Reactor Shutdown

Decay Heat - Nuclear Power - Thermal PowerA reactor is considered to be shut down when it is subcritical and sufficient shutdown reactivity exists so there is no immediate probability of secondary criticality. Despite the fact the chain reaction is quickly broken down, all of the fissioning in the core is not stopped. There are always neutrons in the core and therefore a residual fissioning occurs. On the other hand, the amount of heat being generated due to the residual fissions (initiated by source neutrons) which are not stopped and the decay heat is much less than that which can be removed by the plant systems.

In general, there are two types of reactor shutdown:

  • Emergency Shutdown – SCRAM. A reactor “SCRAM” (or “reactor trip”) is the rapid insertion or fall of the control rods into the core to stop the fission chain reaction. It is a key safety function of each nuclear reactor and usually does not need any power supply, since control rods fall only by gravitational force.
  • Gradual Shutdown. Usually there is no need to stop the chain reaction promptly. Shutdown is normally accomplished by gradual insertion of some (or all) of the control rods, or by introduction of soluble neutron poison (boric acid in case of PWRs) into the reactor coolant. Reactor thermal power is in the first instance decreased from 100% of rated power to about 10 – 30% of rated thermal power. At this power level the turbine is disconnected from the electrical grid, while all the steam production is taken over by the turbine bypass system (TBS). The reactor is now at zero electrical load and the thermal power can be decreased under 1% of rated power (to the Startup Mode) and the reactor dynamics changes to zero-power reactor dynamics. Further introduction of negative reactivity causes controlled exponential decrease in neutron power and after reaching shutdown margin (e.g. keff = 0.98) the reactor is in safe shutdown state.
Every steam turbine is also provided with emergency governors which come into action under specific conditions. In general, an unplanned or emergency shutdown of a turbine is known as a “turbine trip”.  The turbine trip signal initiates fast closure of all steam inlet valves  (e.g. turbine stop valves – TSVs) to block steam flow through the turbine.

The turbine trip event is a standard postulated transient, which must be analyzed in the Safety Analysis Report (SAR) for nuclear power plants.

See also: Turbine Trip

Emergency Shutdown – SCRAM

A reactor “SCRAM” (or “reactor trip”) is the rapid insertion or fall of the control rods into the core to stop the fission chain reaction. At PWRs, all control rods are usually inserted within two to four seconds. Control rods are an important safety and control system of nuclear reactors. Their prompt action and prompt response of the reactor is indispensable. Control rods are also used for maintaining the desired state of chain reaction within a nuclear reactor (i.e. subcritical state, critical state, supercritical state). Reactor Trip System (RTS), which actuates a reactor trip, is a part of a Reactor Protection System (RPS).
In a pressurized water reactor, the control rods are inserted (dropped) from the top of the reactor vessel into the core. In a boiling water reactor, the control rods are inserted from the bottom of the reactor vessel into the core.

SCRAM from Hot Zero Power – Dynamics

Assume a reactor trip from hot zero power (from “zero power criticality”). The power decrease, in this case, does not influence the criticality (keff). In this case, the reactivity of the multiplying system becomes quickly negative and higher than β (i.e.that ρ is negative and ρ β).

The rate at which the reactor fission rate decays immediately following shutdown is similar for all reactors provided a large amount of negative reactivity is inserted. After a large negative reactivity addition the neutron level undergoes a rapid decrease of about two decades (prompt drop) until it is at the level of production of delayed neutrons.

But even if it were possible to insert an infinite negative reactivity, the neutron flux would not immediately fall to zero. Prompt neutrons will be absorbed almost immediately. It is consistent with the prompt drop formula. Therefore the resulting neutron flux will be:

negative period - prompt drop

It is obvious, the neutron flux cannot drop below the value βn1. The real values are much higher. The integral worth of all control and emergency rods (PWRs) is for example -9000pcm. It is equal to ρ = -9000/600 = -15β = -0.09. (β= 600pcm = 0.006)

For this negative reactivity the prompt drop is equal to:

n2/n1 = 0.006/(0.006+0.09)=0.063

which is about ten times higher than in case of an infinite negative reactivity insertion.

reactor trip - scram - point kineticsThe neutron flux then continues to fall according to stable period. The first root of reactivity equation occurs at s0 = – λ1the decay constant of the long-lived precursors group. The shortest negative stable period is then e = – 1/λ1 = -80s. The neutron flux cannot be reduced more rapidly than this period. On the other hand, the prompt drop causes an immediate drop to about 6% of rated power and within few tens of seconds the thermal power which originates from nuclear fission is below the thermal power which originates from decay heat.

