The Fluence Prediction of Reactor Pressure Vessel Materials Due to Neutron Irradiation by Ultrasonic Characteristic Analysis

2002 ◽  
Author(s):  
Sam Lai Lee ◽  
Byoung Chul Kim ◽  
Kee Ok Chang
Keyword(s):  
Weld Metal ◽  
Neutron Irradiation ◽  
Neutron Fluence ◽  
Ultrasonic Attenuation ◽  
Charpy Impact ◽  
Reactor Vessel ◽  
Characteristic Analysis ◽  
Operating Reactor ◽  
Vessel Material ◽  
Reactor Pressure

Ultrasonic attenuation analysis from Charpy impact specimens withdrawn from operating reactor vessel during plant outages have been performed in order to predict the amount of neutron fluence directly in a reactor vessel material by neutron irradiation. The results showed that systematic changes in ultrasonic attenuation versus neutron fluence and operating frequencies were observed for the base and weld metal and a possibility to predict neutron fluence in a specimen which has no information on the amount of neutron fluence using such consistent relationship exists.

PREDICTION OF A NEUTRON FLUENCE BY AN ULTRASONIC MEASUREMENT IN AN IRRADIATED MATERIAL USED FOR SURVEILLANCE TESTING OF AN OPERATING REACTOR

2008 ◽  
Vol 22 (11) ◽  
pp. 1063-1068
Author(s):  
SAMLAI LEE
Keyword(s):  
Impact Test ◽  
Neutron Fluence ◽  
Ultrasonic Attenuation ◽  
Charpy Impact ◽  
Charpy Impact Test ◽  
Reactor Vessel ◽  
Ultrasonic Measurement ◽  
Systematic Relationships ◽  
Operating Reactor ◽  
Vessel Material

The ultrasonic attenuation versus neutron fluence relationship from Charpy impact test specimens withdrawn from an operating reactor vessel during plant outages has been evaluated in order to predict the neutron fluence through a direct attenuation measurement for a reactor vessel material irradiated by neutrons. The results showed that systematic relationships for the ultrasonic attenuation versus neutron fluence at specific operating frequencies were observed for the base and weld metal and the possibility to predict neutron fluence in a specimen, that has no information on the amount of neutron fluence, by using such a consistent relationship.

Complex Monitoring Programme for WWER Reactor Pressure Vessels Lifetime Assessment in the Czech Republic

2008 ◽  
Author(s):  
Milan Brumovsky ◽  
Milos Kytka ◽  
Petr Novosad ◽  
Jiri Brynda
Keyword(s):  
Czech Republic ◽  
Weld Metal ◽  
Heat Affected Zone ◽  
Neutron Fluence ◽  
Pressure Vessels ◽  
Surveillance Program ◽  
The Czech Republic ◽  
Complex Monitoring ◽  
Metal Weld ◽  
Reactor Pressure

Lifetime of reactor pressure vessels practically depends on a level of degradation of RPV material properties during operation. The most important degradating mechanism of RPV materials is usually radiation damage, characterized by values on neutron fluence on one side and radiation embrittlement of RPV materials on the second side. WWER reactor pressure vessels in the Czech Republic are a subject of a very thorough and complex monitoring program, that includes: • Standard material surveillance program containing of WWER-440 RPV materials — base metal, weld metal, heat affected zone, but irradiated with high lead factor (13 to 18), • Supplementary surveillance program of WWER-440 RPV materials, including additionally austenitic cladding materials, IAEA reference material JRQ irradiated with low lead factor (2 to 3) with parts subjected to annealing and re-irradiation after annealing, • Modified surveillance program of WWER-1000 RPV materials — base metal, weld metal, heat affected zone, cladding materials, IAEA reference JRQ material irradiated in low lead factor (2 to 3) near RPV inner beltline region, • Integrated surveillance specimen program for WWER-1000 reactor including materials from NPP Temelin (Czech Republic), Belene (Bulgaria), Kalinin (Russia) and Ukranian NPPs, • Continous exvessel monitoring of neutron fluence on outer RPV surface for both WWER-440 and WWER-1000 plants, • Neutron fluence determination on inner RPV surface (austenitic cladding) using special technique for removal of specimens from cladding for Nb activity measurements, • Ex-vessel temperature measurements during RPV operation. All these programs serve for precision of operation conditions and determination of degradation of RPV materials for RPV integrity and lifetime assessment.

