Reactor power distribution

More generally, the neutron number density and the reactor power distribution are both time- and space-dependent. Also, there is a complex relation between reactor power, heat removal, and reactivity. 

It also became necessary to consider interactions between the nuclear and the thermal-hydraulic performance. Iterative calculations, notably between the AIMAZ (nuclear) and CUSH (core and primary circuit hydraulic models) were needed to reach final estimates of overall reactor power distributions and void coefficients. 

The Model 412 PWR uses several control mechanisms. The first is the control cluster, consisting of a set of 25 hafnium metal rods coimected by a spider and inserted in the vacant spaces of 53 of the fuel assembhes (see Fig. 6). The clusters can be moved up and down, or released to shut down the reactor quickly. The rods are also used to (/) provide positive reactivity for the startup of the reactor from cold conditions, (2) make adjustments in power that fit the load demand on the system, (J) help shape the core power distribution to assure favorable fuel consumption and avoid hot spots on fuel cladding, and (4) compensate for the production and consumption of the strongly neutron-absorbing fission product xenon-135. Other PWRs use an alloy of cadmium, indium, and silver, all strong neutron absorbers, as control material. 

Use of a saturable reactor (choke) to balance a large unbalanced power distribution system... 

Figure 28.27 Design of a reactor for a power distribution bus system. Illustrating Example 28.10...

Incidence Models. First attempts to describe the radiant power distribution in photochemical reactors can be summarized under the heading of the RI model (radial incident model, Figure 25a). This model is based on the hypothesis of a radial radiation field [2, 3], that is, that all the light striking the reactor wall will be directed radially inward. Corresponding profiles of radiant power or of irradiance are strongly dependent on the radius of the cylindrical reactor (Eq. 68). 

These conditions are (i) it is needed to have a validated kinetic scheme (a detailed mechanism or a precise empirical representation), (ii) it is needed to have a validated, intrinsic reaction kinetic expression as a function of position and time [R( r, t)], (iii) it is needed to use in both reactors the same spectral radiation output power distribution [A for monochromatic radiation and/(A) for the polychromatic cases], and (iv) it is needed to apply and correctly solve a rigorous mathematical model to both the laboratory and the large-scale reactor. 

Rieck [Rl] has used the computer codes LEOPARD IBl] and SIMULATE [FI] to predict the power distribution in the fuel and poison arrangement shown in Fig. 3.19 for the first fuel cycle for this reactor, and the amount of thermal energy produced by each assembly up to the time when the reactor ceases to be critical with all soluble boron removed from the cooling water. Figure 3.20 is a horizontal cross section of one-quarter of the core of this reactor. Each square represents one fuel assembly. The core arrangement has 90° rotational symmetry, about the central assembly 1AA at the upper left of the figure. 

Hie preceding example shows that accurate representation of fuel-cyde performance requires such detailed examination of the time-varying dianges in power distribution and fuel composition at many points in a nuclear reactor as to necessitate use of a digital computer. The purpose of this section is to develop an approximate procedure for calculating fuel-cycle performance that, although complicated, can be carried out by hand calculation. The FWR described in Sec. 4.1 will be used as example, except that its rated electric output is taken as 10S4 MWe instead of 1060 MWe. 

Fission counters are used extensively for both out-of-core and in-core measurements of neutron flux in nuclear flux in nuclear reactors. In out-of-core situations, they monitor the neutron population during the early stages of power ascension when the neutron flux level is very low. For in-core measurements, fission counters are used for flux mapping (and consequently, determination of the core power distribution). They are manufactured as long thin q lindrical probes that can be driven in and out of the core with the reactor in power. Typical commercial fission counters for in-core use have diameters of about 1.5 mm (0.06 in), use uranium enriched to at least 90 percent in as the sensitive material, and can be used to measure neutron fluxes up to 10 neutrons/(m s) [10 neutrons/(cm s)]. 

Rhodium, vanadium, cobalt, and molybdenum have been used as emitters for SPNDs. Since rhodium SPNDs are the main in-core instruments for the determination of power distribution in pressurized-water reactors (PWR), they are discussed first and in greater detail than the others. 

The reflector is installed inside the reactor vessel and the heat generated in the reflector is cooled by sodium. The equivalent core diameter is 0.8m which satisfy negative void reactivity requirements. The reflector length is 1.5m and the reflector gradually moves up to control the reactivity leading to bum-up. The axial power distribution changes as shown in Fig. 3 according to the reflector position. 

