Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurred at PWR Plants

2006 ◽  
Author(s):  
Jong Chull Jo ◽  
Myung Jo Jhung ◽  
Seon Oh Yu ◽  
Hho Jung Kim ◽  
Young Gill Yune
Keyword(s):  
Branch Pipe ◽  
Hot Water ◽  
Thin Wall ◽  
Reactor Operation ◽  
Support Condition ◽  
Primary Coolant ◽  
Lower Mode ◽  
Root Cause ◽  
Main Pipe ◽  
Flow Situation

Thermal sleeves in the shape of thin wall cylinder seated inside the nozzle part of each safety injection (SI) line at pressurized water reactors (PWRs) have such functions as prevention and relief of potential excessive transient thermal stress in the wall of SI line nozzle part which is initially heated up with hot water flowing in the primary coolant piping system when cold water is injected into the system through the SI nozzles during the SI operation. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the in the junction of primary coolant main pipe and SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in details by using both computational fluid dynamic (CFD) code and structure analysis finite element code. As the results, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15 to 18, which coincide with the lower mode natural frequencies of thermal sleeve having a pinned support condition on the circumferential prominence on the outer surface of thermal sleeve which is put into the circumferential groove on the inner surface of SI nozzle at the mid-height of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yield alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.

Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurring at PWR Plants

2008 ◽  
Vol 131 (1) ◽  
Author(s):  
Jong Chull Jo ◽  
Myung Jo Jhung ◽  
Seon Oh Yu ◽  
Hho Jung Kim ◽  
Young Gill Yune
Keyword(s):  
Cold Water ◽  
Operation Mode ◽  
Branch Pipe ◽  
Thin Wall ◽  
Reactor Operation ◽  
Support Condition ◽  
Primary Coolant ◽  
Root Cause ◽  
Main Pipe ◽  
Flow Situation

At conventional pressurized water reactors (PWRs), cold water stored in the refueling water tank of emergency core cooling system is injected into the primary coolant system through a safety injection (SI) line, which is connected to each cold leg pipe between the main coolant pump and the reactor vessel during the SI operation, which begins on the receipt of a loss of coolant accident signal. In normal reactor power operation mode, the wall of SI line nozzle maintains at high temperature because it is the junction part connected to the cold leg pipe through which the hot main coolant flows. To prevent and relieve excessive transient thermal stress in the nozzle wall, which may be caused by the direct contact of cold water in the SI operation mode, a thermal sleeve in the shape of thin wall cylinder is set in the nozzle part of each SI line. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the junction of primary coolant main pipe-SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in detail by using both computational fluid dynamics code and structure analysis finite element code. As a result, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15Hzto18Hz. These frequencies coincide with the lower mode natural frequencies of thermal sleeve, which has a pinned support condition on the outer surface with the circumferential prominence set into the circumferential groove on the inner surface of SI nozzle at the midheight of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yields alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.

Numerical Estimation of Flow Penetrating Depth Into Stagnant Branch Pipes for Thermal Fatigue Prediction

2014 ◽  
Author(s):  
Yoshihiro Ishikawa ◽  
Yukihiko Okuda ◽  
Naoto Kasahara
Keyword(s):  
Thermal Stress ◽  
High Temperature ◽  
Temperature Fluctuation ◽  
Thermal Stratification ◽  
Branch Pipe ◽  
Large Temperature ◽  
Bend Pipe ◽  
Main Pipe ◽  
High Temperature Water ◽  
Temperature Water

In the nuclear power plants, there are many branch pipes with closed-end which are attached vertically to the main pipe. We consider a situation in which the high temperature water is transported in the main pipe, the branch pipe is filled with stagnant water which has lower temperature than the main flow, and the end of the branch pipe is closed. At the branch connection part, it is known that a cavity flow is induced by the shear force of the boundary layer which separates from the leading edge of the branch pipe along the main pipe wall. In cases where the high temperature water penetrates into the branch pipe, there is a possibility that a steep and large temperature gradient field, called “thermal stratification layer” is formed at the boundary between high and low temperature water in the branch pipe. If the thermal stratification layer is formed in a bend pipe, which is used for connecting the vertical branch pipe and to a horizontal pipe, at the same time, the temperature fluctuation by the thermal stratification layer motion occurs, there may cause the thermal stress in the piping material. Furthermore, keeping the piping material under the thermal stress, there might be a possibility of a crack on the surface of the bend pipe. For this reason, the evaluation of the position where the thermal stratification layer reaches is very important during early piping design process. And, deeply understanding regarding the phenomena, is also important. However, because of the complexities of the phenomena, it is difficult to immediately clarify the whole mechanisms of the thermal stress arising due to the temperature fluctuation by the thermal stratification layer change. The complete prediction method for the position of the thermal stratification layer based on the mechanisms that is able to be applied to any piping system, any temperature and any velocity conditions, is also difficult. Therefore, a practical approach is required. The authors attempt to develop the practical estimation method for the thermal stratification layer position using the three-dimensional Navier-Stokes simulation which was based on the Reynolds-average in order to reduce the computational costs. In this paper, three different configurations of the piping were simulated and the simulation results were compared with the experimental results obtained by the other research group.

