Analysis of thermal stratification in the primary circuit of a VVER-440 reactor with the CFX code

The thermal stratification can lead an important role in the aging of the NPP piping because of the stresses caused by the temperature differences and the cyclic temperature changes. These stresses can limit the lifetime of the piping, or lead to penetrating cracks. For the stress analyses, the determination of the thermal hydraulic parameters of the stratified flow is necessary, which can be simulated by computational fluid dynamics (CFD) codes. The results of the simulation show the time development and the breaking up of the stratification and the temperature distribution of the stratified flow. The main difficulty of these CFD simulations is the uncertainty of the boundary conditions because of the unknown flow circumstances. In this paper, some results of CFX simulations are presented concerning the pressurizer surge line, and the injection pipe of the HPIS for VVER-440 type reactors.     

Monitoring MTC using cold leg thermocouple and background neutron detector noise signals in a VVER-440 reactor during a whole fuel cycle

Monitoring of the Moderator Temperature Coefficient (MTC) was performed from the noise signals of cold leg thermocouples and background neutron detectors in a VVER-440 type reactor during a whole fuel cycle. A modified traditional noise based estimator was applied: the estimator was extended in order to take into account the effects of measurement geometry, coolant velocity and the relatively long time constant of the thermocouples. A systematic evaluation of measurement settings and evaluation parameters was carried out in order to determine optimal parameters. Optimal evaluation parameters were determined by considering the frequency dependence of the estimator, and by minimizing the statistical and systematic errors of the results. It can be concluded that the modified estimator provides adequate results which are close to the MTC given by the core design code calculations. It was found that relatively long FFT window sizes are needed to obtain correct results. The method needs long but industrially acceptable measurements for robust operation.     

Analysis of loss-of-coolant and loss-of-flow accidents in the first wall cooling system of NET/ITER

This paper presents the thermal-hydraulic analysis of potential accidents in the first wall cooling system of the Next European Torus or the International Thermonuclear Experimental Reactor. Three ex-vessel loss-of-coolant accidents, two in-vessel loss-of-coolant accidents, and three loss-of-flow accidents have been analyzed using the thermal-hydraulic system analysis code RELAP5/MOD3. The analyses deal with the transient thermal-hydraulic behavior inside the cooling systems and the temperature development inside the nuclear components during these accidents. The analysis of the different accident scenarios has been performed without operation of emergency cooling systems. The results of the analyses indicate that a loss of forced coolant flow through the first wall rapidly causes dryout in the first wall cooling pipes. Following dryout, melting in the first wall starts within about 130 s in case of ongoing plasma burning. In case of large break LOCAs and ongoing plasma burning, melting in the first wall starts about 90 s after accident initiation.     

Crack initiation and arrest in a pressurized thermal shock test for a model pressure vessel made of VVER-440 reactor pressure vessel steel

A joint pressure vessel integrity research programme involving three partners is being carried out during 1990–1995. The partners are the Central Research Institute of Structural Materials “Prometey” from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Centre of Finland (VTT). The main objective of the research programme is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The programme is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing programme comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material.     

Detailed CFX-5 study of the coolant mixing within the reactor pressure vessel of a VVER-1000 reactor during a non-symmetrical heat-up test

As part of the reactor dynamics activities of FZK/IRS, the qualification of a detailed 3D CFD model of a reactor pressure vessel is a key step in safety evaluations for improving predictive capabilities and acceptability of commercial CFD tools in reactor physics. The VVER-1000 Coolant Transient Benchmark, initiated by OECD, represents an excellent opportunity for validation. In this work a CFD model for the complete VVER-1000 reactor pressure vessel is presented. Due to computational limits simplifications of the core and of some other geometrical details are introduced. The simulated scenario is the heat-up of one coolant loop in case of the isolation of a steam generator while the reactor is operating at a low power level. Two transient runs with a first and second order approximation of the spatial discretization are performed. Unexpectedly, the first order method reveals better agreement with measured data.     