Trips shall be provided to annunciate (audible and visual) and scram the reactor for exponential power level rate Increases having periods of 15 seconds or less. Trips and instrumentation ranges shall be fixed when the reactor power level is in this range (lO" to 10 ). The xesponse tlme of the system shall be k seconds or less. Geometric distribution of transducers is not important in first four decades of this range (10 7 to 10 3), All transducers shall be located in one general area to monitor ths same power level. 

In total, 104 abnormal operation events have occurred during power unit operation (as of December 2001) resulting in the unscheduled reactor power decrease. Distribution of these events with respect to years, components (irrespective of the cause) and causes, respectively, is shown in Figs 4-6. 

The entire active zone is divided arbitrarily into groups of cells containing 12 fuel channels with fuel elements, two channels with auxiliary absorber rods and two channels with control and safety rods. This grouping is varied at the reactor periphery by reducing the number of control and safety rods to flatten the radial power distribution. 

The control rod system provides for automatic control of the required reactor power level and its period reactor startup manual regulation of the power level and distribution to compensate for changes in reactivity due to burn-up and refuelling automatic regulation of the radial-azimuthal power distribution automatic rapid power reduction to predetermined levels when certain plant parameters exceed preset limits automatic and manual emergency shutdown under accident conditions. A special unit selects 24 uniformly distributed rods from the total available in the core as safety rods. These are the first rods to be withdrawn to their upper cut-off limit when the reactor is started up. In the event of a loss of power, the control rods are disconnected from their drives and fall into the core under gravity at a speed of about 0-4 m/s, regulated by water flow resistance. 

The power density distribution is controlled by the 12 LAC and 24 LS rods. The average power control system is used as standby in the 20-100% power range and is switched on automatically when the LAC system malfunctions. The automatic control system holds reactor power to within 1 % of the required output in the range 20-100% full power and to within 3% in the range 3-5-20% full power. 

Print-out of reactor flux distribution and reactivity as forerunner to test. Indicated reactivity reserve margin too low and this should have brought about an immediate shutdown of the reactor by the operators. " Power distribution distorted—some channels producing much more steam than others... 

Criteria for acceptable values and uses of uncertainties in operation, instrumentation numerical requirements, limit settings for alarms or scram, frequency and extent of power distribution measurements, and use of excore and incore instruments and related correlations and limits for offsets and tilts, all vary with reactor type. Guidance will need to be developed for each specific reactor type. 

Anticipated operational occurrences are off-normal events, usually plant transients, which can be coped with by the plant protection systems and normal plant systems but which could have the potential to damage the reactor if some additional malfunction should happen. Their typical frequency of occurrence may be more than 10 year Some of the anticipated occurrences (PIEs - postulated initiating events) are due to the increase of reactor heat removal (as might occur for an inadvertent opening of a steam relief valve, malfunctions in control systems, etc.). Some are due to the decrease of reactor heat removal (such as for feed-water pumps tripping, loss of condenser vacuum and control systems malfunctions). Some are due to a decrease in reactor coolant system flow rate, as in the case of a trip of one or more coolant pumps. Some are connected with reactivity and power distribution anomalies, such as for an inadvertent control rod withdrawal or unwanted boron dilution due to a malfunction of the volume control system for a PWR. Events entailing the increase or decrease of the reactor coolant inventory may also happen, due to malfunctions of the volume control system or small leaks. Finally, releases of radioactive substances from components may occur. 

II. 1 Neutron physical calculations of selected core configurations for burner reactors (CAPRA type) During the determination of the neutronic input data for safety related SAS4A studies systematic deviations in the power distribution for the CAPRA reference core were recognised. The detailed investigations showed that for the more complex types of CAPRA cores more refined calculational treatment is required, i.e. either a fully 3-dimensional diffusion method or even the VARIANT transport method. In the latter case, the considerable influence on control rod worths and on the sodium void effect - especially for configurations with control rods inserted (i.e. not fully withdrawn) - must be studied. 

Statistical distribution of the main performance indicators of the BN-600 reactor power unit operation and the range of the variation are represented in Fig. 3 -11. 

II. Definitions and general formalism. There probably exist as many notational systems in the field of reactor kinetics as there are authors. The following treatment (taken from [1]) has been found to be reasonably flexible.2 First we note that there is usually little need to characterize the reactor energy production by more than the total instantaneous power. In principle, a description in terms of neutron normal modes is required for completeness, but only the fundamental mode can be critical, and a description involving only the total power, the power distribution for the fundamental mode, and the effectiveness function for various perturbations will be essentially complete. ... 

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