Study by Positron Annihilation of Neutron Damage in a Pressurized Water Reactor (PWR) Pressure Vessel Steel After a 13-Year Irradiation in the CHOOZ A Reactor Surveillance Program

2009 ◽  
pp. 132-132-7 ◽  
Author(s):  
JC Van Duysen ◽  
J Bourgoin ◽  
P Moser ◽  
C Janot
Keyword(s):  
Positron Annihilation ◽  
Pressure Vessel ◽  
Surveillance Program ◽  
Pressure Vessel Steel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Vessel Steel ◽  
Water Reactor ◽  
Neutron Damage

Neutron Damage in Reactor Pressure-Vessel Steel Examined with Positron Annihilation Lifetime Spectroscopy

1996 ◽  
Vol 439 ◽  
Author(s):  
Stephen E. Cumblidge ◽  
Arthur T. Motta ◽  
Gary L. Catchen
Keyword(s):  
Positron Annihilation ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Positron Annihilation Lifetime Spectroscopy ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Positron Annihilation Lifetime ◽  
Neutron Damage ◽  
Positron Lifetimes ◽  
Reactor Pressure

AbstractWe have used positron annihilation lifetime spectroscopy to study the development of damage and annealing behavior of neutron-irradiated reactor pressure-vessel steels. We irradiated samples of ASTM A508 nuclear reactor pressure-vessel steel to fast neutron 172 fluences of up to 1017 n/cm2, and we examined these samples using positron annihilation lifetime spectroscopy (PALS) to study the effects of neutron damage in the steels on positron lifetimes. Non-irradiated samples show two positron lifetimes: a 110 ps component corresponding to annihilations in the bulk material, and a 165 ps lifetime corresponding to annihilations in dislocation defects. The irradiated samples show an additional lifetime component of 300 ps in the PAL spectra and an increase in the proportion of annihilations with a 165 ps lifetime, suggesting that vacancies and vacancy clusters are present in the material after room temperature irradiation. The samples were then annealed to temperatures ranging from 210° C to 450° C. The positron lifetimes introduced by neutron damage disappear after annealing the samples at 280° C.

Examination of Neutron-Irradiated Pressure-Vessel Steel Using Positron Annihilation Lifetime Spectrosopy

1998 ◽  
Vol 540 ◽  
Author(s):  
Stephen E. Cumblidge ◽  
Arthur T. Motta ◽  
Gary L. Catchen
Keyword(s):  
Positron Annihilation ◽  
Pressure Vessel ◽  
Positron Lifetime ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Positron Annihilation Lifetime ◽  
Annealing Temperatures ◽  
Neutron Damage ◽  
Reactor Temperature ◽  
Positron Lifetimes

AbstractOn a variety of pressure-vessel (PV) steels, we have observed changes in the average positron lifetime with increasing (near end-of-life) neutron fluences. Samples were irradiated at reactor-temperature and subjected to post-irradiation annealing, and they were examined using positron annihilation lifetime spectroscopy (PALS). The measured average positron lifetimes in high-temperature (2900 C-300° C) irradiated PV steels decrease with increasing neutron damage up to fluences of 8.5×1018 cm−2 and increase again at higher fluences. Annealing of high-fluence, 300° C irradiated ASTM A508 PV steel samples produces an initial decrease in average positron lifetimes with increasing annealing temperatures of up to 400° C, followed by an increase in average positron lifetime with higher annealing temperatures, when samples were annealed in successive 24-hour steps. A sample of weld steel, irradiated to 2.2×1019 cm−2 at 290° C, shows similar behavior in which the minimum lifetime occurs at ≈ 450° C. These trends are similar to those seen in previous studies performed on VVER and other ferritic steels.

Assessment of U.S. Embrittlement Trend Equations Considering the Latest Available Surveillance Data

2004 ◽  
Author(s):  
William L. Server ◽  
Randy G. Lott ◽  
Stan T. Rosinski
Keyword(s):  
Statistical Test ◽  
Surveillance Data ◽  
Surveillance Program ◽  
Correlation Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Correlation Models ◽  
Pressurized Thermal Shock ◽  
Surveillance Programs

