ORNL-6763 Advanced Neutron Source (ANS) Project Progress Report FY 1993 Principal Authors D. L. Selby R.M. Harrington Oak Ridge National Laboratory P. B. Thompson Engineering Division Martin Marietta Energy Systems, Inc. Editor J. H. Campbell Oak Ridge National Laborato_ Electronic Publishers L. W. Davis S. C. Lyttle Oak Ridge National Laboratory Date Published--January 1994 Prepared by the Oak Ridge National Laboratory Oak Ridge, Tennessee 37831-6285 managed by Martin Marietta Energy Systems, Inc. U.S. DEPARTMENT OF ENERGY under Contract No.forDE-ACO5-84OR21400the _j_STE _ otS"rmeOTJI1F,T.eO__P.Ol:P._.tS.r,'ru_IJ_P Contributing Authors D. J. Alexander A.W. Longest W.K. Sartory J. L. Anderson A.T. Lucas H.B. Shapira R. L. Battiste C.R. Luttrell M. Siman Toy R. E. Battle M.F. Marchbanks H.G. Smith N. C. J. Chen J.A. March-Leuba T.K. Stovall (3. L. Copeland J.T. Mihalczo W.F. Swinson W. W. Engle H.A. Mook R.P. Taleyarkhan D. K. Felde R.M. Moon M.W. Wendel J. C. Gehin D.G. Morris C.D. West M. L. Gildner R.E. Pawel B.A. Worley W. A. Hamilton C.T. Ramsey G.T. Yahr W. R. Hendrich J.P. Renier G.L. Yoder, Jr. R. A. Lillie Oak Ridge National Laboratory R. R. Allen J.R. DeVore F.J. Peretz R. S. Booth C.C. Eberle C.C. Queen W. W. Bowman T.J. McManamy J.P. Schubert R. A. Brown G.R. McNutt B.R. Smith P. F. Cento T.F. Orlin D.W. Theisen K. K. Chipley H.R. Payne D.M. Williams J. E. Cleaves Engineering Division, Martin Marietta Energy Systems, hw. R. C. Birtcher J.L. Snelgrove G. L. Hofman T.C. Wiencek J. Rest Argonne National Lzlboratorv C. D. Fletcher J.M. Ryskamp L. S. Ghan C.A. Wemple A. M. Ougouag Idaho National Engineering Laboratot3, H. N. Goldstein P.J. Shipper W. E. Meek P.T. Talarico GilbertCommonwealth, Inc. L. Crow C.R. Bass A. Steyerl H.G. Wood, III University of Rhode Island University of Virginia G. K. Carlough M. Ibn-Khayat DRS/Hundley Kling Gmitter Oak Ridge Associated Universities C. F. Majkrzak M. Kaminaga National Institute of Standards and Technology Japan Atomic Energy Research Institute CONTENTS Figures .............................................................................................................. vii Tables ................................................................................................................ xi Acronyms .......................................................................................................... xiii Foreword ........................................................................................................... xv 1. Project Management ................................................................................. I 1.1 ANS Project Quality Program ............................................................................... 2 2. Research and Development ....................................................................... 3 2.1 Reactor Core Development .................................................................................. 3 2.1.1 Development of New Fuel Grading .......................................................... 3 2.1.2 Analysis of the FOEHN Criticals .............................................................. 4 2.1,3 Treatment of Uncertainties ........................................................................ 5 2.1.4 Impact of Photoneutron Production on Neutron Spectrum in the Reflector .......................................................................... 6 2.2 Fuel Development ................................................................................................ 6 2.2.1 Irradiation of Fuel Capsules ...................................................................... 7 2.2.2 Fuel Performance Studies and Modeling .................................................. 7 2.2.3 Fabrication Development .......................................................................... 9 2.3 Corrosion Loop Tests and Analyses ..................................................................... 9 2.4 Thermal-Hydraulic Loop Tests ............................................................................ i() 2.4.1 Test Loop Facility ..................................................................................... !() 2.4.2 Data Reduction, Results, and Analysis ..................................................... 12 2.4.3 Future Plans .............................................................................................. 16 2.5 Reactor Control and Shutdown Concepts ............................................................ 17 2.6 Critical and Subcritical Experiments .................................................................... 17 iii 2.7 Material Data, Structural Tests, and Analyses .................................................... 18 2.7.1 Code Case for 606 l-T6 Aluminum ......................................................... 18 2.7.2 Fracture Toughness of Irradiated 6061-T6 Aluminum ........................... 18 2.7.3 Core Pressure Boundary Tube Fracture Evaluation ................................ 19 2.7.4 Advanced Neutron Source Materials Information System ..................... 20 2.7.5 Fuel Plate Stability Tests and Analysis ................................................... 20 2.7.5.1 Experimental Studies of the Structural Response of the Fuel Plates to Coolant Flow ........................................... 20 2.7.5.2 Fuel Plate Thermal Deflection ................................................. 21 2.7.6 Inner Control Element Structural Stability .............................................. 21 2.7.6.1 Flow-Induced Vibrations .......................................................... 22 2.7.6.2 Scram Performance Studies ..................................................... 22 2.7.6.3 Control Element Test Facility .................................................. 22 2.8 Cold Source Development .................................................................................. 23 2.8.1 Design Principles of the Moderator Unit ................................................ 25 2.8.2 Cold Source Neutronics .......................................................................... 26 2.8.3 Liquid Deuterium Loop Design .............................................................. 26 2.8.4 Cold Source Review ................................................................................ 26 2.8.5 General Design Parameters ..................................................................... 27 2.9 Beam Tube, Guide, and Instrument Development .............................................. 27 2.9.1 Polarizer Development ............................................................................ 27 2.9.2 Neutron and Gamma Filters for Straight Neutron Guides ...................... 28 2.10 Hot Source Development .................................................................................... 29 2.11 Neutron Transport and Shielding ........................................................................ 29 2.11.1 Cold Neutron Guide Tube Tunnel Shielding .......................................... 29 2.11.2 Shielding Cross-Section Library ............................................................. 30 2.11.3 Flux Characterization in the Heavy Water Vessel .................................. 30 2.11.4 Analysis of Beam Tube Shield Plugs ...................................................... 30 2.12 I&C Research and Development ......................................................................... 31 2.12.1 Neutron Kinetics Modeling ..................................................................... 31 2.I2.2 ANSDM, PRSDYN, RELAP5/ANS Benchmark ................................... 31 2.12.3 Reactor Inlet Temperature Control ......................................................... 31 2.13 Facility Concepts ................................................................................................. 33 2.13.1 Test Plan Development ........................................................................... 33 2.13.2 Inner Control and Shutdown Rod Tests .................................................. 34 2.13.3 Other Tests .............................................................................................. 35 iv Advanced Neutron Source (ANS) Progress Report 3. Design ...................................................................................................................... 37 3.1 Systems Integration .............................................................................................. 37 3.1.1 Configuration Management Philosophy .................................................. 37 3.1.2 Configuration Management Experience .................................................. 37 3.2 Reactor Systems ................................................................................................... 38 3.2.1 Reactor Primary Pressure Boundary Design ............................................ 38 3.2.2 Reflector Vessel ....................................................................................... 40 3.2.3 Control and Shutdown Rod Assembly ..................................................... 41 3.2.4 Refueling .................................................................................................. 44 3.2.5 Reactor Maintenance ................................................................................ 45 3.2.6 Modeling of the Reactor Assembly .......................................................... 45 3.2.7 Thimbles and Light Water Transition Assemblies ................................... 46 3.3 Experiment Systems ............................................................................................