Integrity of feedwater and main steam piping in KWU light water reactor plants

New standard catalogs for piping, supports, and valves have been introduced by Kraftwerk Union (KWU) for the first time in its Convoy series of PWR plants. These catalogs, underlying regulatory codes, and newly developed KWU specifications are described. Feedwater and main steam piping systems within the containment, including pipe supports and valves, are used to demonstrate the high quality level of piping technology achieved in the Federal Republic of Germany. Such quality standards ensure the integrity of single components as well as of the entire system, so that, under certain conditions, pipe whip restraints against postulated breaks have become unnecessary. The quality aspects apply basically for both PWR and BWR plants of KWU.     

Unsteady piping forces caused by hot water discharge from suddenly opened safety/relief valves

Various fluid transients in nuclear pressure vessels create the possibility of liquid discharge from safety/relief valves designed for steam release. Piping systems which carry the flow from safety/relief valves to an appropriate discharge environment are designed to withstand the unsteady reaction forces associated with steam flows. If liquid discharge occurs, the piping system may be subject to different unsteady reaction forces. This study provides a means of estimating discharge rates for saturated and subcooled water and resulting pipe forces, for the limiting case of a suddenly opened valve. It was found that pipe forces caused by the moving shock are about the same for either steam or water discharge. Pipe forces caused by a moving steam-water/air interface during water discharge may exceed those forces caused by the steam-air interface resulting from steam discharge. It was determined that water discharge would result in a longer opening time for a piston-type valve. It is therefore expected that with realistic estimates of valve opening time, the magnitude of water discharge forces will be approximately the same as steam discharge piping forces.     

Proactive approaches to assure the structural integrity of Boiling Water Reactor components

Pressure vessel components in operating Boiling Water Reactor (BWR) plants are subjected to a variety of loading and environmental conditions which could lead to degradation over time. The significant damage mechanisms such as fatigue, stress corrosion cracking (SCC) and irradiation embrittlement are considered in the design basis of the reactor components and thus provide adequate structural margins over the operating life of the plant. Nevertheless, when the design basis assumptions are exceeded, e.g., thermal cycles, vibratory loading or chemistry transients, cracking may occur in pressure boundary components. Several proactive measures are being implemented to address this concern and assure the structural margins in BWR plants. These measures include: (i) control of materials and design to mitigate SCC and improvement of the environmental conditions through the implementation of Hydrogen Water Chemistry, (ii) advances in automated ultrasonic inspection of the BWR pressure vessel and piping, (iii) improved monitoring techniques for tracking fatigue usage and SCC effects in the piping and in the core, and (iv) development and qualification of durable repairs and specialized techniques such as use of high purity materials and temper bead repair. This paper describes current progress in implementing these proactive approaches for Boiling Water Reactors.     

Precracked pipe under waterhammer action

We numerically simulate a full scale test in several computational steps with the finite element method and compare all calculated data with the experimental findings. First, we compute the deflection under static loading and the spectrum of eigenfrequencies of an integer piping, attached to a nuclear reactor pressure vessel (RPV). Then we consider a sudden pipe break at some distance from the vessel, immediately followed by an undamped closure of a check valve close to the break on the RPV side, and calculate the elastic and plastic transient dynamic response of the integer piping part between the RPV and the break. Finally, we consider a circumferential internal surface crack, fairly close to the vessel; after extensive testing of our fracture mechanics calculation procedure we investigate the stress in the crack region under the waterhammer action.     

The experiences of non-code repairs for nuclear safety-related piping in Taiwan

Based on USNRC GL 90-05 and 91-18, the implementations of non-code repairs to stop the leakage of the class 3 moderate piping without the plant shutdown for code repairs have been established in Taiwan. This paper presents the experiences of the implements of the non-code class 3 piping repairs in Taiwan Power Company's BWR and PWR plants. The comprehensive procedures provide a sound and effective managing strategy to repair the leakage of the safety-related piping.     