Effect of Neutron Irradiation on the Mechanical Properties of an A508 CL2 and 15Kh2NMFA Irradiated in the NOMAD_3 Rig in the BR2 Cooling Water

2021 ◽  
Author(s):  
Inge Uytdenhouwen ◽  
Rachid Chaouadi
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Cooling Water ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Significant Loss ◽  
Irradiation Temperature ◽  
Irradiation Damage ◽  
Transition Curves ◽  
Reactor Pressure

Abstract The typical operating temperatures of a nuclear reactor pressure vessel in a PWR are between 290°C and 300°C. However, many BWRs and some PWRs operate at slightly lower temperatures down to 260°C. Most of the literature and neutron irradiation damage is therefore focused on those irradiation temperatures. It is well-known that the lower the irradiation temperature, the more neutron irradiation damage occurs, because no appreciable annealing happens below approximately 230°C. The NOMAD_3 irradiation consisted in total of 24 Charpy sized samples from an A508 Cl.2 forging and a 15Kh2NMFA material. They were irradiated to three various fluences between 1.55 and 7.90 × 1019 n/cm2 (E > 1MeV) at approximately 100°C. The hardening of the A508 Cl.2 was between 260 and 400 MPa which was much higher than the NOMAD_0 properties which were irradiated at approximately 280°C. The tensile tests of irradiated materials are all characterized by a significant loss of work hardening capacity leading to plastic flow localization promptly after the yield strength is reached. This affects also the shape of the Charpy impact transition curves. The radiation embrittlement derived from Charpy impact tests, ΔT41J, is up to 156°C for the highest fluence. For this irradiation, the embrittlement to hardening ratio was also around 0.43 +/−0.2°C/MPa as it was found in the previous campaign NOMAD_0. This paper discusses the tensile, hardness and impact properties of the NOMAD_3 irradiation campaign. It is compared to the NOMAD_0 with respect to effect of irradiation temperature and annealing recovery.

An Update on the Investigation of Fracture Toughness Properties of the High Flux Reactor Vessel From Surveillance Test Campaign in 2017

2019 ◽  
Author(s):  
M. Kolluri ◽  
F. H. E. de Haan – de Wilde ◽  
H. S. Nolles ◽  
A. J. M. de Jong
Keyword(s):  
Mechanical Properties ◽  
Fracture Toughness ◽  
Fracture Surface ◽  
Neutron Fluence ◽  
Reactor Vessel ◽  
Surveillance Program ◽  
High Flux ◽  
High Flux Reactor ◽  
Flux Reactor ◽  
Vessel Material

Abstract The reactor vessel of the High Flux Reactor (HFR) in Petten has been fabricated from Al 5154-O alloy grade with a maximum Mg content of 3.5 wt. %. The vessel experiences large amount of neutron fluences (notably at hot spot), of the order of 1027 n/m2, during its operational life. Substantial damage to the material’s microstructure and mechanical properties can occur at these high fluence conditions. To this end, a dedicated surveillance program: SURP (SURveillance Program) is executed to understand, predict and measure the influence of neutron radiation damage on the mechanical properties of the vessel material. In the SURP program, test specimens fabricated from representative HFR vessel material are continuously irradiated in two specially designed experimental rigs. A number of surveillance specimens are periodically extracted and tested to evaluate the changes in fracture toughness properties of the vessel as function neutron fluence. The surveillance testing results of test campaigns performed until 2015 were published previously in [1, 2]. The current paper presents fracture toughness and SEM results from the recent surveillance campaign performed in 2017. The fracture toughness specimen tested in this campaign received a thermal neutron fluence of 13.56 x1026 n/m2, which is ∼8.9 × 1025 n/m2 more than the thermal fluence received by the specimen tested in SURP 2015 campaign. These results from this campaign have shown no change in the fracture toughness from the values measured in the previous SURP campaign. The SEM observations are performed to study the fracture surface, to measure (by WDS) the transmutation Si formed near crack tip and to investigate various inclusions in the microstructure. SEM fracture surface investigation revealed a tortuous (bumpy) fracture surface constituting micro-scale dimples over majority of the fracture area. Islands of cleavage facets and secondary cracks have been observed as well. EDS analysis of various inclusions in the microstructure revealed presence of Fe rich inclusions and Mg-Si rich precipitates. Additionally, inclusions rich in Al-Mg-Cr-Ti were identified. Finally, changes in mechanical properties of Al 5154-O alloy with an increase in neutron fluence (or transmutation Si) are discussed in correlation with SEM microstructure and fracture morphology observed in SEM. TEM investigation of precipitate microstructure is ongoing and those results will be published in future.

