Monte Carlo N-Particle Simulations for Nuclear Detection and Safeguards: An Examples-Based Guide for Students and Practitioners

Description

This  open access book is a pedagogical, examples-based guide to using the Monte Carlo N-Particle (MCNP®) code for nuclear safeguards and non-proliferation applications. The MCNP code, general-purpose software for particle transport simulations, is widely used in the field of nuclear safeguards and non-proliferation for numerous applications including detector design and calibration, and the study of scenarios such as measurement of fresh and spent fuel. This book fills a gap in the existing MCNP software literature by teaching MCNP software usage through detailed examples that were selected based on both student feedback and the real-world experience of the nuclear safeguards group at Los Alamos National Laboratory. MCNP input and output files are explained, and the technical details used in MCNP input file preparation are linked to the MCNP code manual. Benefiting from the authors’ decades of experience in MCNP simulation, this book is essential reading for students, academic researchers, and practitioners whose work in nuclear physics or nuclear engineering is related to non-proliferation or nuclear safeguards.  Each chapter comes with downloadable input files for the user to easily reproduce the examples in the text. Preface Acknowledgments About This Book Contents About the Authors Chapter 1: Introduction 1.1 Nuclear Safeguards 1.2 Monte Carlo N-Particle Transport: MCNP Code 1.3 Objective References Chapter 2: Basic Concepts 2.1 Geometry 2.1.1 Simplest Possible Input File 2.1.2 Running MCNP: The Simplest Case 2.1.3 Simple Input File 2.1.4 Running and Plotting MCNP Geometries 2.1.5 Surfaces and Complicated Cells: Intersections and Unions 2.1.6 Duplicate Cells, Compliments, and Translations: LIKE n BUT and TRCL 2.1.7 Filled Cells: Universes 2.1.8 Lattice Geometries 2.1.9 Fully Specified Lattice Geometries 2.2 Materials and Cross Sections 2.2.1 Specifying Materials 2.2.2 Neutron Cross Sections 2.2.3 Low-Energy Neutron Problems: Thermal Free Gas Treatment 2.2.4 Low-Energy Neutron Problem Data: S(α,β) Thermal Treatment 2.2.5 Photon Cross Sections 2.2.6 Electron-Stopping Powers for Coupled Photon and Electron Problems 2.2.7 Data and Models for Ions and Charged Particles 2.2.8 Additional Data Diagnostics and Recommendations 2.3 Sources 2.3.1 SDEF Fixed Sources 2.3.2 SDEF Source Distributions 2.3.3 SDEF Dependent Distributions: DS 2.3.4 Criticality Sources 2.3.5 Surface Source Write and Read (SSW, SSR) 2.3.6 Checking sources 2.4 Output and Tallies 2.4.1 Output Files 2.4.2 MCNP Estimators and Tally Types 2.4.3 Basic Tally Format 2.4.3.1 The Eight Tally Dimensions: FDUSMCET 2.4.3.2 F: Surface and Cell Tallies 2.4.3.3 F: Tallies in Lattices and Repeated Structures 2.4.3.4 C: Angle Bins and FC Tally Comments 2.4.3.5 E, T: Energy and Time Bins, Default Bins, and FQ and TF Format Control 2.4.3.6 M: Multiplier Bins and SD Divisors 2.4.3.7 D: Direct and SF, CF Flagging Bins 2.4.3.8 U: User Bins 2.4.3.9 S: Segment Bins 2.4.4 Special Tally Treatments 2.4.4.1 FRV: Fixed Arbitrary Reference Direction for Tally F1 or F2 Cosine Binning 2.4.4.2 GEB: Gaussian Energy Broadening 2.4.4.3 TMC: Time Convolution 2.4.4.4 INC: Identify the Number of Collisions 2.4.4.5 ICD: Identify the Cell from