The mechanistically-guided embrittlement correlation model adopted in ASTM E 900-02 was based on a database of U.S. surveillance results current through calendar year 1998. There exists now an extensive amount of new surveillance data that includes a large amount of boiling water reactor (BWR) results from an integrated, supplemental surveillance program designed to augment the plant-specific BWR surveillance programs. These recent data allow a statistical test of the ASTM E 900-02 embrittlement correlation, as well as the NRC correlation model currently being used in the pressurized thermal shock (PTS) re-evaluation effort and the older Regulatory Guide 1.99, Revision 2 correlation. Even though the ASTM E 900-02 embrittlement correlation is a simplified version of the NRC model, a comparison of the two embrittlement correlation models utilizing the new database has proven to be revealing. Based on the new BWR data, both models are inadequate in their ability to predict BWR results; this inadequacy has even more significance for extrapolation outside of the database for BWR heat-up and cool-down curves, as well as some pressurized water reactor (PWR) heat-up curves. Other aspects of the two models, as revealed from this preliminary look at the new data, are presented.

Assessment of Pressure Vessel Steel Irradiation Embrittlement Up to 40 Years Using Local Approach to Fracture Modelling: Application to the French Surveillance Program

2011 ◽  
Author(s):  
B. Tanguy ◽  
A. Parrot ◽  
F. Cle´mendot ◽  
G. Chas
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Chemical Elements ◽  
Experimental Results ◽  
Physical Models ◽  
Surveillance Program ◽  
Pressure Vessel Steel ◽  
Irradiation Embrittlement ◽  
Vessel Steel ◽  
Different Levels

For western pressure vessel reactors, assessment of pressure vessel steels irradiation embrittlement due to neutron irradiation is based on a semi-empirical formulae which predicts the shift of a reference lower bound fracture toughness curve as a function of fluence and embrittlement-involved chemical elements. Periodically, in order to monitor the embrittlement of each RPV, the predictions of the formulae is confronted to experimental results obtained from Charpy specimens located in surveillance capsules irradiated with a higher fluence level than the pressure vessel itself. Historically only the shift of the temperature index defined for a given level of energy, e.g. 56J in the French surveillance program, is used. In support to the French surveillance program methodology, for some of the French RPVs, physical models of fracture (for both cleavage and ductile fracture) are used to analyse in details the whole experimental basis available at different levels of fluence. This study presents the methodology developed in order to analyse the experimental results of a RPV steel from the french surveillance program, including Charpy and fracture toughness tests at different levels of fluence i.e. of embrittlement. The methodology applied aims to use the numerous Charpy tests results available in order to assess, at the same fluence levels, the fracture toughness embrittlement. The results are then compared to available fracture toughness results for a given level of embrittlement.

Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant

2012 ◽  
Vol 9 (4) ◽  
pp. 104016 ◽  
Author(s):  
D. A. Thornton ◽  
D. A. Allen ◽  
A. P. Huggon ◽  
D. J. Picton ◽  
A. T. Robinson ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

Fracture Toughness of a Highly Irradiated Pressure Vessel Steel in Warm Pre-Stress Loading Conditions (WPS)

2011 ◽  
Author(s):  
Guillaume Chas ◽  
Eric Molinie´ ◽  
Eric Garbay ◽  
Francois Cle´mendot ◽  
Dominique Moinereau ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Load Level ◽  
Experimental Program ◽  
Favourable Effect ◽  
Surveillance Program ◽  
Pressure Vessel Steel ◽  
Loading Conditions ◽  
Load History ◽  
Vessel Steel

The warm pre-stress (WPS) of a flawed structure occurs when it is pre-loaded at high temperature in the ductile domain then cooled and loaded up to fracture in the brittle to ductile transition temperature domain. This load history is a feature of RPV accidental transients of LOCA type. Numerous tests on non irradiated specimens and structures have shown the favourable effect of WPS on fracture behaviour. Theorical knowledge let expect that the WPS effect occurs by the same way on irradiated material, but experimental approach had to be completed in such conditions. The experimental program presented in the present article consists in fracture toughness tests under WPS loading conditions performed on two RPV steels irradiated up to a fluence of 6,5.1019 n/cm2. The CT12.5 specimens used for these tests had been irradiated in the capsules of the pressure vessel surveillance program of two french reactors. Different types of WPS load history have been applied to cover typical accidental transients. All the results obtained confirmed for an irradiated steel the two assumptions generally made about the WPS effect: no fracture occurred during the cooling step of the loading even at high load level and the mean fracture toughness value is higher than that measured with conventional mono-temperature tests.

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