Transient deformation of LMFBR cores due to local failure: Experimental and theoretical investigation

This paper describes an effort to predict the mechanical core deformation caused by local failure within an LMFBR core. These activities are intended to cover all the potential core damage possibilities currently under discussion and analysis. In particular it is shown that the reactor can be scrammed in time under pessimistic-realistic pressure transients and that the damage does not exceed tolerable limits.A special gas generator technique to simulate a fuel coolant explosion has been developed at AWRE Foulness. This has been used to perform the explosion tests needed to demonstrate the safety of the SNR 300 core. A molten fuel—coolant interaction (MFCI) experimental facility, and a drop tower to carry out sub-assembly crushing tests are described. Theoretical studies are presented which use mass-spring-dashpot, lumped parameter-hinge or micro-rigid-lumped-mass models. They simulate the crushing and bending of a single sub-assembly interacting with the coolant as well as the behaviour of a multirow “spoke” model.For the core analysis only preliminary computational results are presently available which can be compared with the full scale tests in which the fluid pressure did not exceed a “threshold” of about 100 bar. Parameter studies show the influence of pulse shape, material properties as well as the time integrator.Some of the unanswered question concern the dydrodynamic feedback of the deformations on the pressure distribution in space and time. Also the behaviour of the highly irradiation-embrittled cores is poorly understood today. Finally, an enhanced energy release package to describe the MFCI must still be added to the reactivity calculation module of a future fast reactor dynamics code.     

Heat transfer from a partially uncovered reactor core

This paper documents a model which has been developed for predicting the temperature distribution along a “flow channel” of a pressurized water reactor during simulated, uncovered core conditions. In the model, heat conduction along the fuel element, convection from the surface to the coolant, radiation exchange between the clad surface and steam, and surface exchange between adjacent fuel rods are considered. Variations of the thermophysical properties of the fuel road and of the coolant with temperature are accounted for, but oxidation of Zircaloy is not modeled. Extensive sensitivity studies on the effects of heat generation in the core, steam velocity, pressure level, uncovered core height, presence of hydrogen gas in the coolant, power skew, clad emissivity, and convective heat transfer correlations have been examined. The results show that the importance of radiation in comparison with convection increases with an increase in the fuel rod temperature, pressure, and clad emissivity.     

Two-phase flow models in unbounded two-phase critical flows

Referring to a Loss-of-Coolant Accident situation in LWRs, an analysis of the two-phase region just downstream from the broken pipe, in which a two-phase critical flow takes place, has been performed. A characterization of the flow pattern inside the unbounded two-phase jet is given considering:

• - jet's external shape, obtained by means of photographic pictures;

• - pressure profiles inside the jet, obtained by means of a movable “Pitot” gauge;

• - jet phase's distribution information, obtained by means of X-ray pictures.

Jet's X-ray pictures show the existence of a central high-density “core” gradually evaporating all around, which gives place to a characteristic “dart flow” the length of which depends on stagnation thermodynamic conditions.     

Fluid-structure interaction analysis of a hypothetical core disruptive accident in LMFBRs

To ensure safety, it is necessary to assess the integrity of a reactor vessel of liquid-metal fast breeder reactor (LMFBR) under HCDA. Several important problems for a fluid-structural interaction analysis of HCDA are discussed in the present paper. Various loading models of hypothetical core disruptive accident (HCDA) are compared and the polytropic processes of idea gas (PPIG) law is recommended. In order to define a limited total energy release, a “5% truncation criterion” is suggested. The relationship of initial pressure of gas bubble and the total energy release is given. To track the moving interfaces and to avoid the severe mesh distortion an arbitrary Lagrangrian–Eulerian (ALE) approach is adopted in the finite element modeling (FEM) analysis. Liquid separation and splash from a free surface are discussed. By using an elasticity solution under locally uniform pressure, two simplified analytical solutions for 3D and axi-symmetric case of the liquid impact pressure on roof slab are derived. An axi-symmetric finite elements code FRHCDA for fluid-structure interaction analysis of hypothetical core disruptive accident in LMFBR is developed. The CONT benchmark problem is calculated. The numerical results agree well with those from published papers.     

