Vessel Head Research Articles

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.

Predictive Displays for Survey Vessels

Inexperienced helmsmen often oversteer because of the lag that occurs between changing the rudder angle and a change in the vessel's heading. Predictive displays are a common way of mitigating the effect of lag on human control. Accordingly we developed a predictive display to show the position and heading of a vessel a short time in the future. With this display, the helmsman's task becomes that of keeping the predictor on the planned path. In effect, the predictor is steered, not the vessel. Our predictive model was statistical and based on data gathered from a 40 foot survey vessel carrying out a variety of maneuvers while the position, heading, speed through the water and rudder angle were continuously recorded. The advantage of such a predictor is that it can, in principle, be generated automatically, without any need for a model of hull shape or vessel dynamics. We evaluated the predictor by having both experienced and inexperienced helmsmen steer a pre-defined figure-of-eight course. The results showed substantial reduction in cross track error for inexperienced participants to the point that their performance was indistinguishable from those that were more experienced.

Gross Plastic Deformation of Axisymmetric Pressure Vessel Heads

The gross plastic deformation and associated plastic loads of four axisymmetric torispherical pressure vessels are determined by two criteria of plastic collapse: the ASME twice elastic slope (TES) criterion and the recently proposed plastic work curvature (PWC) criterion. Finite element analysis was performed assuming small and large deformation theory and elastic-perfectly plastic and bilinear kinematic hardening material models. Two plastic collapse modes are identified: bending-dominated plastic collapse of the knuckle region in small deformation models and membrane-dominated plastic collapse of the cylinder or domed end in large deformation models. In both circumstances, the PWC criterion indicates that a plastic hinge bending mechanism leads to gross plastic deformation and is used as a parameter to identify the respective plastic loads. The results of the analyses also show that the PWC criterion leads to higher design loads for strain hardening structures than the TES criterion, as it takes account of the effect of strain hardening on the evolution of the gross plastic deformation mechanism.

Simulation of melt jet breakup and debris bed formation in water pools with IKEJET/IKEMIX

The objective of the development of the code system KESS is simulating the processes of core melting, relocation of core material to the lower head of the reactor pressure vessel (RPV) and its further heatup, modelling of fission product release and coolability of the core material. In the scope of the code development, IKEJET and IKEMIX were designed as key models for the breakup of a molten jet falling into a water pool, cooling of fragments and the formation of particulate debris beds. Calculations were performed with these codes, simulating FARO corium quenching experiments at saturated (L-28) and subcooled (L-31) conditions, as well as PREMIX experiments, e.g. PM-16. With the assumption of a reduced interfacial friction between water and steam as compared to usually applied laws, the melt breakup, energy release from the melt and pressurisation of the vessel observed in the experiments are reproduced with a reasonable accuracy. The model is further applied to reactor conditions, calculating the relocation of a mass of corium of 30 t into the lower plenum, its fragmentation and the formation of a particle bed.

Validation and application of the WABE code: Investigations of constitutive laws and 2D effects on debris coolability

