# Method to measure doppler coefficient of reactivity

FIELD: power engineering.

SUBSTANCE: time-series data by reactivity is produced from time-series data by a neutron bundle by the method of reverse dynamic characteristic in respect to a single-point kinetic equation of the reactor. Time-series data by fuel temperature exposed to previously determined averaging is produced using time-series data by power output of the reactor and pre-determined dynamic model. The component of contribution to feedback by reactivity is determined using time-series data by reactivity and introduced reactivity. The Doppler coefficient of reactivity is determined using the received time-series data by average temperature of a moderator in the reactor, time-series data by fuel temperature exposed to previously determined averaging, isothermic temperature coefficient of reactivity and component of contribution to feedback by reactivity.

EFFECT: increased accuracy and simplicity of measurements of the Doppler coefficient and possibility of its usage in case of use of discrete data.

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The technical field

The present invention relates to a method of measuring the Doppler coefficient of reactivity and, in particular, to a method of measuring the Doppler coefficient of reactivity for direct measurement of the Doppler coefficient of reactivity using data from physical testing of a nuclear reactor.

Prior art

At commercial nuclear power plants, for example, in the reactor with pressurized water (hereinafter, the "PWR"), to ensure safe and economical operation, the design of the reactor core is carried out before each cycle to resolve various issues, such as how to accommodate fuel assemblies, each of which has a different fading and, therefore, different reactivity in the reactor core, and sufficient characteristics of self-regulation of the reactor core.

During periodic inspections during each cycle, performed physical testing of a nuclear reactor (physical testing of the reactor at the start) for measurement and evaluation of physical characteristics of the reactor, related to the reactor core for the current cycle of operation. For example, during testing, determination is made, is there a critical condition designed reactor core, through the observe is the R operations, and measured the change in reactivity when moving control rods relative to the reactor core and reactivity change when the temperature moderator to confirm the correctness of the design of the reactor core.

Here, characteristics of self-regulation are characteristics in which, when the reactivity of the reactor core is changed for one reason or another, as a result, in the reactor, naturally, is a phenomenon that acts in the opposite direction, i.e. so-called negative feedback reactivity, which is a very important factor for ensuring the safe operation of a nuclear reactor. In the PWR, the temperature coefficient of reactivity of the fuel, i.e. the change in the reactivity of a nuclear reactor, caused by the change of fuel temperature, and the temperature coefficient of reactivity of the inhibitor, i.e. changing the reactivity of a nuclear reactor, caused by the change of temperature moderator, both negative (if temperature rise is added to the negative reactivity), and, thus, PWR demonstrates characteristics of self-regulation. In boiling water reactor (hereinafter referred to as "BWR"), characteristics of self-regulation is additionally manifested in the form of the phenomenon (effect), which consists in the fact that the number of Nate the ones, slowing down the cooling water is reduced, while the number of bubbles in the cooling water increases with the temperature.

The above-mentioned temperature coefficient of reactivity of the nuclear fuel is determined by a phenomenon called the Doppler effect. The Doppler effect is that, with increasing temperature, the nuclide present in the fuel, increases the resonance absorption of neutrons, resulting in the number of neutrons that are involved in nuclear fission is reduced and, as a result, the reactivity of the active zone is reduced. The ratio of the changes in the reactivity to a single temperature change is called the Doppler coefficient of reactivity. In U238, which is a significant part of the uranium fuel used in existing light water reactors, this effect is significant, because U238 shows a strong resonance absorption of neutrons. This effect realizes the function of reducing the power of the reactor with increasing temperature nuclear fuel. In addition, this effect implements a fast response time, because it reflects the change in fuel temperature, not the moderator. Therefore, it is assumed that in a PWR, this effect plays a key role in ensuring safe operation, because PWR lack of effect achieved by increasing the bubbles in the cooling water to the th is achieved in a BWR.

The ratio between the change in fuel temperature and the reactivity at commercial nuclear plant, is estimated using, for example, data of the Doppler coefficient of reactivity of the nuclear fuel obtained by measuring the ratio between the change of the temperature of nuclear fuel and nuclear reaction, for example, by absorption of neutrons. Note, however, that when the physical testing of nuclear reactor temperature coefficient of reactivity of the nuclear fuel is measured indirectly, since it is difficult to directly measure the temperature of the fuel and if the fuel temperature is changed, other parameters, including the temperature of the moderator, also change. Instead, it is an analytical proof, based on a combination of measurement tests isothermal temperature coefficient and values analysis of neutron characteristics during construction (paragraphs 0003 and 0004 JP 2006-84181A, patent document 1).

To ensure the safe operation of a nuclear reactor with a high degree of reliability, it is preferable to directly measure the Doppler coefficient of reactivity, in order to verify the design of the active zone. This is especially important for PWR in which the planned full-scale utilization of MOX fuel and fuel with high burnup within a few the fir years.

In countries outside of Japan, direct measurement of the ratio between the change in fuel temperature and the change in reactivity, i.e. the temperature coefficient of reactivity of the nuclear fuel, was carried out in research reactors several times in the 1950s In such measurements, the temperature of the small balls made of uranium metal or uranium oxide was increased under soft neutron spectrum with a few fast neutrons, and measured the change in reactivity (non-patent documents 1 and 2, see below).

In Japan, using FCA (Fast Critical Assembly: a very small nuclear reactor) of the atomic energy Agency, Japan, in 2005, only fuel based on uranium oxide or MOX fuel was loaded in the field of soft neutron spectrum, increased temperature and measured the change in reactivity (non-patent document 3, see below).

Data obtained by the above measurements using actual reactors, etc. that are important for the expansion of the database and scan codes of the nuclear design of General purpose. However, nuclear reactors are used for the actual measurements are very small and vary in form and structure from commercial nuclear power plants. Thus, the correctness of the core design and validation of design codes intended for krupnomas is of major industrial nuclear power plants, which require high precision is achieved, not directly, but indirectly.

For this reason, the necessary technology, which provides a direct measurement of the Doppler coefficient of reactivity, especially for PWR.

Was recently developed method (see patent document 1), according to which, in General, the method of measuring the isothermal temperature coefficient of reactivity and dynamic identification are used together to measure the temperature coefficient of reactivity of the nuclear fuel. In General, the method includes the steps described below.

Here, "isothermal temperature coefficient of reactivity" means the sum of coefficient of reactivity associated only with fuel temperature (obtained by defining a private derivative of the temperature of the fuel), and coefficient of reactivity associated only with a temperature moderator (obtained by defining a private derivative with respect to a temperature moderator).

First of all, from the above measurements, isothermal temperature coefficient of reactivity is obtained as the ratio of the changes in the reactivity to changes in the temperature of the moderator.

Then, the extracted control rods to add external reactivity, making the reactor power increases, for example, about 1% but analnoe power.

This measured the time variation of the added external reactivity, the response of the neutron detector outside the active zone, the temperature of the coolant (moderator) at the entrance and the average temperature of the coolant (moderator), and accumulate their time-series data.

