Penetration Nozzles Research Articles

Finite Element Welding Residual Stress Analysis of CRDM Penetration Nozzles

Abstract This article investigates numerically welding residual stress distributions of a tube with J-groove weld in control rod drive mechanisms of a pressurized nuclear reactor vessel. Parametric study is performed for the effect of the tube location, tube dimensions, and material's yield strength. It is found that residual stresses increase with increasing the inclination angle of the tube, and the up-hill side is the most critical. For thicker tube, residual stresses decrease. For material's yield strength, both axial and hoop residual stresses tend to increase with increasing the yield strength of Alloy 600. Furthermore, axial stresses tend to increase with increasing yield strength of Alloys 82/182.

Break location influence in pressure vessel SBLOCA scenarios

The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused on simulating different hypothetical accidental scenarios in the PV using the thermal-hydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios.

PWSCC growth rate model of alloy 690 for head penetration nozzles of Korean PWRs

Abstract This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.

PWSCC initiation and propagation in a CRDM penetration nozzle

In the winter of 2012, several cracks were detected in the penetration nozzles of the reactor pressure vessel closure head in a pressurized water reactor in Korea. Primary water stress corrosion cracking in the CRDM penetration nozzle is usually initiated due to a combination of a particular alloy, temperature, boric acid and weld residual stress. Garud et al. suggested an empirical equation for PWSCC initiation time based on a strain rate damage model. In this study, the authors introduce a user subroutine to apply Garud’s simplified SRDM model and XFEM to simulate PWSCC crack initiation location and growth. The suggested model includes a new algorithm to track cracks and update crack domains by simulating primary water contact conditions. A finite element model of the CRDM penetration nozzle were constructed to simulate such phenomena. Only one quarter of the nozzle were modeled because of its geometrical symmetry. In order to calculate the welding residual stress, a heat transfer analysis with a model change function followed by a thermal stress analysis were performed. Adopting the calculated welding residual stress as initial condition, the PWSCC crack initiation and propagation analysis were done for over 11 cycles of normal operating conditions. After 11 cycles of normal operation, the crack depth and length were measured and compared with the reported cracks from operating nuclear power plants in Korea. A comparison of PWSCC driving force among hoop stress, von Mises stress and maximum principal stress were investigated. Also a comparison of PWSCC dimension between two different Alloy 600 material conditions (MA, CW) were investigated.

Effect of chevron nozzle penetration on aero-acoustic characteristics of jet at M = 0.8

Aero-acoustic characteristics of a high-speed jet with chevron nozzles are experimentally investigated at a Mach number of 0.8. The main focus is to examine the effects of the extent of chevron penetration and its position in the mixing layer. Chevron nozzles with three different levels of penetration employed at three different longitudinal locations from the nozzle lip are tested, and the results are compared with those of a plain baseline nozzle. The chevrons are found to produce a lobed shear layer through the notched region, thereby increasing the surface area of the jet, particularly in the close vicinity of the nozzle, which increases the mixing and reduces the potential core length. This effect becomes more prominent with increasing penetration closer to the nozzle lip in the thinner mixing layer. Near field and far field noise measurements show distinctly different acoustic features due to chevrons. The chevrons are found to effectively shift the dominant noise source upstream closer to the nozzle. Present investigation proposes a simpler method for locating the dominant noise source from the peak of the centerline velocity decay rate. The overall noise levels registered along the jet edge immediately downstream of the chevrons are higher, but further downstream they are reduced in comparison with the plain baseline nozzle. Also, the chevrons beam the noise towards higher polar angles at higher frequencies. At shallow polar angles with respect to the jet axis in the far field, chevrons suppress the noise at low frequencies with increasing penetration, but for higher polar angles, while they continue to suppress the low frequency noise, at higher frequencies the trend is found to reverse. The noise measured in the near field close to the jet edge is composed of two components: acoustic and hydrodynamic. Of these two components, the chevrons are found to reduce the hydrodynamic component in comparison with the acoustic one.

An extended finite element method‐based representative model for primary water stress corrosion cracking of a control rod driving mechanism penetration nozzle

AbstractPrimary water stress corrosion cracking incidents have been reported in nuclear reactors over the past several decades. Garud et al developed an empirical equation to express primary water stress corrosion cracking (PWSCC) initiation time by using experimental data. This strain rate damage model has been used in multiple simulation studies. Some of these studies used the extended finite element method (XFEM) to simulate the PWSCC propagation in Alloy 600. However, several studies showed that the accuracy of XFEM depends on the mesh quality. Different mesh qualities can change the heat flux of a welding procedure, leading to different weld residual stresses. We performed a parametric study on PWSCC initiation and propagation of a control rod driving mechanism by using different mesh qualities. The major variables explored here are number of elements per bead, number of circumferential elements, and number of weld beads. Finally, an XFEM‐based representative model was suggested for PWSCC initiation and propagation simulation.

