IMPROVING THE RELIABILITY AND PERFORMANCE OF A NUCLEAR POWER

PLANT BY USING COMPUTERS IN CONTROL SYSTEMS

F. M. Mitenkov, V. V. Vasil'eva, V~ S. Vostokov, V. S. Gorbunov, N. P. Zaets, and O. B. Samoilov

UDC 621.039.526:621.039.5-19

Recently plane have appeared for using computers for direct process control at nuclear power plants: the Remikont R-100 system, the plans for BN control and safety rods, the VG-400, etc. The problems that can be solved by computers usually do not go beyond the framework of functions handled by traditional systems as they apply to process control. A paradoxical situation is taking shape in which the computer equipment being employed, with its enormous capabilities, performs control functions, ensuring only with difficulty the required continuity of their execution (through redundancy, majorization, etc.), and therefore merely backs up control systems constructed from traditional tools. This last factor also has evidently de- termined the lack of a formulation of substantive control problems that can be solved only by using computer technology.

We believe that the fundamental possibility exists for simplifying the controlled system itself by using computer technology for control problems with some increased complexity. Let us note that we believe this formulation of the problem to be the only realistic source of a payback on computer technology used to control a nuclear power plant.

Two ways of elevating the technical and economic indicators of a plant's operation are of interest in the construction of a safety and control system utilizing the capabilities of computer technology: i) simplification of reactor design, and 2) an increase in the opera- tional reliability of the nuclear power plant through a reduction of the number of trips of the emergency protection system, and optimization of accident conditions with a view to reduc- ing thermal stresses in structural elements, including the reactor core.

Obviously, these two directions are competitive, and the question of the optimal propor- tion of each is a subject for an independent study. In this paper we set the task and, using fast reactors as an example, illustrate the possibility of carrying it out without considering weighting factors for each direction.

SIMPLIFICATION OF REACTOR DESIGN

BY INCREASING THE COMPLEXITY

OF THE CONTROL SYSTEM

Such simplification can be accomplished by reducing the number or variety of operating controls. Let us consider in this regard the possibility of doing away with automatic control rods.

The control of nuclear power plants is presently based on maintaining power in accordance with a preset level or according to the coolant flow rate through the core, by means of high- speed control and shim rods operating on a relay principle. Hunting during the interaction of both types of rods is precluded by the excess velocity worth of control rods relative to shim rods [i].

With the transition to relay control (with a constant velocity of the actuators) based on coolant temperature, the question arises of ensuring the stability of the system [2]. Ac- cording to [3], in this case the criterion for occurrence of oscillations will be the excess reactivity introduced by the regulator over a time period equal to the time constant of the thermocouple, beyond the temperature effect of reactivity in an interval of the inert zone of the regulator, i.e., rb/~a>~/2, where T is the time constant of the thermocouple, b is the rate of reactivity insertion by the shim rod, ~ is the temperature coefficient of reactivity, and a

P

Translated from Atomnaya Energiya, Vol. 66, No. 6, pp. 425-427, June, 1989. Original article submitted November ii, 1987.

0038-531X/89/6606-0479512.50 �9 1989 Plenum Publishing Corporation 479

Input )

t

circle of first loop I , �9

No of pump of first

No valve

V Yes

Yes

first-circuit ' pump speeds

e,,~, ~

T

I

Output

>

-1=4- input )

I Calc. of I f~ K

I I

�9 ~': ~ ~ Engagement of L ~o I 1-18 shim rods I

Y ~ ~ ~ E~agemenh of ~ 1-20 shim rods F

o f ~ Engag~ent of

Fig. i. Structure of control program for nuclear power plant using shim rods with the loop disconnected (p, 6K is the reactivity; H is the coordinate of the rod; N is the neutron power; and the subscripts have the following meanings; pre) preset; cur) current; 0, in) initial; i) rod number).

Fig. 2. Module for selecting the num- ber of control rods.

is the inert zone of the regulator. For example, after inserting the numerical values of the parameters (T = 5 sec, b = 0.i x i0 -~ sec-*, ~ = 0.57 x i0 -~~ C-*), for a fast reactor we find the stability of the control system may be manifest if the accuracy of coolant-temperature maintenance at the outlet is less than 0.6=C. With consideration for a channel of power regu- lation (in lieu of supplementation of the channel of temperature control) that has a shorter time constant (TN<I sec), we may assume that the real system will be stable.

Because of the relatively low rate of the process, the implementation of regimes of a planned change in power by using the shim rods does not pose any difficulty. The most stringent requirements for a control system are imposed in accident situations that do not require reac- tor shutdown, e.g., upon disconnection of the heat-transfer loop (Fig. I). The algorithm for selecting the required number of shim rods involved in responding to conditions with considera- tion for both the change in velocity worth from the location in the core and the previously calculated reactivity required to ensure performance, 6Kpr e = f(Nin), seems simple (Fig.2). After the loop-disconnect algorithm is worked out, we turn to the program for maintaining the new specified power. A program that analyzes the positions of shim rods in the core is used to ensure the required distribution of the neutron field about the radius of the core. Depend- ing on the sign of the control signals, the program issues a "move" signal to a maximally or minimally inserted rod. For example, equalization of their positions in the core is achieved in this way.