Gas-cooled fast reactors—Status of CEA preliminary design studies

The gas-cooled fast reactor (GFR) is one of the six reactor concepts selected in the frame of the Generation IV initiative. The most significant GFR option is the use of a helium high temperature primary coolant. The helium option is very attractive (chemical inertness, neutron transparency, etc.) but it leads to very specific thermal-hydraulic issues.As far as the reactor core design is concerned, a ceramic fuel concept with a good thermal conductivity has been chosen. The main requirement is to obtain an average exit core temperature of 850 °C (energy conversion efficiency) with a maximum fuel temperature of about 1200 °C and with a low core pressure drop (in order to ease the decay heat removal). The main core characteristics have been determined for two reactor powers: a medium one (600 MWth) and a large one (2400 MWth). For various reasons, this latter became the CEA reference choice. A consistent set of core parameters has been determined taking into account the different constraints including the thermal-hydraulics. The reference arrangement proposed is based on plate fuel elements.A significant issue for the GFR is the decay heat removal. An innovative approach has been chosen in case of loss of coolant accidents (LOCAs). A “guard containment” enclosing the primary system is used to maintain a medium gas pressure (10 bar) in order to remove the decay heat by low power forced convection systems in the short term and natural convection systems in the long term. This guard containment is not pressurized during normal operations and can be a metallic structure.As far as the energy conversion system is concerned, an indirect-combined cycle has been chosen. The significant advantages of this choice are: a moderate core inlet temperature (400 °C instead of 480 °C for the direct cycle) and an attractive compactness of the primary system (facilitating the guard containment design).Due to the novelty of these options, a significant effort of components pre-sizing and design calculations has been achieved. Following this effort, a CATHARE model of the reactor system has been made and the calculation of the reactor steady-state confirms the consistency of the overall system pre-sizing. This model has been used for a first transient calculation. Other types of transients have to be analyzed, however, it is thought that the proposed GFR design can reach the safety requirements of Generation IV systems.     

Heat flux distribution from a volumetrically heated pool with high Rayleigh number

Experimental results are presented on the heat flux distribution at the boundaries of volumetrically heated pools at high enough Rayleigh numbers to be directly relevant to the problem of retention of a molten corium pool inside the lower head of a reactor pressure vessel. The experimental facility, named COPO, is a 2-dimensional “slice”, Joule-heated and geometrically similar in shape (torispherical at 1/2-scale) to the lower head of a VVER-440 reactor. The results show that: the heat flux on the side wall (vertical portion) is essentially uniform; the downward heat flux strongly depends on position along the curved wall; and average fluxes on the side in the downward direction are in agreement with existing correlations, but somewhat underestimated in the upward direction. For the shape considered, the heat flux along the lower curved wall seems to be independent of the presence and extent of the liquid pool (contained by the vertical sidewalls) portion above it.     

Inherently safe containments for nuclear power plants

Nuclear energy cannot be avoided in the near future. To regain public acceptance the safety of nuclear power plants has to be increased. Consequently, feasibility studies have been carried out for a containment proposal for future pressurized water reactors which will keep people unharmed even in the case of severe nuclear accidents under the assumption “all that can go wrong, will go wrong”. The main features of the design concept are a core melt cooling and retention device, a passively acting cooling system to remove the decay heat and a double-wall containment which is able to withstand high static and dynamic internal pressures due to hydrogen detonation. Internal structures are designed to resist extreme loadings resulting from various accident scenarios including in-vessel steam explosion and vessel failure under high system pressure.     

Heat transfer during molten corium-concrete interactions

Severe accidents in light water-cooled nuclear power plants involved heat transfer from molten reactor core materials or “corium” penetrating the reactor pressure vessel and coming to rest upon the containment building concrete floor covered by water. This paper discusses the difficulties of getting good information about the properties of the components and the flow structure during molten corium-concrete-water interactions. Also, potential heat transfer mechanisms are described and available prototypical tests are utilized to show that the enhancement in heat transfer by rising gas bubbles is the most likely mechanism, particularly if heat transfer by iradiation across the gas bubbles is included.     

