Reactor Power Level Calculator

At the reactor power levels defined In C.l above the calculated fuel surface heat flux at the point closest to burnout In any process channel shall, not exceed 70 per cent of the burnout heat flux as predicted from established burnout correlations. 

Another important property of the reactor is its reactivity behavior with changes of the thermal power output. In Table VI the essential parameters are listed for a few power levels (calculated values). 

After a few subcritical runs have been made, the reactor can be brought to delayed critical and pulsed. Care must be taken at this point, since the delayed-neutron background will yield a steady-state neutron level in the reactor. The background will steadily increase each time more neutrons are injected into the system. After enough data have been taken, the pulsing is stopped, and the reactor power level should level off and remain at steady state, since the reactor is in a delayed critical condition. The decay constant calculated for this run will yield the constant of proportionality between the decay constant and the negative reactivity (in dollars) of the system. This constant is equal to jSgff/ip as pointed out in Section II. 

Given a reactor where the time between production and absorption of neutrons is 1 ms and the power level is 1 MWe, calculate the number of free neutrons in the reactor during operation. 

This section will cover some simple calculations related to the reactor. The reactor had a cold, beginning of life, neutron multiplication factor of 1.037 0.001, which corresponds to an excess positive reactivity of 5.7 based on a delayed neutron fraction of 0.0065. The burnup for the reactor was determined using a fairly simple set of equations. The consumption over 10 years at a power level of 200 kWth was 0.8 kgs of and at 400 kWth, 1.6 kgs of would be consumed. Given that the fuel loading is 186 kgs of the burnup is -0.86% for the upper end of the uranium consumed. This burnup results in a loss of 1 reactivity. 

Subsequent calculations gave the air req.uirements as 1526 Ib/min to cool the graphite, And experimental facilities, and 174 Ib/min to cool the top thermal shield, for a total air flow of 1700 Ib/min. Also, critical experiments conducted at ORNL indicated that the heat generation in the reflector would be somewhat less than that used for the above calculations and the air requirements would therefore be less. However, the design figures were maintained at 2000 Ib/min flow and 55 in. of water pressure drop since it may be desirable and possible to operate the reactor at a power level as high as 60,000 kw at some time in the future. 

Using of the calculation code, that substantiate the allowable power levels of the reactor BN-350 in conditions of absence of the seismic resistant system for emergency cooling of an active zone, the EPPE experts had conducted precursory calculations of the reactor s parameters in case of active zone cooling by natural circulation of the heat-carrier. 

After the types of reactor assemblies have been characterized by equivalent few-group cross sections as a function of exposure, these assemblies are used to analyze the reactor core as a whole. Each of the assemblies is placed in its appropriate location and is then allowed to be exposed based on the calculated assembly power levels across the core. This allows the analyst to accurately predict the Ic-effective of the core and the varying spatial power shape across the core and the assembly contents as the reactor ages. 

Solution of the space-time Equation 21.24 is usually done either at the assembly level (Step 6), where the fission powers of each of the assemblies are the variables of the equation, or at the reactor level, where the power level of the reactor as a whole is calculated and then the powers are "distributed" among the assemblies based on the calculated power profiles from Step 6. To show how the equations work, we will assume that the second path is taken and a reactor-wide kinetic power calculation is performed. 

The major concern of safety when working at critical facilities was the risk of criticality accidents. These facilities were meant to operate at zero or at very low power level. That is why the shielding and provisions for heat transport were often very limited. The safety was based on design calculations, reliable monitoring of power level, reactor period, and distance from critical condition. 

The reactor will be pulsed with reactivity insertions from 1% to 1.4%(or the maximum available) in 0.1% increments. Calculate the transient rod fired position to obtain the desired reactivity insertions. One pulse will be fired for each reactivity step. A second pulse will be fired for one of the steps to provide an indication of the reproducibility of the measurements. The signal from the gamma ion chamber is sampled at one millisecond intervals and stored in the TestLab. The TestLab internal program calculates the information needed for the reactor log and records two channels of power-level data derived from the gamma-ray dose from the reactor core. The first channel, labeled "pulsetrace", records the full pulse. The second, labeled "pulserise", uses a smaller range and thus records the early part of the pulse most useful for reactor period measurements. In both cases the raw data is normalized such that the output is in MW. 

Compare the movement of the transient rod with the prediction from your pre-lab calculations. Suggest methods of assuring all reactivity insertion has been completed before the reactor begins to experience reactivity feedback from the increasing power level. 

Equation (4.6) can be used to calculate reactor power in a prompt critical reactor at any time following the start of a transient. Use the example from an earlier chapter P(0) = 3300MW, Keff = 1.001, AK =. 001, t = 1 second. The power level after a one second transient is ... 

There are a number of techniques for meastuing subcritical reactivity relative to a calibrated reference control rod, in addition to soxuce multiplication. These include rod drop, rod jerk, source jerk, pulsed source and reactor power noise. Account must be taken of spatial flux transients (either by calculating them or measuring them with arrays of covmters) and of the spatial distribution of natural neutron sources due to spontaneous fission and (a,n) reactions and any fixed sources introduced to increase the subcritical flux level. The different methods have been reviewed and intercompared, for example, at the 1976 Specialists Meeting [4.87]. 

From the information given in Paper 6, It Is difficult to Judge the suitability of the SGHWR for use as spinning reserve. Are the curves of Pig. 5 based on calculation or tests What was the power level prior to the transient Have the Authors calculated the total heat stored in the steam mains and drum of the SGHWR at, say, 75% power What Increase in drum size do they think would be needed to maintain a 10 increase in power output for 20 s without reliance on increased power from the reactor, at 75% load ... 

The outer radius of the cadmium-poisoned zone is set by the criteria of operating for 10 years at a power level of 500 kW. Note that the reactor must be critical at operating temperature—nominally 1000 K—and dry yet be subcritical when flooded at room temperature—300 K. Calculations with SCALE/KENO-V.a (238-group, ENDF/B-V cross sections) showed that an outer radius of the poisoned zone of 75 cm was required. Under the assumption that a right, circular cylinder yields the minimum critical mass in cylindrical geometry, a height of 150 cm was determined. 

The 1962 siting criteria reflected and were consistent with prior site decisions on commercial power reactors. This was shown through the application of the criteria s calculations to several reactor projects that had been proposed or were authorized for construction. The Commonwealth Edison Dresden reactor, for example, with a power level of 630 megawatts thermal, had a calculated exclusion distance of 0.5 miles, which corresponded exactly with the actual exclusion distance at the... 

The main input parameters required for carrying out the options given in the above menu are Reactor power. Moderator level. Coolant inlet temperature to the core. Secondary coolant flow rate, temperature rise of secondary coolant across the heat exchangers, heavy water level in the Dump tank and Drop time of individual shut off rods. All these data items are stored continuously in two files. The programme when invoked will access the required data from these files and does the individual operation/function as demanded. It keeps updating the current status of any calculated parameter by scanning the data in the file every 2 minutes. 

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