Corium melt researches at VESTA test facility

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging ZrO2 melt jet on a sacrificial material were performed to investigate the ablation characteristics. ZrO2 melt in an amount of 65–70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5–2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. ZrO2 melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is UO2 60%, Zr 10%, ZrO2 15%, SUS304 14%, and B4C 1%, was melted in a cold crucible using an induction heating technique.

Structural Integrity Analysis of Nuclear Power Plant Pressure Vessel Penetration Nozzle Repaired

Primary water stress corrosion cracking (PWSCC) has been observed in CRDM nozzles, BMI nozzles and other penetration nozzles. The industry has used the repair method of replacement of nozzles fabricated of Alloy 690. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design requirement. In this paper, the pressurizer top head instrument nozzle of PWR nuclear power plant is evaluated as a typical pressure vessel penetration nozzle. The results showed that the repaired nozzle satisfies the ASME Code design requirement and the crack growth of the postulated flaw in 40 years of the nuclear plant life is acceptable.PWSCC degradation mechanism has been observed in CRDM nozzles, BMI nozzles and other penetration nozzles [1]. In some nuclear power plants built in China earlier, such as DAYABAY nuclear power plant and QINSHAN nuclear power plant, PWSCC degradation mechanism has been found in CRDM nozzle welds which manufactured of Alloy 600 and welded of Alloy 82/182[2]. The repair of the degraded nozzles is the popular choice for the nuclear power plant owners. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design requirement. In this paper, the pressurizer top head nozzle of PWR nuclear power plant is evaluated as a typical pressure vessel penetration nozzle. Stress intensities were conservatively determined for pressure and applicable thermal transients and compared to the allowable values of the ASME Code, Section III. Thermal stress of the transients was obtained from 3D finite element model (FEM). Residual stress of J-groove weld was obtained from 2D FEM analysis and used for fracture mechanics analysis. All of the analysis showed that the repaired nozzle satisfies the ASME Code design requirement and the crack growth of the postulated flaw in 40 years of the nuclear plant life is acceptable.

PWSCC Assessment by Using Extended Finite Element Method

The head penetration nozzle of control rod driving mechanism (CRDM) is known to be susceptible to primary water stress corrosion cracking (PWSCC) due to the welding-induced residual stress. Especially, the J-groove dissimilar metal weld regions have received many attentions in the previous studies. However, even though several advanced techniques such as weight function and finite element alternating methods have been introduced to predict the occurrence of PWSCC, there are still difficulties in respect of applicability and efficiency. In this study, the extended finite element method (XFEM), which allows convenient crack element modeling by enriching degree of freedom (DOF) with special displacement function, was employed to evaluate structural integrity of the CRDM head penetration nozzle. The resulting stress intensity factors of surface cracks were verified for the reliability of proposed method through the comparison with those suggested in the American Society of Mechanical Engineering (ASME) code. The detailed results from the FE analyses are fully discussed in the manuscript.

Residual stress assessment of nickel-based alloy 690 welding parts

Failures of welding parts in nuclear power plants have increased significantly. One of main degradation mechanisms was initiation and growth of primary water stress corrosion cracking accelerated by residual stress at dissimilar metal welds of piping, reactor head penetrations and nozzles connecting major components. Therefore, estimation of the welding residual stress became an important issue for ensuring reliability of them. The present study deals with verification of welding simulation techniques through sensitivity analyses of bottom-mounted instrumentation nozzle mock-up and control rod drive mechanism (CRDM) nozzle mock-up, examination of residual stress effects for a welded cylindrical block and application of the residual stress evaluation method to real penetration nozzles of CRDM made of nickel-based alloy 690.

Evaluation of J-groove weld residual stress and crack growth rate of PWSCC in reactor pressure vessel closure head

Over the last decade, primary water stress corrosion cracking (PWSCC) has been frequently found in pressurized water reactor (PWR) applications. Especially, PWSCC has occurred in long-term operated PWRs. As this phenomenon leads to serious accidents, we must be beforehand with the anticipated problems. A typical PWR consists of J-groove welded components such as reactor pressure vessel closure head and nozzles. Reactor pressure vessel closure head is made of SA508 and it is covered by cladding. Alloy 600 is used for nozzles. And J-groove weld is conducted with alloy 82/182. Different material properties of these metals lead to residual stress and PWSCC consequentially. In this study, J-groove weld residual stress was investigated by a three-dimensional finite element analysis with an actual asymmetric J-groove weld model and process of construction. Also crack growth rate of PWSCC was evaluated from cracks applied on the penetration nozzles. Based on these two values, one cannot only improve the structural integrity of PWR, but also explain PWSCC behavior such that high residual stress at the J-groove weld area causes crack initiation and propagation through the surface of nozzles. In addition, crack behavior was predicted at the various points around the nozzle.

