ASTM E2006

1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E 706-IIE1) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactory fluences with a higher degree of confidence.

1.2 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 limitations prior to use.

2. Referenced Documents

E1018 Guide for Application of ASTM Evaluated Cross Section Data File, E706 (IIB)
E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Fields, E706(IIE-1)
E261 Practice for Determining Neutron Fluence Rate, Fluence, and Spectra by Radioactivation Techniques
E262 Test Method for Determining Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques
E706 Master Matrix for Light Water Reactor Pressure Vessel Surveillance Standards, E706 (O)
E844 Guide for Sensor Set Design and Irritation for Reactor Surveillance
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E706 (IIA)

Index Terms

benchmark testing; calculational methods; least-square adjustment; neutron transport calculations; nuclear data; reactor pressure vessel; uncertainty estimates; ICS Number Code 27.120.10

Citing ASTM Standards

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