Modeling of flashing-induced instabilities in the start-up phase of natural-circulation BWRs using the two-phase flow code FLOCAL

This paper reports on the modeling and simulation of flashing-induced instabilities in natural-circulation systems, with special emphasis on natural-circulation boiling water reactors (BWRs). For the modeling the 4-equation two-phase model FLOCAL [Rohde, U., 1986. Ein teoretisches Modell fur Zweiphasen-stromungen in wassergekulthen Kernreaktoren und seine Anwendung zur Analyse des Naturumlaufs im Heizreaktor AST-500. Ph.D. dissertation, Akademie der Wissenschaften der DDR, Dresden], developed at the Forschungszentrum Rossendorf (FZR, Germany), has been used. The model allows for the liquid and vapor to be in thermal non-equilibrium and, via drift-flux models, to have different velocities.The phenomenology of the instability has been studied and the dominating physical effects have been determined. The results of the simulations have been compared qualitatively and quantitatively with experiments [Manera, A., van der Hagen, T.H.J.J., 2003. Stability of natural-circulation-cooled boiling water reactor during start up: experimental results. Nuc. Technol., 143] that have been carried out within the framework of a European project (NACUSP) on the CIRCUS facility. The facility, built at the Delft University of Technology in The Netherlands, is a water/steam 1:1 height-scaled loop of a typical natural-circulation-cooled BWR.     

Validation of the engineering plant analyzer methodology with peach bottom 2 stability tests

The Engineering Plant Analyzer (EPA) had been developed in 1984 at Brookhaven National Laboratory to simulate plant transients in boiling water reactors (BWR). Recently, the EPA with its High-Speed Interactive Plant Analyzer code for BWRs (HIPA-BWR) simulated for the first time oscillatory transients with large, non-linear power and flow amplitudes; transients which are centered around the March 9, 1988 instability at the LaSalle-2 BWR power plant.The EPA's capability to simulate oscillatory transients has been demonstrated first by comparing simulation results with LaSalle-2 plant data (Wulff et al., NUREG/CR-5816, BNL-NUREG-52312, Brookhaven National Laboratory, 1992). This paper presents an EPA assessment on the basis of the Peach Bottom 2 instability tests (Carmichael and Niemi, EPRI NP-564, Electric Power Research Institute, Palo Alto,CA, 1978). This assessment of the EPA appears to constitute the first validation of a time-domain reactor systems code on the basis of frequency-domain criteria, namely power spectral density, gain and phase shift of the pressure-to-power transfer function.The reactor system pressure was disturbed in the Peach Bottom 2 power plant tests, and in their EPA simulation, by a pseudo-random, binary sequence signal. The data comparison revealed that the EPA predicted for Peach Bottom tests PT1, PT2, and PT4 the gain of the power-to-pressure transfer function with the biases and standard deviations of (−10 ± 28)%, (−1 ± 40)% and (+28 ± 52)%, respectively. The respective frequencies at the peak gains were predicted with the errors of +6%, +3%, and −28%. The differences between the predicted and the measured phase shift increased with increasing frequency, but stayed within the margin of experimental uncertainty.The code assessment presented here is valid only for small-amplitude oscillations, but it encompasses neutron kinetics, fuel thermal response, coolant thermohydraulics and control-system dynamics. To our knowledge, this assessment of the time-domain HIPA-BWR code by frequency-domain methods and spectral plant data demonstrates for the first time the feasibility of such an assessment.     

Dynamic analysis of multiple nuclear-coupled boiling channels based on a multi-point reactor model