Study on Characteristics of Wall Temperature Fluctuation at a Mixing Tee With an Upstream Elbow

2018 ◽  
Author(s):  
Koji Miyoshi ◽  
Akira Nakamura
Keyword(s):  
Wall Temperature ◽  
Temperature Fluctuation ◽  
Hot Spot ◽  
Branch Pipe ◽  
Circumferential Direction ◽  
Wall Jet ◽  
Main Pipe ◽  
Inner Surface ◽  
Fluctuation Range ◽  
Dominant Frequencies

The characteristics of wall temperature fluctuation at the mixing tee with an upstream elbow were investigated and compared to those of the case without the elbow. The elbow of 90 degrees was installed in the inlet of the horizontal main pipe. The inlet flow velocities in the main and branch pipes were set to about 1.0 m/s and 0.7 m/s, respectively, to produce a wall jet pattern where the jet from the branch pipe was bent by the main pipe flow and made to flow along the pipe wall. A total of 148 thermocouples were installed near the pipe inner surface to measure the temperature distribution in the mixing tee. The upstream elbow decreased the temperature fluctuation intensity and the temperature fluctuation range at the inner surface. On the other hand, the distribution profiles and the dominant frequencies of temperature fluctuations were similar. The temperature fluctuation was also caused by the movement of a hot spot in the circumferential direction for both cases with and without the upstream elbow. The reduction of the movement of the hot spot in the circumferential direction decreased the temperature fluctuation for the case with the upstream elbow.

Buildup and Decay Analysis of Corrosion Products Activity in Primary Coolant Loop of AP-1000

2018 ◽  
Author(s):  
Fiaz Mahmood ◽  
Huasi Hu ◽  
Liangzhi Cao
Keyword(s):  
Radiation Exposure ◽  
Operating Conditions ◽  
Corrosion Products ◽  
Normal Operation ◽  
Reactor Operation ◽  
Primary Coolant ◽  
Reactor Control ◽  
Power Regulation ◽  
Coolant Loop ◽  
Control Regime

The broad half-life range of Activated Corrosion Products (ACPs) results in major radiation exposure throughout reactor operation and shutdown. The movement of unpredicted activity hot spots in coolant loop can bring about huge financial and dosimetric impacts. The PWR operating experience depicts that activity released during reactor operation and shutdown cannot be estimated through a simple correlation. This paper seeks to analyze buildup and decay behavior of ACPs in primary coolant loop of AP-1000 under normal operation, power regulation and shutdown modes. The application of a well-tested mathematical model is extended in an in-house developed code CPA-AP1000, to simulate the behavior of dominant Corrosion Products (CPs), by programing in MATLAB. The MCNP code is used as a subroutine of the program to model the reactor core and execute energy dependent neutron flux calculations. It is observed that short-lived CPs (56Mn, 24Na) build up rapidly under normal operation mode and decay quickly after the reactor is shutdown. The long-lived CPs (59Fe, 60Co, 99Mo) have exhibited slow buildup under normal operating conditions and likewise sluggish decay after the shutdown. To analyze activity response during reactor control regime, operating power level is promptly decreased and in response specific activity of CPs also followed decreasing trend. It is noticed that activity of CPs drops slowly during reactor control regime in comparison to emergency scram. The results are helpful in estimating radiation exposure caused by ACPs during accessibility of the equipment in coolant loop, under normal operation, power regulation and shutdown modes. Moreover, current analyses provide baseline data for further investigations on ACPs in AP-1000, being a new reactor design.

Numerical Simulation on Internal Flows of Reducing Cross

2014 ◽  
Vol 945-949 ◽  
pp. 980-986
Author(s):  
Jian Ping Yuan ◽  
Wen Ting Sun ◽  
Yin Luo ◽  
Bang Lun Zhou
Keyword(s):  
Numerical Simulation ◽  
Flow Rate ◽  
Branch Pipe ◽  
Three Dimensional ◽  
Uniform Flow ◽  
Hydraulic Loss ◽  
Steady Flows ◽  
Internal Flows ◽  
Inlet Velocity ◽  
Main Pipe

In order to study the internal flows and hydraulic loss of reducing cross, numerical simulation was carried out on a horizontally installed reducing cross. Three schemes of pipe diameters were studied. The time-averaged N-S equations of three-dimensional steady flows in the reducing pipe were calculated by CFX 14.5 based on the standard - two equation turbulence model together with standard wall function. The results show that the higher the inlet velocity, the hydraulic loss become larger when the split ratios are same for the reducing cross. With the uniform inlet velocities the higher the inlet velocity, the quicker the increasing rate of the hydraulic loss in main pipe, as well as the branch pipe. The integral change rules of hydraulic loss are similar with the condition of uniform flow rate inflow when the flow patterns at inlet are uniform. But with the same spilt ratio, the hydraulic loss of uniform velocity inflow is markedly less than that of uniform flow rate inflow in both main pipe and branch pipe. The bigger the differences of the diameters between the main pipe and the branch pipe, the larger the hydraulic loss of the branch pipe.