Analysis on transient piping pressure and force caused by high pressure steam flow disturbances

This paper presents a quantitative analysis on transient responses of pressure and its resulting force to sudden operation of a pressure relief valve on the main steam line of a pressurized water reactor plant, which is initially in a closed system at high pressure. When the valve at closed position changes to the full open position suddenly, the shock can be generated by the sudden steam flow which may reach critical velocity and then travel through the pipe line causing some damages to the piping system. The major characteristics of shock are quantitatively investigated on the basis of one dimensional pressure wave theory. Through the application to the actual transient incident occurred at the Kori unit 1 on 16 April 2005, it is found that the present approach is quite practical to predict the transient behavior of shocks caused by sudden operation of valves.     

Opening and extension of circumferential cracks in a pipe subject to dynamic loads

Nuclear power plants are presently designed to withstand instantaneous pipe severance in combination with the maximum seismic loads. The hypothetical combination of these two unlikely events leads to system designs which are very expensive and require dynamic event devices such as pipe whip restraints which have the potential for deleterious interaction with the piping system during normal operations. These present pipe rupture criteria are based on the a priori hypothesis that the instantaneous guillotine pipe break is possible, rather than from a consideration of the manner in which cracks might open or extend in a real piping system. The objective of this study is to help establish the basis for understanding how cracks which might exist in the primary piping of a pressurized water reactor (PWR) would open and extend so that improved criteria can be developed based on this information.One of the regions where loss of pressure boundary integrity must be postulated is the terminal end of the cold leg at the reactor vessel inlet nozzle. This region (including the effects of the reactor vessel and the primary pump) is modelled for analysis with the MARC general purpose finite element program. A circumferential crack, one-half circumference long, is considered to suddenly occur around the outside of the elbow when the pipe is at normal operating pressure. The most severe part of the safe shutdown earthquake (SSE) loading transient is applied simultaneously with the initiation of the crack.The plastic dynamic analysis of the crack opening effects in the discharge leg pipe is performed using the MARC program until the maximum opening occurs. The J-integral plastic crack extension criterion is computed for all times during the transient. The results indicate that none of the cracks will extend significantly and that the opening areas are small fractions of the flow area of the pipe.     

Formation and growth of cracking in feed water pipes and RPV nozzles

The observation of numerous small and large cracks in ferritic feed water pipes of boiling (BWR) and pressurized water reactors (PWR) in the last few years has led to basic research into the causes of cracking and the crack growth mechanisms.In horizontal feed water pipe sections connected to nozzles of reactor pressure vessels (RPV) of BWR's as well as of steam generators (SG) of PWR's, circumferential macro and micro cracks were detected. These cracking phenomena could be observed in base material of pipes as well as in weld seam regions. The examination of the stress state displayed that the cracked pipe regions have been exposed to a number of cyclic thermal transients (thermal shock, flow stratification) during start-up (hot stand-by) and shut-down periods of the plants. During thermal transient periods, local and global cyclic stresses in the referred pipe cross sections have been induced which in interaction with the influence from environment (in operation as well as in shut-down periods) and local geometrical imperfections led to the initiation and formation of macro and micro cracks.In the reactor water clean-up system of BWR through which reactor water is fed from the RPV to the main feed water line, two longitudinally welded elbows have been detected to be severely cracked. Both elbows have been subjected to an internal pressure corresponding to RPV and additionally to a relevant “in-plane” bending moment. These longitudinal cracks were found to be started from the inner elbow surface. In one case the longitudinal crack was situated in the base material and was enlarged to leakage. In the second elbow the longitudinal crack was located in the heat affected zone (HAZ) of a longitudinal weld. In both cases the macro cracks started either from corrosion pits located in defective areas of the magnetic protection layer or from geometrical notches (weld root). The semi-elliptic small cracks got linked to more extended shallow cracks.Formation and growth mechanism of these cracks have been studied at the MPA Stuttgart in laboratory under simulated operation conditions which were held as realistic as possible compared with those in nuclear power plants.The results of experimental studies in laboratory as well as conclusions based on the above mentioned cracking phenomena in piping have been used as basic information for a realistic design of large scale (RPV) thermal shock experiments under operation conditions. The formation and growth mechanism of these cracks and their detection by means of NDE during thermal transients at the inner surface of RPV nozzle and at the adjacent cylindrical areas of RPV shell will be described.     

