Effect of Normal Operating Condition Analysis Method for Weld Residual Stress of CRDM Nozzle in Reactor Pressure Vessel

A reactor pressure vessel (RPV) head of PWR has penetration holes for the CRDM nozzles, which are connected with the vessel head by J-shaped welds. It is well-known that there is high residual stress field in vicinity of the J-shaped weld and this has potentiality of PWSCC degradation. For assuring stress integrity of welding part of the penetration nozzle of the RPV, it is necessary to evaluate precise residual stress and stress intensity factor based on the stress field. To calculate stress intensity factor K, the most acceptable procedure is numerical analysis, but the penetration nozzle is very complex structure and such a direct procedure takes a lot of time. This paper describes applicability of simplified K calculation method from handbooks by comparing with K values from finite element analysis, especially mentioning crack modeling. According to the verified K values in this paper, fatigue crack extension analysis and brittle fracture evaluation by operation load were performed for initial crack due to PWSCC and finally structural integrity of the penetration nozzle of RPV head was confirmed.

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Assessment of a Stress-Free Temperature Model for Residual Stresses in Surface Cladding of a Reactor Pressure Vessel

Volume 6B: Materials and Fabrication ◽

10.1115/pvp2017-65255 ◽

2017 ◽

Author(s):

B. Richard Bass ◽

Paul T. Williams ◽

Terry L. Dickson ◽

Hilda B. Klasky

Keyword(s):

Residual Stress ◽

Finite Element ◽

Thermal Expansion ◽

Residual Stresses ◽

Base Metal ◽

Pressure Vessel ◽

Reactor Pressure Vessel ◽

Normal Operation ◽

Inner Surface ◽

Reactor Pressure

This paper describes further results from an ongoing study of a simplified engineering model that is intended to account for effects of clad residual stresses on the propensity for initiation of preexisting inner-surface flaws in a commercial nuclear reactor pressure vessel (RPV). The deposition of stainless steel cladding during fabrication of an RPV generates residual stresses in the cladding and the heat affected zone of the under-lying base metal. In addition to residual stress, thermal strains are generated by the differential thermal expansion (DTE) of the cladding and base material due to temperature changes during normal operation. A simplified model used in the ORNL-developed FAVOR probabilistic fracture mechanics (PFM) code accounts for the clad residual stress by incorporating a stress-free temperature (SFT) approach. At the stress-free temperature (Ts-free), the model assumes there is no thermal strain, i.e., the thermal expansion stresses and clad residual stresses offset each other. For normal cool-down transients applied to the RPV, interactions of the latter stresses generate additional crack driving forces on shallow, internal surface-breaking flaws near the clad/base metal interface; those flaws tend to dominate the RPV failure probability computed by FAVOR. In a previous report from this study (PVP2015-45086), finite element analysis was used to compare the stresses and stress-intensity factors (SIF) during a cool-down transient for two cases: (1) the existing SFT model of FAVOR, and (2) directly applied RPV clad residual stress (CRS) distribution obtained from empirical (hole-drilling) measurements made at room temperature on an RPV that was never put into service. However, those analyses were limited in scope and focused on a single flaw orientation. In this updated study, effects of CRS on the SIF histories computed for both circumferential and axial flaw orientations subjected to a cool-down transient were determined from an extended set of finite element analyses. Specifically, comparisons were made between results from applying CRS experimental data to ABAQUS two-dimensional, inner-surface flaw models and those generated by the FAVOR SFT model. It is demonstrated that the FAVOR-recommended SFT value of 488 °F produces conservatively high values of SIF relative to the use of CRS profiles in the ABAQUS models. For the vessel and flaw geometry and transient under study, the circumferential flaw (360° continuous) required a decrease of SFT down to 390 °F to match the CRS SIF histories. For the infinite axial flaw model, a decrease down to 300 °F matched the CRS SIF histories. Future plans are described to develop more general conclusions regarding the FAVOR model.

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Residual Stresses in Clad Nuclear Reactor Pressure Vessel Steels: Prediction, Measurement and Reconstruction

Volume 6B: Materials and Fabrication ◽

10.1115/pvp2015-45224 ◽

2015 ◽

Cited By ~ 1

Author(s):

Karim Serasli ◽

Harry Coules ◽

David Smith

Keyword(s):