The WABE-2D model aims at the problem of coolability of degraded core material during a severe accident in a light water reactor (LWR) and describes the transient boil-off and quenching behavior of debris beds. It is being developed in the frame of the KESS code system, which is considered to describe the processes of core heatup, melting, degradation and relocation to the lower plenum as well as the subsequent behavior. The models developed in this frame are being integrated in the German system code ATHLET-CD. An emphasis of the present contribution lies on multidimensional aspects of the cooling behavior. From multidimensional features a significant improvement of overall coolability is expected compared to what is concluded based on classical one-dimensional analyses. Such analyses – also mainly oriented at top cooling conditions – additionally miss the expected importance of interfacial drag which should support coolability in co-current flow situations due to bottom flooding. The latter situation plays a role in the multidimensional behavior expected under realistic conditions. Thus, a further emphasis in the present contribution lies on the constitutive drag laws and their effects in such configurations. Calculations comparing top and bottom flooding and the influence of interfacial friction are presented. An explanation for effects observed in related experiments at Forschungszentrum Karlsruhe is provided based on this influence. The significant increase of dryout heat flux with water inflow from below, driven by a lateral water column, is reproduced and understood. Enhanced cooling due to this and in general by lateral inflow is also demonstrated for reactor scenarios, considering particulate debris in the lower head of the reactor pressure vessel (RPV) of a LWR or in a deep water pool in the reactor cavity of a boiling water reactor (BWR). Cooling by steam flow through local dry zones can establish under lateral water supply to regions below and yield a further extension of coolability. Quenching of hot material is also analyzed. Finally, cases with loss of coolability, dry zone formation and melting are considered, especially in the perspective to analyze melt pool formation in the lower head of the RPV and the history of thermal interaction with the lower head wall. The latter will determine failure possibilities and modes of the RPV.

米国の加圧水型原子力発電所におけるAlloy 600製圧力バウンダリー構成機器の1次冷却水応力腐食割れ事例の傾向分析

Primary Water Stress Corrosion Cracking (PWSCC) which occurs on Nickel based alloy (Alloy 600) is a worldwide concern since early 1980's. Recently several significant degradations that originate from PWSCC in the reactor coolant pressure boundary (RCPB)components have been observed at U.S. PWR plants (e.g. Oconee-3, Davis Besse). The United States Nuclear Regulation Commission (NRC) has issued generic communications to address this problem and, in response to the Davis Besse event in 2002, gave the inspection order EA-03-009 for the PWR licensees to implement the inspection of the reactor vessel heads depending upon the effective degradation years. As well, in Japan, PWSCC is considered one of the safety issues, in particular, for aged nuclear power plants and actually, some plants have experienced PWSCC on RCPB components. In the present study, we analyzed the U.S. experience with Alloy 600 degradation by reviewing the licensee event reports from 1999 to 2005 and examined the trend of them mainly focusing on affected components, characteristics of cracking and inspection approaches for detecting the PWSCC. This study indicates that PWSCC is found to be occurred on the RCPB components exposed to the environment with high temperature such as the reactor vessel head, and has the tendency to happen for specific manufactures and material according to the RCPB components. As well, it is shown that for several components, the non-destructive examination is generally needed to detect and/or confirm the PWSCC after the visual inspection and different repair techniques are applied depending on the components affected.

Development of an Enhanced Core Catcher for Improving In-Vessel Retention Margins

In-vessel retention (IVR) of core melt that may relocate to the lower head of a reactor vessel is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for several advanced light water reactors. A U.S.-Korean International Nuclear Energy Research Initiative project has been initiated to explore design enhancements that could increase the margin for IVR for advanced reactors with higher power levels [up to 1500 MW(electric)]. As part of this effort, an enhanced in-vessel core catcher is being designed and evaluated. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary). The first is a base material that has the capability to support and contain the mass of core materials that may relocate during a severe accident; the second is an oxide coating on top of the base material, which resists interactions with high-temperature core materials; and the third is an optional coating on the bottom side of the base material to protect it from oxidation during the lifetime of the reactor. This paper summarizes results from the in-vessel core catcher design and evaluation efforts, focusing on recently obtained results from materials interaction tests and prototypic testing activities.

Comparison of Severe Accident Results among SCDAP/RELAP5, MAAP, and MELCOR Codes

This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP.

Assessment on Integrity of BWR Internals against Impact Load by Water Hammer under Conditions of Reactivity Initiated Accidents

The integrity of the reactor pressure vessel (RPV) head and reactor internals was assessed by means of fluid and fluid-structural coupled analyses to evaluate the water hammer phenomenon arising from postulated high burnup fuel failure under reactivity initiated accident (RIA) conditions. The fluid viscosity effect on the water column burst as well as the complex three-dimensional flow paths caused by a core shroud and standpipes were considered in this study. It is shown that fluid viscosity becomes an influential factor to dissipate impacting kinetic energy. Integrity of the RPV head and the shroud head was ensured with a sufficient level of margin even under these excessively conservative RIA conditions.

