Measurement reactors

Solution Example 4.5 was a reverse problem, where measured reactor performance was used to determine constants in the rate equation. We now treat the forward problem, where the kinetics are known and the reactor performance is desired. Obviously, the results of Run 1 should be closely duplicated. The solution uses the method of false transients for a variable-density system. The ideal gas law is used as the equation of state. The ODEs are... 

Whenever it is necessary to measure reactor neutron flux profiles, a section of wire or foil is inserted directly into the reactor core. The wire or foil remains in the core for the length of time required for activation to the desired level. The cross-section of the flux wire or foil must be known to obtain an accurate flux profile. After activation, the flux wire or foil is rapidly removed from the reactor core and the activity counted. 

The determination of rate change of the logarithm of the neutron level, as in the source range, is accomplished by the differentiator. The differentiator measures reactor period or startup rate. Startup rate in the intermediate range is more stable because the neutron level signal is subject to less sudden large variations. For this reason, intermediate-range startup rate is often used as an input to the reactor protection system. 

Figure 2,13a and b shows two control structures that work (CS4 and CS1), Both of these provide a mechanism for adjusting the fresh feed reactant flowrates so that the overall stoichiometry can be satisfied. In CS4 this is accomplished by measuring reactor composition. In CS1 it is accomplished by deducing the amounts of the reactants in the process from two levels in the two recycle loops. 

In the reference test, the low reactor exit temperature at the constant plant-energy input conditions indicates the expected higher heat losses in a short-duration reactor. The corresponding lower overall temperature profile through the test-reactor length reduces the process kinetic time-at-temperature. The associated gas-phase chemical kinetics at the lower residence times are believed to be responsible for the slight discrepancies in the reference test gas yields. Also, the "true enthalpy used for cracking is lower than that indicated by the measured reactor temperature. 

The plasma operating conditions performed in the study are summarized in Table 19.1. The monomer flow rates were determined by measuring the system pressure increase over the given time interval from the isolation of vacuum pump and then converted to flow rate (seem). Each reactor volume was measured by gas expansion method small size = 557 cm, medium size = 1278 cm, and large size = 2357 cm. (The measured reactor volume includes the volume of connecting tubes beyond the center part glass tube.)... 

Figure 4 shows the experimental data taken for hydrogen-air combustion. The solid squares are the result of a complete kinetic calculation (not assuming partial equilibration) (26), treating the jet-stirred reactor as a well stirred system and inputing measured reactor temperature or heat loss rates. The reaction scheme included the following ... 

Fig. 3 Measured reactor bed temperatures. Locations of thermocouples 3 - points (+), 2 - points ( ), 1 - points (o) - see Fig. 2.

For this high throughput serial screening approach to succeed, the observed discrepancy between geometrically calculated and experimentally measured reactor volumes must be resolved. If this issue cannot be resolved, quantitative reaction rate data cannot be generated using the pulse injection mode of operation. 

Figure 2. Measured reactor outlet temperatures vs. oxygen fed per...

In defense of the simplemixing models, however, they do provide another important insight. We can determine the limits of achievable mixing consistent with a measured reactor residence-time distribution. These mixing limits do provide some insight into the limits of achievable reactor performance, although this connection remains an active area of research as discussed in section 8.4. 

Three extended wide range nuclear channels combine signals from the three in-vessel detectors with signals from three ex-vessel detectors to measure reactor power on a logarithmic scale and rate of power change from startup source level to full power. These signals are used for rod drive control as the reactor is started up and brought to power. 

Coinciding with the above measures. Reactors 1 and 2 were shut down in the early hours of Sunday 27 April 1986. Dosimetric and plant damage surveys were carried out in Reactor 4 building, as far as this was possible in the severely hazardous conditions prevailing. Most of the installed instrumentation in the reactor had been destroyed. Aerial surveys from helicopters, using infrared imaging equipment and lowering radiation detectors over the core, enabled the Commission to decide a course of action in an attempt to blanket the fire. 

The reactor, plates as well as the housing, is constructed of stainless steel. Each plate contains 34 rectangular channels of 300 pm width, 240 pm depth and 20 mm length. It contains S reaction plates (coated with catalyst) and 5 thermostatisation plates glued together. The deposited catalyst mass is 75 mg, the measured reactor volume, Vr, is about 0.5 cm. For the experiments described here, the reactor block is heated from the outside by an electrical tape. As the enthalpy of the model reaction is small and low reactant concentrations are applied, a homogeneous temperature distribution is obtained. The temperature inside the reactor is measured using a K-type thermocouple (Philips AG, Dietikon, Switzerland) introduced into the outlet tube. 

