Loss Of Coolant Research Articles

Effect of the loss of coolant on the resonance escape probability

In the analysis of the loss of coolant accident in natural uranium reactors of the Atucha or Candu type, the influence of the coolant voiding on the resonance escape probability appears to be of great importance. In this paper we study this effect on the resonance absorption, analyzing its behaviour vs coolant density. We use the codes COCAP and WIMS, ∗ ∗ We have used WIMS D version of this code. in which the shadowing among pins in the cluster, or the Dancoff effect, is taken into account with different models. In COCAP, the Dancoff factor calculation based on Carlvik's method shows a remarkable dependency on the coolant density, and its values at very low densities lead to an overestimation of the cell reactivity. It is shown that the inclusion of a Bell factor in the definition of the cluster effective surface improves the agreement with experimental results.

Experience with the Fort St. Vrain reactor

Public Service Company of Colorado began electric power generation in late 1976 at its Fort St. Vrain Nuclear Generating Station located near Platteville, Colorado, 35 miles north of Denver. The plant utilizes the advanced high temperature gas cooled reactor (HTGR) concept. General Atomic Company designed and constructed the reactor using Sargent & Lundy as architect-engineer and EBASCO as general construction contractor. The research and development associated with this endeavor was undertaken under the former Atomic Energy Commission Power Reactor Demonstration Program. Preliminary low power and intermediate power tests have proved the design concept features and calculations and most of the components. Safe conditions following loss of coolant were confirmed. Release of radioactivity from the fuel was less than predicted up to 70% full power. Modifications and repairs caused delays and included, chiefly: Pelton wheel emergency cooling drives, helium circulator bearings, the pressure-regulating system for water lubricant in circulator bearings, liner cooling lines, removal of moisture from the vessel, electrical cable rerouting, helium seals around control rod drives, and a leak in one steam generator. Core temperature oscillation problems are being analyzed before passing the currently imposed 70% power level. Fort St. Vrain has demonstrated that the concept of a large HTGR system incorporating a prestressed concrete reactor vessel is sound. The station has produced more than 500 million kilowatt hours of electric energy to date.

Cooling mechanism during transient reflooding of a reactor fuel bundle after loss of coolant

Abstract Cooling efficiency during transient reflooding under loss of normal coolant conditions has been examined with a 7 × 7 simulated fuel rod bundle and jet pump bypass. The bundle contains 49 electrically heated rods with 3600 mm heated length and a pseudo cosine axial power distribution. Water is injected into the lower plenum and the superheated bundle is reflooded from the bottom with some flow diverted to the simulated jet pump bypass. The results show that effective cooling can be maintained.

An integrated safe shutdown heat removal system for light water reactors

A design concept is introduced for an integrated safe shutdown heat removal system (ISSS) for light water reactors that is independent of all components and systems outside the primary containment (other than the elements of the ISSS itself) and which includes an integral stored water supply. The principal purpose is to obtain a simple, reliable, and highly protected means for water makeup and steam release for basic post-scram cooling. The objective of the system is to provide safe shutdown cooling under virtually all emergency conditions which do not involve the loss of coolant caused by piping failures within the primary containment. Thus, the system is designed for the hazards of fire and sabotage as well as numerous other potential accidents, all of them considered as a set of events to be handled by the same system, rather than as discrete problems — each with its own, sometimes contradictory, solution.

The collapse of heat exchanger tubes with ovality and simulated defects

Heat exchanger tubes of a pressurised water reactor steam generator form part of the boundary between the primary and secondary circuits of the reactor. It is important that no gross breach of this boundary occurs at anytime. Normally the tubes are subjected to internal pressure. However, should a loss of coolant occur in the primary circuit, these tubes are subjected to external pressure. This investigation compares theoretical and experimental results for such tubes, firstly with varying amounts of ovality and secondly with defects artificially induced to simulate: 1, Longitudinal tube cracking; 2, Fretting at tube supports; 3, Erosion. The collapse is compared with the strength of an original perfect tube. Collapse was found to be of a ductile nature in every instance, and in no case did the defect propagate through the wall of the tube. The findings are also of general interest for any externally pressurised tubing.

