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    STP1490

    Reactor Dosimetry: 12th International Symposium

    Vehar David, Gilliam David, Adams James
    Published: 2007


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    The latest edition of this popular ASTM series provides an extensive overview of the latest advances in reactor dosimetry.

    As operating nuclear power reactors have aged and continue to operate on extended operating licenses, new pressure vessel surveillance techniques have been required. Eastern European pressurized water reactors, especially those of the VVER-440 type, continue to have greater concerns about steel embrittlement, because of higher neutron radiation exposures than most Western European and US reactors. Accordingly, broader dosimetry studies are being made on the VVER reactors through retrospective dosimetry, ex-vessel dosimetry, and careful re-analysis of previously reported data.

    60 peer-reviewed papers cover:

    • Power Reactor Surveillance

    • Test Reactors, Accelerators, and Advanced Systems

    • Benchmarks and Intercomparisons

    • Cross Sections, Nuclear Data; Damage Correlations

    • Transport Calculations

    • Adjustment Methods

    • Experimental Techniques


    Table of Contents

    Feasibility Study on a Simple Method of Retrospective Neutron Dosimetry for Reactor Internals and Reactor Vessel

    Reactor Dosimetry Issues During Justification of Extension of Service Life of Nonrestorable Equipment of Russian VVER

    Validation of the Neutron Fluence Calculation on the VVER-440 RPV support structure

    Reactor Dosimetry Study of the Rhode Island Nuclear Science Center

    Comparison of the Results of the Calculational and Experimental VVER-440 Pressure Vessel Dosimetry at Paks NPP

    Characterizing the Time- and Energy-Dependent Reactor n/γ Environment

    Determination of Adjusted Neutron Spectra in Different MUSE Configurations by Unfolding Techniques

    Characterization of Neutron Fields Using MCNP in the Experimental Fast Reactor JOYO

    Benchmark on the 3-D VENUS-2 MOX-Fueled Reactor Dosimetry Calculations by DANTSYS Code System

    Experimental and Calculation Investigations of the Space-Energy Neutron and Photon Distribution in the Vicinity of Reactor Pressure Vessel and Surveillance Specimen Box of New Type in the WWER-1000 Mock-Up

    Characterization of the Neutron Field in the HSSI Reusable Irradiation Facility at the Ford Nuclear Reactor

    Benchmark Experiments/Calculations of Neutron Environments in the Annular Core Research Reactor

    Measurement of Helium Generation in AISI 304 Reflector and Blanket Assemblies after Long-term Irradiation in EBR-II

    Verification of MultiTrans Calculations by the VENUS-3 Benchmark Experiment

    The Neutron Spectrum of NBS-1

    Radiation Dosimetry in the BNCT Patient Treatment Room at the Brookhaven Medical Research Reactor

    TRIM Modeling of Displacement Damage in SiC for Monoenergetic Neutrons

    Thermal and Epithermal fluence Rate Measurements in Multipurpose Reactors: Application of a Least-Squares Fitting Code RESDET to Obtain Thermal and Epithermal Fluence Rates from Measured Reaction Rates

    Neutron and Photon Dosimetry at the LR-0 Reactor Using Paired Detectors

    Survey of the Latest Evaluated Nuclear Data

    Gas Production in Reactor Materials

    Dosimetry Requirements for Pressure Vessel Steels Toughness Curve in the Ductile to Brittle Range

    Mesh Tally Radiation Damage Calculations and Application to the SNS Target System

    Attenuation of Radiation Damage and Neutron Field in Reactor Pressure Vessel Wall

    An International Evaluation of the Neutron Cross Section Standards

    Precision Neutron Total Cross-Sectional Measurements for Natural Carbon at Reactor Neutron-Filtered Beams

    Spallation Radiation Damage Calculations and Database: Cross-Section Discrepancies between the Codes

    Proton Induced Activation in Mercury: Comparison of Measurements and Calculations

    Measurements and Monte Carlo Calculations of Gamma and Neutron Flux Spectra Inside and Behind Iron/Steel/Water Configurations

    Advances in Calculation of Fluence to Reactor Structures

    Deterministic and Monte Carlo Neutron Transport Calculation for Greifswald-1 and Comparison with Ex-vessel Measured Data

    Coarse-Mesh Adjoint Biasing of a Monte Carlo Dose Calculation

    Analysis of the VENUS-3 PWR Pressure Vessel Surveillance Benchmark Using TRIPOLI-4.3 Monte Carlo Code

    Influence of the Multigroup Approximation on VVER-1000 RPV Neutron/Gamma Flux Calculation

    A New Derivation of the Perturbation Operator Used in MCNP

    Calculation of Neutron Fluxes for Radioactive Inventory Assessment of Magnox Power Plant

    Extensive Revision of the Kernel-Based PREVIEW Program and Its Input Data

    Investigation of Radiation Transport Modeling Trends in the WSMR Fast Burst Reactor Environments

    Coupled Neutron-Gamma Calculations for the LR-0 Experimental Benchmark

    Use of CPXSD for Generation of Effective Fast Multigroup Libraries for Pressure Vessel Fluence Calculations

    Benchmarking of PENTRAN-SSN Parallel Transport Code and FAST Preconditioning Algorithm Using the VENUS-2 MOX-Fueled Benchmark Problem

    Sensitivity Analysis and Neutron Fluence Adjustment for VVER-440 RPV

    Generalized Linear Least-Squares Adjustment, Revisited

    Retrospective Measurement of Neutron Activation within the Pressure Circuit Steelwork of a Magnox Reactor and Comparison with Prediction

    Comparison of Predicted and Measured Helium Production in U.S. BWR Reactors

    Shielding Calculations for the Upgrade of the HFIR HB 2 Beam Line

    An Approach to Determining the Uncertainty in Reactor Test Objects Using Deterministic and Monte Carlo Methods

    Development of a Silicon Calorimeter for Dosimetry Applications in a Water-Moderated Reactor

    Measurement and Calculation of WWER-440 Pressure Vessel Templates Activity for Support of Vessel Dosimetry

    Fast Neutron Dosimetry and Spectrometry Using Silicon Carbide Semiconductor Detectors

    Retrospective Dosimetry of Fast Neutrons Focused on the Reactions 93Nb(n,n')93Nbm and 54Fe(n,p)54Mn

    Spent Fuel Monitoring with Silicon Carbide Semiconductor Neutron/Gamma Detectors

    Digital Multiparameter System for Characterizing the Neutron-Gamma Field in the LR-O Experimental Reactor

    A Beam-Monitor System for Neutrons and Gamma Rays in the Medical Irradiation Facility of the Kyoto University Research Reactor

    Application of a Silicon Calorimeter in Fast Burst Reactor Environments

    The Validity of the Use of Equivalent DIDO Nickel Dose for Graphite Dosimetry

    Neutron Damage in SiC Semiconductor Radiation Detectors in the GT-MHR

    Evaluation of Diamond Detectors for Fast Neutron Fluence Measurements in WWER-1000 Surveillance Assemblies

    A New Methodology for Adjustment of Iron Scattering Cross Sections Using Time-of-Flight Spectroscopy

    Reactor Dosimetry with Niobium

    Author Index

    Subject Index


    Committee: E10

    DOI: 10.1520/STP1490-EB

    ISBN-EB: 978-0-8031-6242-6

    ISBN-13: 978-0-8031-3412-6

    ASTM International is a member of CrossRef.

    0-8031-3412-6
    978-0-8031-3412-6
    STP1490-EB