This work investigates the non-linear dynamics and stabilities of a multiple nuclear-coupled boiling channel system based on a multi-point reactor model using the Galerkin nodal approximation method. The nodal approximation method for the multiple boiling channels developed by Lee and Pan [Lee, J.D., Pan, C., 1999. Dynamics of multiple parallel boiling channel systems with forced flows. Nucl. Eng. Des. 192, 31–44] is extended to address the two-phase flow dynamics in the present study. The multi-point reactor model, modified from Uehiro et al. [Uehiro, M., Rao, Y.F., Fukuda, K., 1996. Linear stability analysis on instabilities of in-phase and out-of-phase modes in boiling water reactors. J. Nucl. Sci. Technol. 33, 628–635], is employed to study a multiple-channel system with unequal steady-state neutron density distribution. Stability maps, non-linear dynamics and effects of major parameters on the multiple nuclear-coupled boiling channel system subject to a constant total flow rate are examined. This study finds that the void-reactivity feedback and neutron interactions among subcores are coupled and their competing effects may influence the system stability under different operating conditions. For those cases with strong neutron interaction conditions, by strengthening the void-reactivity feedback, the nuclear-coupled effect on the non-linear dynamics may induce two unstable oscillation modes, the supercritical Hopf bifurcation and the subcritical Hopf bifurcation. Moreover, for those cases with weak neutron interactions, by quadrupling the void-reactivity feedback coefficient, period-doubling and complex chaotic oscillations may appear in a three-channel system under some specific operating conditions. A unique type of complex chaotic attractor may evolve from the Rossler attractor because of the coupled channel-to-channel thermal-hydraulic and subcore-to-subcore neutron interactions. Such a complex chaotic attractor has the imbedding dimension of 5 and the fractal dimension ranging from 1.26 to 1.35.     

Closure of the direct containment heating issue for Zion

This paper, which was originally published in more detail (M.M. Pilch, M.D. Allen, D.L. Knudsen, D.W. Stamps and E.L. Tadios, Rep. NUREG/CR-6075, Supplement 1, 1994b (Sandia National Laboratories, Albuquerque, NM)), provides closure of the direct containment heating (DCH) issue for the Zion plant. It incorporates the comments and suggestions of the peer reviewers of NUREG/CR-6075 (M.M. Pilch, H. Yan, and T.G. Theofanous, Rep. NUREG/CR-6075, SAND93-1535, 1994a (Sandia National Laboratories, Albuquerque, NM)) and specifically includes assessments of four new splinter scenarios defined in working group meetings and modeling enhancements recommended by the working groups. In the four new scenarios, consistency of the initial conditions has been implemented by using insights from systems-level codes. was used to analyze three short-term station blackout cases with different leak rates. In all three cases, the hot leg or surge line failed well before the lower head and thus the primary system depressurized to a point where DCH was no longer considered a threat. However, these calculations were continued to lower head failure in order to gain insights that were useful in establishing the initial and boundary conditions. The most useful insights are that the reactor coolant system pressure is low at vessel breach, metallic blockages in the core region do not melt and relocate into the lower plenum, and melting of upper plenum steel is correlated with hot leg failure. The output was used as input to to assess the containment conditions at vessel breach. The containment-side conditions predicted by are similar to those originally specified in NUREG/CR-6075.The methodology originally developed in NUREG/CR-6075 (M.M. Pilch, H. Yan, and T.G. Theofanous, Rep. NUREG/CR-6075, SAND93-1535, 1994a (Sandia National Laboratories, Albuquerque, NM)) was used to analyze the new splinter scenarios. Some modeling enhancements in response to working group discussions were implemented for these analyses. The entrainment of hydrogen pre-existing in the atmosphere into a burning jet was examined more carefully. In addition, the impact of DCH-induced deflagrations on DCH loads was quantified. A new computational tool—the two-cell equilibrium—Latin hypercube sampling (TCE-LHS) code—was developed for this effort to perform Monte Carlo sampling of the scenario distributions. The TCE-LHS code was benchmarked against the original Scenario I calculations in NUREG/CR-6075 performed using the code, which is based on the method of discrete probability distributions. The results were in excellent agreement.The analyses of the new scenarios showed no intersection of the load distributions and the containment fragility curves, and thus the containment failure probability was negligible for each scenario. These supplemental analyses complete closure of the DCH issue for Zion.     

Annular flow entrainment rate experiment in a small vertical pipe

Two-fluid model predictions of film dryout in annular flow, leading to nuclear reactor fuel failure, are limited by the uncertainties in the constitutive relations for the entrainment rate of droplets from the liquid film. The main cause of these uncertainties is the lack of separate-effects experimental data in the range of the operating conditions in nuclear power reactors. An air–water experiment has been performed to measure the entrainment rate in a small pipe. The current data extend the available database in the literature to higher gas and liquid flows and also to higher pressures. The measurements were made with the film extraction technique. A mechanistic model was obtained based on Kelvin–Helmholtz' instability theory. The dimensionless model includes the Weber number of the gas and the liquid film Reynolds number. Kataoka and Ishii's correlation (Kataoka, I., Ishii, M., 1982. NUREG/CR-2885, ANL-82-44) is modified based on this model and the new data. The new correlation collapses the present air–water data and Cousins and Hewitt's data (Cousins, L.B., Hewitt, G.F., 1968. UKAEA Report AERE-R5657) The effects of pressure and surface tension were considered in the derivation so it may be applied for boiling water reactor operating conditions.     

