Change search
ReferencesLink to record
Permanent link

Direct link
Long term irradiation effects on the mechanical properties of reactor pressure vessel steels from two commercial PWR plants
KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).ORCID iD: 0000-0003-1498-5691
Ringhals AB. (RTQM)
Epsilon AB.
2013 (English)In: ASTM Special Technical Publication, ASTM International, 2013, Vol. 1547, 52-68 p.Conference paper (Refereed)
Abstract [en]

The Swedish nuclear power plants all have plant specific surveillance programs which includes samples from all relevant materials that are subjected to a fluence-level that exceeds 1*1017 n/cm2 over the estimated period of operation for the specific power plants. The Swedish pressurized water reactor (PWR)-plants are currently planning for a service period beyond 50 years of operation. As a portion of that, two of the three PWR units at the Ringhals site are conducting a major effort to verify the fitness to service of the reactor pressure vessel (RPV). In this case it is the weld in the belt-line region of the RPV, which is the apparent limiting factor. The weld metal contains high Nickel and high Manganese levels, not commonly used in other PWR-reactors. The effort includes a densified testing of the available surveillance capsule material in order to better understand the degradation phenomena and also an extended testing scope. A spin off effect of this program is that high fluence data for the base material also is made available from the testing. The chemical composition of the base metal is valid for many of the currently operating PWR-vessels. This study is an analysis of both the weld and the base material data extracted from the surveillance program. The results are evaluated against currently available data and correlation curves. In general, the results point out that the current Regulatory Guide 1.99 revision 2-correlation regarding the prediction of as-irradiated transition temperature is under-conservative for the tested material. The transition temperature shift, here evaluated as the temperature shift at 41J, is under-predicted by the correlation by as much as 70°C in some cases and increases with increasing fluences. However, prediction made by the French average irradiation embrittlement prediction formula, FIM-formula, is consistently better but still slightly under conservative.

Place, publisher, year, edition, pages
ASTM International, 2013. Vol. 1547, 52-68 p.
Keyword [en]
Reactor pressure vessel, long term irradiation, irradiation effects, mechanical properties, modell analysis
National Category
Metallurgy and Metallic Materials
URN: urn:nbn:se:kth:diva-78711DOI: 10.1520/STP104004ScopusID: 2-s2.0-84875799317OAI: diva2:492794
25th Symposium on the Effects of Radiation on Nuclear Materials; Anaheim, CA; United States2011

QC 20131200

Available from: 2012-02-08 Created: 2012-02-08 Last updated: 2013-12-11Bibliographically approved

Open Access in DiVA

No full text

Other links

Publisher's full textScopus

Search in DiVA

By author/editor
Efsing, Pål
By organisation
Solid Mechanics (Dept.)
Metallurgy and Metallic Materials

Search outside of DiVA

GoogleGoogle Scholar
The number of downloads is the sum of all downloads of full texts. It may include eg previous versions that are now no longer available

Altmetric score

Total: 60 hits
ReferencesLink to record
Permanent link

Direct link