Analysis of a selected two-phase flow phenomenon in VVER reactors with horizontal steam generators

Since 1984 the thermal-hydraulic code ATHLET has been also applied for the analyses of LOCA and transients in VVER plants. The specific design of these plants especially of the steam generator design requires a specific modeling of the phenomena which may occur under LOCA and transient conditions in these plants. Differences in design compared to the design of western reactors have been briefly listed. Specific phenomena occuring under small leak accidents are shortly described. The consideration of the simulation of the boiler-condensor mode illustrates the modelling requirements for a code which may be applied to the prediction of such a thermalhydraulic behaviour. Facing the lack of experimental data, the reliability of the simulation has been discussed by means of plausibility studies based on the momentum balance for steam and water.     

Water Chemistry Experience of Nuclear Power Plants in Japan

Japanese LWRs have experienced several troubles caused by corrosions of structural materials in the past ca. 20 years of their operational history, among which are increase in the occupational radiation exposures, intergranular stress corrosion cracking (IGSCC) of stainless steel piping in BWR, and steam generator corrosion problems in PWR. These problems arised partly from the improper operation of water chemistry control of reactor coolant systems. Consequently, it has been realized that water chemistry control is one of the most important factors to attain high availability and reliability of LWR, and extensive researches and developments have been conducted in Japan to achieve the optimum water chemistry control, which include the basic laboratory experiments, analyses of plant operational data, loop tests in operating plants and computer code developments. As a result of the continuing efforts, the Japanese LWR plants have currently attained a very high performance in their operation with high availability and low occupational radiation exposures. A brief review is given here on the R & D of water chemistry in Japan     

Erratum

Nuclear power plants are presently designed to withstand instantaneous pipe severance in combination with the maximum seismic loads. The hypothetical combination of these two unlikely events leads to system designs which are very expensive and require dynamic event devices such as pipe whip restraints which have the potential for deleterious interaction with the piping system during normal operations. These present pipe rupture criteria are based on the a priori hypothesis that the instantaneous guillotine pipe break is possible, rather than from a consideration of the manner in which cracks might open or extend in a real piping system. The objective of this study is to help establish the basis for understanding how cracks which might exist in the primary piping of a pressurized water reactor (PWR) would open and extend so that improved criteria can be developed based on this information.One of the regions where loss of pressure boundary integrity must be postulated is the terminal end of the cold leg at the reactor vessel inlet nozzle. This region (including the effects of the reactor vessel and the primary pump) is modelled for analysis with the MARC general purpose finite element program. A circumferential crack, one-half circumference long, is considered to suddenly occur around the outside of the elbow when the pipe is at normal operating pressure. The most severe part of the safe shutdown earthquake (SSE) loading transient is applied simultaneously with the initiation of the crack.The plastic dynamic analysis of the crack opening effects in the discharge leg pipe is performed using the MARC program until the maximum opening occurs. The J-integral plastic crack extension criterion is computed for all times during the transient. The results indicate that none of the cracks will extend significantly and that the opening areas are small fractions of the flow area of the pipe.     

Target Curve with Maintenance Time in Optimal Xenon Shutdown Problems

A series of sparger tests have been conducted to investigate the performance of a steam sparger, which will be used in a Korean Advanced Power Reactor, APR1400. The tests have been conducted at the Blowdown and Condensation Loop in the Korea Atomic Energy Research Institute using a prototypic steam sparger. The major test parameters include amount of air mass and air temperature in the discharge line, valve opening time, steam mass flow rate, and water temperature and level in the In-Containment Refueling Water Storage Tank. The hydrodynamic loads induced by the air clearing phenomenon during typical operating conditions seem to be dependent on the valve opening time, steam mass flow rate, submergence of a sparger, and distance between the sparger and the structure. The effect of the amount of air mass in the discharge line and the water temperature to the peak load seems to be negligible for a given range of parameters. The hydrodynamic load induced by the prototypic steam sparger was less than those expected in a BWR and the steam sparger tested in this program can be used satisfactorily for the APR1400.     