Questions on safety margins of components at loads beyond design — failure modes of the primary system of PWRs

The reliability and load-carrying capabilities of structures are an important part of any risk analysis in two aspects. One is the probability of failure as an initiating event, the other is the probability of or time to failure in response to load situations beyond design conditions. The methods to predict the probability of failure of the primary pressure boundary as an initiating event for a loss-of-coolant accident have already been published by Beliczey and Schulz in 1986.For the analysis of the structural behaviour of components of the primary system at loads beyond design different questions have to be answered, e.g. - most probable sequence of loading; - most probable sequence of failure; - failure loads or times connected to a “high confidence of low probability of failure”. The failure modes of the primary circuit system and the respective times to failure were investigated for core melt-down under high pressure (HPC) and low pressure (LPC) conditions.Particular interest was directed towards the behaviour of steam generator tubes, to surgeline and main coolant piping, to the upper head and flange connections of the reactor pressure vessel, and to the lower head.     

PSB-VVER simulation of Kozloduy NPP “loss of feed water transient”

This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions.RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient.The objective of the experiment “loss of feed water”, which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as “integral system effects” and ”natural circulation“. For assessment of the RELAP5 capability to predict the “Integral system effect” phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the “Natural circulation” phenomenon the hot and cold leg temperatures behavior have been investigated.This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

A bounding technique for dynamic plastic deformations of damped structures

A bounding principle for elastic-perfectly plastic creeping and noncreeping structures subjected to mechanical and/or thermal loads varying below or above the shakedown limit is presented. This principle contains some free “perturbation functions” which, suitably chosen, enable it to specialize, so generating bounds on a variety of deformation measures (such as inelastic work dissipated within any portion of the body, inelastic strains and displacements), some of which are new results, others recover or generalize known results. The resulting bounding technique possesses a quite unified character which is useful for computational purposes. The concept of “pseudo-plastic” strain is shown to be crucial for the derivation of bounds applicable above the shakedown limit.     

The DEEPSSI project, design, testing and modeling of steam injectors

The DEEPSSI project is a steam injector research programme. Among thermal-hydraulic passive systems, the steam injectors (also called “condensing ejectors” or “steam jet pumps”) are very interesting apparatus with very specific characteristics (high velocity, very low pressure). The envisaged reactor application is the Steam Generator Emergency FeedWater System (EFWS) of Pressurised Water Reactors (PWRs). The heart of this project is the development and the testing of an innovative steam injector design. Three experimental facilities are involved: CLAUDIA in France, IETI in Italy and IMP-PAN in Poland. In these facilities, different design options have been tested and some significant improvements of the initial design have been obtained.In addition to the experimental studies, the development of a steam injector computational model has been undertaken in order to model industrial systems based on steam injectors. The one-dimensional module of the system code CATHARE2 has been chosen to be the basis of this model. The first results obtained have confirmed the capabilities of CATHARE2 to describe the steam injector thermal-hydraulics.     

A two-dimensional model for fluid-structure interaction in curved pipes

A “channel” model was developed for the purpose of simulating the interactive fluid-structural response of curved pipes to pressure pulses. Simulation is shown to have been achieved analytically in both the axisymmetric (“breathing”) and transverse (“bending”) modes of interactive behavior.An experimental program which was aimed at the validation of the model is also described. Tests were run in both straight and curved pipe configurations. Comparisons between measurements and model calculations demonstrate the validity of the model within the range of parameters under consideration.The model was implemented into the DISCO code for nonlinear fluid-shell interaction.     

Fuel coolant interaction experiments in TROI using a UO2/ZrO2 mixture

This paper discusses the results of steam explosion experiments using reactor material carried out under “Test for Real cOrium Interaction with water (TROI)” program. About 4–9 kg of corium melt jet is delivered into a sub-cooled water pool at atmospheric pressure. Spontaneous steam explosions are observed in four tests among six tests. The dynamic pressure, dynamic load, and morphology of debris clearly indicate the cases with steam explosion. The initial conditions and results of the experiments are discussed.     

