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.     

UPTF experiment refined PWR LOCA thermal-hydraulic scenarios: conclusions from a full-scale experimental program

A total of 34 tests were performed at upper plenum test facility (UPTF, a 1:1 scale test facility) to investigate the thermal-hydraulic phenomena in a pressurized water reactor (PWR) primary system during end-of-blowdown, refill and reflood phases of a loss-of-coolant accident (LOCA). Separate effect tests as well as integral tests were carried out. After the completion of the program a summary of the basic findings from the full-scale tests is given, focusing on thermal-hydraulic issues related to: two-phase flow phenomena at the ECC injection ports for cold or hot leg injection; the ECC delivery into core area via the downcomer or the tie plate; entrainment de-entrainment phenomena during reflood (i.e. the “steam binding” and driving water head reduction problems).     

A study on the actuator efficiency behavior of safety-related motor operated gate and globe valves

Periodically, the operability of the safety-related motor-operated valves (MOVs) in nuclear power plants must be verified. Because the actuator efficiency is one of the most important factors in the determination of the actuator output, it should be considered in ensuring the operability of MOVs during the verification duration. In particular, special consideration should be paid to its potential degradation, but the design efficiency provided by manufacturers is usually used because the actuator efficiency calculation is difficult and requires considerable time and effort. In this paper, a method is introduced to calculate actuator efficiency by using diagnostic signals acquired in field tests. The actuator efficiency was calculated from the estimated motor torque, the stem thrust measured in field tests, and overall gear ratio provided by manufactures. The motor torque was estimated by using an algorithm, which can calculate electric torque from the three phases of currents and voltages, resistances between phases acquired in field tests. The validation of the design efficiencies was evaluated by comparing those efficiencies with the calculated actuator efficiencies. And, the age-related degradation was analyzed through the behavior analysis over time of the calculated actuator efficiencies. Most of the actuator efficiencies were found not to be degraded over time and kept efficiency greater than the design efficiency. However, two actuator efficiencies with lower motor speed, overall gear ratio, and maximum motor torque rating are susceptible to be lower than the design efficiencies. For the two actuators, threshold efficiencies were calculated and provided to replace their design efficiencies.     

Fission gas disposition in decay heat environment

A computer code is developed to model fission gas disposition in UO₂ fuel during nearly isothermal heating, as would result from decay heat. The intragranular analysis, random diffusion model, uses a spatial solution for random migration and bubble coalescence. Nonequilibrium bubble growth and interactions between bubbles as well as nonequilibrium bubblegrain boundary interactions are considered. In the intergranular analysis, grain growth is allowed until tunnels have formed; this is set at 6% grain edge swelling. Grain face bubbles are assumed uniform in size and distribution. Grain edge tunnels are approximated in toroidal geometry. The model (a system of two grains, one shrinking and one growing but with total volume conserved; each grain originally contained 50% of the total fission gas) is applied to a “postulated” LMFBR accident condition involving a slow “nearly isothermal” heating of the fuel. The intragranular release is computed at 3.4% without grain growth, but at 14% with grain growth. Intragranular release is found to be dominated by grain growth.The analysis was applied also to the FGR-34 transient of HEDL. It is pointed out, however, that in the FGR-34 experiment thermal gradients were present whereas in the present code, only isothermal heating is considered. In spite of this significant difference between the modeled and the observed thermal state of the fuel, the comparison was carried out with a purpose to examine the existence of nonequilibrium attractive forces, between bubbles and grain boundaries, which were suggested by HEDL as perhaps responsible for the bubble denuding observed on both sides of the grain boundary. The computations did demonstrate the existence of nonequilibrium conditions, but the computed intragranular bubble radii, with only random diffusion as the operative mechanism, were well below the reported values. It is likely that this descrepancy between computed and observed bubble radii is due to (1) the presence in FGR-34 tests of thermal gradients, which would make bubble biased migration operative, and/or (2) the possibility of very strong enhancement, significantly more than two orders of magnitude, of the diffusion coefficient due to the prevailing nonequilibrium bubble conditions. The present code treats nonequilibrium conditions, but contains no physical mechanism for diffusion enhancement.     

