Off-normal Conditions Research Articles

Divertor and first wall armor design for the TIBER-II engineering test reactor

TIBER-II is a compact, high-power-density, steady-state current-drive engineering test reactor that uses a double-null divertor configuration, operating in the high recycle mode with low particle temperature. The nominal peak heat flux is 3.3 MW/m 2 , with off-normal condition peaks of up to 6.6 MW/m 2 . Plasma disruptions may cause peak energy deposition of up to 13 MJ/m 2 . The design uses water-cooled copper alloy tubes with brazed-on protective tiles. During early operating phases when disruptions may be frequent, these tiles will be made of carbon. During the nuclear testing phase, disruptions must be limited so that more erosion-resistant tungsten tiles may be used. The first walls of TIBER-II experience a heat flux of 0.23 MW/m 2 with disruption energy density of 2.4 MJ/m 2 . They are protected by carbon-carbon composite armor tiles. The TIBER-II design requirements are challenging, and although the divertor and first wall armor designs successfully meet these requirements, significant issues must be resolved to verify their performance. These critical issues and the R&D required to resolve them are described.

Graphite as plasma facing material

Graphite materials are widely used as plasma facing components for major fusion devices such as TFTR, JET and JT-60, because of their low Z and good high temperature characteristics. However, for their further application in near and middle term machines a comprehensive evaluation has to be undertaken, including vacuum properties, fracture toughness under normal and off-normal conditions, hydrogen recycling, erosion, retention of tritium and effects of radiation by 14 MeV neutrons. The present status of the data base for each item is presented and discussed on the basis of graphite chemistry and physics. Another issue to be considered is the design of plasma facing component such as tile attachment, active cooling capability, etc. These considerations are extremely important for the next generation fusion machines such as ITER.

Automated reasoning with dynamic event trees: a real-time, knowledge-based decision aide

Abstract The models and data contained in a PRA Event Sequence Analysis represent a wealth of information that can be used for dynamic calculation of event sequence likehood. In this paper we report a new and unique computerization methodology which utilizes these data. This sub-system (reffered to as PREDICTOR) has been developed and tested as part of a larger system. PREDICTOR performs a real-time (re)calculation of the estimated likelihood of core-meltsas a function of plant status. This methodology uses object-oriented programming techniques from the Artificial Intelligence discipline that enable one to codify event tree and fault tree logic models and associated probabilities developed in a PRA study. Existence of off-normal conditions is reported to PREDICTOR, which then updates the relevant failure probabilities throughout the event tree and fault tree models by dynamically replacing the ‘off-the-seel’ (or prior) probabilities with probabilities based on the current situation. The new event probabilities are immediately propagated through the models (using ‘demons’) and an updated core-melt probability is calculated. Along the way, the dominant non-sucess path of each event tree is determined and highlighted.

A fast and simple iteration scheme for the temperature calculation in a fuel rod

A new and simple iteration scheme for solving the nonlinear heat conduction in a fuel rod is developed. The method is based on a Newton-Raphson iteration scheme and takes two distinctive characteristics of the problem into account. The result is a scalar Regula Falsi technique which is superior to a relaxation technique. The new iteration scheme has been applied successfully over 2 years in the TRANSURANUS code, analysing several hundred complicated data cases under normal, off-normal and accident conditions. Convergence problems never occurred and the computational costs were significantly reduced.

Safety analysis for the loss-of-flow and loss-of-heat sink without scram tests in EBR-II

This paper discusses the safety considerations and the analysis that was done to support the conduct of the Loss-of-Flow and Loss-of-Heat Sink Without Scram tests in EBR-II. The plant safety limits and the analysis codes and methods are presented. The impact of the tests and off-normal conditions on the fuel and plant structures is evaluated. Conclusions are that the test program had no impact on plant operations and the accumulated fuel damage was negligible.