Quasi 3-D measurements of turbulence structure in horizontal air–water bubbly flow

Quasi 3-D measurements of the turbulence structure of air–water bubbly flow in a horizontal tube with 35 mm i.d. are undertaken with two TSI “X”-type hot-film probes. The turbulent fluctuations, u_f,v_f,w_f, in axial, radial and circumferential directions, respectively, and Reynolds stresses and are obtained. It is found that in the lower portion of the tube, the profiles of turbulent fluctuation and Reynolds stress resemble those of single phase flow; whereas in the upper portion of the tube, where the bubble population is high, the turbulence, especially the circumferential fluctuation w_f, is substantially enhanced, and the radial turbulence assumes highest value in the radial position −0.7<r/R<0.5. The magnitudes of Reynolds stresses and in our measurements are in the same level except in the lower portion of the tube where assumes a value close to zero as is the case in single phase flow and vertical air–water bubbly flow.     

The use of energy absorbers to protect structures against impact loading

When a flying missible impacts a fixed structure, the interface loading is dependent on the deformation characteristics of both impacting and impacted bodies. If both are too rigid to accommodate the amount of gross deformation required to neutralize the incoming kinetic energy, or if such energy absorption has a chance to proceed in uncontrolled and unreliable ways, then there is a need to interpose a specifically designed “energy absorber” between missile and structure, from which a well-defined load time history can be derived during the course of impact.

The required characteristics of such an energy absorption material are:

&#x02022; the capability to accommodate large permanent deformation without structural failure; and

&#x02022; the reliable and controlled “load-deformation” (or “stress-strain”) behaviour under dynamic conditions, with an aim at an optimal square shape curve.

Consideration must also be given to environmental or other disturbing effects, like temperature, humidity, and “out of plane” loading. A short survey is presented of the wide range of energy absorbers already described in technical papers or used in a number of practical safety applications within varied engineering fields (from vehicle crash barriers to high energy pipe whipping restraints). However, with such open a literature, information is usually lacking in the specific data required for design analysis.

The following “energy absorption” materials and processes have thus been further experimentally investigated, with an a aim at pipe whipping restraint application for nuclear power plants:

1. (1) plastic extension of austenitic stainless steel rods;

2. (2) plastic compression of copper bumpers; and

3. (3) punching of lightweight concrete structures.

Dynamic “stress-strain” characteristics have been established for stainless steel bars at several temperatures under representative loading conditions. For this purpose, a test rig has been specifically designed to incorporate a number of adjustable parameters and to behave as a representative “slice” of an actual pipe whipping restraint; typical strain rates are in the 10 sec⁻¹ range. The behaviour of copper bumpers has been compared under static and dynamic conditions (using a conventional drop weight test (DWT) machine); as no significant strain rate effects were emphasized, only static tests have been further developed. The DWT rig was used again to investigate crushing or punching of cellular concrete under varying geometries and loading conditions. To remedy certain deficiencies of the regular commercial grades of cellular concrete, special lightweight mixtures have been studied to optimize material toughness and provide a wider range of specific resistance.Results of this experimental program are presented and discussed. The use of energy absorbers is then illustrated for a few typical pipe whipping restraints. The design of restraints is based on real dynamic characteristics of “energy absorption” material as produced by the test program. To derive design loads of restraints, a number of methods can be used ranging from a simplified “energy balance” graph to sophisticated plastodynamic computer analysis. Typical results are presented and discussed to compare the efficiency of these alternative methods.     

A method for quantifying the reliability of instrumentation for engineered safety features to improve accident sequence recognition in reactor control rooms

This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to quantitatively evaluate the reliability of the instrumentation for engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel through the conversion of system fault trees to alarm trees. In the alarm tree, possible states of each instrumented alarm are identified as “true” or “false”. In addition, a “warning” status is also defined and integrated into the alarm analysis routine. The impact of this additional status condition on the Boolean laws used to evaluate the alarm trees is examined. An application is described for a BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents.     

