Point Kinetics Equations

As we have seen in previous chapters, the number of neutrons is multiplied by a factor keff from one neutron generation to the next, therefore the multiplication environment (nuclear reactor) behaves like the exponential system, that means the power increase is not linear, but it is exponential.

Reactor Period - Inhour EquationIt is obvious the effective multiplication factor in a multiplying system is a measure of the change in the fission neutron population from one neutron generation to the subsequent generation.

  • keff < 1. If the multiplication factor for a multiplying system is less than 1.0, then the number of neutrons is decreasing in time (with the mean generation time) and the chain reaction will never be self-sustaining. This condition is known as the subcritical state.
  • keff = 1. If the multiplication factor for a multiplying system is equal to 1.0, then there is no change in neutron population in time and the chain reaction will be self-sustaining. This condition is known as the critical state.
  • keff > 1. If the multiplication factor for a multiplying system is greater than 1.0, then the multiplying system produces more neutrons than are needed to be self-sustaining. The number of neutrons is exponentially increasing in time (with the mean generation time). This condition is known as the supercritical state.

But we have not yet discussed the duration of a neutron generation, that means, how many times in a one second we have to multiply the neutron population by a factor keff. This time determines the speed of the exponential growth. But as was written, there are different types of neutrons: prompt neutrons and delayed neutrons, which completely change the kinetic behaviour of the system. Therefore such a discussion will be not trivial.

To study the kinetic behaviour of the system, engineers usually use the point kinetics equations. The name point kinetics is used because, in this simplified formalism, the shape of the neutron flux and the neutron density distribution are ignored. The reactor is therefore reduced to a point. In the following section we will introduce the point kinetics and we start with point kinetics in its simplest form.

Let n(t) be the number of neutrons as a function of time t and l the prompt neutron lifetime, which is the average time from a prompt neutron emission to either its absorption (fission or radiative capture) or to its escape from the system. The average number of neutrons that disappear during a unit time interval dt is n.dt/l. But each disappearance of a neutron contributes an average of k new neutrons.

Finally, the change in number of neutrons during a unit time interval dt is:

derivation of point kinetics equation-min

where:

n(t) = transient reactor power

n(0) = initial reactor power

τe = reactor period

The reactor period, τe, or e-folding time, is defined as the time required for the neutron density to change by a factor e = 2.718. The reactor period is usually expressed in units of seconds or minutes. The smaller the value of τe, the more rapid the change in reactor power. The reactor period may be positive or negative.

An equation governing the neutron kinetics of the system without source and with the absence of delayed neutrons is the point kinetics equation (in certain form). This equation states that the time change of the neutron population is equal to the excess of neutron production (by fission) minus neutron loss by absorption in one prompt neutron lifetime. The role of prompt neutron lifetime is evident. Shorter lifetimes give simply faster responses of multiplying systems.

If there are neutrons in the system at t=0, that is, if n(0) > 0, the solution of this equation gives the simplest form of point kinetics equation (without source and without delayed neutrons).

point kinetics equation

This simple point kinetics equation is often expressed is terms of reactivity and prompt generation time, Λ, as:

point kinetics equation_1-min

where

  • ρ = (k-1)/k is the reactivity, which describes the deviation of an effective multiplication factor from unity.
  • Λ = l/keff = prompt neutron generation time, which is the average time from a prompt neutron emission to an absorption that results only in fission.

Both forms of the point kinetics equation are valid. The equation using Λ, prompt neutron generation time, is usually better for calculations. This is because most reactivity transients are induced by changes in the absorption cross-section rather than in the fission cross-section. During this transients the prompt neutron lifetime is not constant whereas the prompt generation time remains constant.

Example:

Let us consider that the prompt neutron lifetime is ~2 x 10-5 and k (k – neutron multiplication factor) will be step increased by only 0.01% (i.e. 10pcm or ~1.5 cents), that is k=1.0000 will increase to k=1.0001.

It must be noted such reactivity insertion (10pcm) is very small in case of LWRs. The reactivity insertions of the order of one pcm are for LWRs practically unrealizable. In this case the reactor period will be:

T = l / (k- 1) = 2 x 10-5 / (1.0001 – 1) = 0.2s

This is a very short period. In one second the neutron flux (and power) in the reactor would increase by a factor of e5 = 2.7185, in 10 seconds the reactor would pass through 50 periods and the power would increase by e50 = ……

Furthermore in case of fast reactors in which prompt neutron lifetimes are of the order of 10-7 second, the response of such a small reactivity insertion will be even more unimaginable. In case of 10-7 the period will be:

T = l / (k- 1) = 10-7 / (1.0001 – 1) = 0.001s

Reactors with such a kinetics would be very difficult to control. Fortunately this behaviour is not observed in any multiplying system. Actual reactor periods are observed to be considerably longer than computed above and therefore the nuclear chain reaction can be controlled more easily. The longer periods are observed due to the presence of the delayed neutrons.