Prediction of pressure vessel failure: A critical review of the probabilistic approach

Considered is a pressure vessel in the presence of crack-like defects. The probability of fast fracture, at any time, is formulated in terms of a characteristic defect size exceeding a critical value. The probability is conditional, given the random event of the annual occurrence of particular loading conditions. Much of the literature is reviewed, and erroneous formulations in previous work are noticed. The concept of probability of annual failure employed in the present account has to be distinguished from the concepts of reliability and failure rate used in reliability theory. Two sets of data are extracted from experimental evidence reported in the literature. One concerns the size distribution and total number of defects. The other pertains to the critical values of defect size for several loading conditions, together with the probability of the annual occurrence of these conditions. The effect of ultrasonic inspection on defect occurence is taken into account. As an illustrative example, an analytic, closed-form solution for the failure probability is presented for a typical, modern, pressurized water reactor vessel under end-of-life conditions. The different contributions to failure probability of various normal and off-normal loading conditions are clearly displayed, and the importance of a judicious choice of both modern low-residual-element pressure vessel steels and an improved ultrasonic inspection technique is demonstrated.

Fracture behavior of a neutron-irradiated stainless steel submerged arc weld cladding overlay

Stainless steel weld overlay cladding was irradiated at temperatures and fluences relevant to power reactor operation. The cladding was applied to a pressure vessel steel plate by the submerged arc, single-wire, oscillating-electrode method. Three layers of cladding were applied. The first layer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by excessive melting of the base plate. Charpy V-notch and tensile specimens were irradiated at 288°C to a fluence of 2 × 10 23 neutrons/m 2 (> 1 MeV). When irradiated, both types 308 and 309 cladding increased 5 to 40% in yield strength and slightly increased in ductility in the temperature range from 25 to 288°C. All cladding exhibited ductile-to-brittle transition behavior during impact testing caused by temperature dependent failure of the δ-ferrite phase. The type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation. Conversely, the impact properties of the specimens containing the highly diluted type 309 cladding, microstructurally similar to that produced during some off-normal welding conditions in existing reactors, experienced significant increases in transition temperature and drops of up to 50% in upper-shelf energy.

Two-phase thermal-hydraulic simulations with COMMIX-2

The COMMIX-2 computer code has been developed for steady/unsteady, three-dimensional thermal-hydraulic analysis of reactor components under normal and off-normal operating conditions. To permit analysis of a wide spectrum of flow conditions, i.e., from homogeneous and equilibrium to nonhomogeneous and nonequilibrium conditions, we have provided two alternative models. One is the two-fluid model, and the other is the homogeneous equilibrium model. The new concept of volume porosity, surface permeability, distributed resistance, and distributed heat source is implemented in the code to permit modeling of a flow domain with stationary solid objects. The present paper briefly describes the unique features of the COMMIX-2 computer code and presents the results of two two-phase numerical simulations with COMMIX-2. One simulation is the transient-boiling experiment in a seven-pin hexagonal fuel assembly. The other is boiling due to blockage in a square fuel assembly. The converged results of numerical simulation and comparison with measurement demonstrate the two phase simulation capability of the COMMIX-2 computer code.

Low enrichment dispersion fuels for research and test reactors

Metallurgical studies have been carried out to develop fuel dispersions that would permit the use of L ow E nrichment U ranium (LEU ~ 20% 235-U) instead of the H igh E nrichment U ranium (HEU ~ 93% 235-U) used uptil now in Research and Test Reactors. Uranium silicides are considered to be the prime candidates as dispersants in an Al-matrix; U 6Fe being a potential alternative on account of its higher U-density. The paper describes the preparation of the compounds U 3Si, U 3Si 2 and U 6Fe and their powder metallurgical processing to miniature fuel element plates with U-densities up to 7.0 Mg U m −3 in the dispersion. The reaction behaviour under equilibrium conditions at 873 K in the relevant portions of the ternary U-Si-Al and U-Fe-Al systems is described along with the dimensional changes that are associated with reaction after prolonged heat treatment of miniature fuel plates. The exothermic reactions that could occur in the dispersions under off-normal conditions at temperatures up to clad melting are described and discussed.