scholarly journals Non-hardening embrittlement mechanism of pressure vessel steel Ni-Cr-Mo-V welds during thermal aging

2020 ◽  
Vol 12 (2) ◽  
pp. 168781402090456
Author(s):  
Guojun Wei ◽  
Chenglong Wang ◽  
Xingwang Yang ◽  
Zhenfeng Tong ◽  
Wenwang Wu
Keyword(s):  
Weld Metal ◽  
Thermal Aging ◽  
Pressure Vessel ◽  
Impact Test ◽  
Charpy Impact ◽  
Charpy Impact Test ◽  
Reactor Pressure Vessel ◽  
Impact Fracture ◽  
The Impact ◽  
Reactor Pressure

The mechanical performance of reactor pressure vessel materials is an important factor in the safety and economics of the operation of a nuclear power plant. The ductile-to-brittle transition temperature tested by Charpy impact test is the key parameter for evaluating the reactor pressure vessel embrittlement. In this article, the study of thermal aging embrittlement of temperature sets of reactor pressure vessel surveillance Ni-Cr-Mo-V steel weld metal was conducted by Charpy impact test. The thermal aging effect on the impact fracture behavior was analyzed. The impact test of the three batches of weld surveillance sample indicated that the weld metal embrittled during thermal aging. The study of impact fracture and Auger electron spectroscopy indicated that the element P segregated to the grain boundaries and lowered their cohesion strength during the long-term thermal aging. Therefore, the non-hardening embrittlement of Ni-Cr-Mo-V steel welds in a reactor pressure vessel caused by segregation of impurity elements P occurs during thermal aging.

Preliminary Results On Ultrasonic Attenuation Detection Of Neutron Irradiation Embrittlement Of Nuclear Reactor Steel

1997 ◽  
Vol 503 ◽  
Author(s):  
A. L. Hiser ◽  
R. E. Green
Keyword(s):  
Fracture Toughness ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Ultrasonic Attenuation ◽  
Reactor Pressure Vessel ◽  
Irradiation Embrittlement ◽  
Preliminary Results ◽  
Nondestructive Determination ◽  
Reactor Pressure

ABSTRACTNeutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper provides preliminary results of an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels.

Irradiation Induced Changes in Mechanical and Microstructural Properties of the High Flux Reactor Vessel: Update of the Results From 2014 and 2015 Surveillance Test Campaigns

2017 ◽  
Author(s):  
M. Kolluri ◽  
F. H. E. de Haan-de Wilde ◽  
H. S. Nolles ◽  
A. J. M. de Jong ◽  
F. A. van den Berg
Keyword(s):  
Mechanical Properties ◽  
Fracture Toughness ◽  
Neutron Fluence ◽  
Reactor Vessel ◽  
Surveillance Program ◽  
High Flux ◽  
High Flux Reactor ◽  
Operational Life ◽  
Flux Reactor ◽  
Vessel Material

The reactor vessel of the High Flux Reactor (HFR) in Petten has been fabricated from Al 5154 - O alloy grade with a maximum Mg content of 3.5 wt. %. The vessel experiences large amount of neutron fluences (notably at hot spot), of the order of 1027 n/m2, during its operational life. Substantial damage to the material’s microstructure and mechanical properties can occur at these high fluence conditions. To this end, a dedicated surveillance program: SURP (SURveillance Program) is executed to understand, predict and measure the influence of neutron radiation damage on the mechanical properties of the vessel material. In the SURP program, test specimens fabricated from representative HFR vessel material are continuously irradiated in two specially designed experimental rigs. A number of surveillance specimens are periodically extracted and tested to evaluate the changes in fracture toughness properties of the vessel as a function neutron fluence. The surveillance testing results of test campaigns performed until 2009 were already published by N. V. Luzginova et. al. [1]. The current paper presents results from the two recent surveillance campaigns performed in 2014 and 2015. The fracture toughness and tensile testing results are reported. Changes in mechanical properties of Al 5154-O alloy with an increase in neutron fluence are discussed in correlation with the irradiation damage microstructure observed in TEM and the fracture morphology observed in SEM. The HFR surveillance testing results are compared to the historically published results on irradiated aluminum alloys and conclusions about the evolution of embrittlement trends in relation with irradiation induced damage mechanisms in HFR vessel are drawn at the end.

