is the total prompt energy deposition in fission for the same three systems. Results are presented in equation form. New measurements are recommended as a consequence of this study. © 2006 Elsevier B.V. All rights reserved. PACS: 24.75.+i; 25.85.Ec; 25.85.Ca; 27.90.+b Keywords: Energy release and energy deposition in neutron-induced fission; Experiment and Los Alamos model; 235U; 238U; 239Pu 1. Introduction This study is a consequence of open questions on the magnitudes of the total prompt energy re- lease in fission, the total prompt energy deposition in fission, the components of these quantities, and their dependencies upon the kinetic energy of the neutron inducing the fission. Our results are given in Eqs. (31)–(33) and Fig. 17 for the total prompt energy release in fission, and in Eqs.
    [Show full text]
  • Systematic of Delayed Neutron Parameters
    RU0110988 40 XIV International Workshop on Nuclear Fission Physics Systematic of Delayed Neutron Parameters S.G. Isaev, V.M. Piksaikin Institute of Physics and Power Engineering, Obninsk, Russia Abstract The experimental studies of the energy dependence of the delayed neutron parameters for various fissioning systems has shown that the behavior of a some combination of delayed neutron parameters has a similar features. On the basis of this findings the systematics of delayed neutron experimental data for thorium, uranium, plutonium and americium isotopes have been investigated with the purpose to find a correlation of DN parameters with characteristics of fissioning system as well as a correlation between the delayed neutron parameters themselves. Below we will present the preliminary results which were obtained during this study omitting the physics interpretation of the results. Introduction The knowledge of the time dependence of P-delayed neutron emission is of great importance for the development of reliable data base required both for reactor-physics analysis and for the investigation of nuclei which are far from the valley of stability. Since discovery of delayed neutrons the time dependence of delayed neutron emission from neutron induced fission was studied in more than 238 experiments [1]. In parallel to the experimental studies of aggregate DN emission from neutron induced fission the group parameters of delayed neutron emission were calculated using microscopic approach (summation technique) [2]. These calculations are based on the knowledge of the emission probabilities values Pn, half-life values and yields of individual precursors of delayed neutrons [3]. One of the main results of the aggregate DN decay curve measurements was that the group periods (or decay constants) for various fissioning systems for fast neutrons induced fission do not differ from each other [4].
    [Show full text]
  • Radiation Detection from Fission
    ORNL/TM-2004/234 RADIATION DETECTION FROM FISSION J. T. Mihalczo DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via the U.S. Department of Energy (DOE) Information Bridge. Web site http://www.osti.gov/bridge Reports produced before January 1, 1996, may be purchased by members of the public from the following source. National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900 E-mail [email protected] Web site http://www.ntis.gov/support/ordernowabout.htm Reports are available to DOE employees, DOE contractors, Energy Technology Data Exchange (ETDE) representatives, and International Nuclear Information System (INIS) representatives from the following source. Office of Scientific and Technical Information P.O. Box 62 Oak Ridge, TN 37831 Telephone 865-576-8401 Fax 865-576-5728 E-mail [email protected] Web site http://www.osti.gov/contact.html This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof.
    [Show full text]
  • Nuclear Fission Dynamics: Past, Present, Needs, and Future
    NT@UW-19-18,LA-UR-19-32211 Nuclear Fission Dynamics: Past, Present, Needs, and Future Aurel Bulgac,1 Shi Jin,1 and Ionel Stetcu2 1Department of Physics, University of Washington, Seattle, Washington 98195–1560, USA 2Theoretical Division, Los Alamos National Laboratory, Los Alamos, NM 87545, USA (Dated: March 23, 2020) Significant progress in the understanding of the fission process within a microscopic framework has been recently reported. Even though the complete description of this important nuclear reac- tion remains a computationally demanding task, recent developments in theoretical modeling and computational power have brought current microscopic simulations to the point where they can pro- vide guidance and constraints to phenomenological models, without making recourse to parameters. An accurate treatment compatible with our understanding of the inter-nucleon interactions should be able to describe the real-time dynamics of the fissioning system and could justify or rule out assumptions and approximations incompatible with the underlying universally accepted quantum- mechanical framework. Of particular importance are applications to observables that cannot be directly measured in experimental setups (such as the angular momentum distribution of the fission fragments, or the excitation energy sharing between the fission fragments, or fission of nuclei formed during the r-process), and their dependence of the excitation energy in the fissioning system. Even if accurate predictions are not within reach, being able to extract the trends with increasing exci- tation energy is important in various applications. The most advanced microscopic simulations of the fission process do not support the widely used assumption of adiabaticity of the large amplitude collective motion in fission, in particular for trajectories from the outer saddle towards the scission configuration.