Comparative chemical composition and biological activity of single- and two-stage coal liquefaction process streams

Samples from several integrated two-stage coal liquefaction (ITSL) process streams were collected under both normal and off-normal operating conditions. Selected crude materials were tested for their ability to initiate tumorigenicity in the initiation/promotion assay in mouse skin. Crude materials and their chemical class fractions were assayed for microbial mutagenic activity in the standard Ames test with S. typhimurium TA98, and selected samples were also subjected to the Chinese hamster ovary mammalian cell assay for mutagenicity. Mass spectral and gas chromatographic—mass spectral (g.c.—m.s.) analyses were carried out on these materials. Results were compared with those for analogous materials from solvent refined coal (SRC) single stage liquefaction processes. In general, the ITSL distillates were of higher molecular weight, higher hydroaromatic content, lower nitrogen content and were somewhat less alkylated than the analogous SRC materials. Compared to the SRC coal liquefaction materials, the ITSL distillates and bottoms materials collected under normal run conditions were substantially less active in both the microbial and mammalian cell mutagenicity assays. These same materials were also less active than ITSL samples collected under off-normal conditions. G.c.—m.s. analyses showed that ITSL materials collected under normal operating conditions were substantially reduced in amino polycyclic aromatic hydrocarbon content when compared to both the off-normal ITSL, and SRC materials. Activity of the ITSL materials as initiators of tumorigenesis was greater than expected, based on results from the in vitro assays and not significantly different than that of the SRC materials. The higher initial boiling point of the ITSL distillates probably accounts for much of their greater than expected initiating activity compared to the SRC materials.

Safety Aspects of Fuel Behaviour in Off-Normal and Accident Conditions

Safety Aspects of Fuel Behaviour in Off-Normal and Accident Conditions

A Model for Gap Conductance in Nuclear Fuel Rods

Computation of nuclear reactor fuel behavior under normal and off-normal conditions is influenced by gap conductance models. These models should provide accurate results for heat transfer for arbitrary gap widths and gas mixtures and should be based on considerations of the kinetic theory of gases. There has been considerable progress in the study of heat transfer in a simple gas for arbitrary Knudsen numbers (Kn = l/similar to d, where l is a meanfree-path and similar d is the gap width) in recent years. Using these recent results, a simple expression for heat transfer in a gas mixture (enclosed between parallel plates) for an arbitrary Knudsen number has been constructed, and a new model for gap conductance has been proposed. The latter reproduces the free molecular (small gap, Kn >> 1) and the jump limits (large gaps, Kn << 1) correctly, and it provides fairly accurate results for arbitrary gap widths. The new model is suitable for use in large fuel behavior computer programs.

Results of the Mission Profile Life Test - First Test Segment: Thruster J1

A series of long term test segments of 30 cm diameter mercury bombardment thrusters is being conducted as the Mission Profile Life Test. The first 4000 hour segment has been completed with the J series thruster, J1. Thruster and power processing units were controlled by computer with software algorithms governing normal functions of startup, throttle, and shutdown as well as automatically handling a variety of off-normal conditions. Thruster operation includes a discussion of the test chronology describing notable events and their significance. Post-test examination provides insight into thruster lifetime. Results are consistent with mission requirements of 15,000 hours at 2A

First wall thermal hydraulic models for fusion blankets

Subject to normal and off-normal reactor conditions, thermal hydraulic models of first walls, e.g., a thermal mass barrier, a tubular shield, and a radiating liner, are reviewed. Under normal operation the plasma behaves as expected in a predicted way for transient and steady-state conditions. The most severe thermal loading on the first wall occurs when the plasma becomes unstable and dumps its energy on the wall in a very short period of time (milliseconds). Depending on the plasma dump time and the area over which the energy is deposited melting of the first wall surface may result, and if the temperature is high enough, vaporization.

Gap Conductance and Temperature Transients in Modified Pulse Design Experiments

Computation of nuclear reactor fuel behavior under normal and off-normal conditions is influenced by gap conductance models. In this paper the gap conductance results obtained by the use of the GAPHTR model and a new model that has a more sound theoretical basis in the jump regime are compared for small gaps with the modified pulse design experimental data. The new model is found to provide a better and more satisfactory prediction of the experimental data.