Which Each F5 Detector Score Is Made 2.4.4.6 SCX: Identify the Sampled Index of a Specified Source Distribution 2.4.4.7 SCD: Identify Which of the Specified Source Distributions Was Used 2.4.4.8 ELC: Electron Current Tally 2.4.4.9 PTT: Put Different Multigroup Particle Types in Different User Bins 2.4.4.10 RES: Residual Nuclei 2.4.4.11 TAG: Tally Tagging 2.4.4.12 LET: Tally Stopping Powers Instead of Energy 2.4.4.13 ROC: Receiver-Operator Characterization 2.4.4.14 PDS: Point Detector Sampling 2.4.4.15 FFT: First Fission Tally 2.4.4.16 COM: Compton Image Tally 2.4.4.17 CAP: Coincidence Capture 2.4.4.18 PHL: Pulse-Height Light Tally 2.4.5 Pulse-Height Tallies 2.4.6 Point Detectors and Next-Event Estimators 2.5 Plotting 2.5.1 Geometry Plotting and Command Files 2.5.2 Cross-Section Plotting 2.5.3 Tally Plotting 2.5.4 Mesh, Radiography, and Ring Tallies 2.6 Statistics and Convergence References Chapter 3: Examples for Nuclear Safeguards Applications 3.1 Fuel Assembly in Water Tank 3.1.1 Problem Description 3.1.2 Geometry Description 3.1.3 Other Data: Sources, Materials, Tallies, and More 3.1.4 MCNP Output 3.2 Coincidence Counter with F4 and F8 Tallies for Coincidence and Multiplicity Counting Rates 3.2.1 Description 3.2.2 Materials 3.2.3 Source 3.2.4 Tallies 3.2.5 Warning Messages 3.2.6 Results 3.2.6.1 From the Point Model 3.2.6.2 Rates Calculated without Point Model Assumptions 3.3 Gamma Pulse Height 3.3.1 Description and Input File 3.3.2 Geometry 3.3.3 Materials 3.3.4 Methods 3.3.5 Results 3.4 Active Neutron Example: Californium (Cf) Shuffler 3.4.1 Description and Input File 3.4.2 Results References Chapter 4: Examples of Advanced Concepts 4.1 Variance Reduction 4.1.1 Introduction 4.1.2 Multigroup Weight Windows and Time Splitting: Lead Slowing-Down Spectrometer 4.1.2.1 Input File Notes 4.1.2.2 Variance Reduction Step 1: Simplify the Problem and Add the Weight Window Generator 4.1.2.3 Iteration 2 4.1.2.4 Iteration 2a 4.1.2.5 Additional Iterations 4.1.2.6 Cylindrical Mesh Weight Window Summary 4.1.3 Cell-Based Weight Windows for the Lead Slowing-Down Spectrometer 4.1.4 Time Splitting 4.1.5 Variance Reduction for the Cf Shuffler 4.1.5.1 Cf Shuffler Modified Input 4.1.5.2 Particle Production Bias, Time Splitting, and Windows 4.1.5.3 Analysis of Cf Shuffler Variance Reduction 4.2 DXTRAN and Other Capabilities for Distributed Source Problems 4.2.1 UF6 Cask Model 4.2.2 DXTRAN 4.2.3 Source Position Biasing 4.2.4 Best Single Detector Solution 4.3 Neutron Detector Operation in More Detail 4.3.1 Introduction 4.3.1.1 Make Reaction Products (Model, Data) and Recoil Nuclei 4.3.1.2 Track Created Particles in Real Gas Composition 4.3.1.3 Tally Energy Deposition of Particles in Active Volume (F8 CAP EDEP for Coincidence/Multiplicity Counting) 4.4 Examples 4.4.1 3He Detector Pulse Height 4.4.2 3He Detector Coincidence Calculation 4.4.3 10B-Lined Detectors References Chapter 5: Additional Topics 5.1 Troubleshooting or “How Can I Be Confident in the Results?” 5.1.1 Geometry and Materials 5.1.2 Detector Modeling 5.1.3 Source Modeling 5.1.4 Sample Modeling 5.1.5 Tracking Limitations 5.1.6 Nuclear Data 5.1.7 Statistics 5.1.8 User 5.1.9 Checking Your Results 5.2 Sampling Collision Progeny 5.2.1 Analysis of Delayed Neutron Production 5.2.2 Comparison of Table Physics vs. Model Physics References Appendix: How to Get MCNP Software Index