Residual Stress ◽

Finite Element ◽

Residual Stresses ◽

Pressure Vessel ◽

Nuclear Reactor ◽

Finite Element Simulations ◽

Reactor Pressure Vessel ◽

Mapping Technique ◽

Pressure Vessel Steels ◽

Reactor Pressure

Most residual stress measurement methods are limited in terms of their stress and spatial resolution, number of stress tensor components measured and measurement uncertainty. In contrast, finite element simulations of welding processes provide full field distributions of residual stresses, with results dependent on the quality of the input conditions. Measurements and predictions are often not the same, and the true residual stress state is difficult to determine. In this paper both measurements and predictions of residual stresses, created in clad nuclear reactor pressure vessel steels, are made. The measurements are then used as input to a residual stress mapping technique provided within a finite element analysis. The technique is applied iteratively to converge to a balanced solution which is not necessarily unique. However, the technique aids the identification of locations for additional measurements. This is illustrated in the paper. The outcomes from the additional measurements permit more realistic and reliable estimates of the true residual state to be made. The outcomes are compared with the finite element simulations of the welding process and used to determine whether there is a need for additional input to the simulations.

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WJP Applicability Study for PWR Components As a PWSCC Mitigation Technique

Volume 1B: Codes and Standards ◽

10.1115/pvp2017-66164 ◽

2017 ◽

Author(s):

Ryuji Kimura ◽

Noboru Saito ◽

Hisamitsu Hato ◽

Akihiro Kanno ◽

Masami Ando ◽

...

Keyword(s):

Residual Stress ◽

Water Flow ◽

Pressure Vessel ◽

Surface Stress ◽

Reactor Pressure Vessel ◽

The Other ◽

Performance Criteria ◽

Head Penetration ◽

Water Chamber ◽

Reactor Pressure

Water Jet Peening (WJP) has been widely applied to nuclear power plants in Japan as one of mitigation techniques against Stress Corrosion Cracking (SCC) initiation [1]. WJP utilizes high pressure water flow including numerous cavitation bubbles and improves surface residual stress of susceptible materials used in reactor internals from tensile stress to compressive stress without significant plastic deformation, hardening, heating and furthermore retrieval of foreign materials. An inspection relief for the Primary Water SCC (PWSCC) concerned components, by means of peening technique application, has been discussed among PWR owners in the US for about last 10 years. The topical report on PWSCC mitigation by surface stress improvement (Material Reliability Program (MRP)-335, revision 3-A) was published through the above activities by Electric Power Research Institute (EPRI) MRP [2]. The target components, where PWSCC is concerned, are listed as Reactor Pressure Vessel Head Penetration Nozzles (RPVHPNs), such as Control Rod Drive Mechanism Nozzle (CRDMN), and dissimilar metal welds (DMWs) of Reactor Coolant System (RCS) nozzles, and performance criteria for peening are defined in the topical report. Moreover, the technical basis for PWSCC mitigation by surface stress improvement (MRP-267, revision 2) was published by EPRI MRP [3].The report details numerous data for each peening technique which show the effectiveness in mitigating the PWSCC initiation and its sustainability, i.e. state of stress. The report also includes the process control; covering nozzle diameter, water flow rate, application time, jet stand-off, impingement angle and stationary nozzle time for WJP [3]. RPVHPNs inner diameter (ID), such as CRDMN ID, is in narrower areas than the other target components of peening techniques. Hence the WJP nozzle should be set appropriate condition, e. g. sufficient stand-off distance or angle of the WJP nozzle, in line with the MRP-267 in order to ensure the stress improvement effect by WJP. Further, the reactor pressure vessel head, which has the RPVHPNs including the CRDMNs, is placed on the refueling floor and under atmosphere condition during outage, and therefore, the CRDMNs have to be filled with water by plugging etc. for WJP application on CRDMN ID. Thus the CRDMN ID becomes a closed narrow chamber. In such a closed narrow chamber, water flow might become complex and disturb the cavitation collapse on the target surface, resulting in decreased stress improvement. Additionally, WJP has been rarely applied in a narrow closed water chamber, and only a few residual stress measurement data are available for such a WJP treated specimen. For the above reason, we has conducted a WJP test utilizing the water chamber and measured the residual stress of the test coupon simulating the CRDMN ID before and after WJP application as our own research. As a result, an improvement in residual stress was ensured even in an application of WJP in a closed narrow water chamber, which assumes CRDMN ID configuration, and created a depth over the performance criteria (0.01” (0.25 mm) in depth) stated in MRP-335 [2]. As an another applicability study, we developed a WJP tool for Bottom Mounted Instrument (BMI) Nozzles and confirmed that the residual stress of BMI ID and Outer Diameter (OD) can be improved . The background of this study is that BMI nozzle is under discussion for inspection relief as one of the components which are concerned about PWSCC. Especially, BMI ID is narrow area for WJP application; on the other hand it does not need to become a closed chamber since the reactor pressure vessel, which has the BMI Nozzles on the bottom head, is filled with water during outage. As a result, it is ensured that the residual stress for BMI ID and OD is improved by WJP to a depth of at least 0.2mm which is deeper than the performance criteria for the depth of compressive residual stress of Austenitic Stainless Steel in Japan (3.9 × 10−3” (0.1mm) in depth).