Figure 3 shows the results of calculations of the dynamics of the disconnect mode of the heat-transfer loop during response to reactivity effects solely by means of shim rods, compared with the planning algorithm using a "loop rod" specially provided for this mode and a control rod. We should note that in this case the need for loop rods disappears; these rods can be fully utilized in the emergency protection system.

480

Input

tel. units: N,G,

0,8 J

9, 610 13- ~0 q(5" z"

I H, N

z0

,, + for seiectin of rods I

f IR [of Hi ..... !~

i I

sec

~ Output

I

Fig. 3 Fig. 4

Fig. 3. Change in power (i), coolant flow rate (2), and coolant temperature (3) for a nuclear power plant with a fast reactor with a heat-transfer loop disconnected, in a treatment using shim rods alone (solid curves) and the planned algorithm (dashed curves).

Fig. 4. Structure of a control program for a nuclear power plant with second-circult pump disengaged.

INCREASING PLANT OPERATIONAL RELIABILITY

Two ways of increasing the reliability of electric-power generation seem feasible with respect to plants with fast reactors:

-Elimination of spurious signals by checking them against the aggregate of plant param- eters. Here the design of protection should not undergo any changes. The readings of sensors incorporated in the emergency protection system should be checked, e.g., by disconnecting faulty channels when the set points of the warning signal system are reached. It is not especially difficult to work up an algorithm for monitoring the reliability of protection signals based on thermophysical equations.

�9 The creation of a more highly branched plant protection system.

For example accident situations due to the failure of a signal characterizing maintenance of the preset coolant flow rate to arrive are possible. This leads to tripping of the scram system, although such accidents would be localized by disconnecting the heat-transfer loop and holding the plant at power-generating levels. Given the existing elementary base of the con- trol and safety system, performing the required analysis is complicated. The use of computers substantially facilitates this undertaking.

An increase in the operational reliability of nuclear power plants also can be achieved by reducing the thermal stresses in structural elements. The main trend of work in this area is the expansion and detailing of protection and interlocks. In case of an emergency signal, the causes of the signal must be analyzed by polling the sensors (the locations of all actua- tors, core inlet temperature, flow rates through the loops, etc.) and the plant must be af- forded differentiated protection, depending on the type of faults.

As we showed above, instead of issuing a scram signal, in some cases it is advisable to protect the plant by emergency disconnection of the heat-transfer loop.

The small number of structural assemblies generally limits the permissibility of riding out conditinns. For example, for nuclear power plants with fast reactors these assemblies are the fuel elements, tube sheets of the intermediate heat exchanger, pressure chamber, and reactor vessel. A reduction of thermal stresses can be achieved through a better-reasoned but

481

more complex protection algorithm. The basic difficulty here lies in the formulation of the control problem: if we know the contribution to the damageability of the structure for each set of conditons, we can clearly define the requirements for the algorithm. For example, thermal stresses in the upper tube sheet of the intermediate heat exchanger can be reduced severalfold through complete or partial tracking of the second-circuit flow rate from the GTsN [primary circulating pump] in coastdown with a corresponding flowrate in the emergency heat-transfer loop of the first circuit. A simplified discrete variation of the algorithm is presented in Fig. 4.

The possibility also exists of improving the course of other accident situations, includ- ing the most severe ones. For example, upon emergency switchover to natural coolant circula- tion in the first circuit, the maximum temperature on the fuel-rod cladding can be lowered significantly through the use of a more complicated transition algorithm.

SAFETY QUESTIONS

Ensuring safety becomes especially important in the rush toward greater complexity of control algorithms under conditions where computer hardware components have relatively low reliability. This contradiction (design simplification, increased reliability of the rep- resentation and perception of information with low reliability of hardware components) must be resolved on a custom basis for each specific problem while the level of safety achieved is preserved without fail.

However, the overall direction of the problem today can be seen to lie in the selection and proper formulation of the problem entrusted to the computer. For example, there is no question of the expediency of [having computers handle] routine information-presentation tasks. The safety of the plant is ensured even in problems of direct control while the tradi- tional implementation of the protective functions of the system is preserved. Nonetheless, it also seems unacceptable to lower the operational reliability of a nuclear power plant with respect to its primary function and to reduce the installed-capacity utilization factor be- cause of computer failures. In this regard we need a careful analysis of the reliability in- dices of the control system and selection of the necessary structure and degree of redundancy, including backup of the control and information functions of the computer system by traditional means.

Let us note that the frequently adopted premise that the nature of failure in computer hardware (or software) in unpredictable (0, i, or any number) has a significant effect on the results of analysis, Obviously, the specific structure of the control system will determine the type and severity of a failure due to the computer system. For example, if computer con- trol of the upward movement of shim rods is not provided for in the system (only downward move- ment is allowed)~ then this situation should not be viewed as a result of hardware failure. Furthermore, the failure generally is not detected by hardware or software and leads to termi- nation of the control action. A failure in the form of a spurious signal also is unlikely in particular because the action of a spurious signal (e.g., noise) will persist only until the next cycle of the control action.

