Experimental study of whole core thermal-hydraulic phenomena in EBR-II

In the analysis of core-wide temperature distributions in a liquid-metal-cooled fast breeder reactor (LMFBR) under both normal and abnormal operating conditions, it is commonly assumed that there is little, if any, thermal interaction between adjacent subassemblies. Since intersubassembly heat transfer tends to reduce the transverse temperature gradients in a reactor, thereby ameliorating the effects of local overheating, this assumption is conservative. In order to assess the importance of this effect as well as of flow redistribution in a reactor core, an experimental study was conducted in EBR-II covering a wide range of operating power and flow conditions, including both forced and natural convection. The results of this study indicate that radial heat transfer and flow redistribution are important mechanisms in the thermal-hydraulics of LMFBR cores, especially at low flow rates.     

Investigation of nonrecoil fission-product release phenomena using multifrequency source-perturbation experiments in EBR-II

A widely-accepted theory that was proposed to explain nonrecoil fission-product release phenomena in LMRs postulated that isotopic transport kinetics are attributable to liquid-state diffusion across a viscous sublayer of Na near the surface of the fuel. A series of multifrequency source-perturbation experiments was recently performed in EBR-II to isolate and quantitatively explore the boundary-layer diffusion mechanism while separating out all other physical variables that affect the isotopic transport kinetics. Detailed analyses of the results using bivariate spectral decomposition and cross-correlation techniques are presented. Results of the analyses provide incontrovertible experimental evidence that boundary-layer diffusion in fact plays no part in the release of short-lived fission products in LMRs. A major conclusion of this investigation is that all nonrecoil fission-product release phenomena originate from mechanisms acting inside the breached element itself. Implications of this and other findings made during this investigation are discussed, and recommendations are made for extending the techniques introduced here in future experiments involving actual breached pins.     

Preliminary thermal-hydraulic phenomena investigation during total instantaneous blockage accident for CEFR

The Chinese Experimental Fast Reactor (CEFR) is under installation and commissioning right now. It is essential to investigate core disruptive accidents (CDAs) for the evaluation of CEFR's safety characteristic. As part-I preliminary investigation, accident of total instantaneous blockage (TIB) in single subassembly scale is modeled and analyzed. The degradation scenario has been calculated by a fluid-dynamics analysis code for liquid–metal fast reactors (LMFRs). For further investigation of accident process and influence to the neighboring bundles, seven subassembly domain is then simulated and calculated as part-II investigation. Total instantaneous blockage is assumed to occur in the center subassembly under normal operating conditions and consequences to neighboring assemblies are studied. The result shows that the key events, such as sodium boiling, clad melting, fuel particles relocation, hexcan melt-through and melt propagation into neighboring six assemblies symmetrically are adequately simulated. From comparison and discussion of the CEFR's results with the SCARABEE tests and Superphenix (SPX1)-type reactor simulation, it is concluded that all the key events appear in the same sequence whereas the propagation is limited in neighboring six assemblies. The discrepancy is probably due to less fuel inventory and better cooling capacity in CEFR subassembly design. TIB calculations help to give a better understanding and prediction of hypothetical accident scenario in subassembly blockage accidents for CEFR.     

Simulation and analysis of thermal-hydraulic phenomena in a PWR hot leg related to SBLOCA

Thermal-hydraulic phenomena in the hot leg of a pressurized water reactor during the small break loss-of-coolant accident (SBLOCA) are simulated and studied in this paper. They include the single-phase flow dynamics, the cocurrent stratified flow during the natural circulation conditions, and the countercurrent stratified flow during the reflux condensation conditions.Satisfactory results were obtained from the computations in comparison with the data from the German Upper Plenum Test Facility. It is revealed that the fluid flow exhibits strong multi-dimensional effects, i.e. an appreciable acceleration and deceleration along different regions of the hot leg, and a four-vortex secondary flow structure in the cross-section of the bend region. Cocurrent stratified flow under the natural circulation conditions is successfully simulated, presenting two different water transport mechanisms. Under the reflux condensation conditions, different countercurrent flow structures are found under the conditions away from and with the countercurrent flow limit.     

Evolution of thermal-hydraulics testing in EBR-II

A thermal-hydraulics testing and modeling program has been underway at the Experimental Breeder Reactor-II (EBR-II) for 12 years. This work culminated in two tests of historical importance to commercial nuclear power, a loss of flow without scram and a loss of heat sink without scram, both from 100% initial power. These tests showed that natural processes will shut EBR-II down and maintain cooling without automatic control rod action or operator intervention. Supporting analyses indicate that these results are characteristic of a range of sizes of liquid metal cooled reactors (LMRs), if these reactors use metal driver fuel. This type of fuel is being developed as part of the Integral Fast Reactor Program at Argonne National Laboratory. Work is now underway at EBR-II to exploit the inherent safety of metal-fueled LMRs with regard to development of improved plant control strategies.     