On using alternative perturbation theory methodologies in the evaluation of reactivity effects in ADS cores

Core-average Doppler and coolant void reactivity coefficients, as well as the kinetic parameters (β_eff and Λ), have been determined for sub-critical accelerator-driven systems employing lead–bismuth eutectic (LBE) and helium gas coolants. To determine these parameters use is made of the standard procedure for analyzing critical reactors, which is based on “perturbation-theory” (PT), while in addition two dedicated methodologies for sub-critical systems, i.e. “inhomogeneous perturbation-theory” (IPT) and “heuristically based generalized perturbation-theory” (HGPT), have been employed to compute these parameters in a more rigorous manner.The two methods (PT and IPT/HGPT) are found to give similar results for each application and despite a smaller target k_eff-value, the sensitivity of the method is small in the case of the gas-cooled system, thus confirming the adequacy of the standard procedure. As regards the coolant void reactivity coefficient in the gas-cooled ADS, this finding can mostly be attributed to the fact that the core is always transparent with respect to the source neutrons, irrespective of the specific helium content.The sensitivity of the Doppler coefficient is also rather low in the case of the LBE cooled system. However, the dedicated methods are needed for the correct prediction of the coolant void reactivity coefficient, especially if minor actinides are introduced into the core. More important, in this case, is the fact that the PT-approach does not produce conservative results. Finally the sensitivity of the reactivity and kinetic parameters to the different methods is of the same order as that due to uncertainties in nuclear data and therefore these will need to be included in any overall evaluation of the impact of uncertainties on steady-state and transient ADS performance.     

Prediction of fuel channel—graphite gas-gap behaviour in RBMK reactors

To maintain thermal contact between the fuel assembly and the graphite moderator, RBMK design reactors employ graphite split rings, which are alternatively tight on the pressure tube or tight on the graphite brick central bore. The split in the graphite rings allows a helium/nitrogen gas mixture to flow up the fuel channel. This prevents oxidation of the graphite and can be sampled to detect pressure tube leaks. The initial clearance between the rings and pressure tube or graphite brick is approximately 2.7 mm (1.35 mm each side). Due to material property changes of the pressure tubes and graphite during operation of the reactor, the size of the clearance between the rings and the pressure tube/brick, called the “gas-gap”, varies. Closure of these gaps has been identified as a possible safety case issue by reactor designers and by independent reviews carried out as part of TACIS reviews and as part of the Ignalina Safety Analysis Report. The reasons for this are that gas-gap closure would cause the pressure tube to be tightly gripped by the graphite bricks via the split rings, which could lead to:

• Extra loading on the upper pressure tube zirconium/steel transition joint, particularly during shut down and emergency transients.

• Splitting of the graphite brick, leading to loss of thermal contact between the pressure tube and graphite. As approximately 5.6% of the heat in graphite-moderated reactor is generated within the moderator through neutron and gamma-heating, loss of thermal contact would result in higher graphite temperatures, accelerating the rate of graphite expansion and hence increasing the loading of the core radial restraint.

• Graphite debris may become lodged in inter-brick gaps, leading to increased axial pressure tube loading during shut down and emergency transients.

The authors have carried out deterministic assessments based on the Ignalina RBMK-1500 reactors in Lithuania, modelling the behaviour of the graphite under irradiation and have predicted graphite bore diameter changes that are in good agreement with the measurements of graphite bore diameters taken at Ignalina Nuclear Power Plant (NPP). A probabilistic model has been developed using the actual results of the deterministic calculations with non-linear graphite behaviour. Statistical analysis of the measurements of tube and graphite diameters taken from Units 1 and 2 at Ignalina NPP has been carried out. Further work has been carried out to try to determine the uncertainty inherent in the predictions of the gas-gap closure from the calculations. The overall objective of the studies is to aid prediction of the gas-gap closure process, and help to identify a suitable monitoring strategy for gas-gap closure that could be used for any RBMK reactor.     

The NUREG-1150 probabilistic risk assessment for the Sequoyah Nuclear Plant

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Sequoyah Nuclear Plant performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, and consequence analyses, and the risk results obtained when the results of these analyses are combined with the accident frequency analysis. The results of this PRA indicate that the offsite risk from internal initiating events at Sequoyah are quite low with respect to the safety goals. The containment appears likely to withstand the loads that might be placed upon it if the reactor vessel fails. A good portion of the risk, in this analysis, comes from initiating events which bypass the containment. These events are estimated to have a relatively low frequency of occurrence, but their consequences are quite large. Other events that contribute to offsite risk involve early containment failures that occur during degradation of the core or near the time of vessel breach. Considerable uncertainty is associated with the risk estimates produced in this analysis. Offsite risk from external initiating events was not included in this analysis.