With respect to the problem of the direct control of a nuclear power plant by means of computer technology, this last consideration seems especially significant. If the failure is identified, even in the most severe accident situation backup analog facilities that ensure the requisite performance in riding out the conditions can be provided -- e.g., through timely preparation actions by attendants.

Thus, in a practical sense the problem reduces to estimating the probability of an unde- tected failure of computer facilities, whose probability is known to be small compared with any computer failures.

i.

2.

LITERATURE CITED

S. I. Bernshtein and M. M. Solov'ev, "Hunting in a reactor power control system with two operating controls," in: Control of Nuclear Power Plants [in Russian], B. A. Kuvshin- nikov (ed.), No. 2, Atomizdat, Moscow (1967), pp. 72-85. F. M. Mitenkov, B. I. Motorov, and I. N. Romanov, "Stability of the temperature-maintenance relay system in the first circuit of a nuclear power plant," in: Control of Nuclear Power Plants [in Russian], B. A. Kuvshinnikov (ed.), No. 2, Atomizdat, Moscow (1967), pp.

19-27.

482

. F. M. Mitenkov and B. I. Motorov, Mechanisms of Unstable Processes in Thermal and Nuclear Power Engineering [in Russian], Energoatomizdat, Moscow (1981).

INCREASING THE SENSITIVITY OF THE X-RAY FLUORESCENCE METHOD

OF DETERMINING THE CONCENTRATION OF A WIDE RANGE

OF ELEMENTS

P. A. GalItsev and B. S. Iokhin UDC 543.426

One possible method of lowering the detection thresholds of energy-dispersion x-ray fluorescence analysis is preselection by the radiant energy of the sample [i, 2]. This method was originally used to analyze the sample content of a number of heavy elements [1-5]. Figure 1 shows a diagram of a cylindrical Bragg reflector for preselection by sample radiant energy.

In this paper we investigate the capabilities of the method in determining small quanti- ties of a wide range of elements in the energy range of analytical lines 4-26 keV.

Preselection by sample radiant energy leads to a situation in which the width of the analysis region AE depends on the tuning energy E of the cylindrical Bragg reflector approxi- mately as E 2 [3]~ In the energy region 4-10 keV, AE<I keV. Here just two or three neighbor- ing elements can be analyzed at once for the K series. Therefore, to extend the method to the entire studied energy region we built an apparatus that includes a mechanical system for high-speed retuning of the cylindrical Bragg reflector.

The selection of the anode material of the x-ray tube is important for effective implemen- tation of preselection. When the radiation is diffracted from the sample on the cylindrical assembly, second and higher orders of reflection with multiple energies are realized in addi- tion to the first-order reflection corresponding to the energy region of the analytical lines of the elements being analyzed. The exciting radiation scattered from the sample must not fall in the energy intervals of either the first or multiple orders; otherwise the positive effect of using this method Will be reduced significantly. If this condition can be strictly fulfilled for the primary line K~ of the anode material (the energy of K~ must be approximately 50% higher than the energy of the analytical line), then at the same time it will be fulfilled for the K B line only approximately (at second order). Because of the lack of a complete set of tubes with the required anode materials, we used a 3.5BKhV7 tube with copper and gold anodes (and copper and gold filters) for the region 4-9 keV, a palladium anode and a molybdenum filter for the region 9-12 keV, a palladium anode and filter for the region 12-17.5 keV, and a gold anode and filter (in order to segregate the spectral region of hard bremsstrahlung) for the region 17.5-26 keV. The values of the Bragg angles for this setup lie in the range 4-28 ~ The required diameter of the bottom of the cell increases with increasing Bragg angle and should correspQnd to the maximum value to standardize the cell for the entire energy range. To reduce the sample volume, we used a different method (different from that of [2, 3]) of selecting sample radiation by means of a collimator consisting of a ring and a disk mounted coaxially with respect to each other (see Fig. i). This makes it possible to reduce the volume of the sample being analyzed to approximately 0.15 cm s.

The attainable detection thresholds are determined by the background over the analytical peaks in the spectra obtained. As has been shown elsewhere [2], the background components in x-ray fluorescence analysis with energy preselection are due to the following effects:

i) The tube bremsstrahlung remaining after filtering, which is scattered from the sample;

2) The detector "tail" that appears primarily during the detection of sample-scattered exciting radiation, which is reflected in multiple orders by pyrographite;

3) Bremsstrahlung from the photoelectrons in the sample;

4) Compton scattering of exciting radiation from bound electrons in the sample.

Translated from Atomnaya Energiya, Vol. 66, No. 6, pp. 427-430, June, 1989. article submitted November 18, 1987.

Original

0038-531X/89/6606-0483512.50 �9 1989 Plenum Publishing Corporation 483