The simplest equation governing the neutron kinetics of the system with delayed neutrons is the simple point kinetics equation with delayed neutrons. This equation states that the time change of the neutron population is equal to the excess of neutron production (by fission) minus neutron loss by absorption in one mean generation time with delayed neutrons (ld).

ld = (1 – β).lp + ∑li . βi => ld = (1 – β).lp + ∑τi . βi

where

  • (1 – β) is the fraction of all neutrons emitted as prompt neutrons
  • lp is the prompt neutron lifetime
  • τi is the mean precursor lifetime, the inverse value of the decay constant τi = 1/λi
  • The weighted delayed generation time is given by τ = ∑τi . βi / β = 13.05 s
  • Therefore the weighted decay constant λ = 1 / τ ≈ 0.08 s-1

The number, 0.08 s-1, is relatively high and have a dominating effect of reactor time response, although delayed neutrons are a small fraction of all neutrons in the core. This is best illustrated by calculating a weighted mean generation time with delayed neutrons:

ld = (1 – β).lp + ∑τi . βi = (1 – 0.0065). 2 x 10-5 + 0.085 = 0.00001987 + 0.085 ≈ 0.085

In short, the mean generation time with delayed neutrons is about ~0.1 s, rather than ~10-5 as in section Prompt Neutron Lifetime, where the delayed neutrons were omitted.

The role of ld is evident. Longer lifetimes give simply slower responses of multiplying systems. The role of reactivity (keff – 1) is also evident. Higher reactivity gives simply larger response of multiplying system.

If there are neutrons in the system at t=0, that is, if n(0) > 0, the solution of this equation gives the simplest point kinetics equation with delayed neutrons (similarly to the case without delayed neutrons):

point kinetics equation with delayed neutrons

Example:

Let us consider that the mean generation time with delayed neutrons is ~0.085 and k (k – neutron multiplication factor) will be step increased by only 0.01% (i.e. 10pcm or ~1.5 cents), that is k=1.0000 will increase to k=1.0001.

It must be noted such reactivity insertion (10pcm) is very small in case of LWRs (e.g. one step by control rods). The reactivity insertions of the order of one pcm are for LWRs practically unrealizable. In this case the reactor period will be:

T = ld / (k-1) = 0.085 / (1.0001-1) = 850s

This is a very long period. In ~14 minutes the neutron flux (and power) in the reactor would increase by a factor of e = 2.718. This is completely different dimension of the response on reactivity insertion in comparison with the case without presence of delayed neutrons, where the reactor period was 1 second.

Both previous simple point kinetics equations are only an approximation, because they use many simplifications. The simple point kinetics equation with delayed neutrons completely fails for higher reactivity insertions, where is significant difference between the production of prompt and delayed neutrons. Therefore a more accurate model is required. The exact point kinetics equations, that can be derived from the general neutron balance equations without making any approximations are:

point kinetics equations_1

In the equation for neutrons, the first term on the right hand side is the production of prompt neutrons in the present generation, k(1-β)n/l, minus the total number of neutrons in the preceding generation, -n/l. The second term is the production of delayed neutrons in the present generation. As can be seen, the rate of absorption of neutrons is the same as in the simple model (-n/l). But a distinction is between the direct channel for prompt neutrons (1-β) production and the delayed channel resulting from radioactive decay of precursor nuclei (λiCi).

In the equation for precursors, there is the balance between the production of the precursors of i-th group and their decay after the decay constant λi. As can be seen, the rate of the decay of precursors is the radioactivity rate (λiCi) and the rate of production is proportional to the number of neutrons times βi, which is defined as the fraction of the neutrons which appear as delayed neutrons in the ith group.

As can be seen, the point kinetics equations include two differential equations, one for the neutron density n(t) and the other for precursors concentration C(t).

Again, the point kinetics equations are often expressed is terms of reactivity (ρ = (k-1)/k) and prompt generation time, Λ, as:

point kinetics equations_2

Both forms of the point kinetics equation are valid. The equation using Λ, prompt neutron generation time, is usually better for calculations. This is because most reactivity transients are induced by changes in the absorption cross-section rather than in the fission cross-section. During this transients the prompt neutron lifetime is not constant whereas the prompt generation time remains constant.