Scaling analysis for the OSU AP600 test facility (APEX)

In this paper, the authors summarize the key aspects of a state-of-the-art scaling analysis (Reyes et al., 1995. Westinghouse Electric Corporation, WCAP-14270) performed to establish the facility design and test conditions for the Advanced Plant Experiment (APEX) at Oregon State University (OSU). This scaling analysis represents the first, and most comprehensive, application of the Hierarchical Two-Tiered Scaling (H2TS) Methodology (Zuber, 1991. US Nuclear Regulatory Commission, Washington DC, NUREG/CR-5809) in the design of an integral system test facility. The APEX test facility, designed and constructed on the basis of this scaling analysis, is the most accurate geometric representation of a Westinghouse AP600 nuclear steam supply system. The OSU APEX test facility has served to develop an essential component of the integral system database used to assess the AP600 thermal hydraulic safety analysis computer codes.     

FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications

This paper is devoted to new numerical methods developed for three-dimensional two-phase flow calculations. These methods are finite volume numerical methods. They are based on an extension of Roe’s approximate Riemann solver to define convective fluxes versus mean cell quantities [Godunov, S.K., 1959, Math. Sb. 47, 217; Roe, P.L., 1981, Approximate Riemanns solvers parameter vectors and difference scheme. J. Comp. Phys. 43, 357–372; Toumi, I., 1992, A weak formulation of Roe’s approximate Riemann solver. J. Comp. Phys. 102, 360–373]. To go forward in time, a linearized conservative implicit integrating step is used [Yee, H.C., 1987. NASA TM-89464], together with a Newton iterative method. We also present here some improvements performed to obtain a fully implicit solution method that provides fast running steady state calculations. This kind of numerical method, which is used widely for fluid dynamic calculations, has proved to be very efficient for the numerical solution to two-phase flow problems. This numerical method has been implemented for the three-dimensional thermal-hydraulic code FLICA-4 that is mainly dedicated to core thermal-hydraulic transient and steady-state analysis [Toumi, I., Caruge, D., 1998. An implicit second order method for 3D two phase flow calculations. Nucl. Sci. Eng. 130, 213–225; Raymond, P., Toumi, I., 1992. Numerical method for three-dimensional steady-state two-phase flow calculation, NURETH-5, Salt Lake City]. Hereafter, elements of physical validation against hydraulic and two-phase flow rod bundle experiments are presented. We will also find some results obtained for the EPR reactor running in a steady-state at 60% of nominal power with three pumps out of four, and a thermal-hydraulic core analysis for a 1300 MW PWR at low flow Steam-Line-Break conditions.     

Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation in an extended operation domain with increased void and thereby increased void reactivity feedback and which often have thinner fuel rods and thereby decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of “unexpected” instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, the subject of BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a “new and improved” state of the art has emerged recently.     

Reactor cavity and ALS thermal-hydraulic evaluation in the case of fuel channels ruptures at Ignalina NPP

This paper presents the work analysis of the thermal-hydraulic parameters behavior in the RBMK-1500 reactor cavity (RC) and other connected volumes in the case of fuel channels ruptures. The analysis is performed with CONTAIN code using the models of accident localization system (ALS) and reactor cavity venting system (RCVS). The RCVS capacity is assessed and expressed as a number of ruptured fuel channels at which the integrity of RC is maintained. The uncertainty analysis of pressure behavior in RC during multiple fuel channel rupture was performed. The initial and boundary conditions and the code models were selected and their influence on the results is estimated.Calculation of coolant mass and energy release to the reactor cavity in case of fuel channels rupture performed using the main circulation circuit model of Ignalina NPP, which was developed by employing state-of-the-art code RELAP5/MOD3.2 [Fletcher et al., RELAP5/MOD3 code manual user’s guidelines, Idaho National Engineering Lab., NUREG/CR-5535 (1992)]. These results were applied further as the initial data for the analysis of the thermal-hydraulic parameters behavior in the affected compartments employing CONTAIN code.     