    [Show full text]
  • Reactor Physics
    REACTOR PHYSICS COURSE INTRODUCTION This Training Manual assumes prior knowledge of Nuclear Theory. It extends this information into a discussion of Reactor Physics, particularly as it relates to CANDU reactors. The course begins with the general principles of reactor configuration required to maintain a self- sustaining chain reaction. It continues with reactor dynamics (in both the critical and subcritical core), reactivity feedback effects (temperature effects, fission product poisoning, and fuel burnup), and ends with operational considerations (at low and high power). The material covers four main areas, subdivided into eight major sections as follows: • The Critical Reactor at Steady Power Output (Section 1) • The Dynamic Reactor (Sections 2 and 3) • Reactivity Feedback Effects (Sections 4, 5, and 6) • Reactor Operations (Sections 7 and 8) CNSC i Science and Reactor Fundamentals – Reactor Physics Technical Training Group TABLE OF CONTENTS Objectives 1 The Critical Reactor at Steady Power Output 9 1.0 INTRODUCTION 9 1.1 Fission 9 1.2 Harnessing Fission 15 1.3 Movement of Neutrons Through the CANDU lattice Lattice 18 1.4 The Finite Reactor 26 Response of The Critical Reactor to a Reactivity Change 31 2.0 INTRODUCTION 31 2.1 Exponential Power Rise 31 2.2 Corrections to Exponential Reactor Response 33 2.3 The Effect of Delayed Neutrons 36 2.4 Prompt Criticality 40 2.5 Power Rundown: The Prompt Drop 42 Responsiveness of The Subcritical Reactor 45 3.0 INTRODUCTION 45 3.1 Neutron Flux in a “Shut Down” Reactor 45 3.2 Dynamics in the Subcritical
    [Show full text]
  • 1- TOPIC: 292001 KNOWLEDGE: K1.02 [3.0/3.1] QID: B45 the Term “Neutron Generation Time” Is Defined As the Average Time Betw
    NRC Generic Fundamentals Examination Question Bank--BWR May 2020 TOPIC: 292001 KNOWLEDGE: K1.02 [3.0/3.1] QID: B45 The term “neutron generation time” is defined as the average time between... A. neutron absorption and the resulting fission. B. the production of a delayed neutron and subsequent neutron thermalization. C. neutron absorption producing a fission and absorption or leakage of resultant neutrons. D. neutron thermalization and subsequent neutron absorption. ANSWER: C. TOPIC: 292001 KNOWLEDGE: K1.02 [3.0/3.1] QID: B174 Which one of the following is the definition of the term, prompt neutron? A. A high-energy neutron emitted from a neutron precursor, immediately after the fission process. B. A neutron with an energy level greater than 0.1 MeV, emitted in less than 1.0 x 10-4 seconds following a nuclear fission. C. A neutron emitted in less than 1.0 x 10-14 seconds following a nuclear fission. D. A neutron emitted as a result of a gamma-neutron or alpha-neutron reaction. ANSWER: C. -1- Neutrons NRC Generic Fundamentals Examination Question Bank--BWR May 2020 TOPIC: 292001 KNOWLEDGE: K1.02 [3.0/3.1] QID: B245 Delayed neutrons are neutrons that... A. have reached thermal equilibrium with the surrounding medium. B. are expelled within 1.0 x 10-14 seconds of the fission event. C. are expelled with the lowest average kinetic energy of all fission neutrons. D. are responsible for the majority of U-235 fissions. ANSWER: C. TOPIC: 292001 KNOWLEDGE: K1.02 [3.0/3.1] QID: B1146 (P1945) Which one of the following types of neutrons has an average neutron generation lifetime of 12.5 seconds? A.