The experimental definition of the acoustic standing waves series shapes, formed in the coolant of the primary circuit of VVER-440 type reactor

On the basis of pressure fluctuation measurements in some primary circuit loops at 2^nd Unit of Kola NPP with VVER-440 type reactors, the shapes of acoustic standing waves (ASW) were determined at frequencies corresponding to four minimal oscillation eigenfrequencies in the primary circuit coolant. On identification of the ASW modes and properties, experimental results based on six circulating loops in symmetric arrangement allowed determination of the three-dimensional space structure of the wave nodes and antinodes inside and outside of the reactor vessel (RV). As part of this analysis, the geometric features of the primary circuit that caused the formation of these standing waves were identified. Differences in each ASW shape were shown to cause different individual effects on the neutron field in the reactor core and on fuel assembly vibration. This has been partially confirmed by ex-core neutron ionization chamber noise analysis. One type of ASW, possessing an antinode inside the RV, can be used for measurement of the pressure coefficient of reactivity. However, this must be done with care to avoid the potential for incorrect results in some cases. The results presented in this paper can be readily extended to other VVER type reactors with both odd and even number of loops.     

Simulation of start-up measurements with the KIKO3D code and its application for determining the scram rod worth uncertainty in the VVER-440 reactor

The reasons for large discrepancies between the computed and measured values of the efficiency of control rods observed during start-up experiments on the Russian pressurized water type VVER reactors are discussed. The numerical simulation of the measurements including the prediction of the ex-core detector signals was used to resolve the discrepancies. The time and space dependent neutron flux in the core during these measurements have been calculated by the KIKO3D nodal kinetic code. For calculating the ionization chamber signals the Green function technique has been applied. The Green functions of ionization chambers have been evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals have been calculated and compared with measured ones using the inverse point kinetics transformation. Large number of asymmetric rod drop measurements (with one rod stuck) and some differential rod worth measurements from the Zero Power Physics Tests were provided by the Paks NPP for validation. The experiments cover different fuels (without and with enrichment zoning) and loading patterns. The intermediate range ionization chambers have been used during the scram measurements. The newly developed method provides fairly sufficient match of measured and calculated results. The time behavior of the detector readings observed in the measurements are described by the code in a consistent manner.As a further application the uncertainty of scram rod worth of the KARATE-440 code system was determined by static calculations and subsequent simulation of rod drop with the KIKO3D code. The calculated results were compared to measurements carried out by the Paks NPP. The uncertainty of scram rod worth is established by statistical analysis.     

Effect of deformation and loss-of-seal on the embrittlement of VVER fuel-element cladding in loss-of-coolant accidents

The structural-phase state and residual lifetime of the mechanical properties in the regions where VVER fuel element cladding has become unsealed after tests on the PARAMETR stand under loss-of-coolant accident conditions are investigated. It is determined that deformation engenders breaks in the oxide film which promote oxidation. The structural and strength characteristics in the region of unsealing exhibit strong azimuthal nonuniformity, which is due to the nature of the deformation, thinning of the cladding wall, and the degree of damage to the oxide as result of deformation. The results presented show that it is very important to examine the processes leading to the deformation of fuel element cladding and the subsequent behavior of the regions of unsealing in a corrosive medium from the standpoint of the formation of the most vulnerable zone, which is responsible for the residual lifetime of the entire fuel element. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 273–276, May, 2008.     

Experimental and analytical studies of melt jet-coolant interactions: a synthesis

Instability and fragmentation of a core melt jet in water have been actively studied during the past 10 years. Several models, and a few computer codes, have been developed. However, there are, still, large uncertainties, both, in interpreting experimental results and in predicting reactor-scale processes. Steam explosion and debris coolability, as reactor safety issues, are related to the jet fragmentation process. A better understanding of the physics of jet instability and fragmentation is crucial for assessments of fuel-coolant interactions (FCIs). This paper presents research, conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning molten jet-coolant interactions, as a precursor for premixing. First, observations were obtained from scoping experiments with simulant fluids. Second, the linear perturbation method was extended and applied to analyze the interfacial-instability characteristics. Third, two innovative approaches to computational fluid dynamics (CFD) modeling of jet fragmentation were developed and employed for analysis. The focus of the studies was placed on (a) identifying potential factors, which may affect the jet instability, (b) determining the scaling laws, and (c) predicting the jet behavior for severe accident conditions. In particular, the effects of melt physical properties, and the thermal hydraulics of the mixing zone, on jet fragmentation were investigated. Finally, with the insights gained from a synthesis of the experimental results and analysis results, a new phenomenological concept, named ‘macrointeractions concept of jet fragmentation’ is proposed.     

Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project “TACIS R2.02/02 - Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.     

Thermal–hydraulic characteristics of a next-generation reactor relying on steam generator secondary side cooling for primary depressurization and long-term passive core cooling

System experiments were conducted at the ROSA-V Large Scale Test Facility (LSTF) for investigation of new safety systems to mitigate consequences of postulated accidents in pressurized water rectors (PWRs). Tested systems included a steam generator (SG) secondary-side automatic depressurization system (SADS) and gravity-driven injection system (GDIS), which are candidates of safety systems for some next-generation PWR designs. The experimental results showed several thermal–hydraulic behaviors typical of these safety systems, including the primary depressurization due to natural circulation cooling, a nonuniform flow behavior among SG U-tubes, an accumulation of the non-condensable gas originally contained in the injected water, liquid holdup in U-tubes due to the countercurrent flow limiting, and long-term passive core cooling with the GDIS injection. From the assessment of the RELAP5/MOD3 code using the present data, it was found that the inability of the code to predict the U-tube nonuniform flow behavior resulted in overprediction of the primary depressurization rate at a pressure less than 1 MPa, and exaggerated oscillation of the natural circulation flow rate in the primary loop.     

Radioecological studies of a construction site in Sayda Guba for long-term storage of reactor compartments

Specialists at the Russian Science Center Kurchatov Institute have performed radioecological studies on a construction site in Sayda Guba (Kol’skiy Zaliv) for long-term storage of reactor compartments. The results of these investigations are presented. Detailed investigations have shown that at the present time all radioecological parameters (γ-radiation dose rate on site, soil radon flux density, specific content of radionuclides in natural objects of plant and animal origin) fall within the natural background range. No technogenic anomalies have been found. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 276–281, March, 2007.     

Experimental studies on the upward fuel discharge for elimination of severe recriticality during core-disruptive accidents in sodium-cooled fast reactors

In order to eliminate the energetic potential in the case of postulated core-disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner-duct structure (FAIDUS) has been considered. Recently, a design option of FAIDUS which leads molten fuel to upward discharge has been considered as the reference core design of the Japan Sodium-Cooled Fast Reactor (JSFR). In this study, a series of experiments which consisted of three out-of-pile tests and one in-pile test were conducted to obtain experimental knowledge of the upward discharge of molten fuel. Experimental data which showed a sequence of upward fuel discharge and effects of initial pressure conditions on upward discharge were obtained through the out-of-pile and in-pile test. Preliminary extrapolation of the present results to the supposed condition in the early phase of the CDA in the JSFR design suggests that the sufficient upward flow rate of molten fuel is expected to prevent the core melting from progressing beyond the fuel subassembly scale and that the upward discharge option will be effective in eliminating the energetic potential.     

Further studies on the inhibition by boric acid of the oxidation of zirconium in high pressure steam

It is known that the rapid oxidation of zirconium in steam at 500° C, 1000 psi is prevented by adding boric acid to the steam. Inhibition has now been obtained for up to 400 days exposure. Zircaloy 2 however does not show any improvement in inhibited steam. Some alloys, having poor oxidation resistance in steam at 500° C, have been improved in inhibited steam, although the effect is not so great as with zirconium. Alloys resistant to steam are not greatly affected.The minimum effective concentration of boric acid could be as low as 0.003 g/g of steam. No inhibition takes place at 500° C, 1 atm, or at 400° C, 1000 psi.Anodising in borate solutions does not inhibit subsequent oxidation in 1 atm or 1000 psi steam at 500° C.