Additionally, the response of the neutron detector outside the active zone is fed to a digital reactivity meter to obtain the transient reactivity of a nuclear reactor.

From the obtained data of the time series, temperature coefficient of reactivity of the nuclear fuel is calculated using the dynamic identification.

In particular, data are numerical Fourier transform and served in the transfer function frequency response, and the Doppler coefficient of reactivity, which satisfies the ratio of the feedback is determined by approximation by the least squares method (basically, using the least squares method).

Note that the temperature coefficient of reactivity of the inhibitor can be calculated by subtracting the Doppler coefficient of reactivity of the isothermal temperature coefficient of reactivity.

Dynamic identification in the frequency domain relates to a method of estimating the frequency transfer function G(s), which sets the ratio of the frequency characteristics between the input and output signals u(t) and x(t), and, thus, find g(t) (the solution of the functional equation), when known function u(t) is injected into a partially or fully unknown function g(t)and the output function x(t) is known.

Patent document 1: JP2006-84181A;

Non-patent document 1: E. Creutz, et al., "Effect of Temperature on Total Resonance Absorption of Neutrons by Spheres of Uranium Oxide," J. Apple. Phys. 26, 276 (1955);

Non-patent document 2: R. M. Pearce et al., "A Direct Measurement of Uranium Metal Temperature Coefficient of Reactivity," Nucl. Sci. Eng., 2, 24 (1957);

Non-patent document 3: JAERI-Research, 2005-026, issued by the atomic energy Agency of Japan.

A brief statement of the substance of the invention

The above method of direct measurement of the Doppler coefficient of reactivity requires numerical Fourier transformation, i.e. a transformation in the frequency data, which complicates the use of discrete data. As a result, when the frequency of the switching range NIS (Neutron Instrumentation system (System neutron instruments): the neutron detector outside the active zone) or move groups of control rods, causing the adulteration of noise and significant fluctuations in the measured values, the application of this method becomes very difficult.

Thus, we need a way of measuring the Doppler coefficient of reactivity of a nuclear reactor, which provides a simple measurement and applies to discrete data.

The present invention nab is alleno at solving the above problems,
and, according to the invention, the initial reactivity ρ_{in}introduced from subcritical, but very close to the critical condition, after which the reactor power increases with a constant period of the reactor. Contribution to feedback reactivity is determined from changes in the reactivity ρ_{p}corresponding to the constant period of the reactor (reactivity with the constant period of the reactor, i.e. the reactivity added when the capacity increases with a constant velocity in the logarithmic scale, when the reactor goes from subcritical to supercritical state in).

"The reactor period" is the period of time for which the nuclear reactor power is increasing in e (approximately 2,718) times.

Here, you need to consider the following points.

In a subcritical condition, when the reactor power is extremely low, the function of self-regulation nuclear reactor is not shown.

Reactivity with the constant period of the reactor, essentially constant, regardless of the reactivity in the subcritical state. This is confirmed by checking combinations that achieve the best match between the actual measurement and analysis through simulation.

Additionally, as a means to determine the contribution to the feedback reactivity, the values of ρ_{in}and ρ_{p}subject analysis is through the simulation in the range of low power,
that gives a small feedback, and are determined by the values that reproduce the actually measured signals NIS.

The specific process for measuring includes six stages, in which: collect data; remove the effect of γ-irradiation from the data collected neutron flux; allocate component contribution to the feedback reactivity; define the upper boundary of the reactor power; calculate the average temperature of the fuel rods; and evaluate the Doppler coefficient of reactivity. Below, each step will be described using mathematical expressions.

As preconditions for measurement, assume that the values obtained by the neutron detector outside the active zone in the power range, the average temperature of cooling water of a nuclear reactor as the object to be measured in the form of data time series, and all the data necessary for analysis of the active zone, is available. Thus, for example, single-point kinetic reactor parameters βi and λi (where i represents one of the six groups of delayed neutrons; i=1,..., 6), averaged over the significance of the power correction factor {factor used to convert averaged over the volume of the temperature of the fuel rod is calculated using single-point kinetic model of the reactor, in rudnyy temperature, weighted distribution of the neutron flux and the conjugate distribution of the neutron flux (the significance of neutrons)} the temperature of the fuel rod, and the history of the exploitation of groups of control rods are quite accurately known or approximately known from a single theoretical analysis or workflow log. Additionally, measurement tests isothermal temperature coefficient, carried out to the current dimension, also known isothermal temperature coefficient of reactivity (= the Doppler coefficient of reactivity + temperature coefficient of reactivity moderator).

In addition to the above, from theoretical analysis during construction and from past experience approximately known various factors, including the Doppler coefficient of reactivity of the fuel as the object of measurement, the initial pocketinet etc. Thus, when the error function is estimated by means of dynamic identification, these estimates are often introduced as initial values, or often introduced similar values.

The data collection phase

From a nuclear reactor (PWR) in the subcritical state, the control rods are extracted by a given amount to achieve the supercritical state with low power, and the number of changes of neutron Patoka average temperature moderator are continuously collected in the form of data time series.
Here, pocketinet ρ_{sub}you can calculate in reverse order of reactivity ρ_{in}entered when removing the control rods to the preset value, and from the reactivity ρ_{p}with the constant period of the reactor.

Industrial reactor provides some power even in the subcritical state.

Additionally, the critical state is called the state of equilibrium in which the number of neutrons emitted in the fission in the reactor is equal to the number of neutrons lost by absorption in the reactor and leaves the reactor (effective multiplication factor equals 1), and (thermal) output in the reactor is determined depending on the level of the number of neutrons, which gives the equilibrium state.

The removal of γ-rays from neutron data flow

As described above, when the collection changes of the neutron flux in a nuclear reactor, the power of which is extremely small compared with the rated power, and, consequently, the neutron flux is small, in the form of data time series, there is a limit of neutron detector, if not undertaken any special measures. In particular, the spent nuclear fuel is in the reactor, and is usually used NIS (system nuclear instrumentation: the neutron detector also responds to γ-radiation emitted worked out the th nuclear fuel. There is always a constant dose of γ-radiation emitted by the spent nuclear fuel, even during the test at zero power. On the other hand, the neutron flux in the reactor during the test at zero power is low, because the reactor power is low. As a result, during the test at zero power, the background component contained in the collected data, or noise due to γ-radiation is not negligible. After a substantial increase in the reactor power, the number of neutrons generated in the reactor increases, and thus, the influence of γ-radiation becomes small enough that they can be neglected.

Thus, when tested at zero power, the influence of γ-radiation, i.e. component (noise)due to γ-radiation, mistakenly recorded as neutrons, is removed from the data for the neutrons obtained with the ionization chamber, through the use of the fact that the component associated with the γ-radiation is almost constant regardless of the power reactor. Note that the perspective of the design of PWR provides the possibility of direct measurement of the neutrons are not affected by γ-radiation, so this operation is not needed.