Investigation on the effects of geometric variables on the residual stresses and PWSCC growth in the RPV BMI penetration nozzles

This study investigated the effects of various geometric variables on the residual stresses and PWSCC growth of RPV BMI penetration nozzles. An FE residual stress analysis procedure was developed and validated from the viewpoint of FFS assessment. The validated FE residual stress analysis procedure and the PWSCC growth assessment procedure in the ASME B&PV Code, Sec.XI were applied to the BMI penetration nozzles with specified ranges of the geometric variables. The total stresses at steady state during normal operation including welding residual stresses increase with increasing inclination angle of the BMI nozzles, and with tilt angle, depth, and root width of the J-groove weld. The lifetime from the assumed initial crack to the acceptance criteria according to the ASME B&PV Code, Sec.XI also decreases under these conditions. The total stresses decrease and the lifetime increases with increasing nozzle thickness, but outer radius of the BMI nozzles has an insignificant effect on both of these factors.

Residual stress reduction in the penetration nozzle weld joint by overlay welding

Stress corrosion cracking (SCC) in the penetration nozzle weld joint endangers the structural reliability of pressure vessels in nuclear and chemical industries. How to decrease the residual stress is very critical to ensure the structure integrity. In this paper, a new method, which uses overlay welding on the inner surface of nozzle, is proposed to decrease the residual stresses in the penetration joint. Finite element simulation is used to study the change of weld residual stresses before and after overlay welding. It reveals that this method can mainly decrease the residual stress in the weld root. Before overlay welding, large tensile residual stresses are generated in the weld root. After overlay weld, the tensile hoop stress in weld root has been decreased about 45%, and the radial stress has been decreased to compressive stress, which is helpful to decrease the susceptibility to SCC. With the increase of overlay welding length, the residual stress in weld root has been greatly decreased, and thus the long overlay welding is proposed in the actual welding. It also finds that this method is more suitable for thin nozzle rather than thick nozzle.

Three-dimensional finite element welding residual stress analysis of penetration nozzles: I – Sensitivity of analysis variables

This paper investigates effects of variables related to 3-D finite element welding residual stress analyses of penetration nozzles on residual stresses. Five variables are considered; (i) number of elements in the circumferential direction, (ii) number of beads in the circumferential direction, (iii) kinematic boundary conditions, (iv) circumferential modelling angle in 3-D finite element models and (v) detailed weld shape. Based on sensitivity analysis results, guidelines for 3-D finite element welding residual stress analyses of penetration nozzles are provided.

원전 정상가동조건 적용 방식이 원자로 압력용기 상부헤드 관통 노즐의 용접 잔류응력에 미치는 영향

가압형 경수로 원자로의 압력용기 상부헤드 관통노즐 J-groove 용접부 주변에서 일차수응력부식균열(PWSCC)로 인한 냉각수 누설사례가 발생하고 있다. 본 연구에서는 PWSCC 의 주요 원인 중 하나인 용접 잔류응력을 유한요소 해석을 이용해 평가하고 원자력 발전소의 정상가동 조건을 해석에 반영하는 방법이 용접잔류응력 분포에 미치는 영향에 대한 분석을 수행하였다. 또한 반복되는 원자력 발전소의 가동 주기가 용접잔류응력 분포에 미치는 영향을 확인하여 정상가동조건에서의 정확한 용접 잔류응력을 예측할 수 있는 방법을 분석하였다.

Review of PWSCC and mitigation management strategies of Alloy 600 materials of PWRs

Primary water stress corrosion cracking (PWSCC) of Alloy 600 penetration nozzles in pressurized water reactors (PWRs) was reported in the control rod drive mechanism (CRDM), pressurizer instrumentation, and pressurizer heater sleeves. Recently, two cases of boric acid precipitation that indicated leaking of the primary cooling water were reported on the bottom head surface of the steam generators (SGs) in Korea. PWSCC crack indication in CRDM was also detected in a Korea plant. It is necessary to set up a rigid maintenance and inspection guidelines for the components. The PWSCC history of Alloy 600 penetration nozzles of PWRs and maintenance strategies are reviewed based upon the open literature and some experiences in Korea. The inspection requirements, repair techniques such as material changes, the isolation, weld overlays, stress improvements, water chemistry changes are reviewed. Management strategies for the Alloy 600 nozzles are also described.

