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.     

Occurrence of critical heat flux during blowdown with flow reversal

A small-scale experiment using Freon-11 at 54°C and 450 kPa in a transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The transients were initiated by simultaneous operation of three air-actuated values. The inner tube of the annulus was uniformly heated over its 0.61-m length while the outer transparent pyrex wall was unheated. Test section instrumentation included seven pressure taps, inlet and outlet capacitance void probes, inlet and outlet turbine flowmeters, inlet and outlet fluid thermocouples, and 22 wall thermocouples. High-speed motion pictures were taken of the lower end of the test section where the intial CHF occurred. From these high-speed pictures, the flow reversal was observed to occur between 60 and 80 ms followed by a rapid thermal excursion at about 400 ms in the lower regions of the test section. Consequently, this measured CHF occurred well after the flow had reversed and re-established itself in the downward direction. At about 300 ms an annuflar flow pattern appeared and was well-developed at 400 ms. Therefore, the early CHF measured were all associated with the transition from bubbly to annular flows. In some cases this early CHF was rewet and the heated section remained in a stable coolable state for a considerable length of time, and experienced a later CHF when the liquid was nearly depleted. This long-term dryout was a function of the liquid volume contained in the system, whereas the early CHF was independent of the system volume. In addition, the early CHF did not show any significant sign of propagation whereas the latter one was observed to propagate smoothly upward.     

谐波传动式核动力装置用闸阀的设计

针对我国现有核阀的不足,提出了一种新型的谐波螺旋式阀门的设计,它可以提高核阀的传动性能和可靠性。     

A study of the critical flow models used in reactor blowdown analysis

Two-phase critical flow models widely used in safety calculations are compared with extensive published experimental measurements of the choked flowrates of steam-water mixtures. Comments are made on the applicability of the models in their supposed regions of validity and a mathematical upper limit to discharge flowrate is derived. The data fall below this limit and generally above a homogeneous isentropic flow model in thermal equilibrium. Suitable measurements of critical flowrates in pipes of diamater in the range of direct relevance to reactor blowdown calculations are still unavailable.     

Experimental and computed results for fluid-structure interactions with impacts in the HDR blowdown experiment

The HDR experimental facility has been used for several blowdown experiments in order to study fluid-structure interactions and loadings on the pressure vessel internal structures of a pressurized water reactor. We have developed the code FLUX to analyse the motions in the initial blowdown period.This paper describes a new type of HDR experiments (V34) and compares the experimental results with the FLUX-code results. As novel feature, the core barrel is not rigidly clamped to the vessel as in earlier experiments but supported with gaps such that the core barrel can move freely upwards for about 2 mm and horizontally for 0.3 mm at the upper flange. At the lower core-barrel edge, snubbers restrict the horizontal motion to about + 1.4 mm and −2.8 mm.The experimental results show that the core barrel is deflected sidewards until it hits the snubber at the lower edge and then swings back to hit the opposite snubber. By this some kinetic energy is lost due to plastic snubber deformations. At the same time, the measurements show that the core barrel lifts rather uniformly from its support upwards until it hits the upper constraint. Several bounces up and down are observed until the core barrel becomes fixed probably due to friction from the side.This situation has been pre- and post-computed with the new FLUX-version which contains a very effective algorithm to treat supports with gaps and resultant impacts. For treatment of plastic supports, a simple model is added. Pre-computations were not meaningful because of large deviations in the pre-estimated initials gaps. However the computed pressure-field is not influenced very much by these parameters and predicted very well. This was favoured by the isothermal fluid initial conditions. Post-computations show sufficient agreement with respect to computed core barrel motion. The axial motion is described very well. Some problems remain which are due to the model for the upper flange support.Impacts do not results in greatly enlarged loadings, strains or accelerations for this situation.     

Dynamic load effects on gate valve operability and snubber performance

The Idaho National Engineering Laboratory (INEL) participated in an internationally sponsored seismic research program conducted at the decommissioned Heissdampfreaktor (HDR) located in the Federal Republic of Germany. An existing piping system was modified by installation of 200-mm, naturally aged, motor-operated gate valve from a U.S. nuclear power plant and a piping support system of U.S. design. Using various combinations of snubbers and other supports, six other piping support systems of varying flexibility from stiff to flexible were also installed and tested. Additional valve loadings included internal hydraulic loads and, during one block of tests, elevated temperature. The operability and integrity of the aged gate valve and the dynamic response of the various piping support systems were measured during 25 representative simulations of seismic events.     

Thermal transient simulation tests of a sodium valve

The feasibility of thermal transient testing of sodium components using a fluid other than sodium is considered. Simulation of thermal transient conditions that may exist in the sodium system is considered to be, in general, achievable in a special test fixture, if the thermal and hydraulic conditions of the fluid to be used are properly selected. This feasibility is demonstrated for the 28 in. FFTF hot-leg isolation valve by introducing a high-speed gaseous nitrogen flow into an annulus formed by the valve body and a cylindrical pipe insert. The structural wall temperature distributions were estimated first for the sodium system, and the simulation was established by equating the transient heat transfer rates as a function of time at key locations in both real sodium and simulating nitrogen systems. Remarkably good simulation test results were achieved.     