Consider a specific way. First of all, neutron flux, suitable for the power reactor,
converted into data current. Then, based on the received data, the error function E(g_{c}, ρ_{P}expressing the error between the numerically evaluated by numerical modeling low transient reactor power and the actually measured value is determined by the following Equation (1). Then, the indicator contact of γ-radiation (peremeshivaemogo noise component to the true neutron signal corresponding to the initial reactor power) g_{c}is deposited on the X-axis, the reactivity with the constant period of the reactor ρ_{p}delayed Y-axis, and the error function E(g_{c}, ρ_{P}) deferred on axis Z. in Addition, using the g_{c}and ρ_{p}as parameters, is determined by the combination of points g_{c}and ρ_{p}that minimizes the value of E(g_{c}, ρ_{P}) (this operation is finding the values of parameters at which the error function reaches a minimum, mainly based on the method of least squares, also referred to as “approximation”). The value of g_{c}found thus serve as an actual indicator of adulteration with γ-radiation.

Here, the error function defined as the logarithm (ln), based on the fact that the reactor power increases exponentially with time.

img src="http://img.russianpatents.com/1147/11474895-s.jpg" height="16" width="90" /> | (1) |

In the above equation, P represents the power of the reactor, the superscripts s and m represent the analytical value and the measured value, respectively, t is time, N is the number of data items, t_{i}represents the time corresponding to the data item i, 0 is the initial value (t=0), and the active area is in a subcritical state. As the measured value of the reactor power signal is used NIS, based on the fact that the signal NIS proportional to the reactor power.

Because the power response of the reactor required for the above analysis, is in the range of low power, in which the contribution to the feedback reactivity is negligible, and because it requires only the relative change of the characteristics of the reactor power in relation to the initial capacity, initial power P^{S}_{0}the reactor can be set arbitrarily in the case of these conditions. The absolute value of the characteristic power of the reactor is determined by the processing method, which is described below.

The output from the remote γ-radiation, converted from an actually measured signal NIS, defined below in Equation (2):

(2) |

The range of change of reactor power, R_{zm}can be calculated in accordance with the following Equation (3), from the initial power P^{m}_{g}(0) and the maximum reactor power P^{m}_{g,max}[max{P^{m}_{g}(t)}], which is achieved just before group of control rods inserted to reduce the reactor power after a large temperature feedback effect on reactivity was observed in the transient reactance (hereinafter will be called the maximum power of the "upper boundary reactor power").

Industrial power reactor provides some level of number of neutrons in a subcritical condition, as described above, and thus, the denominator in Equation (3) is not equal to zero.

(3) |

Knowing the reactivity with the constant period of the reactor ρ_{p}from Equation (4) can be obtained starting pocketinet p^{0}_{sub}.

(4) |

Here, enter the reactivity ρ

The output component in the feedback reactivity

From the data time series reactor power with the remote γ-radiation, time-series data of reactivity ρ are calculated with the method of the inverse kinetics for single-point kinetic equation of the reactor. Component contribution to the feedback reactivity Δρ_{fd}obtained by subtracting ρ_{in}from changes in the reactivity ρ(t)expressed by Equation (5).

(5) |

On the other hand, the component contribution to the feedback reactivity Δρ_{fd}equal to the sum of the contributions of Doppler coefficient feedback reactivity α_{f}and ratio feedback temperature reactivity moderator, and is expressed below in Equation (6):

(6), |

where: α_{itc}represents the isothermal temperature coefficient of reactivity; ∆ T_{f,av}represents the change in the average temperature of the fuel rods; and ΔT_{c,av}represents the change in the Central temperature of the moderator.

Component contribution to the reactivity Δρ_{fc}{component of the first member in the right-hand side of Equation (6)}associated with the Doppler coefficient of reactivity α_{f}, can be expressed below in Equation (7). This component is calculated from Δρ_{fd}obtained from Equation (5), the measured value changes ∆ T_{c,av}the average temperature of the moderator and the measured isothermal temperature coefficient of reactivity α_{itc}.

(7) |

On the other hand, Δρ_{fc}related according to the Equation (8) with an average temperature of fuel rods and, thus, if the average temperature of the fuel rods can be estimated, it is possible to estimate and Doppler coefficient α_{f}.

(8) |

If the actual change of the reactor is known, the change in the average temperature of the fuel rods ΔT_{f,av}(t) can be numerically evaluated from the equation of thermal conductivity of the fuel rod, using the actually measured average temperature of the moderator T_{c,av}(t). In particular, in a conventional PWR, sensors for measuring the temperature of the cooler (which of amentites) installed in the cooling pipes near the entrance and exit,
accordingly, the active zone of the reactor, and the values measured by these sensors, is displayed through the averaging scheme as the temperature of the coolant (moderator).

From the signal obtained by removing γ-radiation of the signal NIS, you can define the ratio R_{zm}the upper boundary of the reactor power to primary power, but it is impossible to determine the absolute value of the power. If the temperature of the moderator at the entrance to the active zone is constant, or if it is measured, the reactor power can be obtained by evaluating the temperature difference at the inlet and exit of the active zone of the actually measured average temperature of the moderator T_{c,av}. However, changing the reactor power balance between the supply of heat from the reactor core and the heat removal by the secondary side of steam generator cooling circuit temporarily violated, which leads to a change in temperature at the entrance to the active zone. Thus, the assumption of constancy of the temperature at the entrance of the active zone is incorrect.

When the temperature sensor is provided at the entrance to the active zone of the nuclear reactor, which is already designed and has a history of exploitation, to measure with high precision the temperature of the moderator at the entrance to the active zone in the form of data time series is required to re-enter the measuring device, etc.

To measure the Oia Doppler coefficient of reactivity using traditional measuring systems, the model of dissipation in the primary cooling loop PWR, for example, shown in figure 1, is introduced in the kinetic simulation model of a nuclear reactor that allows you to get the absolute value of change of reactor power.

The model of heat for the primary cooling circuit

Figure 1 shows the active area 10 of the reactor, the steam generator 20, a circulation pump 30 for cooling water pipe 41 on the outlet side of the reactor tube 42 on the inlet side of the reactor, arrows represent the flow of moderator (coolant water) and a wide white arrows represent over heat.

Simulation model for simulating heat in the primary cooling circuit is constructed on the basis of the equation of heat transfer and the equation of energy conservation for medium temperature cooler inlet and outlet pipes of the reactor, the average temperature of the coolant on the primary side of the steam generator and the average temperature of the cooler compartment cooling pump.

Definition of upper boundary the reactor power

The most important parameter in determining the cooling characteristics of the primary cooling circuit is a time constant τ_{sg 12}in relation to the transfer of heat from the primary side to the secondary side of the steam generator, which needs to be defined.

Unlike small nuclear reactor, for example, R is the actor for testing materials or installation for critical tests,
a large nuclear reactor for power generation, for example, PWR, provided with a heat exchanger, for example, the steam generator. Thus, usually there is a time difference between a peak time data of neutron flux and peak time of the temperature of the retarder, for example, cooling water. Considering the fact that there is a direct relationship between the time difference and the time constant τ_{sg 12}and that there is a strong correlation between the top edge of the reactor power and the maximum average temperature of the moderator T_{c,av}you can define the upper boundary of the reactor power and τ_{sg 12}.

