1. Scope

1.1 Need for Neutronics Calculations - An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E 944 and Practice E 853 define appropriate computational procedures.

1.2 Methodology Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: ( 1 ) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, ( 2 ) determination of the neutron source distribution in the reactor core, and ( 3 ) calculation of neutron fluence rate at the surveillance position and in the pressure vessel.

This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory requirements prior to use.

2. Referenced Documents (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.

ASTM Standards

E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E 706(ID)

E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E 706(0)

E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC)

E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706(IA)

E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)

E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)

E2006 Guide for Benchmark Testing of Light Water Reactor Calculations

Nuclear Regulatory Documents

Keywords

discrete ordinates; dosimetry; exposure parameter; Monte Carlo; neutron fluence; pressure vessel; radiation transport;

ICS Code

ICS Number Code 27.120.20 (Nuclear power plants. Safety)

DOI: 10.1520/E0482-07

ASTM International is a member of CrossRef.