원자로 CRDM 관통노즐 J-Groove 용접부 잔류응력 예측을 위한 유한요소 변수 민감도 해석

최근 원자로 압력용기 상부헤드 관통노즐 J-groove 용접부 주변에서 균열로 인한 냉각수 누출사고가 발행하고 있다. 이러한 사고의 원인은 용접에 의한 인장잔류응력, 농축된 붕산수 및 응력부식에 민감한 재료로 인한 일차수응력부식균열(PWSCC : primary water stress corrosion cracking)인 것으로 판명되었다. PWSCC 평가는 원자로 건전성 평가의 주요 관심사로서 용접에 의해 발생되는 잔류응력을 정확하게 예측함으로써 가능하다. 본 연구에서는 유한요소해석을 이용하여 국내 원자로의 일반적인 J-groove 용접부의 해석절차를 소개하고, 용접해석 관련 변수의 민감도 해석을 통해 잔류응력 예측기법을 제시하고자 한다. 이를 위해 2차원 및 3차원 요한요소해석 방법을 바탕으로 변수 민감도 해석을 수행하였으며, 기존 연구결과와 비교를 통해 해석절차 및 방법의 유용성을 검정하였다.

원전 이종금속 맞대기용접부 PWSCC 균열건전성평가

가압경수로형 원전의 Alloy 600 원자로압력용기헤드 관통노즐 및 Alloy 82/182 이종금속 맞대기 용접부에서 일차수응력부식균열(PWSCC)이 보고된 이후 전 세계적으로 PWSCC 에 의한 용접부 파단을 예방하기 위해 강화검사를 적용하고 있다. 본 이종금속용접부에 대한 가동중검사에서 균열이 발견된 경우 건전성평가 결과가 도출되기까지 발전소가 정지 상태에 있게 됨에 따라 원전 이용율 저하가 발생할 수 있는데, 이를 예방하기 위해서는 균열건전성평가 관련 기술의 정립뿐만 아니라 신속하게 평가 결과를 도출할 수 있는 시스템의 구축이 필요하다. 본 연구에서는 이종금속 맞대기 용접부를 대상으로 진행하고 있는 PWSCC 균열건전성평가 기준 정립 및 전산 시스템 개발 결과를 제시하였다. 본 연구를 통해 이종금속 맞대기 용접부 PWSCC 균열건전성평가 기술이 정립되고 전산 시스템으로 구현되어 원자로압력용기 주변 이종금속 맞대기 용접부에서의 PWSCC 균열에 대한 기술적 건전성평가 수단을 확보하였다.

Decreasing welding distortions in penetration nozzles by applying proper welding sequences

Welding distortions cause problems in assembly and performance of weldments, including nozzles in pressure vessels. In the present work, a three-dimensional thermo-mechanical finite element model was expanded to study the effect of the welding sequence on welding distortions in a pressurized nozzle. Among different types of distortions, shrinkage and angular distortions have been investigated. Since changes in the welding sequence increase the computational time and complicate the welding process, the effects of the welding sequence in different passes have also been studied. For the case studies, a five-pass shielded metal arc welding with V-joint geometry has been selected. The vessel has a diameter of 2000 mm and a thickness of 8 mm, and the nozzle has a diameter of 100 mm and a thickness of 8 mm. The pressure vessel and nozzles are made of 304 stainless steel material with the same chemical compositions. The finite element and experimental results have been compared, and turned out to be in good agreement. The results showed that welding in four sections with a proper sequence has significant influence on decreasing angular distortions, but the welding sequence is not an effective parameter to reduce the shrinkage type distortions. Moreover, sequencing can be used in some passes to simplify the welding process and also to decrease the process time.

An Investigation of the Stress Intensity Factor Along a Corner Crack in Pressurizer Vent Nozzle Penetration Weld

Elastic–plastic fracture mechanics as well as linear-elastic fracture mechanics may be applied to evaluate a flaw in ferritic low alloy steel components for operating conditions when the material fracture resistance is controlled by upper shelf toughness behavior. In this paper, the distribution of the stress intensity factor (SIF) along a corner crack using elastic–plastic fracture mechanics technique is investigated to assess the effect of a structural factor on mechanical loads in pressurizer vent nozzle penetration weld. For this purpose, the stress intensity factor and the plastic-zone correction of a corner crack are calculated under internal pressure, thermal stress, and residual stress in accordance with Electric Power Research Institute (EPRI) equation and Irwin’s approach, respectively. The resulting stress intensity factor and the plastic-zone correction were compared with those obtained from Structural Integrity Associates (SIA) and Kinectrics Inc., and were observed to be in good agreement with Kinectrics results.