To assess the difference between the peak times of measurement values power and temperature by approximation using the least squares method, introduces the error function given in Equation (9), using the initial power P_{0}{=P(0)} and the time constant τ_{sg 12}related to heat transfer, as parameters. In Equation (9), superscripts s and m represent the analytical value and the measured value, respectively.

(9), |

where: t_{p}represents the maximum average temperature of the moderator T_{c,av}; ∆ T_{c,av}presented yet the range of changes in temperature at the maximum temperature (the temperature rises from a subcritical state).

The conditions under which the error function E(τ_{sg 12}P_{0}) reaches its minimum, i.e. the values of the time constant τ_{sg 12}and the initial power P_{0}at which time the maximum temperature of the moderator and the maximum range of the increase in the average temperature of the moderator becomes equal to the actual measured values, are calculated similarly approximation using the least squares method using the above Equation (1). The upper boundary of the reactor power P_{max}is assumed to be the maximum achievable capacity of the reactor, calculated by analyzing the model in (τ_{sg 12}P_{0})that minimizes the error function E. In each iterative process of approximation by the least squares method, P_{0}re-evaluated from Equation (3).

Determination of changes in the average temperature of the fuel rods

Using the maximum achievable power reactor P_{max}and the range of change of reactor power, R_{zm}defined in this way are determined by the characteristics of the reactor power from the primary power to the maximum achievable capacity of the reactor, on the basis of the signal NIS remote noise of γ-radiation. Introducing the characteristics of the reactor power and the actually measured average temperature of the moderator T_{c,av}the EQ is the heat transfer
associated with an average temperature of fuel rods, determine the change in the average temperature of the fuel rods ΔT_{f,av}(t).

The calculation of the effective average fuel temperature

The average fuel temperature increases/decreases in accordance with the capacity of the reactor, and the temperature change is large and is fast compared with the moderator. When using perturbation theory first order to study the impact of changes in fuel temperature on reactivity, it is possible to obtain the average change in fuel temperature ΔT_{f,av}(t) as the average value of the power value of ΔT^{ip}_{f,av}(t)expressed by the following Equation (10).

Perturbation theory of first order is used for the following reason: on theory of perturbations is taken small change, and considers the impact of this change. Thus, it is a good method of making the correction values in the basic formula, free from disturbance, and, with regard to the amendment, the outrage of the first order is usually approximates the term of the first order as the correction member. Thus, the application of perturbation theory are considered suitable, because during physical testing of the reactor the temperature change of the fuel a little.

(10) |

The average change in fuel temperature ΔT_{f,av}(t), calculated according to the single-point kinetic reactor model, is the value averaged over the volume, which is expressed in the following Equation (11):

(11) |

When operating at the rated capacity, the moderator (cooling water) in the upper part of the reactor core has a higher temperature and lower density than the moderator in the lower part of the active zone. Thus, combustion in the upper part of the active zone lasts no longer than the bottom portion. When the reactor power is as small as in the test at zero power, the difference in density of the moderator is small between the upper and lower parts of the active zone, whereas in the upper part of the active zone remains more unburned fuel. As a result, the distribution of the neutron flux (φ) is shifted in the upper part of the active zone, and therefore the distribution of power, which is approximately proportional to the distribution of the neutron flux, is also shifted in the upper part of the active zone. Accordingly, the temperature of the fuel (fuel rod) also varies widely in the top frequent the active zone.
In particular, in the upper part of the active zone of the neutron flux (φ) is larger, and the distribution of the significance of neutrons (ϕ^{T}) is also higher. As a result, the average value of power is estimated higher than the value averaged over the volume. Thus, determine the correction factor given in Equation (12), which is displayed using one-dimensional code (in direction of flow of the stream) kinetic modeling, taking into account the spatial dependence of the distribution of neutron flux (including the paired distribution of the neutron flux in the direction of flow of the coolant (moderator)). Averaged over the significance of power and fuel temperature is obtained from the average volume values by multiplying by a correction factor c^{ip}.

(12) |

The calculation of the Doppler coefficient of reactivity

Equation (13) expresses the error function defined through the Doppler coefficient of reactivity. Doppler reactivity coefficient α_{f}that minimizes the error function, is the measured Doppler coefficient of reactivity.

(13) |

For the evaluation of the error function, uses the values measured in the time period of absence of influence of switching ranges NIS, i.e. the period after confirmation of the value of the upper limit, in which significantly the effect of feedback reactivity, and immediately after the operation of insertion of control rods.

Below the invention will be described in accordance with each claim.

According to paragraph 1, a method for measuring the Doppler coefficient of reactivity containing phases in which:

measure the neutron flux, when the reactor power is increased by a specified amount, and therefore the temperature reactivity effect is large enough to be significant during the transient reactivity, due to the input of reactivity in the reactor core, and neutron flux during this period is measured in the form of data time series;

get time-series data of the average temperature of the moderator in the reactor, when the reactor power is increased to a predetermined value due to the input of reactivity in the reactor core, and the average temperature of the moderator in the reactor during this period to receive data in time series in accordance with the prescribed procedure;

using the method of inverse kinetics to single point kineticheskom equation of the reactor receives time-series data the reactivity of the measured data in the time series of the neutron flux;

get time-series data of reactor power, on the basis of the received data in the time series of the average temperature of the moderator in the reactor and data time series neutron flux, and time-series data of reactor power, consistent with the above two types of data time series, and numerically estimated time-series data, receive according to a given procedure,

using data from the time series of reactor power, and given kinetic models are time-series data of fuel temperature, subjected to a prescribed averaging;

using data from the time series of reactivity and the reactivity with the constant period of the reactor get component contribution to the feedback reactivity; and

according to a given procedure, using data from the time series of the average temperature of the moderator in the reactor, the data of the time series of the temperature of the fuel subjected to the specified averaging, isothermal temperature coefficient of reactivity and component in the feedback reactivity receive Doppler coefficient of reactivity.

The present invention provides a measurement of the Doppler coefficient of reactivity of a nuclear reactor, which is simple to implement and applicable to discretional.

In addition, a given value of the reactor power increases by a specified amount, and the reactor power increases at a constant period of the reactor in the range of low power, and, then, increase rate gradually decreases due to the compensation effect of feedback on the temperature reactivity. The actual value, however, is about 1% of rated power, since the measurement is performed during physical testing of a nuclear reactor, and thus small changes in reactor power, changing various physical constants, such as heat capacities and densities of the moderator and fuel, affecting the dynamics of the reactor induced by raising the temperature remains nearly constant, i.e. constant during the measurement.

In addition, "time series data" is data measured during the time from the beginning until it reaches a given value of power. However, the data do not have to span the entire period of time. Bit data, for example, about 30 seconds after switching the range of measuring the neutron flux and 100 seconds after moving the groups of control rods can be eliminated. As for the sampling interval, although it is desirable to have a sampling interval of 0.001 seconds from the point of view of the trade-off between accuracy of the analysis and on the reception of computing, this is not a limitation, and are not excluded also analog data.