KSNP+ reactor vessel head drop analysis for a 5.5M free fall

The KSNP+ RV closure head drop analysis was carried out to assess the reactor core coolability in case of the RV closure head drop accident during the refueling operation. The analysis consists of a number of different RV head drop scenarios as the postulated accident events during refueling that include a concentric head drop case and three different cases of laterally offset head drop cases. The analysis was initiated due to the adoption of the IHA (Integrated Head Assembly) in the KSNP+ reactor design, which increases the weight of the RV closure head assembly. Four different analysis models were developed that correspond to the RV head drop analysis scenarios. An in-house dynamic analysis code was used for the RV head drop analysis. The entire reactor internals and fuel assemblies are modeled by using lumped masses and spring elements. Because of the extreme load exerted by RV head drop, most members experience stresses that are beyond the elastic limits. A separate elastic—plastic analysis for some members was carried out and the resulting load-deflection curve was used as the stiffness of the element. The effect of water above the reactor vessel in the refueling pool was ignored for the conservative estimation of the analysis. The analysis shows that the concentric head drop is the most severe case of loading condition. It also reveals that the local deformation of some reactor internals and the fuel assemblies is occurred; however the primary membrane stresses are within the bound of allowable stress limits. Consequently the reactor core remains in coolable state.

Consolidation of scientific and technological expertise to assess the reliability of reactor pressure vessel embrittlement prediction in particular for the arctic area plant (COBRA)

The evaluation and prognosis of reactor pressure vessel (RPV) material embrittlement in WWERs and the allowable period of their safe operation are performed on the basis of impact test results of irradiated surveillance specimens. The main problem concerns the irradiation conditions (irradiation temperature, neutron flux and neutron spectrum) of the surveillance specimens that have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV wall condition. In particular, the key issue is the possible difference between the irradiation temperature of the surveillance specimens and the actual RPV wall temperature. It is recognized that the direct measurement of the irradiation temperature by thermocouples during reactor operation is the only way to obtain reliable information. In addition, the neutron field's parameters in the surveillance specimens location have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level. The COBRA project, which started in August 2000 and had a duration of 3 years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contained state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The selected reactor for the experiment was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings, the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5 ± 4 °C. Uncertainties were evaluated also with experimental work carried out in the WWRSZ research reactor and by finite element modelling of surveillance capsules. The results obtained show that there is not need to perform temperature correction when surveillance data of irradiated specimens are used for embrittlement assessment of WWER-440(213) reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals ∼2.7 × 10 12 cm −2 s −1 with E > 0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.

Research and development on passive cooling system

Experiments are carried out to investigate the effects of the natural convection of superheated gas as well as those of the standpipes on the temperature distributions of the components and the heat removal performance in the water-cooling panel system for the MHTGR for decay heat removal, and to verify reliability of the design and evaluation methods. The numerical results of the code THANPACST2 are compared with the experimental data to verify the numerical methods and axi-symmetric model proposed, which can simulate the three-dimensional configuration of the standpipes on the upper head of the pressure vessel by using porous body cells. The experiments revealed that temperatures increased with elevation on the upper head, because the standpipes restrict radiation heat transfer to the upper cooling panel and reduce the heat transfer area on the upper head, which was superheated by natural convection of helium gas in the pressure vessel. In the presumed accident condition in which thermal radiative heat transfer is responsible for the majority of the total heat transfer, the numerical methods were able to closely duplicate the pattern of the rising temperature profile with elevation around the top of the upper head as observed in the experiments.

