Significance and Use
3.1 Prediction of neutron radiation effects to pressure vessel steels has long been a part of the design and operation of light water reactor power plants. Both the federal regulatory agencies (see 2.3) and national standards groups (see 2.1 and 2.2) have promulgated regulations and standards to ensure safe operation of these vessels. The support structures for pressurized water reactor vessels may also be subject to similar neutron radiation effects (1, 2, 3, 4, 5). The objective of this practice is to provide guidelines for determining the neutron radiation exposures experienced by individual vessel supports.
3.2 It is known that high energy photons can also produce displacement damage effects that may be similar to those produced by neutrons. These effects are known to be much less at the belt line of a light water reactor pressure vessel than those induced by neutrons. The same has not been proven for all locations within vessel support structures. Therefore, it may be prudent to apply coupled neutron-photon transport methods and photon induced displacement cross sections to determine whether gamma-induced dpa exceeds the screening level of 3.0 × 10-4, used in this practice for neutron exposures. (See 1.2).
1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic materials in nuclear reactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for:
1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities.
1.1.2 Making appropriate neutronics calculations to predict neutron radiation exposures.
1.2 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence (E > 1 MeV) that exceeds 1 × 1017 neutrons/cm 2 or 3.0 × 10−4 dpa. (See Terminology E170.)
1.3 Exposure of vessel support structures by gamma radiation is not included in the scope of this practice, but see the brief discussion of this issue in 3.2.
1.4 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 (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.
E170 Terminology Relating to Radiation Measurements and Dosimetry
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E 706(ID)
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC)
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706 (IIIA)
E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)
ICS Number Code 27.120.20 (Nuclear power plants. Safety)
ASTM International is a member of CrossRef.
Citing ASTM Standards
[Back to Top]