In addition, the prescribed procedure" on "phase of the receive data in the time series of the average temperature of the moderator in the reactor is the procedure for obtaining the average value (result) of the measurement values of the temperature sensors provided in the pipe cooling on the exhaust side of the reactor and pipe cooling on the inlet side of the reactor, respectively, by passing values through the averaging scheme.

In addition, the "prescribed kinetic model of the" on "phase of the receive data time series reactor power" is a widely used single-point kinetic reactor model or analysis program.

In addition, as fuel temperature, subjected to a prescribed averaging", the value obtained on the basis of theory of perturbations of the first order, or other analysis, the experimental value, etc.

In addition, the "prescribed averaging" can represent, for example, "average importance".

In addition, the isothermal temperature coefficient of reactivity" means the sum of coefficient of reactivity only fuel temperature (obtained by defining a private derivative with respect to a fuel temperature) and coefficient of reactivity only tempera is URS moderator (obtained by defining a private derivative with respect to a temperature moderator).

According to paragraph 2, a method for measuring the above-described Doppler coefficient of reactivity, in which

when the measurement data of the time series of the neutron flux at the stage of measuring the neutron flux measured neutron flux and γ-radiation; and

the step of obtaining data of the time series of reactivity is the procedure of removing the influence of γ-radiation from the measured data in the time series of the neutron flux, and, from the data of the time series with the remote influence of γ-radiation, receives time-series data of reactivity using the method of inverse kinetics for single-point kinetic equation of the reactor.

The invention according to this paragraph facilitates precise measurement of the neutron flux at the stage of low power reactor using a simple measuring device, for example, the ionization chamber, the current PWR.

According to paragraph 3, a method for measuring the Doppler coefficient of reactivity according to claim 2, in which the removal procedure,

(1) the error function evaluated by the method of least squares, and

1) the error function is determined with the use of

a) numerically estimated values changes over time, calculated in accordance with the prescribed kinetic equation of a nuclear reactor using reactivity at constant reactor period is the indicator of adulteration with γ-radiation as parameters, associated with the characteristic power of the reactor in the range of low power, in which the contribution of the feedback reactivity is negligible, and

b) part changes over time, corresponding to the characteristics of the reactor power, the actual measurement data of the time series of the neutron flux, and

2) the error function is the difference between these two values on a logarithmic scale; and

(2) receive a combination of reactivity with the constant period of the reactor and the rate of contact of γ-radiation, which minimizes the value of the error function. The rate of contact of γ-radiation, forming a combination, is seen as the true indicator of adulteration with γ-radiation.

In the invention according to this paragraph, use the indicator contact of γ-radiation, which minimizes the error function associated with the difference between the numerically estimated value and the actual measurement value of the reactor power, is closely related to the neutron flux and, thus, it is possible to obtain an accurate indicator of adulteration with γ-radiation and, consequently, the true power of the reactor.

Here, "time-series data in the range of low power, in which the contribution of the feedback reactivity is small", this time-series data of power within 1% of nominal power, and the reason I use the data in this range, is that an indicator of adulteration with γ-radiation can be accurately obtained without the necessity of amending feedback effects on reactivity.

According to paragraph 4, a method for measuring the above-described Doppler coefficient of reactivity, in which

at the stage of obtaining the average temperature of the moderator in the reactor, get the average temperature of the moderator in the form of data time series, when the reactor power is increased to a predetermined value due to the input of reactivity in the reactor core in a state that is close to critical.

In the invention according to this paragraph, the process is carried out in accordance with a given procedure using, for example, temperature moderator (coolant water) inlet and outlet of the steam generator and the heat supplied from the circulating pump retarder, as data for calculation, and, as a result, receive data time series of temperature moderator in the reactor, the timing data time series neutron flux. This allows you to measure the temperature of the coolant (moderator) in the reactor without the necessity of measuring the temperature of cooling water at the entrance to the inside of the reactor (nuclear reactor), where it is difficult to install the temperature sensor. It also provides a direct measurement of the Doppler coefficient reaktivnosti constructed nuclear reactor.

Note that the heat radiation of the cooling pipes, for example, can not be excluded from consideration as the data mentioned above.

In addition, you can get a different temperature or temperature, for example, the temperature of the retarder inlet (low) and the output (high) of the reactor.

According to paragraph 5, a method for measuring the above-described Doppler coefficient of reactivity, in which

at the stage of obtaining data time series reactor power,

the time constant related to the heat transfer from the primary side to the secondary side of the steam generator, and the initial reactor power is chosen as parameters, and

get a combination of the time constant and the initial reactor power, which minimizes the value of the error function, which is expressed as {1-(numerically estimated from the average temperature of the moderator to the maximum temperatures measured from the average temperature of the moderator to maximum temperature)}^{2}+ {1-(numerically estimated value of the maximum range changes in the average temperature of the moderator/measured value maximum range changes in the average temperature of the moderator)}^{2}.

In the invention according to this paragraph, choose a combination of the time constant and the reactor power, which minimizes the value of the function error is, and, on the basis of the search result, you receive optimal time constant and the upper boundary of the reactor power. Thus, it is possible to accurately estimate the average temperature of the moderator in the reactor according to the invention according to claim 4.

In addition, since it is possible to determine the absolute value of the upper boundary of the capacity and the range of change of power from the primary power to the upper boundary of the capacity, it is possible to obtain characterization of reactor power, expressed in absolute value of the signal NIS remote γ-radiation. In addition, from the obtained characteristics of the reactor power and the actually measured temperature moderator, you can get accurate time-series data of the average temperature of the fuel rods.

According to paragraph 6, a method for measuring the above-described Doppler coefficient of reactivity, in which

at the stage of obtaining the data the time series of fuel temperature, averaged over the volume of the fuel temperature, calculated using the heat equation associated with the average temperature of the fuel rods and the data time series reactor power change using the correction factor obtained taking into account distributions of neutron flux and paired neutron flux (the significance of neutrons) in the direction of the leakage flux is and moderator of the zero power thanks to get time-series data of fuel temperature, subjected to a prescribed averaging, based on the perturbation theory of the first order.

Thus, it is possible to accurately estimate the fuel temperature.

According to paragraph 7, a method for measuring the above-described Doppler coefficient of reactivity, in which

Set the averaging is averaged over the significance of power, and in the prescribed procedure at the stage of obtaining the Doppler coefficient of reactivity, use the following equation: component contribution to the feedback reactivity associated with the Doppler coefficient of reactivity = Doppler coefficient of reactivity * (change in fuel temperature obtained using the data of the time series of fuel temperature, averaged over the significance of power - value : changes in the average temperature of the moderator in the reactor) + isothermal temperature coefficient of reactivity * change value fluctuations in average temperature of the moderator in the reactor.

In the invention according to this paragraph, because this is the exact equation, it is possible to accurately estimate the Doppler coefficient of reactivity.