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Welding Residual Stress in a Large Diameter Nuclear Reactor Pressure Vessel Nozzle

Journal of Pressure Vessel Technology ◽

10.1115/1.4007961 ◽

2013 ◽

Vol 135 (2) ◽

Cited By ~ 1

Author(s):

Tao Zhang ◽

Frederick W. Brust ◽

Gery Wilkowski ◽

Heqin Xu ◽

Alfredo A. Betervide ◽

...

Keyword(s):

Residual Stress ◽

Power Plant ◽

Nuclear Power Plant ◽

Pressure Vessel ◽

Nuclear Power ◽

Large Diameter ◽

Reactor Pressure Vessel ◽

Welding Residual Stress ◽

Stress Fields ◽

Reactor Pressure

The Atucha II nuclear power plant is a pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant was designed by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. Welding residual stresses in nuclear power plant piping can lead to cracking concerns later in the life of the plant, especially for stress-corrosion cracking. Hence, understanding the residual stress distribution from welding is important to evaluate the reliability of pipe and nozzle joints with welds. In this paper, a large-diameter reactor pressure vessel (RPV) hot-leg nozzle was analyzed. This is a nozzle from Atucha II nuclear power plant in Argentina. The main piping material is 20MnMoNi55 with Tenacito 65R weld metal, and inner diameter (ID) welded cladding at the girth weld locations is made of 309L. The special materials and weld geometry will lead interesting welding residual stress fields. In addition, postweld heat treatment (PWHT) of the girth welds and its boundary conditions could also play an important role in determining welding residual stress fields at the plant’s normal operating conditions. Sensitivity analyses were conducted and the technical observations and comments are provided.

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Welding Residual Stress and Multiple Flaw Evaluation for Reactor Pressure Vessel Head Replacement Welds With Alloy 52

Volume 6: Materials and Fabrication, Parts A and B ◽

10.1115/pvp2009-78112 ◽

2009 ◽

Cited By ~ 7

Author(s):

T. Zhang ◽

F. W. Brust ◽

G. Wilkowski ◽

D. L. Rudland ◽

A. Csontos

Keyword(s):

Residual Stress ◽

Finite Element ◽

Limit Analysis ◽

Pressure Vessel ◽

Hot Cracking ◽

Reactor Pressure Vessel ◽

Welding Residual Stress ◽

Multiple Cracks ◽

Head Penetration ◽

Reactor Pressure

Small indications were found in one replacement reactor pressure vessel head (RPVH) mock-up being fabricated from Alloy 690 material and compatible weld metals, Alloy 52/152. The mockups were non-destructively examined and the lowest number of cracks found was five and the highest number was 22. There are numerous indications with some of them quite long (50 mm) in length. The source of these weld fabrication cracks is unknown. However, from experience with other difficult to weld materials, the source can range from slag inclusions in the weld metal to hot cracking during the weld deposition process. Hot cracking includes solidification cracking (weld), liquation cracking (HAZ), and ductility dip cracking (DDC). The indications were mostly circumferential in orientation (with respect to the nozzle axis) but some were axial. This paper includes two parts. The first part includes the welding residual stress analysis of RPVH using Alloy 52/152 metal and provides comparison with similar Alloy 82/182 welds. Alloy 82/182 was the material used in the original dissimilar metal welds in these heads. Primary Water Stress Corrosion Cracking (PWSCC) can occur in the primary coolant system when the welds are exposed to water, tensile stress, and temperature (usually higher than 250 C). PWSCC rates are higher in Alloy 82/182 material due to its lower chromium content compared with the replacement alloy. The results for both center hole (0-degree) and side hill (53-degree) nozzles will be discussed. The second part deals with assessment of multiple small cracks in the reactor pressure vessel head penetration nozzles. The finite element alternating method (FEAM) was used for calculating stress intensity factors for cases where multiple cracks exist. More than twenty cracks, which were inserted based on field measurements, are considered in the analyses for both center hole and side hill nozzles. It is observed that the overall stress trends are similar to those without adding cracks. However, cracks introduce more local stress fluctuations. The magnitude of the local fluctuation can be around 100MPa. Limit analysis was also conducted. A new finite element model with a voided-out weld region was used to simulate loss of structural capacity due to multiple flaws. The voided out volume effects on the structural integrity and future performance of RPVH were examined. Discussions based on weld residual stress, multiple flaw analysis and limit analysis conclude the paper.

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