Experimental results of coupled fluid-structure interactions during blowdown of the HDR-vessel and comparisons with pre- and post-test predictions

Experimental results of the first three blowdown tests with reactor pressure vessel internals (core barrel) at the HDR-facility in Kahl, Federal Republic of Germany, are summarized. The major goal of these and the following experiments to be performed during the first half of 1982 is the determination of the importance of multidimensional fluid-structure interaction phenomena during the initial phase of blowdown depressurization transient following a simulated pipe rupture. As such, the performed Preliminary Test Series provides the first, most realistic data in this area available thus far. The experiments have been accompanied by a series of pre- and post-test predictions with several German and American computer codes specifically developed to account for the phenomena of interest. These codes will be discussed briefly and the results presented in comparison with the data. These comparisons allow the verification of these codes on the basis of multidimensional, measured quantities for the first time. Overall conclusions on the basis of these comparisons will be presented from the point of view of Project HDR. Individual contributions by the various institutions which participated in these computational efforts supplement this information in the following articles of this special issue. The test matrix of the forthcoming Main Test Series (V31.2 through V34), defined on the basis of the foregoing experiments, concludes this overview.     

Computer calculations of air and steam blowdown suppression

We describe a computer code that combines an Eulerian incompressible-fluid algorithm (SOLA) with a Lagrangian finite-element shell algorithm. The former models the fluid and the latter models the containing structure in an analysis of pressure suppression in boiling-water reactors. The code (PELE-IC) calculates loads and structural response from air blowdown and from the oscillatory condensation of steam bubbles in a water pool. The fluid, structure, and coupling algorithms are tested by recalculating problems that have known analytical solutions, including tank drainage, spherical bubble growth, coupling for circular plates, and submerged cylinder vibration. Code calculations are also compared with the results of small-scale blowdown experiments.     

Loss-of-primary-flow-without-scram tests: Pretest predictions and preliminary results

A series of tests in the Experimental Breeder Reactor No. 2 (EBR-II) has been concluded that investigated the effects of a complete loss of primary flow without scram. The development and preliminary study of these events is first discussed, including the test limits and controlling parameters. The results of two of the tests, SHRT 39 and 45, are examined in detail, although a compact summary of all the tests is included. The success in meeting the objectives of the test program served to verify that natural processes will shut down the reactor and maintain adequate cooling without control rod or operator intervention. The good comparison between predicted and measured results confirms that such events can be analyzed without elaborate codes if the basic processes are understood. Furthermore, recent studies suggest that the EBR-II results are characteristic of new innovative LMR designs being pursued in the U.S. that incorporate metallic driver fuel.     

Analysis of blowdown experiment of scale steam generator model

A subcooled blowdown experiment in a scale steam generator (SG) model is analyzed by the use of a fluid-structure computer code (MULTIFLEX). The experimental model simulates the secondary side of a SG with a preheater. The MULTIFLEX code that solves simultaneously a coupled set of one-dimensional hydraulic conservation equations and structural dynamic equations is used to analyze the experiment, taking into account the fluid structure interaction between the secondary coolant and the SG structure, the baffle and tube support plates and the divider plate. The computed values of pressure and wall displacement histories agree well with the experimental data. The success of the analysis supports the use of the one-dimensional MULTIFLEX code to analyses of thermal hydraulic transients in the SG secondary side and the validity of the method for modeling the complicated system of the fluid-structure interactions.     

Transient analysis of blowdown thrust force under PWR LOCA

The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces were obtained by Navier-Stokes momentum equation for a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a crifical flow condition was satisfied.The following results are obtained:

1. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena.

2. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08.

3. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one.

4. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break.

Analysis of HDR blowdown experiment for fluid-structure interactions with multiflex

The HDR experimental facility of Kahlsruhe is comprised of a full-scale pressure vessel, core barrel, and piping systems. In the blowdown experiment, V31.1, the fluid-structure interaction of the core barrel and downcomer water is significant. This experiment is analyzed in the present paper. The HDR downcomer annulus is modeled by the one-dimensional network that is equivalent to two-dimensional fluid-structure interactions. The core barrel is modeled by the projector method for combined beam and shell models. The vessel motion is taken into account by means of the relative modal analysis proposed in this paper. Computed time histories of pressure, pressure differentials, and barrel wall displacements are compared with the experimental data. Fair agreement between experiment and post-test computation is found. Effects of the vessel motion are also discussed.     

A study on a characteristic of stem friction coefficient for motor operated flexible wedge gate valve

Stem friction coefficient is a coefficient that represents friction between thread leads of the stem and stem nut. It is an important factor to determine output thrust delivered from the actuator to the valve stem in assessing performance of motor operated valves. This study analyzes the effects of changes in differential pressure on stem friction coefficient, and determines the bounding value of stem friction coefficient. A dynamic test was conducted on multiple flexible wedge gate valves in various differential pressure conditions, and the test data was statistically analyzed to determine the bounding value. The results show that stem friction coefficient in middle and high differential pressure is influenced by fluid pressure, while stem friction coefficient in low differential pressure is almost not affected by fluid pressure. In addition, it is found that the bounding value of stem friction coefficient is higher in a closing stroke than in an opening stroke.     