Interpretation of the results of the CORA-33 dry core boiling water reactor test

All boiling water reactor (BWR) degraded core experiments performed prior to CORA-33 were conducted under ‘wet’ core degradation conditions, in which water remains within the core and continuous steaming feeds metal-steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ‘dry’ core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ‘dry’ core severe accident scenarios and to resolve partially phenomenological uncertainties concerning the behavior of relocating metallic melts that drain into the lower regions of a ‘dry’ BWR core (the ex-reactor experiments at Sandia National Laboratories will further address these uncertainties). CORA-33 was conducted on 1 October 1992, in the CORA test facility at Karlsruhe. A review of the CORA-33 data indicates that the objectives were achieved; i.e. core degradation occurred at a core heat-up rate (characterized by the absence of any temperature escalation caused by oxidation) and a test section axial temperature profile (at incipient structural melting) that are prototypic of full-core nuclear power plant simulations under ‘dry’ core conditions. Simulations of the CORA-33 test at Oak Ridge National Laboratory (ORNL) have required the modification of existing control blade-canister materials interaction models to include the eutectic melting of the stainless steel-zircaloy interaction products and the heat of mixing of stainless steel and zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the post-test analyses carried out at ORNL based on the experimental data and the post-test examination of the test bundle at Karlsruhe. The implications of these results with respect to degraded core modelling and the associated safety issues are also discussed.     

Investigation of the two-phase flow in a boiling water reactor using neutron-noise technique

In this paper, the fluctuations of the neutron flux (“neutron noise”) of the Mühleberg BWR are investigated. Above 2 Hz, the noise measured by the in-core neutron detectors is driven exclusively by local fluctuations of the void fraction. Characteristic changes of the neutron-noise signature along the axis can be attributed to changes of flow pattern. By measuring the phase lag between pairs of axially placed neutron detectors, the transit time of the steam between the detectors can be evaluated. The measured transit times are applied to the study of two-phase flow in the core. The neutron-noise method has the advantage of providing in-core information under operational conditions.     

Development and benchmarking of a mechanical model for GFR plate-type fuel

The reference fuel design currently being considered within the Generation-IV Gas-cooled Fast Reactor (GFR) project is a ceramic plate matrix with a honeycomb inner structure containing small fuel cylinders. The fuel is mixed plutonium–uranium carbide, while the matrix material is silicon carbide. The present paper describes the mechanical part of a thermal–mechanical model being developed for studying the transient behavior of this highly heterogeneous fuel type. Benchmarking has been carried out against detailed finite-elements modeling (FEM).The resultant thermal–mechanical model can provide reliable fuel and cladding (matrix) stress/strain conditions to evaluate temperatures and neutronic feedbacks. As such, it has been integrated into PSI’s coupled code system “FAST”, which aims at the comprehensive safety analysis of advanced fast reactor systems.The detailed FEM analysis of the GFR fuel has been useful not only for benchmarking the new model, but also for obtaining an in-depth understanding of fuel stress/strain characteristics, which cannot be reproduced with simplified models. Thereby, the range of applicability of the new model has clearly been defined. In particular, the 3D FEM analysis has revealed a concentration of stresses at the pellet corners during pellet/matrix contact, which could lead to fuel element failure. This effect is found to be mitigated considerably, if the fuel pellets are shaped in a manner which enhances the contact area.     

Characterization of industrial NbTi strands at variable field for JT-60SA toroidal field coils

The mission of the JT-60SA Tokamak, to be built in Japan, is to contribute to the early realization of fusion energy by its exploitation in support of the ITER program. JT-60SA project is an important part of the “broader approach” activity as a satellite program for ITER. The toroidal field (TF) coils are a European “in kind” contribution and they will partly be built by France. JT-60SA TF coil uses the Cable In Conduit Conductor (CICC) with NbTi superconductor strands. TF conductors will have to operate at 5.7 T, 5 K and at current density of 450 A/mm² with sufficient margins. In the framework of JT-60SA TF coil manufacture, the variable temperature characterization is an important step to select NbTi strand. At an early stage of design, we had to choose the strand with acceptable performances. During the design qualification and validation stage, it is important to qualify strands in conditions close to the operation conditions. The industry has proposed various strands manufactured with different processes. This work and publication examines a strand with an internal CuNi barrier, which is expected to lead to better current distribution between strands, by more precise calibration and control of the inter-strand resistance. The strands were tested at the Grenoble High Magnetic Field Laboratory facility. The domain (B, T, J) explored was in the range of 4.5–11 T for the magnetic field intensity, 4.2–6.5 K for the temperature and between 40 A/mm² and 1200 A/mm² for the current density. For each strand, “critical current density” and “current sharing temperature” measurements have been carried out, with a temperature precision of few tens of mK. Once the measurements performed, the fitting parameters (of type J_C = f(B, T)) of each strand have been found, by performing regression analysis. This work will lead to select the strand with the best characteristics. In this paper, we present the results of this measurement task, the data and regression analysis (fits, T_cs, etc.) and the conclusion about the strand choice.     

