Reactor Power Control

WWER-1000 reactor

Description of VVER-1000 reactor.
1) Control Element Drive Mechanism
2) Reactor vessel head
3) Reactor pressure vessel
4) Coolant inlet – outlet nozzles
5) Downcomer for coolant
6) Neutron reflector
7) Fuel assemblies
Source: www.wikipedia.org

In general, the reactor thermal power and the outlet temperature of the coolant from the reactor core are controlled by manipulating several factors which affect the core’s reactivity. In PWRs, these factors are especially:

  • position of control rods,
  • concentration of boric acid in the RCS
  • core inlet temperature

Control Rods

Control rods are rods, plates, or tubes containing a neutron absorbing material (material with high absorption cross-section for thermal neutron) such as boron, hafnium, cadmium, etc., used to control the power of a nuclear reactor. Control rods usually constitute cluster control rod assemblies (PWR) and are inserted into guide thimbles within a nuclear fuel assembly. The absorbing material (e.g. pellets of Boron Carbide) is protected by the cladding usually made of stainless steel. They are grouped into groups (banks) and the movement occurs usually by the groups (banks). Typical total number of clusters is 70. This number is limited especially by number of penetrations of the reactor pressure vessel head.

A control rod is removed from or inserted into the reactor core in order to increase or decrease the reactivity of the reactor (increase or decrease the neutron flux). Control rods (insertion/withdrawal) influences the thermal utilization factor. For example, control rods insertion causes an addition of new absorbing material into the core and this causes a decrease in thermal utilization factor.

In comparison with the chemical shim, which offset positive reactivity excess in entire core, with control rods the unevenness of neutron-flux density in the reactor core may arise, because they act locally.

Concentration of Boric Acid

Boron letdown curve (chemical shim) and boron 10 depletion

Boron letdown curve (chemical shim) and boron 10 depletion during a 12-month fuel cycle.

In pressurized water reactors, chemical shim (boric acid) is used to compensate an excess of reactivity of reactor core along the fuel burnup (long term reactivity control) as well as to compensate the negative reactivity from the power defect and xenon poisoning during power increase to nominal power.

The concentration of boric acid diluted in the primary coolant influences the thermal utilization factor. For example, an increase in the concentration of boric acid (chemical shim) causes an addition of new absorbing material into the core and this causes a decrease in thermal utilization factor.

When compared with burnable absorbers (long term reactivity control) or with control rods (rapid reactivity control) the boric acid avoids the unevenness of neutron-flux density in the reactor core, because it is dissolved homogeneously in the coolant in entire reactor core. On the other hand high concentrations of boric acid may lead to positive moderator temperature coefficient and that is undesirable. In this case more burnable absorbers must be used.

Moreover this method is slow in controlling reactivity. Normally, it takes several minutes to change the concentration (dilute or borate) of the boric acid in the primary loop. For rapid changes of reactivity control rods must be used.

During any power increase the temperature, pressure, or void fraction change and the reactivity of the core changes accordingly. It is difficult to change any operating parameter and not affect every other property of the core. Since it is difficult to separate all these effects (moderator, fuel, void etc.) the power coefficient is defined. The power coefficient combines the Doppler, moderator temperature, and void coefficients. It is expressed as a change in reactivity per change in percent power, Δρ/Δ% power. The value of the power coefficient is always negative in core life but is more negative at the end of the cycle primarily due to the decrease in the moderator temperature coefficient.

Let assume that the reactor is critical at 75% of rated power and that the plant operator wants to increase power to 100% of rated power. The reactor operator must first bring the reactor supercritical by insertion of a positive reactivity (e.g. by control rodwithdrawal or boron dilution). As the thermal power increases, moderator temperature and fuel temperature increase, causing a negative reactivity effect (from the power coefficient) and the reactor returns to the critical condition. In order to keep the power to be increasing, positive reactivity must be continuously inserted (via control rods or chemical shim). After each reactivity insertion, the reactor power stabilize itself proportionately to the reactivity inserted. The total amount of feedback reactivity that must be offset by control rod withdrawal or boron dilution during the power increase (from ~1% – 100%) is known as the power defect.