Direct water injection cooling for military engines and effects on the diesel cycle

A study was conducted on the feasibility of totally cooling a single-cylinder diesel engine by direct injection of water into the combustion chamber. The term “total cooling” can be taken to mean stabilized cooling at all loads and speeds so as to eliminate need for conventional cooling jackets, cooling fins, or oil spray jets. The engine used was a CLR Direct Injection Diesel with 42.5 cubic inch displacement and a compression ratio of 16:1. Most of the running was at 1800 rpm and 92 psi IMEP. Separate measurements were made of heat rejection to the cylinder head, liner, and crank-case oil to determine more accurately where the cooling effect was being applied. Water injection was by means of a Bosch pump and various pencil-type nozzles installed, adjacent to the fuel injector in the cylinder head. Port injection and port induction were also briefly investigated. A five-hole, 90° included angle nozzle was used, as was a three-hole, 30° included angle unit. For comparison, a nozzle directing one spray obliquely at the cylinder wall was also tested. Firing pressure was monitored using a piezo-electric transducer; both pressure-time and pressure-volume (indicator) records were obtained. In order to determine timing of both fuel and water injection, needle lift was monitored using a differential transformer pickup. The results of this study indicate: t Optimum total engine cooling by direct water injection was accomplished over a wide range of water injection timings (from 450 to 720 CA degrees after TDC power stroke) at water/fuel ratios of 2.9 to 3.7 with output power and brake specific fuel consumption improved 5 to 20%, respectively, over that with the standard jacket-cooled CLR engine. Emissions are affected in an expected manner by the presence of water: NO x is decreased, sometimes substantially, while the other emissions (HC, CO) tend to increase. When cooling the exhaust, the condensate becomes an effective scrubber of sulfur oxides. NO x was not significantly reduced by scrubbing, but if the condensate is made sufficiently alkaline (pH>8), CO 2 was unintentionally scrubbed out. The quality of the uncondensed exhaust for turbocharging is attractive. A theoretical gain of about 17.5% in available exhaust energy due to generation of steam was calculated, along with a temperature decrease of several hundred degrees Fahrenheit. Water contamination of the lubricating oil varies from negligible to extreme, depending on injection quantity, timing, and spray pattern. By not directing water at the liner wall, and by keeping the oil above 212°F, one can maintain the oil in a dry condition. Based on this work, several pertinent recommendations have been made: (1) utilize water injection for short-duration, very high-output operation which would otherwise be destructive due to thermal overload; (2) use water induction cooling in event of loss of conventional liquid coolant; (3) utilize exhaust scrubbing in stationary applications to permit burning of high-sulfur fuels without producing sulfur oxide emissions; nitrogen oxides could likewise be reduced by the injection of small amounts of water; and (4) since 2-stroke-cycle engines are an important category of diesel engines, some work similar to this effort should be done to this engine type; prospects are good for success, but conditions are apt to be more restrictive.

The case for gas cooling

The objective of this paper is to discuss the technological motivations for gas cooling of nuclear power reactors, as there are: (1) pure single-phase operation; (2) low neutron absorption and moderation; (3) chemical and radioactive inertness in the case of helium; (4) transparency; (5) total coolant loss impossible — only depressurization; and (6) low total stored energy. The present discussion is primarily in the context of the needs of the fast breeder reactor. More generally, it is pointed out that gas cooling allows full utilization of the most modern steam plant development, with substantial advantages both to plant cost and efficiency. It also makes feasible the intergral PCRV type of construction. The prospect of future combination with the gas turbine offers a unique field for future reduction of plant cost.

The AEC and the Loss of Coolant Accident

Professor Wilson gives the background to the concern being expressed about the risks associated with the possible loss of coolant from a nuclear reactor.

Ratings and Applications of Power Thyristors for Resistance Welding

Guidelines are given for applying SCR's in resistance welding controls. The stresses the SCR's are subjected to in resistance welding applications due to the cyclic nature of the load are discussed from a device, design, and application point of view. In addition the special rating of SCR's and cooling requirements for cyclic duty are described along with techniques for protecting the devices from overvoltage, overcurrent, and loss of coolant. Gating requirements are outlined from a circuit view point.

Explosive Disassembly of Fast Reactors

Theoretical analysis of the explosive disassembly of fast reactors, following accidental loss of coolant and melting and gravity compaction of the fuel material,has been undertaken. A general expression for the rate of reactivity feedback dueto material disassembly in a core with more than one enrichment zone has been derived. Energy release in cylindrical geometry has been calculated with the effects due to zonal interfaces properly considered in the limit of zero acoustic speeds. The effect on the energy release of the assumption of fixed fuel density in the fuel equation-of-state has been investigated for oxide-fueled zoned cores.Fuel zoning can result in either greater or less energy release, as compared with a homogeneous core of the same fuel inventory and the same initial conditions, depending on the gradient of the density worth function at the zonal interface. In some large fast breeder reactors the gradient of the density worth function may be negative at the zonal interface. The positive pressure gradient at the interface during the disassembly phase of a transient results in inwardly directed fuel displacement at the interface and in a positive contribution to the disassembly reactivity feedback. Typically, this can increase the energy released in a disassembly transient by as much as 35%. The effect on energy release of the assumption of fixed fuel density in the fuel equation-of-state was not significant for the transients initiated in the completely molten sodium-free reactor cores studied in this investigation.