Experimental study of debris head loss through a pressurized water reactor recirculation sump screen after LOCA

A loss of coolant accident (LOCA) in a PWR (pressurized water reactor) would generate debris from thermal insulation and other materials in the vicinity of the break. A fraction of the LOCA-generated debris and pre-LOCA debris would be transported into the sump and accumulated on the sump screens resulting in adverse blockage effects that include decrease in net positive suction head (NPSH) margin. The NUREG/CR-6224 head loss correlation has been widely used to estimate debris-induced head loss across sump screens, but it has been recommended to be used for debris thicknesses less than 4 in. In order to extend the limit, head loss data are obtained over a wider range of bed thicknesses. Experimental results show that the NUREG/CR-6224 correlation conservatively predicts the head loss across NUKON™ debris beds with theoretical thicknesses up to 6 in. Head loss measurements in mixed debris beds show that fibrous debris with calcium-silicate and/or fine particulates such as coatings mainly deteriorates the NPSH.     

高温气冷堆停堆保护系统的多样性分析

为评价高温气冷堆(HTR)停堆保护系统的多样性特征，基于NUREG/CR-6303的分析方法，通过导则中D3评估方法来确定必需的多样性，并采用NUREG/CR-7007的多样性量化评估方法，分析并识别出停堆保护系统7大多样性属性的25条因素值，计算出标准化的多样性量化值。针对系统多样性存在的薄弱点及工程的实际情况，提出了可行的改进方案。重新核算结果表明，改进方案能有效提升系统的多样性量化值。     

Small-scale simulation of vapor discharges into subcooled liquid pools

The methodology proposed by Sonin [Nucl. Engrg. Des. 65 (1981) 17–21] for small-scale modeling of high-flux vapor discharges into subcooled pools is verified experimentally using sonic steam discharges into water. Small-scale simulation is used to show that the dynamic pressures induced by sonic steam discharges increase with decreasing subcooling when the subcooling is high, reach a maximum at a pool subcooling of 20–30 K or greater, and then decrease with further reductions in subcooling, becoming very low as the subcooling approaches zero. Condensation remains complete to very low subcooling. Data are presented for both straight pipe discharges and model devices simulating a BWR quencher.     

Experimental data on burn-out in a simulated BWR fuel bundle

Critical heat flux (CHF) experiments have been carried out on a 16-rod test section having the typical geometry of boiling water reactor (BWR) fuel elements and in particular a 366 cm length. Heat fluxes were uniform, both axially and radially. The tests were carried out for the CNEN Plutonium Program on CISE's 8 MW IETI-3 facility, at 71 kg/cm² abs, mass velocities of 12–200 g/cm² s and inlet sub-cooling of 15–180°C. Each corner rod was instrumented with four separate thermocouples to detect nnd locate the initiation of CHF, while the other rods were instrumented with four-junction thermopiles.     

基于R2S方法的反应堆结构材料活化剂量计算研究

反应堆结构材料在堆芯中子辐照下由于中子活化反应而产生大量的放射性核素,其衰变光子是反应堆停堆检修、换料、退役过程中工作人员职业照射剂量的重要来源。本文基于严格两步法（R2S）,研究了反应堆结构材料栅元活化计算方法,并基于蒙卡粒子输运程序（MCNP）与点活化计算程序（ORIGEN）建立了反应堆结构材料活化剂量计算软件（MOCA）。通过开发功能接口与数据接口程序实现输运程序与活化计算程序的自动耦合,进而实现“中子输运-活化分析-剂量计算”全自动耦合分析。利用M5包壳活化计算模型、不锈钢活化计算模型和NUREG/CR-6115压水堆模型对MOCA进行基准验证,证明了MOCA的正确性与可靠性。      