The experimentally measured reactor inlet concentrations were used to produce model predictions for the concentration responses at the reactor outlet. The reactor inlet concentrations of N2, CO2, N2O and NO2 were always kept zero. 

J. F. WALTER and A, F. HENRY, The Asymmetric Source Method of Measuring Reactor Shutdown, NucL /. n .,32,332(I968). 

The nuclear instrumentation systems are designed to measure reactor neutron flux over the full operating range of the reactor. These measurements are required as inputs to the reactor regulating system and safety systems. The instrumentation for the safety systems is independent of that used by the reactor regulating system. 

Work executed since 1990, even core unloading, resulted in yearly collective doses of personnel less than those in the years of operation (FIG. 5). The highest value during decommissioning occurred in 1995 due to the hands-on decommissioning of the small bura-up measuring reactor and the enclosing of the PCRV-penetrations. 

Loading and unloading of the THTR reactor core was carried out while the bumup of the pebble-shaped fuel elements was monitored by means of a bumup measuring system. The main component of this system was the bumup measuring reactor (Solid Moderated Reactor) in which a reactivity effect is caused by operating elements as they pass through the reactor. Evaluation of this effect permits determination of the type of element and, in case of fuel elements, the bumup of the element. 

Unloading of the THTR reactor core was completed by 28 October 1994 with the establishment of the state "reactor core free of nuclear fuel". By then, the task of the bumup measuring reactor was completed so that dismantling of the bumup measviring reactor could be initiated in order to remove the nuclear fuel contained therein. 

The bumup measuring reactor was a graphite moderated thermal reactor with a rated output of 500 W, arranged in the reactor hall below the prestressed concrete reactor vessel. The reactor core (1.0 m 1.2 m 1.1 m) consisted of various graphite plates provided with grooves for accommodation of the 767 strip-shaped fuel elements. The fuel elements have a rectangular cross-section (15 mm 1.1 mm) and a length of between 89 and 711 nun. They contain 93% enriched uranium in a U-Al alloy (20% uranium, 80% aluminum). Total uranium content of the core was 3.9 kg. 

The bumup measuring reactor had to be dismantled only to the extent necessary for removal of the nuclear fuel. 

Cold reference leg" water level instruments measure reactor vessel water level by measuring the differential pressure of two columns of water — the variable leg and the reference leg. The reference leg is maintained filled to a constant height of water by a condensate chamber located at the top of the reference leg. Non-condensible gases can collect in the condensate chamber and can become dissolved in the water at the top of the reference leg. These dissolved gases can be transported down the reference leg by small leaks in valves and fittings at the bottom of the reference leg. 

After the pre-core reactor coolant system flow rate measurement is taken, analytical adjustments are made to the pre-core measured reactor coolant system flow rate to predict a post-core reactor coolant system flow rate. Calculations of the reactor coolant system flow rate with and without the core loaded are performed. The calculation of the pre-core load reactor coolant system flow rate is compared with results of the pre-core load flow testing and this information will be used in the calculation of the post-core load reactor coolant system flow rate as appropriate. The predicted post-core load reactor coolant system flow rate is checked to verify that it satisfies Technical Specification 3.4.1. Verifications are also made that the post-core load reactor coolant system flow rates satisfy LCO 3.4.1 of the Tech-Specs flow limits during start-up testing. 

Each of these lags forms one link in the chain from average coolant temperature to the measured reactor temperature. But since the manipulated variable is the flow of water added to the coolant stream, a suitable equation converting water flow to coolant temperature must be included. Adding a stream Fw at temperature Tw to the coolant recycle stream F — Fw at temperature 7 2 produces a mixture F at temperature Td, returning to the reactor. The heat balance is... 

Measurements of moderator height coefficients have been made by measuring reactor periods with excess moderator height. The coefficients are 2 % smaller than predicted but are within control system tolerance. Smaller discrepancies of about Q% had previously been observed in zero energy measurements in DIMPLE. 

Experimental Breeder Reactor Shield Measurements," Reactor Shielding Information Meeting, Chicago, November 12 and 13, 1953, WASH-152, March, 1954, pp. 34-35. 

During the period of core damage, there was virtually no information on conditions in the core. Incore thermocouples, which measure reactor coolant temperature at the exit from the core, could only show temperatures as high as 700°F due to limits imposed by the signal conditioning and data logging equipment, not by the thermocouples themselves. 

Measure reactor coolant temperatures for plant and component control... 

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