Vessel Head Research Articles

Analysis of an RPV upper head SBLOCA at the ROSA facility using TRACE

Inspections of existing nuclear power plants have pointed out the possibility that small break loss-of-coolant accidents (SBLOCAs) could be initiated by a small break located in the upper head (UH) of the reactor pressure vessel (RPV). Such type of breaks has been the subject of investigation in some of the tests carried out in the framework of the OECD/NEA ROSA test program for safety research and safety assessment of light water reactors. The ROSA/LSTF test facility simulates a Westinghouse design PWR with a four-loop configuration and 3423 MWth. Areas, volumes and power are scaled down by a factor of 1:48 while the elevations are kept at full height. Only two loops, sized to conserve the volume scaling (2:48), are simulated. The present paper is focused on test 6-1 that simulated a RPV upper head SBLOCA with a break size equivalent to 1.9% cold leg break. The experiment assumes a total failure of the high pressure injection system (HPIS) and a loss of off-site power concurrent with the scram. The main purpose of the present study is the assessment of the capabilities of the best estimate system code, TRACE, to reproduce and understand the physical phenomena involved in this type of SBLOCA scenarios. Special attention was dedicated to the modelling of the leakage flows, necessary to correctly simulate the distribution of the water inventory in the primary side. In addition, the particular location of the break in test 6-1 allows the verification of the chocked flow model in the same way as for a separate-effect test.

Nuclear Power – A Panacea for Future Energy Needs?

Each week seems to bring further evidence that the Earth is warming at a faster rate than previously estimated. Pressure is building to replace power sources that emit carbon dioxide with those tha...

Study on Spinning Process of a Thin-Walled Aluminum Alloy Vessel Head With Small Ratio of Thickness to Diameter

The thin-walled vessel head with the ratio of thickness to diameter less than 3‰ has long been considered to be difficult to be spun because wrinkling is very likely to occur during the thin-walled vessel head spinning process when the thickness is far smaller than the diameter. Based on process experiments and finite element method, the spinning failure of thin-walled vessel head with a small ratio of thickness to diameter is analyzed in the present research. The mechanism of wrinkling is identified and some effective solutions are discussed to prevent the failure. The results show that the feed ratio, the blank geometry, and the roller trajectory are the main factors influencing the spinning qualities. In the shear spinning, the feasible roller feed ratio is found to be within a very small range because of the thin thickness of blanks. Wrinkling will occur if the feed ratio is slightly outside the operation range. Bending the edge of blank or enlarging the blank size can effectively prevent wrinkling at a larger feed ratio, which would increase the operation range of roller feed ratio. Due to the fact that the conventional spinning is a process of multiple passes, there are many factors affecting the forming quality of thin-walled aluminum alloy vessel head. Wrinkling is likely to happen by the influence of roller trajectory in the first pass due to the fact that the thickness of blank is far smaller than the diameter. The straight-line trajectory is the worst trajectory under which wrinkling is most likely to occur.

Stress Intensity Effect on Solid State Oxidation of Ni-Cr Alloy with Different Chromium Concentrates

The life extension of aging nuclear power plant is a challenge due to the influence of various factors on the materials properties. Ni-base alloys are widely used in light water reactor component and stress corrosion cracking (SCC) has been found in pressurized water reactor (PWR) components such as vessel head penetrations and bottom mounted instrument nozzles fabricated with Alloy 600 and its weld metals [1,2]. Ni-base alloys composed of several elements among them nickel and chromium are the major ingredients. X-ray photoelectron spectroscopy (XPS) study of Ni-Cr alloy examined that Cr is oxidized preferentially to Cr2O3 initially but that Ni in the alloy is oxidized more rapidly to NiO than Ni metal [3]. Alloy 600 XPS study found the chromite-rich inner layer due to inward oxygen penetration and simultaneously with outward migration of nickel [4]. However, some theoretical and experimental studies have been carried out on Ni/H2O and oxygen preadsorbed Ni/H2O to understand water dissociation and adsorption phenomena [5,6].Over the years various attempts have been made to obtain mechanism of SCC but it still require more fundamental study to interpret clearly. This study presents an approch based on the multiscale modeling, to assess the influence of alloy composition and stress intensity (K) on the initial stage solid state oxidation of the Ni-Cr binary alloy. The multiscale modeling considers different length scales such as finite element method (FEM) – quasicontinuum (QC)-quantum chemical molecular dynamics (QCMD), for analyzing crack tip molecular domain.