An advanced reduced order model for BWR stability analysis

An advanced reduced order model was developed and qualified in the framework of a novel approach for nonlinear stability analysis of boiling water nuclear reactors (BWRs). This approach is called the RAM-ROM method where RAM is a synonym for system code and ROM stands for reduced order model. In the framework of the RAM-ROM method, integrated BWR (system) codes and reduced order models are used as complementary tools to examine the stability characteristics of fixed points and periodic solutions of the nonlinear differential equations describing the stability behaviour of a BWR loop. This methodology is a novel one in a specific sense: we analyse the highly nonlinear processes of BWR dynamics by applying validated system codes and by the sophisticated methods of nonlinear dynamics, e.g. bifurcation analysis. We claim and we will show that the combined application of independent methodologies to examine nonlinear stability behaviour can increase the reliability of BWR stability analysis.This work is a continuation of previous work at the Paul Scherrer Institute (PSI, Switzerland) of the second author and at the University of Illinois (USA) in this field. In the scope of a PhD work at the Technical University Dresden (Germany), the current ROM was extended to an advanced ROM by adding a recirculation loop model, a quantitative assessment of the necessity for consideration of the effect of sub-cooled boiling and a new calculation methodology for feedback reactivity. A crucial point of ROM qualification is a new calculation procedure for ROM input data based on steady-state RAM (ONA) results. The modified ROM is coupled with the BIFDD bifurcation code which performs a semi-analytical bifurcation analysis (see Appendix C). In this paper, the advanced ROM (TU Dresden ROM, TUD-ROM) is briefly described and the results of a nonlinear BWR stability analysis based on the RAM-ROM method are summarised for NPP Leibstadt, NPP Ringhals and NPP Brunsbüttel. The results show that the TUD-ROM including the new approach for ROM input data calculation is qualified for BWR stability analysis in the framework of the RAM-ROM method.     

Analysis of large break LOCA experiment at ROSA-III

Analysis of the ROSA-III test RUN 704 was performed by using the computer codes RELAP4J, RELAP4/MOD6 and RELAP5/MOD0 to verify the predictive capability of the codes for a BWR LOCA. The ROSA-III facility is a volumetrically scaled (1/424) BWR system with an electrically heated core, designed for in tegral LOCA/ECCS tests. The RUN 704 experiment at the ROSA-III test facility simulated a 200% double-ended offset shear break on the inlet side of the pump in the recirculation loop. From present analyses, key parameters which are important to predict major behavior during a BWR large break LOCA have been clarified and the promising predictive capability of the advanced code RELAP5 has been verified.     

Prediction of turbulent mixing rates of both gas and liquid phases between adjacent subchannels in a two-phase slug-churn flow

This paper presents a slug-churn flow model for predicting turbulent mixing rates of both gas and liquid phases between adjacent subchannels in a BWR fuel rod bundle. In the model, the mixing rate of the liquid phase is calculated as the sum of the three components, i.e. turbulent diffusion, convective transfer and pressure difference fluctuations between the subchannels. The components of turbulent diffusion and convective transfer are calculated from Sadatomi et al.'s [Nucl. Eng. Des. 162 (1996) 245–256] method, applicable to single-phase turbulent mixing, by considering the effect of the increment of liquid velocity due to the presence of gas phase. The component of the pressure difference fluctuations is evaluated from a newly developed correlation. The mixing rate of the gas phase, on the other side, is calculated from a simple relation of mixing rate between gas and liquid phases. The validity of the proposed model has been confirmed with the turbulent mixing rates data of Rudzinski et al. [Can. J. Chem. Eng. 50 (1972) 297–299] as well as the present authors.     

A new finite cloud method for calculating external exposure dose in a nuclear emergency

A new finite cloud method (5/μ method) for calculating external exposure dose in a nuclear emergency is presented in this paper. The method calculates external exposure dose over a specially constructed three-dimensional columned space, whose underside center is the location of the receptor and underside radius and height are both five times mean free path of a gamma-photon. Then, the space is divided into many grid cells for integral to calculate external exposure dose (or dose rate). The calculation values of air external exposure dose rate conversion factors and air-absorbed dose rate conversion factors by the 5/μ method are accordant with the values presented in related references. Comparing with the discrete point approximation method (DPA) [USNRC, The MESORAD Dose Assessment Model. NUREG/CR-4000 Vol. 1, 1986] and the Nomogram method [USNRC, Nomogram for Evaluation of Doses from Finite Noble Gas Clouds, NUREG-0851, 1983], which are two traditional finite cloud methods for calculating external exposure dose, the 5/μ method has a distinct advantage of more fast calculation speed, which is very important in a nuclear emergency. What is more, the 5/μ method can be applied together with three-dimensional atmospheric dispersion models.