Accumulative test data indicate that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of primary pressure boundary components materials to be significantly reduced. Environmentally assisted fatigue (EAF) is the abbreviation of the environmentally assisted fatigue. In 2007, Nuclear Regulatory Commission (NRC) issued RG. 1.207. It was updated in 2014. And, it requires that the effects of LWR environment on the fatigue life reduction of metal components should be considered for new design plants. And it suggests to use environmental correction factor, Fen, to account for EAF. NRC regulation (NUREG), NUREG/CR-6909 (NRC, 2013, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,” U.S. Nuclear Regulatory Commission, Argonne, IL, Standard no. NUREG/CR-6909), presents the detail Fen calculation formula. Fen is a function of temperature, strain rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depending on the experience of the primary pressure boundary piping transient operation, Fen varies during each transient. More uncertainty and confusion are raised during the application of the Fen method. The research work in this paper includes: first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depending on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multitransient loading conditions. And the results are compared with those of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test plan is presented.

References

References
1.
NRC
,
2013
, “
Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials
,” U.S. Nuclear Regulatory Commission, Argonne, IL, Standard no.
NUREG/CR-6909
.https://www.nrc.gov/docs/ML0706/ML070660620.pdf
2.
NRC
,
1995
, “
Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components
,” U.S. Nuclear Regulatory Commission, Washington, DC, Standard no.
NUREG/CR-6260
.https://inis.iaea.org/search/search.aspx?orig_q=RN:26063712
3.
NRC
,
2001
, “
Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants
,” U.S. Nuclear Regulatory Commission, Washington, DC, Standard No.
NUREG-1800
.https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1800/r2/index.html
4.
ASME
,
2013
, “
ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components
,” American Society of Mechanical Engineers, New York, Standard No.
BPVC-III-5-2017
.https://www.asme.org/products/codes-standards/bpvciii5-2017-bpvc-section-iiirules-construction
5.
ASTM
,
2004
, “
Standard Practice for Strain-Controlled Fatigue Testing
,” American Society for Testing and Materials, West Conshohocken, PA, Standard No.
ASTM E606-04
.https://www.astm.org/Standards/E606.htm
6.
Xinqiang
,
W. U.
,
Tan
,
J.
,
Song
,
X. U.
,
Han
,
E. H.
, and
Wei
,
K. E.
,
2015
, “
Corrosion Fatigue Mechanism of Nuclear-Grade Low Alloy Steel in High Temperature Pressurized Water and Its Environmental Fatigue Design Model
,”
Acta Metall. Sin.
,
51
(
3
), pp.
298
306
.
7.
IAEA
,
2003
, “
Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety Primary Piping in PWRs
,” International Atomic Energy Agency, Vienna, Austria, Standard No.
IAEA-TECDOC-1361
You do not currently have access to this content.