Experiments with bar graph process supervision displays on VDUs

Bar graph representations of process variables are compared with alternatives (stroke-type and ‘T’-type, combined bar and stroke) in experiments with human subjects, using either an automatic slide projector or a closed-circuit TV system. The stroke-type appears to give superior results when used for detection of off-normal conditions.

Important material considerations in INTOR

A number of important material-related problems were identified and analyzed during the Phase-I study for INTOR. The first wall and dlvertor collector plate are subjected to severe normal and off-normal conditions. A melt layer is predicted to develop in a bare stainless steel wall under plasma disruptions. Graphite tiles will not melt but they introduce other serious uncertainties into the design. The design strategy for the divertor collector plate focused on separating the surface and high heat flux problems and on utilizing a novel mechanical design concept for attaching tungsten tiles to a stainless steel (or copper) heat sink.

Comparisons of the RETRAN two-phase flow model with experimental data

The RETRAN computer program was designed for thermal-hydraulic analysis of nuclear steam supply systems at steady-state and off-normal conditions. A homogeneous equilibrium mixture model (HEMM) of two-phase flow is used in the program. The constitutive models for wall friction, two-phase friction multiplier, heat transfer coefficient, and critical heat flux are compared with steady-state experimental data in this report. The void fraction predicted by the HEMM is also compared with data. The results of the data comparisons indicate that the constitutive models are accurate representations and that the RETRAN numerical scheme provides an excellent framework for data comparisons.

Molten fuel radial motion and cladding melting during a PCM event in LWRs

The thermal behavior of the fuel and cladding during off-normal operating conditions, generally termed power-cooling-mismatch (PCM), are of great interest to light water reactor (LWR) safety analysis. During a power-cooling-mismatch event, fuel melting may begin at the center of the rods propagating radially outward. The induced pressure at the center of the rod due to fuel melting, fission gas release, and UO 2 fuel vapor would tend to force such molten fuel to flow through radially open cracks in the outer unmelted portion of the pellet and relocate in the fuel-cladding gap. The zircaloy cladding, which is at high temperature during film boiling, may undergo melting at its inside surface upon being contacted by the extruded molten fuel, eventually resulting in a thermal failure of the cladding. Three topics of interest are analyzed in this paper. First, fuel conditions during a hypothesized PCM accident are assessed with regard to pellet cracking and central fuel melting. Secondly, the transient freezing of a superheated liquid penetrating an initially empty crack, maintained at constant subfreezing temperatures, is studied analytically. The analysis is presented in a dimensionless form, illustrating the effect of the governing parameters, namely the driving pressure, crack shape (that is, a divergent, a parallel wall, or a convergent crack), density ratio, Stefan number for freezing, and steady state crust thickness. The calculational results are used to assess the radial extrusion of molten UO 2 fuel observed in some in-pile tests, in which PCM conditions in a pressurized water reactor were simulated. Thirdly, conditions for potential melting of zircaloy cladding upon being contacted by the extruded molten fuel are investigated analytically. The analytical predictions were consistent with the experimental results from PCM in-pile tests.

Steady-State Natural Circulation Performance of the Experimental Breeder Reactor II Primary Heat Transport Circuit

The Experimental Breeder Reactor II is a sodium-cooled fast breeder reactor and is designed to operate at a thermal power of 62.5 MW and an electrical generation rate of 20 MW. In a continuing program devoted to the understanding of the thermal, hydraulic, and neutronic behavior of this reactor under both normal and off-normal operating conditions, a series of steady-state natural convection tests have been conducted. Instrumentation utilized for the control and observation of the reactor behavior during these experiments included both the normal plant sensors as well as those located in-core within a special fueled subassembly. The results of these measurements have been compared to the predictions of an analytical model of the entire primary heat transport circuit and satisfactory agreement was obtained.