A simple model for predicting energy dissipation of thin plates being perforated by “hard” missiles

For the case of low velocity impact a simple model is derived for the determination of energy dissipation of thin plates being perforated by “hard” missiles. The predicted residual energy of the missile having passed through the target is compared with test results. The tests were carried out with plates made of wood-chips (a rather homogeneous and cheap material). For a projectile with large diameter relative to the thickness of the target it is shown that the energy absorption of the plate is essentially influenced by the fracture type.     

Piping research overview

The USNRC Piping Review Committee (PRC) was formed in 1983 with a charter to review NRC piping criteria, to recommend changes to this criteria, and to identify areas that would benefit from future research. This overview will outline the NRC-sponsored research being conducted to address those PRC recommendations concerning the design of nuclear piping systems to withstand dynamic loads. A key element of this research is the joint EPRI/NRC “Piping and Fitting Reliability Research Program.” This program consists of dynamic capacity testing of piping at the system, component, and specimen levels, plus analyses needed to support recommendations for changes to the ASME Code. As part of NRC's contribution to the EPRI/NRC program, a pipe system capacity test will be conducted at ETEC. The “Nonlinear Piping Response Prediction” project at HEDL is evaluating nonlinear response prediction techniques with differing degrees of complexity and will compare the various analytical results both with each other and with physical benchmarks such as the ETEC test. An ORNL project is developing nozzle design guidance that will provide a more realistic basis for evaluating the higher nozzle loads that will result from the more flexible piping systems design that are being considered. INEL will evaluate high frequency damping by considering the existing high frequency data and by conducting high frequency/high stress tests on two piping systems. LLNL is now conducting studies to more completely assess the uncertainties in the seismic response of building structures and piping systems. As a follow-on to the research efforts reported in NUREG/CR-3811, BNL will conduct additional studies to improve combinational procedures for piping response spectra analyses.     

A study on the stem friction coefficient behavior of motor-operated valves

This paper describes how to calculate the stem friction coefficient of safety related motor operated valves (MOVs) that reflects potential degradation with time by using diagnostic signals acquired in static field tests that have been conducted more than two times per valve. Based on the calculated stem friction coefficients, their behaviors with time were analyzed considering various parameters that could cause potential degradation. Most friction coefficients change randomly rather than increasing or decreasing continuously over time. From those trends, a threshold coefficient, which represents the highest expected value of the friction coefficient, was calculated and provided.     

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.     

Advances in two-phase flow modeling for LMFBR applications

This paper summarizes the development of numerical models for analysis of sodium boiling phenomena in LMFBR which has been carried out at M.I.T. over the last five years.With regard to the degree of spatial averaging, our models use the porous body approach, in two and three-dimensional configurations. One important advantage of this model is the ability to accommodate homogenization of arbitrary-sized regions of interest.From a numerical point of view our basic approach is a semi-implicit method in which pressure pulse propagation and local effects characterized by short time constraints are treated implicitly, while convective transport and diffusion heat transfer phenomena, associated with longer time constants, are handled explicitly. This method remains tractable and efficient in multi-dimensional applications.Both a six-equation (“two-fluid”) model and a four-equation (“mixture”) model have been pursued. A considerable effort has been devoted to the development of constitutive relations. Our current package provides an adequate simulation capability for a wide range of applications.This paper will present the general physical formulation of the codes, the constitutive relations, the general numerical approach, applications, and finally some concluding remarks based on our experience with these codes.     