Let assume:

  • the power coefficient:                 Δρ/Δ% = -20pcm/% of rated power
  • differential worth of control rods:    Δρ/Δstep = 10pcm/step
  • worth of boric acid:                                      -11pcm/ppm
  • desired trend of power decrease:              1% per minute

75% → ↑ 20 steps or ↓ 18 ppm of boric acid within 10 minutes → 85% → next ↑ 20 steps or ↓ 18 ppm within 10 minutes → 95% → final ↑ 10 steps or ↓ 9 ppm within 5 minutes → 100%

reactor power - 75 to 100 of rated power

Power increase. Let assume that the reactor is critical at 75% of rated power and that the plant operator wants to increase power to 100% of rated power.

In general, we have to distinguish between three types of power outputs in power reactors.

  • Nuclear Power. Since the thermal power produced by nuclear fissions is proportional to neutron flux level, the most important, from reactor safety point of view, is a measurement of the neutron flux. The neutron flux is usually measured by excore neutron detectors, which belong to so called the excore nuclear instrumentation system (NIS). The excore nuclear instrumentation system monitors the power level of the reactor by detecting neutron leakage from the reactor core. The excore nuclear instrumentation system is considered a safety system, because it provide inputs to the reactor protection system during startup and power operation. This system is of the highest importance for reactor protection system, because changes in the neutron flux can be almost promptly recognized only via this system.
  • Thermal Power. Although the nuclear power provides prompt response on neutron flux changes and it is irreplaceable system, it must be calibrated. The accurate thermal power of the reactor can be measured only by methods based on energy balance of primary circuit or energy balance of secondary circuit. These methods provide accurate reactor power, but these methods are insufficient for safety systems. Signal inputs to these calculations are, for example, the hot leg temperature or the flow rate through the feedwater system and these signals change very slowly with neutron power changes.
  • Electrical Power. Electric power is the rate at which electrical energy is generated by the generator. For example, for a typical nuclear reactor with a thermal power of 3000 MWth, about ~1000MWe of electrical power is generated in the generator.
See also: Power Distribution

axial-temperature-profile

It follows the temperature of top half of the core increases more than the temperature of bottom half of the core. Since the moderator temperature feedback must be negative, the power from top half will shift to bottom half.

The power distribution significantly changes also with changes of thermal power of the reactor. During power changes at power operation mode (i.e. from about 1% up to 100% of rated power) the temperature reactivity effects play very important role. As the neutron population increases, the fuel and the moderator increase its temperature, which results in decrease in reactivity of the reactor (almost all reactors are designed to have the temperature coefficients negative). The negative reactivity coefficient acts against the initial positive reactivity insertion and this positive reactivity is offset by negative reactivity from temperature feedbacks.

This effect naturally occurs on a global scale, and also on a local scale.

During thermal power increase the effectiveness of temperature feedbacks will be greatest where the power is greatest. This process causes the flattening of the flux distribution, because the feedbacks acts stronger on positions, where the flux is higher.

It must be noted, the effect of change in the thermal power have significant consequences on the axial power distribution (global power distribution).

In reality, when there is a change in the thermal power and the coolant flow rate remains the same, the difference between inlet and outlet temperatures must increase. It follows from basic energy equation of reactor coolant, which is below:

P=↓ṁ.c.↑∆t

Power increase. Let assume that the reactor is critical at 75% of rated power and that the plant operator wants to increase power to 100% of rated power.

The inlet temperature is determined by the pressure in the steam generators, therefore the inlet temperature changes minimallyduring the change of thermal power. It follows the outlet temperaturemust change significantly as the thermal power changes. When the inlet temperature remains almost the same and the outlet changes significantly, it stands to reason, the average temperature of coolant (moderator) will change also significantly. It follows the temperature of top half of the core increases more than the temperature of bottom half of the core. Since the moderator temperature feedback must be negative, the power from top half will shift to bottom half. In short, the top half of the core is cooled (moderated) by hotter coolant and therefore it is worse moderated. Hence the axial flux difference, defined as the difference in normalized flux signals (AFD) between the top and bottom halves of a two section excore neutron detector, will decrease.

AFD is defined as:

AFD or ΔI = Itop – Ibottom

where Itop and Ibottom are expressed as a fraction of rated thermal power.

Nuclear and Reactor Physics:

  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2. 
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.