    [Show full text]
  • Module 2: Reactor Theory (Neutron Characteristics)
    DOE Fundamentals NUCLEAR PHYSICS AND REACTOR THEORY Module 2: Reactor Theory (Neutron Characteristics) NUCLEAR PHYSICS AND REACTOR THEORY TABLE OF CONTENTS Table of Co nte nts TABLE OF CONTENTS ................................................................................................... i LIST OF FIGURES .......................................................................................................... iii LIST OF TABLES ............................................................................................................iv REFERENCES ................................................................................................................ v OBJECTIVES ..................................................................................................................vi NEUTRON SOURCES .................................................................................................... 1 Neutron Sources .......................................................................................................... 1 Intrinsic Neutron Sources ............................................................................................. 1 Installed Neutron Sources ............................................................................................ 3 Summary...................................................................................................................... 4 NUCLEAR CROSS SECTIONS AND NEUTRON FLUX ................................................. 5 Introduction .................................................................................................................
    [Show full text]
  • (I) CORRELATION PROPERTIES of DELAYED NEUTRONS from FAST NEUTRON INDUCED FISSION
    XA9848517-/? International Atomic Energy Agency INDC(CCP)-415 Distr. CI + Special INTERNATIONAL NUCLEAR DATA COMMITTEE TWO REPORTS: (i) CORRELATION PROPERTIES OF DELAYED NEUTRONS FROM FAST NEUTRON INDUCED FISSION V.M. Piksaikin, S.G. Isaev Institute of Physics and Power Engineering, Obninsk, Russia (ii) METHOD AND SET-UP FOR MEASUREMENTS OF TRACE LEVEL CONTENT OF HEAVY FISSIONABLE ELEMENTS BASED ON DELAYED NEUTRON COUNTING V.M. Piksaikin, A.A. Goverdovski, G.M Pshakin Institute of Physics and Power Engineering, Obninsk, Russia October 1998 IAEA NUCLEAR DATA SECTION, WAGRAMERSTRASSE 5, A-1400 VIENNA 29-49 Printed by the IAEA in Austria October 1998 INDC(CCP)-415 Distr. CI + Special TWO REPORTS: (i) CORRELATION PROPERTIES OF DELAYED NEUTRONS FROM FAST NEUTRON INDUCED FISSION V.M. Piksaikin, S.G. Isaev Institute of Physics and Power Engineering, Obninsk, Russia Abstract The experimental studies of the energy dependence of the delayed neutron parameters for various fissioning systems has shown that the behavior of a some combination of delayed neutron parameters (group relative abundances a, and half lives T,) has a similar features. On the basis of this findings the systematics of delayed neutron experimental data for thorium, uranium, plutonium and americium isotopes have been investigated with the purpose to find a correlation of DN parameters with characteristics of fissioning system as well as a correlation between the delayed neutron parameters themselves. Below we will present the preliminary results which were obtained during this study omitting the physics interpretation of the results. (ii) METHOD AND SET-UP FOR MEASUREMENTS OF TRACE LEVEL CONTENT OF HEAVY FISSIONABLE ELEMENTS BASED ON DELAYED NEUTRON COUNTING V.M.
    [Show full text]
  • 24)Ay Nuclear Criticality Safety Course
    NUCLEARCRITICALITYSAFETY: 2-DAYTRAININGCOURSE Offeredat o TheLosAlamosNationalLaboratory EditedandCompiledby JohnA. Schlesser LosAlamos NATIONAL LA80RATORV LosMnos, NO.Jhhxieo87S45 4P . , t . “ AnAffirmativeAction/EqualOpportunityEmpbyer TheLosAiamosNationalLaboratoryk operatedbytheUniversityofCaliforniafortheUnitedStatesDepartmentofEnergy undercont~t W-7405ENG-35. Theillustrationonthecoverwasdesi nedbyJohnSchlesserandGaryWebbanddrawn byAnnMarieDyson.Itdepictsoneof thecourse’stwocriticalexperiments. DISWNMER ThismanuscriprwaspraparadasanaccourrtofwwksponsoradbyanagencyoftfmUnitedStatesGovernment.Neitfwhe ntsofthaUniversityofCaMqnia,,theUnitadStatesGovarnqrepInoranyageqthemof, noranyoftheirempbyaes, 3% anywarrantyexpqmwxforI bed,orassumesanyI a!habdttyorresponshIty@rtheaccaracy,compfetenass,or usefufney30faniM’ormatmn,apa ,p*c~tip_%-,u_* tiWtmumwould~tkh~tim~y ownednghk ke~ herwrrto anyapqifc commercialprodurxprocess,or servioeby trqcfename, rnamkturar,orotherwrae,doasnotnemssanIyconstituteor&n yb endrxaementorecommendabon,orfavoringbyUp Un”ti StaresGOmrnmantorfyy ~ thereof.Theviews!snJ”q“nbnsofauthorsexpressedhereindonot~Y stateormflacttfmseoftheUrrned?&esGovernmentoranyagencyUmeof. This com ilation of notes is presented as a source re?erence for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course’s primary instructor. Tableof Contents I. General 2-Day Course Objectives Course Critique 11. History Early History of Criticality Safety Criticality Control in Operations with