The Development of Large High Efficient CNC Gantry Grinder for Vessel Head in Nuclear Power

This paper describes the current processing problem of China's nuclear power vessel head, bring forward a method which can process groove and welding seam, eliminate welding root and polish surface with high efficiency deep grinding wheel and heavy duty abrasive belt grinding technology by CNC in a grinder, expatiates this new large gantry grinding machine structure design and CNC key techniques, finally summarizes this new grinding techniques and machine.

On the effect of the Poisson's ratio (positive and negative) on the stability of pressure vessel heads

AbstractMaterials with negative Poisson's ratio (auxetic) exhibit an unusual property of expanding when stretched in the direction normal to an applied uniaxial tension and vice versa when compressed, a phenomenon which is known to result in many beneficial effects in the performance of materials in various practical situations. In particular, it has long been suggested that spherical shells made from isotropic materials having Poisson's ratio less than −0.5 exhibit enhanced resistance to buckling as a result of externally applied normal pressure. This work uses finite elements (FEs) modelling to assess the behaviour of various spherical, torispherical and ellipsoidal shells when they are subjected to external (in all cases) or internal (in the case of torispherical and ellipsoidal shells) pressures. We find that to a first approximation, the critical buckling pressures scale linearly with (1 − ν2)−1/2 thus suggesting that the critical buckling pressures tends to infinity as ν tends to −1, this being in accordance to what was known for spherical shells. We also find that the Poisson's ratio has an effect on the amplitude and the number of buckling wavelengths that occur when the shells buckle.

Optimal Offloading Configuration of Spread-Moored FPSOs

As the oil offloading operations of floating production storage and offloading (FPSO) units become more routine, the desire grows to increase the availability for offloading and thus decrease production downtime. Experience with these operations is the main tool available to increase the efficiency of this aspect of deepwater production. However, it is clear that a formal optimization approach can help to fine tune design parameters so that not only is availability increased but the significance of each design parameter can be better understood. The key issue is to define the environmental conditions under which the vessels involved in offloading are able to maintain position. By this, we reduce the notion of availability to a set of operating criteria, which can or cannot be met for a particular set of environmental conditions. The actual operating criteria such as relative vessel heading depend on selection of design parameters, such as the direction and magnitude of external force applied by thrusters or tugs. In the earliest offloading operations, engineering judgment was used to determine the feasibility of offloading at a particular time. For example, if wind and current were not expected to exceed a 1year return period, offloading may be considered safe. This approach can be both conservative and unconservative, depending on the nuances of the particular environmental conditions. This study will propose a formal approach to choosing the design parameters that optimize the availability of a FPSO for offloading. A simple analysis model will be employed so that optimization can be performed quickly using a robust second order method. The proposed analysis model will be compared to model test data to demonstrate its agreement with the more complex system.

Spinning of thin metal plate with diode laser

Spinning with diode laser is newly proposed and experiments on AISI304, magnesium and titanium alloy plates are carried out. The sizes of the plates are 0.3, 0.4 and 0.5mm in thickness and 70 mm in diameter. The laser is scanned on lines of 1, 2, and 3mm away from the circumference. An AISI304 plate of 0.3mm thick can be formed into a shape of head of vessel of the diameter of 66mm, and a height of 2.0mm. The bending angle is about 90 degrees. Titanium alloy plates are difficult to form. Magnesium alloy plates are easy to form; however, the surface layer occurs to melt.

Study of spray cooling of a pressure vessel head of a boiling water reactor

The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in a Boiling Water Reactor (BWR). To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate.