Equipment Development and Nondestructive Examination Techniques for Inspection of Reactor Pressure Vessel Head Penetration

Equipment Development and Nondestructive Examination Techniques for Inspection of Reactor Pressure Vessel Head Penetration

Development of circumferential through-wall crack K-solutions for control rod drive mechanism nozzles

The US Nuclear Regulatory Commission (USNRC) has undertaken a program to assess the integrity of control rod drive mechanism nozzles in existing plants that are not immediately replacing their reactor pressure vessel heads. This paper summarizes some of the efforts undertaken on behalf of the USNRC for the development of circumferential crack-driving force solutions to be used in probabilistic determinations of the time from detectable leakage to failure. Detailed three-dimensional elastic–plastic finite element analyses of these nozzles were conducted to determine the effects of operating conditions, tube-to-head interference fit, tube yield strength, and nozzle-to-head angle on the crack-driving force. Three major observations to date are: • Conducting simple tube-only analyses can severely overpredict the actual crack-driving force in the control rod drive mechanism nozzles. • The crack-driving force is a strong function of the nozzle yield strength, nozzle-to-head angle, interference fit, and weld residual stress; and • In many cases, when the crack is made to be perpendicular to the wall of the nozzle, the operating conditions and residual stresses appear to create some crack closure along the crack faces, suggesting that cracks in service will grow at an angle through the thickness.

Conceptual design of an in-vessel core catcher

An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (US)–Korean International Nuclear Energy Research Initiative (INERI) investigating methods to insure retention of materials that may relocate to the lower head of a reactor vessel under severe accident conditions in advanced reactors. This enhanced core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulator coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. This paper summarizes results from thermal, flow, and structural analyses as well as initial scoping materials interaction tests that were completed to support the conceptual design of the core catcher.

Corium retention for high power reactors by an in-vessel core catcher in combination with External Reactor Vessel Cooling

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 (TMI-2) accident. If it is possible to ensure that the vessel lower 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 Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, 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. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.

Strain rate effects in nuclear steels at room and higher temperatures

An investigation of strain rate, temperature and size effects in three nuclear steels has been conducted. The materials are: ferritic steel 20MnMoNi55 (vessel head), austenitic steel X6CrNiNb1810 (upper internal structure), and ferritic steel 26NiCrMo146 (bolting). Smooth cylindrical tensile specimens of three sizes have been tested at strain rates from 0.001 to 300 s −1, at room and elevated temperatures (400–600 °C). Full stress–strain diagrams have been obtained, and additional parameters have been calculated based on them. The results demonstrate a clear influence of temperature, which amounts into reducing substantially mechanical strengths with respect to RT conditions. The effect of strain rate is also shown. It is observed that at RT the strain rate effect causes up shifting of the flow stress curves, whereas at the higher temperatures a mild downshifting of the flow curves is manifested. Size effect tendencies have also been observed. Some implications when assessing the pressure vessel structural integrity under severe accident conditions are considered.

Late-phase melt conditions affecting the potential for in-vessel retention in high power reactors

If cooling is inadequate 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. In such a case, concerns about containment failure and associated risks can be eliminated if it is possible to ensure that the lower head remains intact so that relocated core materials are retained within the vessel. Accordingly, in-vessel retention (IVR) of core melt as a key severe accident management strategy has been adopted by some operating nuclear power plants and planned for some advanced light water reactors. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements can provide sufficient heat removal to assure IVR for high power reactors (i.e., reactors with power levels up to 1500 MWe). Consequently, a joint United States/Korean International Nuclear Energy Research Initiative (I-NERI) has been launched to develop recommendations to improve the margin of success for in-vessel retention in high power reactors. This program is initially focussed on the Korean Advanced Power Reactor—1400 MWe (APR1400) design. However, recommendations will be developed that can be applied to a wide range of existing and advanced reactor designs. The recommendations will focus on modifications to enhance ERVC and modifications to enhance in-vessel debris coolability. In this paper, late-phase melt conditions affecting the potential for IVR of core melt in the APR1400 were established as a basis for developing the I-NERI recommendations. The selection of ‘bounding’ reactor accidents, simulation of those accidents using the SCDAP/RELAP5-3D © code, and resulting late-phase melt conditions are presented. Results from this effort indicate that bounding late-phase melt conditions could include large melt masses (>120,000 kg) relocating at high temperatures (∼3400 K). Estimated lower head heat fluxes associated with this melt could exceed the maximum critical heat flux, indicating additional measures such as the use of a core catcher and/or modifications to enhance external reactor vessel cooling may be necessary to ensure in-vessel retention of core melt.