    [Show full text]
  • Neutron Numbers 8, 20 and 28 Were Considered As Magic Over Several Decades
    Introduction to nuclear physics with accelerated beams O. Sorlin (GANIL) What are limits of stability ? Which new phenomena emerge at the drip lines (halo, clustering, breaking mirror symetry… ) ? How do nuclear structure and shape evolve along the chart of nuclides ? How does nuclear structure change with Temperature and Spin value ? How to unify nuclear structure and reaction approaches ? How to probe the density and isospin dependence of the nuclear equation of state ? What are nuclear processes that drive the evolution of stars and galaxies in the universe ? How / where are nuclei synthesized in the universe ? Find a universal interaction, based on fundamental principles, that can model nuclear structure and reactions in nuclei and in stars (i.e. from the fm to 104 km, over 1022 orders of magnitude) Scientific council IN2P3- June 26th Accelerators, reactions and instrumentation Detectors: Charged particles: GRIT, ACTAR-TPC g-rays: AGATA, nu-ball, PARIS Facilities: spectrometer GANIL, GSI, Dubna, RIKEN, Licorne Neutron wall Stable / radioactive beams EXPAND 5 - 500 MeV/A Reactions: Fusion, transfer, Fission, knockout... A wide variety of phenemona to understand Magic nuclei Fission, pear shapes GDR Soft GMR n p Implantation Residual nuclei b, a-decay, isomeric decay Exotic decays 2n halo molecular cluster rotation- deformation Scientific council IN2P3- June 26th Nuclear physics impact many astrophysical processes X-ray bursts normal star H surface Neutron star Neutron star merger Price & Rossworg ) r s r s Abundance ( solar Sneden & Cowan 2003 Log Mass Number H. Schatz (2016) Nuclear astrophysics Energy profile of X-ray burst and nuclear physics X-ray bursts X-ray burst normal star Neutron star 10s (p,g) (a,p) b- 3a Departure from Hot CNO cycle depends crucially on 15O(a,g)19Ne reaction It is followed by (a,p), (p,g) reactions and b-decays Determination of the 15O(a,g)19Ne reaction rate Cyburt et al.
    [Show full text]
  • Chicago Pile-1 - Wikipedia, the Free Encyclopedia
    Chicago Pile-1 - Wikipedia, the free encyclopedia https://en.m.wikipedia.org/wiki/Chicago_Pile-1#Later_operation Chicago Pile-1 (CP-1) was the Site of the First Self Sustaining Nuclear world's first nuclear reactor to Reaction achieve criticality. Its construction U.S. National Register of Historic Places was part of the Manhattan U.S. National Historic Landmark Project, the Allied effort to create Chicago Landmark atomic bombs during World War II. It was built by the Manhattan Project's Metallurgical Laboratory at the University of Chicago, under the west viewing stands of the original Stagg Field. The first man-made self-sustaining nuclear chain reaction was initiated in CP-1 on 2 December Drawing of the reactor 1942, under the supervision of Enrico Fermi, who described the apparatus as "a crude pile of black bricks and wooden timbers".[4] The reactor was assembled in November 1942, by a team that included Fermi, Leo Szilard, discoverer of the chain reaction, Location Chicago, Cook County, and Herbert L. Anderson, Walter Illinois, USA Zinn, Martin D. Whitaker, and Coordinates 41°47′32″N 87°36′3″W George Weil. It contained 45,000 Built 1942[2] graphite blocks weighing 400 NRHP Reference # 66000314 [1] short tons (360 t) used as a neutron moderator, and was Significant dates fueled by 6 short tons (5.4 t) of Added to NRHP 15 October 1966 [1] uranium metal and 50 short tons (66000314) (45 t) of uranium oxide. In the Designated NHL 18 February 1965[2] pile, some of the free neutrons Designated CL 27 October 1971[3] produced by the natural decay of uranium were absorbed by other uranium atoms, causing nuclear fission of those atoms, and the release of additional free neutrons.
    [Show full text]