Here, the average value of power" means the averaging, weighted by the distribution of the significance of neutrons and distribution is not the throne of flow, on the assumption that the distribution of the neutron flux is essentially proportional to the power distribution.

According to paragraph 8, a method for measuring the above-described Doppler coefficient of reactivity, in which

additionally, in the prescribed procedure at the stage of obtaining the Doppler coefficient of reactivity,

as an option, choose the Doppler coefficient of reactivity, and

the Doppler coefficient of reactivity, which minimizes the value of the error function, assessed as the actual Doppler coefficient of reactivity, and the error function is defined as {1,0 - Doppler reactivity coefficient * (change in fuel temperature obtained using the data of the time series of fuel temperature, averaged over the significance of power - value : changes in the average temperature of the moderator in the reactor)/component contribution to the reactivity associated with the Doppler coefficient of reactivity}^{2}on the basis of the received data.

In the invention according to this paragraph, the Doppler coefficient of reactivity is used as a parameter, and get the Doppler coefficient of reactivity, which minimizes the error function. Thus, the Doppler coefficient of reactivity has high accuracy.

The present image is the buy provides a measurement of the Doppler coefficient of reactivity of a nuclear reactor, which is simple to implement and applicable to discrete data.

Brief description of drawings

The invention is further explained in the description of the preferred variants of the embodiment with reference to the accompanying drawings, in which:

Figure 1 depicts a schematic representation of the heat balance in the primary cooling loop PWR;

Figure 2 depicts a schematic diagram of the measuring system;

Figure 3 depicts a diagram showing actually measured time-series data of the neutron flux;

Figure 4 depicts a chart showing the actual measurement data of the time series of the average temperature of the moderator;

Figure 5 depicts a chart showing time-series data of power before and after the removal of γ-radiation;

6 depicts a chart showing time-series data of the two components of contribution in the feedback reactivity;

7 depicts a chart showing the change with time of the average temperature of the moderator.

Description of the preferred embodiments of the invention

Further, the present invention will be described based on the preferred options of its implementation. Note that the present invention is not limited to the following embodiment. The following variant of the wasp is estline allows various modifications, meet the scope of the present invention and its equivalents.

In the present embodiment, data obtained by actual measurement in existing PWR, are processed to measure the Doppler coefficient of reactivity of a nuclear reactor.

System analysis

Characteristics of the reactor core are analyzed using single-point kinetic reactor model and one-dimensional (in the direction of flow of cooling water) kinetic simulations for analysis, the radial power distribution in the reactor core is assumed flat, because the capacity of the reactor is small.

The measuring system

Figure 2 schematically shows a measuring system that includes a device configuration in accordance with the present embodiment. Figure 2 shows the ionization chamber 11 and 12 for detecting neutron flux outside the active area; the sensors 21 and 22 temperature, for example, resistive temperature detector (RTD); ammeter 31; ammeter 50 weak current; an amplifier 51 DC; terminal base 52; and the Board 53 A/d Converter.

Dotted lines represent lines of the signal to measure.

A/d Converter has a resolution of -10 to +10 In / 16 bits.

The time interval of sampling data of 0.001 seconds, and the time measurement is s equal to 2600 seconds.

In addition, we use a low pass filter, amplifier, etc.

The choice of measurement data

For up to 100 seconds after moving the groups of control rods, is spatial variation in the distribution of power due to the movement of control rods, which affects the signal NIS. Thus, the characteristic of this period of time, with such influence, is excluded from the object approximation (analysis).

When the reactor power increases from subcritical up to about 1% of rated power, the measurement range NIS (number of digits as the measurement object) should switch. This affects the measurement within about 30 seconds after the switch, and therefore, this time period is excluded from the object of approximation.

Using data up to the peak, which recognized the value of the upper boundary, where the conditions of operation of the secondary system steam generator are considered to be relatively stable.

The measured time-series data

Control rods operate so that the PWR, which initially is in a subcritical state, reaches a critical state, and power is additionally slightly increased during measurement of neutron flux and temperature moderator.

Figure 3 shows time-series data of the neutron flux, which received the actual measurement. The y-axis represents the neutron flux is converted to a current (A), and the abscissa represents the time elapsed from the beginning of data collection. Charts showing time-series data of physical quantities, as discussed below, the y-axis represents the physical value, and the abscissa represents the time elapsed from the beginning of data collection, as before.

Figure 4 also shows the time-series data of the average temperature of the moderator. Compared to figure 3, it can be seen that there is a delay of about 50 seconds between the peaks of the maximum values. From theoretical analysis it follows that the time delay caused by the characteristics of the steam generator heat sink, which is mainly determined by the time constant τ_{sg 12}related to heat transfer (heat transfer) from the primary side to the secondary side of the steam generator, and that the more constant τ_{sg 12}the more the delay. Additionally, it is determined that τ_{sg 12}should be about 34 seconds.

The removal of γ-radiation

From the data obtained neutron flux, numerical values of contact of γ-radiation g_{c}=0,78 and reactivity at a constant period of reactor ρ_{p}=46,6 pcm calculated by approximation using the least squares method using Equation (1). Figure 5 shows the data of the temporary number of the power
calculated from Equation (2) using the calculated indicator of adulteration with γ-radiation g_{c}. In figure 5, the ordinate axis represents the ratio of the reactor power P to the nominal power, the solid line represents time-series data of reactor power, P, is obtained on the basis of the neutron flux after the removal of γ-radiation, and the dotted line represents the time-series data of power reactor based neutron flux prior to the removal of γ-radiation. Note that the dotted line mainly coincides with the data in figure 3, expressed in the form of a current value detector.

From figure 5 it follows that in the range where the reactor power P is small, the influence of γ-radiation is significant.

Additionally, the reactor power was changed approximately 220 times.

Presents two components contribution to the feedback reactivity calculated from Equations (4) and (6)corresponding to the time series reactor power P, as shown in figure 5. In the range from 600 to 800 seconds at 6, the upper line shows the component contribution to feedback Δρ_{fc}associated with the Doppler coefficient of reactivity, and the bottom line shows the component contribution to the feedback component reactivity moderator Δρ_{fd}.

Definition of upper boundary the reactor power

According to SP is soba,
described in the section "Model of heat for the primary cooling circuit"applied to the model shown in figure 1, as an object, it was found that τ_{sg 12}=34, P_{0}=(of 3.77*10^{-4})% of rated power, and, therefore, the upper boundary of the reactor power P_{max}=(8,35*10^{-2})% of rated power. Additionally, data were obtained time series of the average temperature of the moderator T_{c,av}. The results are shown in Fig.7, where the solid line represents the calculated value, and the dotted line represents the measured value. In response to a peak time as the object approximation, between the measured value and calculated value, including peak time and the peak value, there was no difference.