KAPOOL experiments on melt-through of a metal plate by an overlying melt pool

In the event of a hypothetical severe accident in a Light Water Reactor (LWR) it is possible that the core melts and relocates to the lower head of the reactor pressure vessel (RPV). If the severe accident management measures are not successful, the RPV can fail and the core melt is discharged onto the basement of the containment. To prevent a significant release of radioactivity to the environment due to the melt-through of the basement, LWR's might incorporate the ability to retain and cool the core melt on the basement of the containment (core catcher concept). Before the melt arrives on the cooling device on the basement, the melt has to penetrate the reactor cavity. The reactor cavity may be equipped with a metallic plate at its bottom, covered by a layer of sacrificial concrete, to collect the melt over a certain time period. To achieve a complete and homogeneous spreading of the melt over the total area of the cooling device a fast and sufficiently wide melt-through of the metallic plate is required. To investigate the important processes concerning the melt-through of a metallic plate by a simulated core melt, a series of transient KAPOOL experiments has been performed in the years 1998 to 2000. The corium melt in the KAPOOL tests is simulated by an alumina/iron melt, produced by a thermite reaction. Therefore it was possible to investigate the interaction of both parts of the corium melt (metallic and oxidic) with a metal plate. The experiments show that metallic melts in contact with a steel plate that simulates the melt gate between the reactor cavity and the core catcher compartment lead to fast melting of the steel plate (KAPOOL 9). Oxide melts in contact with a steel or an aluminium plate lead to oxide crust formation at the steel or aluminium interface, and subsequent formation of gaps between the crust and the metal plate. In case of a steel plate no ablation of the plate by the oxide melt could be observed (KAPOOL 11 and 12). Due to the lower melting temperature of the aluminium one could detect and analyse failure modes of the plate due to high thermal loadings (KAPOOL 13 and 15 to 17). In addition to the experiments, heat transfer calculations with the HEATINGS code have been performed. The results of the calculations helped to reproduce the experimental results and to identify the important processes concerning the melt-through of a metal plate. A drawback of the KAPOOL tests is the transient character: Volumetric heat production in the melt was not possible to simulate the nuclear decay heat. In the case of real corium melts with decay heat a later thermal erosion of a melt gate made of steel would be expected in contrary to the outcomes of the tests KAPOOL 11 and 12. From the results of the KAPOOL tests 15 to 17 it is advantageous to use aluminium as the material for the bottom of the reactor cavity, as for this material the tests have proved a sufficient fast and large opening of the plate. A very important advantage of using aluminium is the high thermal conductivity of the material, which supports a fast opening of the gate over a large area. When steel should be used as material for the melt gate, further investigations deem necessary to prove a sufficiently large opening of this gate by an oxide melt.

Optimal design of an ellipsoidal head of a pressure cylindrical vessel

AbstractThe subject of analysis is a stress concentration problem in a circular cylindrical vessel with ellipsoidal heads. The region of the junction of ellipsoidal and cylindrical shells is subject to stress concentration. Intensity of the concentration depends, first of all, on relative convexity of the ellipsoidal head, and the ratio of thicknesses of the shells. Based on the boundary disturbance theory the stress state in this cylindrical vessel was analytically described. Optimal ellipsoidal heads were determined on the grounds of three optimisation criteria.