Theoretical spectrum-threshold detector data correlation in EBR-II

An experimental and computer program to further examine the neutron environment in the Experimental Breeder Reactor-II (EBR-II) has been completed. Monte Carlo and S₄ Transport methods were used to determine the neutron spectrum at various positions in the EBR-II core and blanket regions. Response functions for the threshold detectors ⁵⁸Ni (n, p) ⁵⁸Co and ⁵⁴Fe (n, p) ⁵⁴Mn were determined for each position and the corresponding predicted induced activities are compared with experimental results. Based on combinations of calculated neutron spectra, experimental detector responses, and cross section end points an empirical differential cross section was determined for the ⁴⁶Ti (n, p) ⁴⁶Sc threshold reaction. Spectrum averaged cross sections for the three threshold reactions which have been determined at various positions in this facility suggest that significant errors in fast neutron fluences will result if the usual fission spectrum averaged cross sections are used.     

Five percent break BWR LOCA/ECC test at ROSA-III without HPCS actuation - two-dimensional core thermal-hydraulic phenomena

The ROSA (Rig of Safety Assessment)-III facility is a volumetrically scaled (1/424) simulated boiling water nuclear reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. A recirculation pump suction line break test with a five percent break area was conducted with the assumption of high pressure core spray system (HPCS) failure. The simulated peripheral fuel rods facing the channel box wall had a tendency to be rewetted temporarily at the upper part of the core by falling water from the upper plenum before low pressure core spray system (LPCS) actuation, while the rods in the central region were not rewetted but quenched mainly from the bottom of the core after low pressure coolant injection system (LPCI) actuation. Therefore, the peak cladding temperatures of the simulated high power fuel rods were limited to lower values since they were located in the peripheral region and the temporary rewetting before LPCS actuation occurred mainly in the peripheral region. The ROSA-III five percent break test and a BWR counterpart were analyzed with the RELAP5/MOD1 (cycle 018) code. Similarity between the ROSA-III small break test and a BWR small break LOCA has been confirmed through comparison of the calculated results.     

Comparisons of PRD components for various EBR-II configurations

Comparison of detailed calculations of contributions by region and component of the power-reactivity-decrements (PRD) for four differing loading configurations of the Experimental Breeder Reactor-II (EBR-II) are given. The linear components and Doppler components are calculated. The non-linear (primarily subassembly bowing) components are deduced by differences relative to measured total PRD values. Variations in linear components range from about 10% to as much as about a factor of two depending upon the component. The deduced non-linear components can differ both in magnitude and sign. Effects of differing assumptions of the nature of the fuel-to-clad interactions upon the PRD components are also calculated.     

Fatigue testing of metallurgically-bonded EBR-II superheater tubes

Fatigue crack growth tests were performed on 2¼Cr–1Mo steel specimens machined from ex-service experimental breeder reactor-II (EBR-II) superheater duplex tubes. The tubes had been metallurgically-bonded with a 100 μm thick Ni layer; the specimens incorporated this bond layer. Fatigue crack growth tests were performed at room temperature in air and at 400 °C in air and humid Ar; cracks were grown at varied levels of constant ΔK. In all conditions the presence of the Ni bond layer was found to result in a net retardation of growth as the crack passed through the layer. The mechanism of retardation was identified as a disruption of crack planarity and uniformity after passing through the porous bond layer. Full crack arrest was only observed in a single test performed at near-threshold ΔK level (12 MPa√m) at 400 °C. In this case the crack tip was blunted by oxidation of the base steel at the steel–nickel interface.     

The Performance of a Counting-Mean-Square Voltage Channel in the EBR-II

A wide range neutron flux monitoring test channel based on counting and mean-square voltage techniques has been installed at the EBR-II. The measured responses to neutron flux are presented and compared to those predicted. Data taken while the reactor is on an essentially constant period during typical re-starts are shown. The gamma discrimination ratio for this system has been determined and is discussed.     

Numerical diffusion in single-phase multi-dimensional thermal-hydraulic analysis

Higher-order difference schemes, including QUICK and its related schemes, are discussed through several numerical experiments with the focus on their performance to reduce numerical diffusion and to suppress local unphysical oscillations. Finally, these schemes are applied to in-vessel thermal-hydraulic analysis and their effectiveness on the analysis is demonstrated.     

Demonstration of EBR-II power maneuvers without control rod movement

A group of five plant inherent control tests was successfully conducted in November 1987 in the Experimental Breeder Reactor II. These tests demonstrated that the plant power of a metal-fueled reactor can be passively controlled over a large power range by slowly changing the primary flow and the reactor inlet temperature. These variables are, in turn, regulated by the primary pump speed, the secondary flow, and the turbine inlet pressure. In all tests, control rods were not used to regulate power. It was demonstrated that the plant power can be controlled with reasonable accuracy without using control rods when the reactivity feedback characteristics of the reactor are well understood and the plant controllers are adequately designed.     

A thermal-hydraulic code (THYD) for the miniature neutron source reactor thermal-hydraulic transients

The transient thermal-hydraulic problem of MNSR is represented by ten differential equations solved numerically using Runge–Kutta method.Computational results are then compared with experimental measurements. Fuel grids and cooling coil models are incorporated in the model too. Radiating energy from the clad is taken into account in the energy balance in the reactor. The pool is divided into three sections in the model. The effect of the cooling coil of the pool upper section on reactor thermal-hydraulic parameters is discussed. The only input parameter of the reactor is the power temporal distribution. Good agreement between calculated and measured data was obtained.     