Thermal Fatigue Analysis at a Mixing Tee by a Fluid-Structural Simulation

2011 ◽  
Author(s):  
Masayuki Kamaya ◽  
Yoichi Utanohara ◽  
Akira Nakamura
Keyword(s):  
Thermal Stress ◽  
Fatigue Damage ◽  
Cold Water ◽  
Axial Stress ◽  
Branch Pipe ◽  
Three Dimensional ◽  
Dynamics Simulation ◽  
Cold Spot ◽  
Current Case ◽  
Main Pipe

In this study, the thermal stress at a mixing tee was calculated by the finite element method using temperature transients obtained by a fluid dynamics simulation. The simulation target was an experiment for a mixing tee, in which cold water flowed into the main pipe from a branch pipe. The cold water flowed along the main pipe wall and caused a cold spot, at which the membrane stress was relatively large. Based on the evaluated thermal stress, the magnitude of the fatigue damage was assessed according to the linear damage accumulation rule and the rain-flow procedure. Precise distributions of the thermal stress and fatigue damage could be identified. Relatively large axial stress occurred downstream from the branch pipe due to the cold spot. The position of the cold spot changed slowly in the circumferential direction, and this was the main cause of the fatigue damage. In the thermal stress analysis for fatigue damage assessment, it was concluded that the detailed three-dimensional structural analysis was not required. Namely, for the current case, a one-dimensional simplified analysis could be used for evaluating the fatigue damage without adopting the stress enhancement factor Kt quoted in the JSME guideline.

scholarly journals Measurement of vertical strain and velocity at Siple Dome, Antarctica, with optical sensors

2002 ◽  
Vol 48 (161) ◽  
pp. 217-225 ◽  
Author(s):  
Mark A. Zumberge ◽  
Daniel H. Elsberg ◽  
William D. Harrison ◽  
Eric Husmann ◽  
John L. Morack ◽  
...  
Keyword(s):  
Optical Fibers ◽  
Optical Sensors ◽  
Ice Sheet ◽  
Hot Water ◽  
Constitutive Relationship ◽  
Strain Rates ◽  
Thin Wall ◽  
Vertical Strain ◽  
The North ◽  
Siple Dome

AbstractAs part of a larger program to measure and model vertical strain around Siple Dome on the West Antarctic ice sheet, we developed a new sensor to accurately and stably record displacements. The sensors consist of optical fibers, encased in thin-wall stainless-steel tubes, frozen into holes drilled with hot water, and stretched from the surface to various depths (up to 985 m) in the ice sheet. An optical system, connected annually to the fibers, reads out their absolute lengths with a precision of about 2 mm. Two sets of five sensors were installed in the 1997/98 field season: one set is near the Siple Dome core hole (an ice divide), and a second set is on the flank 7 km to the north (the ice thickness at both sites is approximately 1000 m). The optical-fiber length observations taken in four field seasons spanning a 3 year interval reveal vertical strain rates ranging from −229 ± 4 ppm a−1 to −7 ± 9 ppm a−1. In addition to confirming a non-linear constitutive relationship for deep ice, our analysis of the strain rates indicates the ice sheet is thinning at the flank and is in steady state at the divide.

Effects of Hot Water and Holds on Fatigue of Stainless Steel

2016 ◽  
Author(s):  
Jussi Solin ◽  
Jouni Alhainen ◽  
Ertugrul Karabaki ◽  
Wolfgang Mayinger
Keyword(s):  
Elevated Temperatures ◽  
Water Environment ◽  
Fatigue Tests ◽  
Hot Water ◽  
Primary Circuit ◽  
Primary Coolant ◽  
Fatigue Assessment ◽  
Atom Diffusion ◽  
Environment Temperature ◽  
Circuit Components