Design by analysis of ductile failure and buckling in torispherical pressure vessel heads

Thin shell torispherical pressure vessel heads are known to exhibit complex elastic–plastic deformation and buckling behaviour under static pressure. In pressure vessel Design by Analysis, the designer is required to assess both of these behaviour modes when specifying the allowable static load. The EN and ASME boiler and pressure vessel codes permit the use of inelastic analysis in design by analysis, known as the direct route in the EN Code. In this paper, plastic collapse or gross plastic deformation loads are evaluated for two sample torispherical heads by 2D and 3D FEA based on an elastic-perfectly plastic material model. Small and large deformation effects are considered in the 2D analyses and the effect of geometry and load perturbation are considered in the 3D analysis. The plastic load is determined by applying the ASME twice elastic slope criterion of plastic collapse and an alternative plastic criterion, the Plastic Work Curvature criterion. The formation of the gross plastic deformation mechanism in the models is considered in relation to the elastic–plastic buckling response of the vessels. It is concluded that in both cases, design is limited by formation of an axisymmetric gross plastic deformation in the knuckle of the vessels prior to formation of non-axisymmetric buckling modes.

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.

Results of the QUENCH-L2, DISCO-L2, and COMET-L2 experiments performed within the LACOMERA project at the Forschungszentrum Karlsruhe

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany (FZK) is a 4-year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU Member Countries and Associated States access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities are being used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarizes the main results obtained in the following three experiments: QUENCH-L2: boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. DISCO-L2: fluid-dynamic, thermal, and chemical processes during melt ejection out of a breach in the lower head of a pressure vessel of the VVER-1000/320 type of reactor. COMET-L2: investigation of long-term melt–concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase.

In-Vessel Retention of Molten Corium: Lessons Learned and Outstanding Issues

In-vessel retention (IVR) of core melt is a key severe-accident-management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (LWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the advanced 600 MW(electric) pressurized water reactor (AP600) designed by Westinghouse, which relied upon external reactor vessel cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission approving the design without requiring that certain features common to existing LWRs, such as containment sprays, be safety related. Clearly, ERVC offers the potential to reduce the AP600’s construction and operating costs. However, it is not clear that the ERVC proposed for the AP600 could provide sufficient heat removal for higher-power reactors [up to 1500 MW(electric)] without additional enhancements. This paper reviews efforts made and results reported regarding the enhancement of IVR in LWRs. Where appropriate, the paper identifies what additional data or analyses are needed to demonstrate that there is sufficient margin for successful IVR in high-power thermal reactors.

B202 原子力の安全規制の最適化に関する研究会 : (2) 第2次米国調査の概要(安全規制の最適化,OS-8 軽水炉・新型炉・核燃料サイクル(2),一般講演,地球温暖化防止と動力エネルギー技術)

This technical research section is surveying the nuclear regulation in the major countries. The first one was to visit Finland and France on January 2007, the second one was to visit US-NRC Head Quarter, Browns Ferry NPS and Hatch NPS on Aug 2007, the third one was to visit Switzerland and Sweden on January 2008, the fourth one was to visit US-NRC Region III, Davis Besse NPS and Quad Cities NPS on Oct 2007. This report summarized the survey results of the fourth visit (second US visit) Davis Besse NPS is so famous about one of the most scary incident which was the big hole in the reactor vessel head by the corrosion by the leakage of the boron. But the plant manager said they completely changed the safety culture and it is the safest nuclear power plant in the world right now. We found their sincere change of the culture and amazed the good results.

Extreme sloshing and whipping-induced pressures and structural response in membrane LNG tanks

The long-term extreme pressure in the membrane LNG tank and structural response of the Mark III containment system are addressed. The effect of hull slamming-induced vibrations on the vertical acceleration is investigated and found to be important in certain situations. Determining structural response due to sloshing requires a stepwise approach considering long-term variation of the sea state, ship motion in a stochastic seaway, fluid motion in the tanks, pressures acting on the tank structure, structural load effects, and their comparison with the appropriate resistance. This paper is focused on the determination of the critical conditions to limit the number of combinations of sea states, vessel heading, and speed that need to be considered in estimating extreme sloshing pressures. In this context, an approximate method is introduced and validated. These sea states are determined by a full linear long-term analysis. The response quantiles suitable for the contour line approach are also found using simplified criteria. The effect of avoiding heavy weather in vessel operation on sloshing and whipping-induced pressures is investigated. Sloshing experiments and various scaling approaches are described. Application of the generalized Pareto distribution and correctly determining its shape parameter is discussed. Structural response of the LNG membrane tank wall is investigated. The importance of accounting for the steel flexibility in calculating sloshing response is shown. A new method for approximating the time history of sloshing impact pressure is suggested.