EBR-II unprotected loss-of-heat-sink predictions and preliminary test results

Two unprotected (i.e., no scram or plant protection system action) loss-of-heat-sink transients were performed on the Experimental Breeder Reactor-II in the Spring of 1986. One was initiated from full power (60 MW) and the other from half power. The loss of heat sink was accomplished in each test by essentially stopping the secondary-loop sodium coolant flow. Pretest predictions along with preliminary test results demonstrate that the reactor shuts itself down in a benign and predictable manner in which all of the reactor temperatures approach a quenching (or smothering) temperature at which the fission power goes to zero.     

Calculation of temperature coefficients of reactivity for EBR-II kinetic analyses

Component and regional temperature coefficients of reactivity for four loading configurations of the Experimental Breeder Reactor-II (EBR-II) are compared. The coefficients are calculated by summations of microcoefficients obtained by fine axial delineations of every subassembly. A special-sum method for obtaining effective coefficients for use in kinetics code channels representing subassembly groupings is described. Evaluations of rod-bank suspension coefficients and of grid-plate radial-expansion coefficients are also presented.     

Modification of EBR-II plant to conduct loss-of-flow-without-scram tests

This paper describes the details of and the philosophy behind changes made to the EBR-II plant in order to conduct loss-of-flow-without-scram tests. No changes were required to conduct loss-of-heat-sink-without-scram tests.     

Implications of the EBR-II inherent safety demonstration test

Extensive thermal-hydraulics testing at EBR-II culminated in the Inherent Safety Demonstration Test on April 3, 1986. This work may well lead to fundamental changes in the approach to the design and licensing of liquid-metal-cooled reactor (LMR) power plants. The EBR-II test program has thus far demonstrated (1) passive removal of decay heat by natural circulation, (2) passive reactor shutdown for a loss of flow without scram, and (3) passive reactor shutdown for a loss of heat sink without scram. Supporting analyses indicate that these characteristics can be incorporated into larger commercial LMRs and be used as the basis for a totally new passive control strategy. Analyses and tests are now in progress to show that LMRs with these characteristics and the passive control strategy are also inherently safe for unprotected overpower accidents.     

Some challenges to the thermal-hydraulic codes

The thermal-hydraulic codes were developed with the data and correlations obtained from separate effect tests. As such. There are some system-related phenomena which cannot be depicted properly by the codes. In this paper we discuss the difficulties encountered by code modeling for the following systems: feedback loop, multichannel system, multidimensional flow and multiloop circulation. The discussion shows that codes can only give probable answers; the difficulties encountered are due to maldistribution of heat and flow, primary-secondary interaction, feedback effect, instrumentation-control interaction and other unknown factors.     

On the dynamic response of EBR-II ducts to pressure pulses

An important consideration for the continued irradiation of experiments to and beyond pin failure in EBR-II is the response of the duct to a pressure pulse that may result from the rapid release of gas from a failed pin. Although all fuel element failures to date in EBR-II have been benign, typically pjn-holes or hairline cracks, the possibility of cladding bursts during transients cannot be precluded. Thus, the EBR-II project has performed numerous pressure pulse tests on subassembly duct and is continuing with analytical studies as well as additional testing. Investigations are being conducted within the fail-safe analysis framework. Correlation of the duct test results yielded an empirical relationship between the permanent deformations of a duct, and the pressure and volume of the fission gas in the fuel element. Results of the correlation are given for a number of variables including the type of gas, the cladding and duct material, the presence of internal restriction to gas flow, and whether predefected tubes of rupture discs are used to produce the pressure pulse.Preliminary dynamic stress analyses have been conducted using the one-dimensional finite element code, STRAW (Report ANL-8065, June 1974). A comparison of analytical predictions of maximum permanent duct deflection between flats and test results is presented in this paper. The analytical results are shown only for the deflection of a single flat. Since the analysis assumes an axisymmetric pressure pulse, opposing flats have equal deformation. However, in the tests the duct flat closest to the opening in the pressurized tube experienced the largest deformation. Parametric analyses have been conducted to study the influence of pulse shapes and material properties on duct deformation. Initial studies using a triangular pulse have shown that the pulse decay time and plastic hardening slope of the material have negligible influence on duct deformation. Consequently, subsequent studies used a constant decay time of 2 msec and a slope of 80 psi. Among the results discussed in the paper are the influence of peak pressure and yield strength on duct deformation for two rates of pressure rise and the influence of non-uniform pressure on duct deformation. The pressure is assumed to vary linearly from the middle of the flat to the corner. A pressure increase from a uniform 200 to 300 psi at the middle of a flat, increases the deformation by only 14 mil. These results demonstrate the importance of the pulse shape and its distribution over the duct and the material properties of the latter.Additional parametric investigations of duct deformation and dynamic stresses are being conducted. Two-dimensional finite element codes are being employed to study the fracture resistance of irradiated ducts.