PWR safety/relief valve blowdown analysis experience

The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively.General approaches for generating forcing functions from thermalfluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the “acceleration or wave force” method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawback are discussed.Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.     

Blowdown thrust force under pipe rupture accident: I. Experimental evaluations of blowdown thrust force and decompression characteristics

Blowdown thrust forces and decompression characteristics were evaluated concerning the jet discharge or pipe whip tests with a 4-inch or 6-inch diameter pipe under PWR LOCA or BWR LOCA conditions related to pipe rupture accidents in nuclear power plants. This paper presents experimental evaluations of time-dependent and maximum blowdown thrust forces, and evaluations of decompression characteristics under instantaneous pipe rupture conditions.The following items are discussed: the peak value of the blowdown thrust force, the jet thrust coefficient for the maximum blowdown thrust force, the pressure recovery after break, and the relationship between the pressure undershoot of the sudden decompression and the decompression rate.     

Blowdown thrust force under pipe rupture accident: II. Analytical evaluations of partial break and ramped opening break

Blowdown thrust forces and decompression characteristics were evaluated concerning the jet discharge or pipe whip tests with a 4-inch or 6-inch diameter pipe under PWR LOCA or BWR LOCA conditions related to pipe rupture accidents in nuclear power plants. This paper presents experimental evaluations of time-dependent and maximum blowdown thrust forces, and evaluations of decompression characteristics under instantaneous pipe rupture conditions.The following items are discussed: the peak value of the blowdown thrust force, the jet thrust coefficient for the maximum blowdown thrust force, the pressure recovery after break, and the relationship between the pressure undershoot of the sudden decompression and the decompression rate.     

Thermal-hydraulic transient analysis of light water reactors using moving boundaries

This paper examines the applicability of a mathematical dynamic model developed here for the simulation of the thermal-hydraulic transient analysis for light water reactors (LWRs). The thermal-hydraulic dynamic modeling of the fuel pin and adjacent coolant channel in LWRs is based on the moving boundary concept. The fuel pin model (FUELPIN) with moving boundaries is developed to accommodate the core thermal-hydraulic model, with detailed thermal conduction in fuel elements. Some results from transient calculations are examined for the first application of the thermal-hydraulic model and the fuel pin model with moving boundaries in a boiling water reactor (BWR). An accurate minimum departure from nucleate boiling ratio (MDNBR) and its axial MDNBR boundary versus time within the fuel channel are predicted during transients. Transient analysis using a known thermal-hydraulic code, COBRA and FUELPIN linked with a PWR systems analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a pressurized water reactor (PWR). The studies of the overall nuclear reactor system show that moving boundary formulation provides an efficient and suitable tool for thermal transient analysis of LWRs.     

Soil-structure interaction for transient loads due to safety relief valve discharges

Dynamic responses of BWR Mark II containment structures subjected to axisymmetric transient pressure loadings due to simultaneous safety relief valve discharges were investigated using finite element analysis, including the soil-structure interaction effect. To properly consider the soil-structure interaction effect, a simplified lumped parameter foundation model and an axisymmetric finite element foundation model with viscous boundary impedance are used. Analytical results are presented to demonstrate the effectiveness of the simplified foundation model and to exhibit the dynamic response behavior of the structure as the transient loading frequency and the foundation rigidity vary. The impact of the dynamic structural response due to this type of loading on the equipment design is also discussed.     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10^?2 (mean) and 3.9x 10^?2 (median) for the BWR suppression pool case, 2.2x10^?3 (mean) and 2.8x10^?10 (median) for the BWR pedestal case, and 6.8X10^?2 (mean) and 1.4x10^?2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.