Fitness-for-service assessment of spherical pressure vessels with hot spots

Spherical shapes are used in industry as hemispherical vessel heads or high-pressure storage vessels due to the inherent strength associated with the shape. Structural integrity of such components needs to be evaluated periodically to prevent failure of the vessels under operating conditions. The paper develops a method for Level 2 (as defined by API 579 [(2000). Fitness-for-service, API 579. Washington, DC: American Petroleum Institute]) fitness-for-service estimation of spherical shapes subject to local hot spots where the temperatures are elevated due to local damage. The decay length for spherical shells is determined, and the size of hot spot to be identified as local is proposed. A lower bound “remaining strength factor” (RSF) for spherical pressure vessels containing hot spots is formulated by the application of Mura's variational formulation and the m α -multiplier method. The effectiveness of the proposed Level 2 method is evaluated and demonstrated through an example.

Thermodynamic Analysis for Molten Corium Stratification Test MASCA with Ionic Liquid U-Zr-Fe-O-B-C-FPs Database

The molten corium stratification tested in the OECD MASCA project was analyzed with our thermodynamic database and the database was verified to be effective for the stratification analysis. The MASCA test shows that the molten corium can be stratified with the metal layer under the oxide when sub-oxidized corium including iron was retained in the lower head of the reactor vessel. This stratification is caused by the increased density of the metal layer attributed to a transfer of uranium metal that was reduced from uranium oxide by zirconium. Thermodynamic equilibrium calculations with the database, which was developed for the corium U-Zr-Fe-O-B-C-FPs system using the ionic two-sublattice model for liquid, show quantitative agreements with the MASCA test, such as the composition of each layer, fission product (FP) partitioning between the layers and B4C effect on the stratification.

Experience on ultrasonic inspection of control rod nozzles of pressurised water reactor pressure vessel

An ultrasonic inspection technique developed by VTT to assess the integrity of control rod nozzles to the reactor pressure vessel head is described. A variant of wall thickness measurement, the technique is based on detecting the presence of water in the air gap between the ferritic nozzle and the austenitic steel corrosion protection tube, which can lead to corrosion of the nozzle. The presence of water allows propagation of the pulse across the gap, generating multiple echos, allowing leaks to be identified. The basis and practical application of the technique is described and its limitations discussed.

원자로 용기 외벽냉각을 위한 1차원 이상유동 실험 및 해석

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests and the simple analysis were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The calculated circulation flow rate was similar to experimental ones within about ±15% error bounds and depended on the form loss due to the inlet/outlet area.

Considerations in the Design and Analysis of an ASME Section VIII, Div. 2 Reactor Support Skirt

This paper describes the considerations employed in the finite element analysis of a relatively “short” support skirt on a hydrocarbon reactor vessel. The analysis is accomplished in accordance with ASME B&PV Code, Section VIII, Division 2 alternate rules in conjunction with the guidelines outlined in WRC Bulletin 429. This provides a sound basis for the classification of the calculated stress intensities. The support skirt is capable of sustaining the deadweight load in addition to resisting the effects of thermal displacements, wind loadings, overturning moments from external piping loads on the attached hydrocarbon reactor vessel, and friction between the skirt base plate and concrete foundation. The displacement and thermal boundary conditions are well defined and discussed in detail. The effects of multiple scenarios for the displacement boundary conditions are examined. The skirt design also employs a hot-box arrangement whereby the primary mode of heat transfer is by radiation. A discussion of the two-part analysis is included and details the interaction between the heat transfer analysis and the subsequent structural analysis. The heat transfer finite element analysis is utilized to determine the temperatures throughout the bottom of the vessel shell and head, as well as the integrally attached support skirt. Of prime importance during the analysis is the axial thermal gradient present in the skirt from the base plate up to and slightly beyond the skirt-to-shell junction. While the geometry of the subject vessel and skirt is best described as axisymmetric, the imposed loadings are a mixture of axisymmetric and non-axisymmetric. This combination lends itself to the judicious selection and utilization of the harmonic finite element and properly chosen Fourier series representation of the applied loads. Comparison of the thermally induced axial stress gradient results from the FEA to those obtained by the closed form beam-on-elastic-foundation are also tendered and discussed. Finally, recommendations are included for the design and analysis of critical support skirts for large, heavy-wall vessels.