There is an exception to this behavior and it is a SCRAM of heavy water reactor. In a heavy water reactor, the photoneutron source is extremely large after shutdown due to the amount of deuterium in the moderator and the large number of high energy gammas from short-lived fission product decay. This high gamma flux from short-lived fission products will decrease rapidly after shutdown. The photoneutron source is large enough to have a significant impact on neutron population immediately after shutdown. The photoneutron source has the result of flux levels decreasing more slowly so that a heavy water reactor will have a significantly larger negative reactor period after a shutdown.

SCRAM from Hot Full Power – Dynamics

For power reactors at power conditions the reactor can behave differently as a result of the presence of reactivity feedbacks. Power reactors are initially started up from hot standby mode (subcritical state at 0% of rated power) to power operation mode (100% of rated power) by withdrawing control rods and by boron dilution from the primary coolant. During the reactor startup and up to about 1% of rated power, the reactor kinetics is exponential as in zero power reactor. This is due to the fact all temperature reactivity effects are minimal.

During further power increase 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 total amount of feedback reactivity that must be offset by control rod withdrawal or boron dilution during the power increase is known as the power defect. The power defects for PWRs, graphite-moderated reactors, and sodium-cooled fast reactors are:

  • about 2500pcm for PWRs,
  • about 800pcm for graphite-moderated reactors
  • about 500pcm for sodium-cooled fast reactors

See also: Operational factors that affect the multiplication in PWRs

This exactly but in the opposite direction acts after a reactor trip (SCRAM) from the Hot Full Power state (HFP). It is logical, as power defects act against power increase, they act also against power decrease. When reactor power is decreased quickly, as in the case of reactor trip, power defect causes a positive reactivity insertion, and the initial rod insertion must be sufficient to make the reactor safe subcritical. Similarly as in the HZP state, the integral worth of all control and emergency rods (PWRs) is for example -9000pcm. It is equal to ρ = -9000/600 = -15β = -0.09. (β= 600pcm = 0.006). Also in this case a prompt drop occurs. The prompt drop is quicker than the deaccumulation of heat from fuel pellets. For this negative reactivity the prompt drop is equal to:

n2/n1 = 0.006/(0.006+0.09)=0.063

but the subsequent power decrease is strongly influenced by changes in core (fuel and moderator) temperatures. After reaching stable temperature, the neutron flux may continue to fall (when subcritical) according to stable period. It is obvious, if the power defect for PWRs is about 2500pcm (about 4 βeff), the control rods must weigh more than 2500pcm to achieve the subcritical condition. To ensure the safe subcritical condition, the control rods must weigh more than 2500pcm plus value of SDM (SHUTDOWN MARGIN). The total weigh of control rods is design specific, but, for example, it may reach about 6000 to 9000pcm. To ensure that the control rods can safe shut down the reactor, they must be maintained above a minimum rod height (rods insertion limits) specified in the technical specifications.

A distinction should be made between indicated reactor power level measured by excore neutron detectors after shutdown and the actual thermal power level. The indicated reactor power level, usually known as the nuclear power, is the power produced directly from fission in the reactor core, but the actual thermal power drops more slowly due to decay heat production as previously discussed. Decay heat, although approximately 5 to 6% of the steady state reactor power prior to shutdown, diminishes to less than 1% of the pre-shutdown power level after about one hour.

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.
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.
  9. Paul Reuss, Neutron Physics. EDP Sciences, 2008. ISBN: 978-2759800414.

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.