Direct strain controlled LCF data for solid specimens is still very rare. In PVP2013-97500 and PVP2014-28465 we reported results for niobium stabilized X6CrNiNb1810mod steel (type 347) fatigued in 325°C and 200°C PWR water according to VGB water chemistry specification. New data in this paper further confirms the conclusions: we are unable to repeat as high Fen factors or short lives as predicted according to NUREG/CR-6909. The slowest strain rate used 4·10−6 in 325°C water would predict Fen > 12, i.e. laboratory specimen data below the current ASME design curve, but our results are superior for this steel generally used in German NPP’s. However, the difference is not necessarily grade specific. Use of 100% relevant fabricated material batch and standard LCF methodology are regarded to play an important role. Notable hardening can be measured, when long duration holds in elevated temperatures are introduced between blocks of cyclic strains at lower temperatures. This is the case for thermal gradient loaded primary circuit components, e.g. the PWR pressurizer spray lines or surge line, which connects the pressurizer to primary coolant line. In PVP2011-57942 we reported improved endurances in fatigue tests aiming to roughly simulate steady state operation between fatigue transients in such NPP components. New test types have been introduced to generalize the results. Mechanisms of time and temperature dependent relaxation of fatigue damage and/or improvement of material fatigue performance during holds are not yet fully revealed, but the rate controlling thermal activation energy is below shown to be near that for vacancy and interstitial atom diffusion. This allows us to draft a thermodynamic prediction model. Improved accuracy of fatigue assessment helps in focusing optimally scheduled nondestructive testing to the most relevant locations and maintaining high level of reliability without excessive cost and radiation doses for inspection personnel. This paper provides previously unpublished experimental results and proposes methods to improve transferability of laboratory test data to fatigue assessment of NPP components. The effects of material, water environment, temperature and service loading patterns are discussed.

Evaluation of plastic limit load of piping branch junctions subjected to out-of-plane moment loadings

2003 ◽  
Vol 38 (5) ◽  
pp. 395-404 ◽  
Author(s):  
F-Z Xuan ◽  
P-N Li ◽  
S-T Tu
Keyword(s):  
Failure Mode ◽  
Limit State ◽  
Branch Pipe ◽  
Limit Load ◽  
Internal Force ◽  
Plastic Limit ◽  
Main Pipe ◽  
Out Of Plane ◽  
Plastic Limit Load ◽  
Global Collapse

Under out-of-plane moment loadings, the piping branch junctions (also called tees in engineering) exhibit three kinds of failure mode, namely collapse failure of the branch pipe, global collapse of the intersection due to plastic hinges forming along the intersection line and local instability of the main pipe at the flank. In this work, the common piping branch junctions utilized in petrochemical and power industries with a failure mode of global collapse were investigated, and a new approximate formula for an out-of-plane plastic limit moment was presented. The formula was built on the following process: firstly, an equation between the out-of-plane limit moment and internal force of the branch pipe along the intersection is set up on the basis of the force equilibrium condition. Regarding this internal force as an external load for the main pipe shell, the internal force and moment along the intersection of the main pipe, under the plastic limit state, are then obtained. Finally, referring to the von Mises yield criterion, the approximate plastic limit load of the piping branch junctions subjected to the out-of-plane moment is derived. The accuracy of the new formula is validated by comparison with finite element analysis and experimental results.

scholarly journals A technique for detection of WWER fuel failures by activity of Xe radionuclides during reactor operation

2018 ◽  
Vol 4 (4) ◽  
pp. 263-270 ◽  
Author(s):  
Peter Kalinichev ◽  
Igor Evdokimov ◽  
Vladimir Likhanskii
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Noble Gases ◽  
Fuel Pellet ◽  
Economic Losses ◽  
Reactor Operation ◽  
Primary Coolant ◽  
Practical Applications ◽  
Fuel Rod ◽  
Fuel Failure

Fuel failures during operation of Nuclear Power Plants (NPPs) may lead to substantial economic losses. Negative effects of reactor operation with leaking fuel in the core may be reduced if fuel failures are detected in due time of the cycle. At present time, the ratio of the normalized release rates of 131I and 134I is used to detect fuel failures in WWERs during steady state operation. However, based on the activity of iodine radionuclides, it is not always possible to detect the fuel failure. This situation may occur in case of a small defect in cladding of a leaking fuel rod or for high burnup fuel if the defect is overlapped by the surface of the fuel pellet. If it is so, fuel deposits may be the dominant contributor to iodine activity, and the fuel failure may not be noticeable. In PWRs, fuel failures are detected by activity of radioactive noble gases. Noble gases are not adsorbed on cladding inner surface, as distinct from iodine radionuclides. Release of noble gases from the leaking fuel rod may be considerable even though defect in cladding is small. In this paper, a technique is proposed for detection of fuel failures at WWER reactors by activity of radioactive noble gases in the primary coolant. It is shown that radioactive noble gases may be a more sensitive indicator of fuel failures than iodine radionuclides. Detection of fuel failures is based on monitoring of the ratio between 133Xe and 135Xe activity. Some examples of practical applications are given.

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