Assessment of Weld Heat-Affected Zones in a Reactor Vessel Material

1978 ◽  
Vol 100 (3) ◽  
pp. 267-271 ◽  
Author(s):  
T. U. Marston ◽  
W. Server
Keyword(s):  
Mechanical Properties ◽  
Base Material ◽  
Charpy Impact ◽  
Dynamic Fracture Toughness ◽  
Reactor Vessel ◽  
Heavy Section ◽  
Vessel Material ◽  
Submerged Arc ◽  
Weld Heat

The mechanical properties of weld heat-affected zones (HAZ’s) associated with the heavy section, nuclear quality weldments are evaluated and found to be superior to those of the parent base material. The nil ductility transition temperature (NDTT), Charpy impact and static and dynamic fracture toughness properties of a HAZ associated with a submerged arc weld and one associated with a manual metal arc weld are directly compared with those of the parent base material. It is concluded that the stigma normally associated with HAZ is not justified for this grade and quality of material and weld procedure.

Effects of Major Parameters on the P-T Limit Curve Construction of Embrittled Reactor Vessel

2004 ◽  
Vol 261-263 ◽  
pp. 1647-1652
Author(s):  
Sung Gyu Jung ◽  
In Gyu Park ◽  
Chang Soon Lee ◽  
Myung Jo Jhung
Keyword(s):  
Fracture Toughness ◽  
Neutron Fluence ◽  
Crack Depth ◽  
Reactor Vessel ◽  
Sensitivity Analyses ◽  
Crack Orientation ◽  
Limit Curve ◽  
Asme Code ◽  
Potential Failure ◽  
Reactor Pressure

To prevent the potential failure of the reactor pressure vessel (RPV), it is requested to operate RPV according to the pressure-temperature (P-T) limit curve during the heat-up and cool-down process. The procedure to make the P-T limit curve was suggested in the ASME Code but it has been known to be too conservative for some cases. In this paper, the conservatism of the ASME Code Sec. XI, App. G was investigated by performing a series of sensitivity analyses. The effects of six parameters such as crack depth, crack orientation, clad thickness, fracture toughness, cooling rate, and neutron fluence were analyzed. The results of P-T limit curves are compared to one another.

Evaluation of Nozzle P-T Limit Curves for Korea Optimized Power Reactor

2020 ◽  
Author(s):  
Hyunchul Lee ◽  
Choonsung Yoo ◽  
Youngjae Maeng ◽  
Sunghoon Yoo
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Power Reactor ◽  
Nuclear Safety ◽  
Reactor Vessel ◽  
Limit Curve ◽  
Reactor Coolant ◽  
Regulatory Issue ◽  
Coolant System ◽  
Reactor Pressure

Abstract The purpose of the Pressure – Temperature (P-T) limit curve is to prevent a failure of reactor pressure vessel during operation of reactor coolant system. In Korea, P-T limit curves have to meet 10CFR50 Appendix G [1] according to Nuclear Safety and Security Commission Notification 2017-20. The P-T limit curves have been traditionally evaluated based on the beltline region which is most affected by neutron irradiation. However due to the geometric discontinuity, the inside corner regions of the vessel nozzles are the most highly stressed regions of reactor vessel. These higher stresses can potentially result in more restrictive P-T limits, even if the reference nil ductility transition temperatures (RTNDT) for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries. In 2014, the NRC issued Regulatory Issue Summary (RIS) 2014-11 [2], which require the consideration of reactor vessel nozzles in P-T limits curve generation. In this paper, P-T limit curves for Korea optimized power reactor (OPR-1000) nozzles at the end of license were evaluated. And then these curves were compared to the traditional beltline region P-T limit curves in order to verify which curve is more limiting.

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