Applications of similitude to the structural mechanics of reactors

The structural mechanics of a reactor involves the prediction of the behavior of the components of the reactor and the entire reactor system under a combination of gravitational and other force fields, electrical and magnetic fields, thermal fields and radiation fields. The nature of the materials and components is such that under the operational conditions imposed, the behavior is often nonlinear, with the result that the application of the usual techniques of mathematical analysis to predict behavior often results in lengthy computations based on assumptions that are somewhat subjective. Although computer techniques have proven very helpful in many instances, they are often limited in scope through size of available computer storage. The direct approach, that of cut-and-try design, is often not appropriate because of the possibility of release of radioactive materials in the event of a faulty design or a failure of the containment or other materials. In addition, economics, as well as safety, usually precludes construction of a complete operating unit without preliminary analysis Under these conditions of complex mathematical analysis and expensive construction the use of small scale models is suggested as a method of obtaining data from which realistic predictions may be made. In the design of small scale models the theory of similitude is used in the same way that it has been used in many fields of engineering. It is also pertinent to consider current power reactors in the 400-MW range as models of projected reactors in the 2000-MW range. Direct application of the principles of similitude leads to a series of design conditions that cannot be satisfied simultaneously. In such a situation the application of distorted model theory is mandatory. The application of distorted model theory requires the determination of the prediction factors associated with the pertinent distortion factors derived from the significant parameters that are distorted. Evaluation of the prediction factors as functions of the distortion factors may be accomplished experimentally or analytically, and the remainder of the presentation is devoted to consideration of the applicable methods and procedures for determining the prediction factors. The analytical procedures may be classed as those based on well established laws and those based on fundamental principles. Specific illustrations of the application of the principles includes fuel element containment, distortion of primary containment, performance of containment shells as a result of sudden release of energy within or from external disturbances, and thermal strains resulting from loss of coolant and subsequent sudden cooling. As the numerical examples will illustrate, distorted model theory often enables one to make practical judgments concerning which distortions are significant and which will have a trivial influence upon the results.

A gas-cooled fast reactor core model for the analysis of severe accidents, with applicability to LMFBRS

A new fast-reactor core model has been developed which allows simulation of the behavior of a multizone core during accidental transients (loss of coolant or external reactivity insertions) leading from steady-state operation to fuel-rod failure and then to gross core damage. The model for the rod failure and for the relocation of the fuel debris is based on an interpretation of the meltdown experiments performed in the TREAT reactor. In the presented core model, the thermal response of the fuel rods is determined using a two-dimensional heat-transfer code in r-z geometry. In addition to the standard kinetics feedback relations, the reactivity effects of the gradual removal of the rod cladding are included. Taking into account the addition and removal of the heat of fusion allows the amount of fuel that resolidifies in the lower core part to be estimated. The presence of rod spacer arrangements is included as additional barriers for the debris relocation. The thermal response of the subassembly walls is calculated taking into account the contact with the high-temperature fuel as well as radiative heat transfer.

A simple coolant blowdown model for the evaluation of safety problems in water-cooled reactors with pressure vessels

A simple coolant blowdown model applicable to water-cooled pressure vessel reactors has been developed. It is used to predict the reactor pressure-temperature transients during postulated loss of coolant, loss of heat sink, pump failure and similar accidents, with the aid of a small capacity digital computer. A three-zone one-dimensional presentation of reactor and pressure vessel is adopted. The physical conservation laws are solved for the reference BHWR, BWR and PWR designs by introducing a number of pessimistic system assumptions. The rates of heat generation, feed water inflow and enthalpy and several other quantities were allowed to vary. Moody's model was used to calculate the discharge rate of coolant from the pressure vessel after a rupture. The model is then applied both to a direct cycle BHWR and to an indirect cycle BHWR with integrated steam generators. Numerical solutions to specific postulated situations are presented. The influence of major assumptions and principal input parameters has been investigated.

Velocity of Sound and Critical Discharge Pressure in Annular Two-Phase Flow

The effects of a rapid loss of coolant caused, for example, by the fracture of a pipe in the primary coolant circuit play an important part in the safety assessment of a water-cooled nuclear reactor. In particular, it is essential to know the rate of decrease of coolant density in the core and the rate of discharge of coolant into the containment space. It is uneconomic to obtain this information from experiments on each new reactor. Consequently computational methods, supported by experiments on basic elements of a coolant circuit, are being developed which will be applicable to a wide variety of reactors. An experiment is described in which fundamental information on fluid flow dynamics was obtained. In the experiment, a flow of two-phase water at high pressure was established in a straight pipe 15 ft long and 0–621 in i.d. At the downstream end of the pipe a rupture disc was suddenly burst. The pressure versus time histories at three locations along the pipe were recorded, using fast response pressure transducers and a multichannel oscilloscope. Experimental results are given for pre-blowdown pressures of 500 and 1000 lb/in2 (abs.) and steam qualities of 10–30 per cent. The mass flux was 0·6 times 106 lb/ft2 h, and the flow regime was annular. From the recorded pressure traces, the speed of the head of the rarefaction wave travelling upstream into the fluid was measured and also the pressure at critical flow conditions as functions of the pre-blowdown conditions. These measurements are compared with values calculated from steam tables.