The results of 5% small-break LOCA tests and natural circulation tests at the ROSA-IV LSTF

Five 5% small-break loss-of-coolant accident (SBLOCA) experiments and two natural circulation experiments were conducted at the ROSA-IV Large Scale Test Facility (LSTF). The liquid holdup in the upflow side of steam generator (SG) U-tubes temporarily depressed the core collapsed liquid level below the bottom of core during the loop seal clearing in the cold-leg break SBLOCA tests. This phenomena was affected by the core power and core bypass but was affected little by the actuation of the high pressure injection system. Overall thermal-hydraulic phenomena in a loop seal line break test was similar to that of cold-leg break tests, however, the liquid holdup phenomena played a little role. In a hot-leg break test a temporary but rapid depression of the core liquid level was observed immediately after the initiation of accumulator injection which caused condensation and depressurization in the cold leg. The change of natural circulation flow rate with the decrease of primary system mass inventory was qualitatively the same as observed in Semiscale, LOBI and PKL. The SG effective overall heat transfer coefficient below the secondary-side collapsed liquid level was weakly dependent on the secondary side liquid level and the core power. The measured minimum heat transfer coefficient was 1.7 kW/m²K for the full secondary side mass inventory.     

组合阀流动阻力数值实验

组合阀的流动阻力是控制棒水压驱动系统设计的关键参数之一.利用计算流体力学软件 CFX-5.6 对组合阀进行数值模拟实验和三维流场计算,得到在不同差压下所对应的稳定流量,进而算出各流道阻力系数.通过比较流动阻力的计算结果,证实φ9阀的工作性能比φ6阀得到了提高.利用后处理工具分析φ9阀流道的流速、压力等流场分布特性,指出阻力损失的主要部分位于来水-侧壁-阀芯区这一段,并提出进一步的结构优化改进意见.     

UPTF test results: First three separate effect tests

The first separate effect tests were run in the Upper Plenum Test Facility — a 1:1 representation of a PWR primary system. These tests were focusing the simultaneous steam up- and water down flow phenomena at the upper tie plate, the fluid-fluid mixing in the cold leg and downcomer and the countercurrent flow conditions of steam and saturated water in a PWR-hot leg.     

Hydraulic flow tests of APWR reactor internals for safety analysis

A concept of radial neutron reflector of APWR brings about safety problems relevant to the flow induced vibration and thermal deformation. The CFD code has been expected to solve them by calculating pressure fluctuations of turbulent flow in the downcomer and the flow distribution into the neutron reflector. A series of hydraulic flow tests was conducted by NUPEC from 1998 to 2002 to demonstrate the new design of the neutron reflector and to obtain test data for validating the CFD code. The measured pressure fluctuations in the downcomer and their statistics were utilized for validating the specific turbulent model to be able to calculate a spectrum of pressure fluctuation such as the LES model. The measured flow rates at inlet holes of the lower core plate were utilized for validating for the general turbulent model, for example, the k– turbulent model. The calculated results with the LES model agreed well with the measured pressure fluctuations and their spectrum, but did not agree with the correlation between adjacent pressure fluctuations. On the other hand, the calculation results with the k– turbulent model agreed well with the measured flow rates at inlet holes of the lower core plate.     

Assessment of RAMONA-3B methodology with oscillatory flow tests

The instability event at the LaSalle County Plant (GE BWR-5) imposed a new challenge on the computer codes available for reactor transient analysis. While the codes were originally designed to predict non-oscillatory transients, the new requirement on the code is to model limit cycle oscillations with large amplitudes, where feed-back effects from the core and the balance of plant, and the nonlinear effects are significant. Two of the United States Nuclear Regulatory Commission's (USNRC) computer codes, namely RAMONA-3B/MODO [1] and HIPA-BWR of Engineering Plant Analyzer [2] were expected, and are shown in part in this paper, to meet the above demands.The RAMONA-3B/MOD1 has now been upgraded from the RAMONA-3B/MODO. It has a three dimensional neutron kinetics model, coupled to multi-channel nonequilibrium drift-flux formulation, and an explicit integration scheme for the thermal hydraulics.The accuracy of the thermohydraulics in the RAMONA-3B code was assessed for the new application by modelling oscillatory transients in the FR1GG test facilty. Nodalization studies showed that twenty-four axial nodes are sufficient for a converged solution; calculations with twelve axial nodes produce, in comparison to the 24-node calculation, the deviation of 4.4% in the peak gain of the power to flow transfer function.The code predicted correctly the effects of power and inlet subcooling on the transfer function gain and the system resonance frequency. For the six available tests modeled, the code-predicted peak gain differs from the experimentally obtained gain on the average by +7%, with the standard deviation of ±30%. The uncertainty in the experimental data lies between −11% and +12%. The difference between predicted and measured frequency at the peak gain on the average is −6%, with the standard deviation of ±14%.