Ultimate strength of reinforce concrete shear walls under multi-axes seismic loads

“The model test on multi-axes loading on RC shear walls” had been carried out as for the 10-year project aiming at comprehension of the earthquake response behavior of three-dimensional (3D) reinforced concrete (RC) shear walls under the 3D of multi-axes earthquake loading condition. The motivation of the project building-up is that the current seismic design of nuclear power plant building is carried out by applying one-dimensional (1D) dynamic earthquake load to an analytical building model in each direction independently whereas actual earthquake jolts the building in the three directions simultaneously. Therefore, there were opinions requesting some testing confirm whether or not the current seismic design methodology is reliable for the input motions exceeding the design earthquake ground motion moreover for the input motions of the 3D. The project had completed with various valuable outcomes that can reply to the opinions. Moreover, the outcomes will play an important role in evaluating seismic margins of important structures in a nuclear power plant. In this paper, based on the published documents relating to this test project, the author describes a review of the whole testing and summarizes the major outcomes extracted by the test project.     

Basic risk analyses for high-temperature reactors

The methodology of PSA/PRA is available for the HTR and has already been applied to various plant concepts. The results are predictive and generic in nature; the analyses have to struggle with less detailed technical information (paper design instead of real operated plants) and little experience from plant practice. The overall degree of uncertainty is similar to studies for LWRs mainly because operating experience can be transferred to some extent and the physical phenomena are much easier to describe. Therefore, the topology of design and beyond-design accidents has been established.For medium-sized HTRs (e.g. HTR-500) of current design failure of active systems for decay heat removal, resulting in core heatup, clearly dominates the risk and leads to the largest releases of radioactive nuclides into the environment. For small-sized HTRs (e.g. HTR-Module) temperature-induced releases from the fuel are insignificantly low for all types of accident; plate-out activities on the steam generator surfaces remobilized in the course of water ingress accidents can be regarded as the main contribution to the comparatively small source term.The largest releases are so low for all HTR concepts that early health effects can be ruled out in any case, including no evacuation. For small HTR plants even late cancer effects need practically not to be expected.A comparison with licensed released values has shown that the applicable current requirements are met by all HTR concepts examined. However, small HTRs especially offer an additional potential for compliance with more stringent safety requirements, “taking the fear out of hypothetical accidents”, by limiting maximum releases. Incidentally, the classically defined “risk” to the population from both plants is generally very low.     

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.     

基于RELAP5平台的LOCA评估程序中喷放模型与CHF模型的修改与验证

为开发核电厂保守性的冷却剂丧失事故(LOCA)安全评审用程序,本文按照10CFR50附录K的核电厂失水事故评估模式要求,对RELAP5 / MOD3.3的两相喷放模型与临界热流密度(CHF)模型进行了修改,并与Marviken Test-22和ORNL THTF相关试验结果分别进行比较,探讨了模型修改的合理性与保守性,为进一步完成认证级LOCA安全评估程序奠定了初步基础.     

Flooding and flow reversal of two-phase annular flow

The flooding and flow reversal conditions of two-phase annular flow are mathematically defined in terms of a characteristic function representing a force balance. Sufficiently below the flooding point in counter-current flow, the interface is smooth and the characteristic equation reduces to the Nusselt relationship. Just below the flooding point and above the flow reversal point in cocurrent flow, the interface is “wavy”, so that the interfacial shear effect plays an important role. The theoretical analysis is compared with experimental results by others. It is suggested that the various length effects which have been experimentally observed may be accounted for by the spatial variation of the droplet entrainment.     

Angular distributions of Ni and Ti atoms sputtered from a NiTi alloy under He and Ar ion bombardment

The angular distributions of sputtered components were measured for NiTi polycrystalline alloy under 9 keV Ar⁺ and He⁺ ions bombardments with various fluences in ultrahigh vacuum. Combination of Rutherford Backscattering Spectrometry (RBS) and Auger Electron Spectrometry (AES) techniques allowed us to observe enhanced concentration of Ni over a layer with thickness comparable to a primary He⁺ ions penetration depth due to selective sputtering of Ti atoms and radiation-induced diffusion processes. A preferential emission of Ni atoms towards the surface normal was observed during bombardment by both He⁺ and Ar⁺ ions. More forward-peaked “over-cosine” angular distributions of sputtered Ni in comparison with those for Ti atoms have been measured. Nonstoichiometric sputtering of NiTi alloy dependent on emission angle was observed for bombardment fluence of He⁺ well below that needed for the steady-state altered layer formation. To explain the peculiarities of NiTi sputtering, an interpretation is discussed in terms of sputtering due to backscattered He⁺ ions.