The calculation of the average fuel temperature

The distribution of the neutron flux correction factor of c^{ip}changes in the average temperature of fuel (fuel rod), averaged over the significance was calculated using Equation (12), and a result was obtained 1,296. When using high-speed bulk neutron flux, because the interest is a feedback effect on the Doppler coefficient of reactivity, the results did not differ from the case of using thermal neutrons. Additionally, with the use of the this value was calculated averaged value for the temperature of the fuel.

Evaluation of the Doppler coefficient of reactivity

From Equation (13) estimate the Doppler coefficient of reactivity α_{f}that minimizes the error function, and the result was α_{f}=-3,2 (pcm/K), and is the same as the constructive value with the approach to two significant digits.

The list of symbols

10 | reactor core |

11 | ionization chamber |

12 | ionization chamber |

20 | the steam generator |

21 | the temperature sensor |

22 | the temperature sensor |

30 | circulating pump for cooling water |

31 | ammeter |

41 | pipe on the exhaust side of a nuclear reactor |

42 | the pipe on the inlet side of a nuclear reactor |

50 | Ammeter low current |

51 | The DC amplifier |

52 | Terminal base |

53 | Charge the a/d Converter (note PC) |

1. The method of measuring the Doppler coefficient of reactivity containing phases in which

measured time-series data of the neutron flux, when the capacity of the reactor is increased to a predetermined value due to the input of reactivity in the reactor core to achieve the supercritical state with low power, while the neutron flux and γ-radiation during this period is measured in the form of data time series,

get time-series data of the average temperature of the moderator in the reactor in which the reactor power increase to a predetermined value due to the input of reactivity in the reactor core to achieve the supercritical state with low power, and high temperature moderator in the reactor during this period to receive data in time series according to a given procedure,

get time-series data the reactivity of the measured data in the time series of the neutron flux, the influence of γ-radiation, remote through the procedure of removing effect is I γ-radiation from the measured data in the time series of the neutron flux,
using the method of inverse kinetics for single-point kinetic equation of the reactor receives time-series data of reactivity

on the basis of the received data in the time series of the average temperature of the moderator in the reactor and data time series neutron flux, as determined by the procedure receives time-series data of reactor power, consistent with the above two types of data time series,

using data from the time series of reactor power, obtained previously, and given kinetic model receives time-series data of fuel temperature, subjected to a given averaging,

using data from the time series of reactivity obtained previously, and the reactivity with the constant period of the reactor receives time-series data component in the feedback reactivity, and using the data in the time series of the average temperature of the moderator in the reactor, the data of the time series of the temperature of the fuel subjected to the specified averaging, isothermal temperature coefficient of reactivity and data time series component in the feedback reactivity, as determined by the procedure, receive the Doppler coefficient of reactivity.

2. The method of measuring the Doppler coefficient of reactivity according to claim 1, in br/>
when the measurement data of the time series of the neutron flux at the stage of measuring the neutron flux measured neutron flux and γ-radiation, and

at the stage of obtaining data time series reactance perform the removal procedure of the influence of γ-radiation from the measured data in the time series of neutron flux and of the data time series with the remote influence of γ-radiation, while time-series data of reactivity receive using the method of inverse kinetics for single-point kinetic equation of the reactor.

3. The method of measuring the Doppler coefficient of reactivity according to claim 2, in which at the stage of removal

evaluate the error function using the least squares method, and define the error function using

time-varying numerically evaluated values calculated according to the kinetic equation of a nuclear reactor using reactivity with the constant period of the reactor and the rate of contact of γ-radiation as parameters related to the power characteristic of the reactor in the range of low power, in which the contribution of the feedback reactivity is negligible, and the time-varying part, the corresponding power characteristics of the reactor, the actual measurement data of the time series of the neutron flow,

the function error which is the difference between these two values on a logarithmic scale,
and determine the combination of reactivity with the constant period of the reactor and the rate of contact of γ-radiation, which minimizes the value. the error function.

4. The method of measuring the Doppler coefficient of reactivity according to any one of claims 1 to 3, in which

at the stage of obtaining the data the time series of the average temperature of the moderator in the reactor get the average temperature of the moderator in the form of data time series, when the reactor power is increased to a predetermined value due to the input of reactivity in the reactor core, which is in a subcritical state or reaches a supercritical state.

5. The method of measuring the Doppler coefficient of reactivity according to claim 4, in which

at the stage of obtaining data time series reactor power

choose as parameters the time constant τ_{sg 12}in relation to the transfer of heat from the primary side to the secondary side of the steam generator associated with the reactor, and the initial reactor power P_{about}and

get a combination of the time constant τ_{sg 12}and the initial reactor power P_{about}that minimizes the value of the error function E(τ_{sg 12}, R_{about}), expressed as a

where t^{s}_{p}is numerically estimated time of transition from the average temperature of the moderator to the maximum the Oh temperature,
t^{m}_{p}is the measured transit time of the average temperature of the moderator to the maximum temperature, ΔT^{s}_{c,av}is numerically estimated value of the maximum range changes in the average temperature of the moderator and ΔT^{m}_{c,av}is the measured value maximum range changes in the average temperature of the moderator.

6. The method of measuring the Doppler coefficient of reactivity according to claim 1, in which

at the stage of obtaining data time series fuel temperature averaged over the volume of the fuel temperature, calculated using the heat equation associated with the average temperature of the fuel rods and the data time series reactor power change using the correction factor obtained taking into account distributions of neutron flux and paired neutron flux (the significance of neutrons) in the direction of flow of the flow moderator in a state of zero power and thereby gain time-series data of fuel temperature, subjected to a given averaging, based on the perturbation theory of the first order.

7. The method of measuring the Doppler coefficient of reactivity according to claim 1, in which

set the averaging is averaged over the significance of power, and

at the stage of obtaining the Doppler is led coefficient of reactivity for obtaining Doppler reactivity coefficient using the following equation:

where Δρ_{fd}(t) represents the component contribution to the feedback reactivity corresponding to the Doppler feedback reactivity, α_{f}is the Doppler reactivity coefficient, ∆ T_{f,av}(t) represents the change in the average temperature of the fuel rods, ∆ T_{c,av}(t) represents the change in the average temperature of the moderator and α_{itc}represents the isothermal temperature coefficient of reactivity.

8. The method of measuring the Doppler coefficient of reactivity according to claim 7, in which

at the stage of obtaining the Doppler coefficient of reactivity

the parameter of the Doppler coefficient of reactivity, which minimizes the value of the error function, assessed as the actual Doppler coefficient of reactivity, and the error function E_{rdf}defined as

where N represents the number of data, t_{i}represents the time corresponding to the i-th data Δρ_{fc}(t) represents the component contribution to the reactivity and^{ip}is a correction factor, defined as

where ΔT^{ip}_{f,av}(t) represents the change in the values averaged over the significance of power and fuel temperature.

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2 cl, 1 dwg

FIELD: analog computer engineering; verifying nuclear reactor reactivity meters (reactimeters).