An Experimental Study of Spray Cooling in Nuclear Reactor Containers

As a contribution to the safety analyses of condition following loss-of-coolant accidents in light water reactor, experiments were performed to obtain data on container spraying and on accompanying heat transfer characteristics, by means of a one-tenth scale model container. The loss of coolant was simulated injecting high temperature saturated water into the model container in which air had been enclosed at atmospheric and room temperature, and then the steam-air mixture was cooled by sprays of water. The rate of heat absorption by the spray was obtained from both stationary and transient data, and it was confirmed that the absorption becomes incomplete as the container pressure is lowered. The evaporation of the water collected at the bottom and of that from sprayed walls of the container was derived from container pressure variation in time, by measuring its rate of change immediately following stoppage of the spray; the rate of evaporation was found more active as the rate of spray cooling increases. The evaporation from the sprayed walls was obtained from thermal data taken on a copper cylinder placed in the container, and it was found that the evaporation becomes more active as the pressure increases.

An Experimental Study of Spray Cooling in Nuclear Reactor Containers

As a contribution to the safety analyses of condition following loss-of-coolant accidents in light water reactor, experiments were performed to obtain data on container spraying and on accompanying heat transfer characteristics, by means of a one-tenth scale model container. The loss of coolant was simulated injecting high temperature saturated water into the model container in which air had been enclosed at atmospheric and room temperature, and then the steam-air mixture was cooled by sprays of water. The rate of heat absorption by the spray was obtained from both stationary and transient data, and it was confirmed that the absorption becomes incomplete as the container pressure is lowered. The evaporation of the water collected at the bottom and of that from sprayed walls of the container was derived from container pressure variation in time, by measuring its rate of change immediately following stoppage of the spray; the rate of evaporation was found more active as the rate of spray cooling increases. The evaporation from the sprayed walls was obtained from thermal data taken on a copper cylinder placed in the container, and it was found that the evaporation becomes more active as the pressure increases.

Loss of Coolant from a Simulated Reactor Loop

Experimental data on the loss of coolant from a pressurized-water test loop are presented. The test loop simulated the components of a full-sized reactor plant with an electrically heated test section, a pressurizer, a heat exchanger, and a circulating coolant pump. Leaks in the coolant system were simulated by rupture disks at various locations.The test results are compared with predictions using the FLASH computer program. The calculated system-pressure transients showed good agreement with the tests. The pressure differential across the test section exhibited an oscillatory behavior, as did the predictions in one of the three tests. Measured test-section temperatures showed a considerable variation from predictions; however, these temperature predictions represent an improvement over those using previous calculation methods.

Behavior of Cladding Tube under Coolant-Loss Accident Conditions

The behavior of cladding tubes, both stainless steel and zircaloy under the post-accidental conditions of coolant loss has been discussed, based on the results of three experiments performed this time. The effect of oxidation in post-accidental steam atmosphere up to temperatures where the cladding tube should burst with the pressure of entrapped fission product gas was found to be negligible. The bursting behavior of tubes containing pre-determined amounts of inert gas corresponding to the amounts of released fission product gas that should be present in various states of fuel usage was also examined. No effect of corrosion in zircaloy-2 that should affect its strength could be found up to a weight gain corresponding to several years of use under normal operating conditions, either in the burst strength or creep burst behavior.

Behavior of Cladding Tube under Coolant-Loss Accident Conditions

The behavior of cladding tubes, both stainless steel and zircaloy under the post-accidental conditions of coolant loss has been discussed, based on the results of three experiments performed this time. The effect of oxidation in post-accidental steam atmosphere up to temperatures where the cladding tube should burst with the pressure of entrapped fission product gas was found to be negligible. The bursting behavior of tubes containing pre-determined amounts of inert gas corresponding to the amounts of released fission product gas that should be present in various states of fuel usage was also examined. No effect of corrosion in zircaloy-2 that should affect its strength could be found up to a weight gain corresponding to several years of use under normal operating conditions, either in the burst strength or creep burst behavior.

The Comparative Study of the Sodium Loss Reactivity Effect of Large Pu-U and Pu-Th Fast Reactors

Fast reactors having a 3,000 l core volume fueled with ceramic Pu-U and Pu-Th are compared with each other for their reactivity effects due to large-scale loss of coolant. Using one-dimensional multi-group calculation, the maximum volume fractions of fuel for both fast systems are determined on the assumption that total loss of sodium in the core should not result in positive reactivity effect. It is concluded that Pu-Th fast reactors can be charged with more than several volume percent of fuel above the limit applicable to Pu-U fast reactors.