SUBSTANCE: proposed simulator has m threshold devices, m threshold selector switches, m series-connected decade amplifiers, m electronic commutators, n - m - 1 series-connected decade frequency dividers, first group of m parallel-connected frequency selector switches, second group of n - m frequency selector switches, and group of n - m parallel-connected mode selector switches. Integrated inputs of threshold selector switches are connected to output of high-voltage amplifier and output of each threshold selector switch, to input of respective threshold device; output of each threshold device is connected to control input of respective electronic commutator; inputs of electronic commutators are connected to outputs of decade amplifiers and outputs are integrated with output of group of mode selector switches and with input of voltage-to-frequency converter; output of inverting amplifier is connected to input of first decade amplifier and to that of group of mode selector switches; input of first group of frequency selector switches is connected to output of voltage-to-frequency converter and to input of first decade frequency divider and output, to integrated outputs of first group of frequency selector switches and to input of division-chamber pulse shaper input; each of inputs of second group of frequency selector switches is connected to input of respective decade frequency divider except for last one of this group of switches whose input is connected to output of last decade frequency divider; threshold selector switches and frequency selector switches of first group, as well as m current selector switches have common operating mechanism; mode selector and frequency selector switches of second group have common operating mechanism with remaining n - m current selector switches. Such design makes it possible to realize Coulomb law relationship at all current ranges of simulator for current and frequency channels.

EFFECT: ability of verifying pulse-current input reactimeters by input signals adequate to signals coming from actual neutron detector.

2 cl, 1 dwg

FIELD: atomic industry.

SUBSTANCE: proposed line is provided with computer-aided system for contactless control of flaw depth and profile on surface of fuel element can and on end parts including sorting-out device that functions to reject faulty fuel elements. This line is characterized in high capacity and reduced labor consumption.

EFFECT: enlarged functional capabilities, improved quality of fuel elements.

1 cl, 2 dwg

FIELD: nuclear fuel technology.

SUBSTANCE: invention relates to production of pelleted fuel and consists in controlling nuclear fuel for thermal resistance involving preparation for selecting pellets from nuclear fuel lot for measuring diameter, which preparation consists in dedusting. Selected pellets are placed in temperature-stabilized box together with measuring instrument. Diameter of each pellet is them measured and measurement data are entered into computer. Thereafter, pellets are charged into heat treatment vessel, wherein pellets are heated in vacuum at residual pressure not exceeding 7·10^{-2 }Pa at heating velocity not higher than 10°C/min to 100-160°C and held at this temperature at most 2 h, whereupon heating is continued under the same conditions to 1470-1530°C and this temperature is maintained for a period of time not exceeding 4 h, after which hydrogen is fed with flow rate 2-6 L/min. Humidity of gas mix is measured in the heat treatment outlet. If humidity of gas mixture in the heat treatment outlet exceeds 800 ppm, hydrogen feeding is stopped and material is subjected to additional vacuum degassing at residual pressure below 7·10^{-2} Pa and held at 1470-1530°C in vacuum for further 4 h. Hydrogen feeding is the repeated at 2-6 L/min. If humidity of gas mixture in the heat treatment outlet is below 800 ppm, preceding temperature is maintained not longer than 2 h and raised to 1625-1675°C at velocity 40-60°C/h and then to 1700-1750°C at velocity 15-45°C/h. When outlet humidity of mixture is 500-750 ppm, hydrogen feeding is lowered to 1 L/min. Temperature 1700-1750°C is maintained during 24±2 h, after which pellets are cooled to 1470-1530ºC at velocity not higher than 10°C/min. Hydrogen is replaced with argon and cooling is continued to temperature not higher than 40°C, which temperature is further maintained. Outside diameter of each pellet from the selection is measured to find average diameter of pellets before and after heat treatment in order to calculate residual sintering ability. When this parameter equals 0.0-0.4%, total lot of pellets is used in fuel elements and in case of exceeding or negative residual sintering ability the total lot of pellets is rejected.

EFFECT: improved pellet quality control.

2 dwg

FIELD: power engineering; evaluating burnout margin in nuclear power units.

SUBSTANCE: proposed method intended for use in VVER or RBMK, or other similar reactor units includes setting of desired operating parameters at inlet of fuel assembly, power supply to fuel assembly, variation of fuel assembly power, measurement of wall temperature of fuel element (or simulator thereof), detection of burnout moment by comparing wall temperatures at different power values of fuel assembly, evaluation of burnout margin by comparing critical heat flux and heat fluxes at rated parameters of fuel assembly, burnout being recognized by first wall temperature increase disproportional relative to power variation. Power is supplied to separate groups of fuel elements and/or separate fuel elements (or simulators thereof); this power supplied to separate groups of fuel elements and/or to separate fuel elements is varied to ensure conditions at fuel element outlet equal to those preset , where G is water flow through fuel element, kg/s; i_{out}, i_{in} is coolant enthalpy at fuel element outlet and inlet, respectively, kJ/kg; N_{δi} is power released at balanced fuel elements (or simulators thereof) where burnout is not detected, kW; n is number of balanced fuel elements; N_{brn.i }is power released at fuel elements (or element) where burnout is detected; m is number of fuel elements where burnout is detected, m ≥ 1; d is fuel element diameter, mm.

EFFECT: enhanced precision of evaluating burnout margin for nuclear power plant channels.

1 cl, 2 dwg

FIELD: analytical methods in nuclear engineering.

SUBSTANCE: invention relates to analysis of fissile materials by radiation techniques and intended for on-line control of uranium hexafluoride concentration in gas streams of isotope-separation uranium processes. Control method comprises measuring, within selected time interval, intensity of gamma-emission of uranium-235, temperature, and uranium hexafluoride gas phase pressure in measuring chamber. Averaged data are processed to create uranium hexafluoride canal in measuring chamber. Thereafter, measurements are performed within a time interval composed of a series of time gaps and average values are then computed for above-indicated parameters for each time gap and measurement data for the total time interval are computed as averaged values of average values in time gaps. Intensity of gamma-emission of uranium-235, temperature, and pressure, when computing current value of mass fraction of uranium-235 isotope, are determined from averaged measurement data obtained in identical time intervals at variation in current time by a value equal to value of time gap of the time interval. Computed value of mass fraction of uranium-235 isotope is attached to current time within the time interval of measurement. Method is implemented with the aid of measuring system, which contains: measuring chamber provided with inlet and outlet connecting pipes, detection unit, and temperature and pressure sensors, connected to uranium hexafluoride gas collector over inlet connecting pipe; controller with electric pulse counters and gamma specter analyzer; signal adapters; internal information bus; and information collection, management, and processing unit. Controller is supplemented by at least three discriminators and one timer, discriminator being connected to gamma-emission detector output whereas output of each discriminator is connected to input of individual electric pulse counter, whose second input is coupled with timer output. Adapter timer output is connected to internal information bus over information exchange line. Information collection, management, and processing unit is bound to local controlling computer network over external interface network.

EFFECT: enabled quick response in case of emergency deviations of uranium hexafluoride stream concentration, reduced plant configuration rearrangement at variation in concentration of starting and commercial uranium hexafluoride, and eliminated production of substandard product.

24 cl, 5 dwg