Reactor Kinetics - change in reactivity

Solution of point kinetics equations for a cyclic change in reactivity

The previous equation defines the reactivity of a reactor, which describes the deviation of an effective multiplication factor from unity. For critical conditions the reactivity is equal to zero. The larger the absolute value of reactivity in the reactor core, the further the reactor is from criticality. In fact the reactivity may be used as a measure of a reactor’s relative departure from criticality. According to the reactivity, we can classify the different reactor states and the related consequences as follows:

The prompt critical state is defined as:

keff > 1; ρ ≥ βeff, where the reactivity of a reactor is higher than the effective delayed neutron fraction. In this case, the production of prompt neutrons alone is enough to balance neutron losses and increase the neutron population. The number of neutrons is exponentially increasing in time (as rapidly as the prompt neutron generation lifetime ~10-5s).

The prompt subcritical and delayed supercritical state is defined as:

keff > 1; 0 < ρ < βeff, where the reactivity of a reactor is higher than zero and lower than the effective delayed neutron fraction. In this case, the production of prompt neutrons alone is insufficient to balance neutron losses and the delayed neutrons are needed in order to sustain the chain reaction. The neutron population increases, but much more slowly (as the mean generation lifetime with delayed neutrons ~0.1 s).

The prompt subcritical and delayed critical state is defined as:

keff = 1; ρ = 0, where the reactivity of a reactor is equal to zero. In this case, the production of prompt neutrons alone is insufficient to balance neutron losses and the delayed neutrons are needed in order to sustain the chain reaction. There is no change in neutron population in time and the chain reaction will be self-sustaining. This state is the same state as the critical state from basic classification.

The prompt subcritical and delayed subcritical state is defined as:

keff < 1; ρ < 0, where the reactivity of a reactor is lower than zero. In this case, the production of all neutrons is insufficient to balance neutron losses and the chain reaction is not self-sustaining. If the reactor core contains external or internal neutron sources, the reactor is in the state that is usually referred to as the subcritical multiplication.

Approximate Solution of Point Kinetics Equations

Sometimes, it is convenient to predict qualitatively the behaviour of a reactor. The exact solution can be obtained relatively easy using computers.  Especially for illustration, the following approximations are discussed in the following sections:

  • Prompt Jump Approximation
  • Prompt Jump Approximation with One Group of Delayed Neutrons
  • Constant Delayed Neutron Source Approximation
As can be seen from the solution of exact point kinetics equation, any reactivity insertion (ρ < β) causes at first a sharp change in prompt neutrons population and then the neutron response is slowed as a result of the more slowly changing number of delayed neutrons. The rapid response is a result of the small value of prompt neutron generation time in the denominator of point kinetics equation.

If we are interested in long-term behaviour (asymptotic period) and not interested in the details of the prompt jump, we can simplify the point kinetics equations by assuming that the prompt jump takes place instantaneously in response to any reactivity change. This approximation is known as the Prompt Jump Approximation (PJA) in which the rapid power change due to prompt neutrons is neglected, corresponding to taking dn/dt |0 = 0 in the point kinetics equations. That means the point kinetics equations are as follows:

prompt jump approximation - equation

From the equation for neutron flux and the assumption, that the delayed neutron precursor population does not respond instantaneously to a change in reactivity (i.e. Ci,1 = Ci,2), it can be derived that the ratio of the neutron population just after and before the reactivity change is equal to:

prompt jump approximation - prompt jump

The prompt-jump approximation is usually valid for smaller reactivity insertion, for example, for ρ < 0.5β. It is usually used with another simplification, the one delayed precursor group approximation.

In the previous section we have simplified the point kinetics equation using prompt jump approximation (PJA). This eliminated the fast time scale due to prompt neutrons. In this section we consider that delayed neutrons are produced only by one group of precursors with the same decay constant (averaged) and delayed neutron fraction. Point kinetics equation using PJA and one group of delayed neutrons becomes:

PJA with one group

This simplification then leads to:

solution - PJA - one group

Assuming that the reactivity is constant and n1/n0 can be determined from prompt jump formula this equation leads to very simple formula:

solution - PJA - one group2

As can be seen from the solution of exact point kinetics equation, any reactivity insertion (ρ < β) causes at first a sharp change in prompt neutrons population and then the neutron response is slowed as a result of the more slowly changing number of delayed neutrons. The rapid response is a result of the small value of prompt neutron generation time in the denominator of point kinetics equation.

If we are interested in short-term behaviour and not interested in the details of the asymptotic behaviour, we can simplify the point kinetics equations by assuming that the production of the delayed neutrons is constant and equal to the production at the beginning of the transient. This approximation is known as the Constant Delayed Neutron Source Approximation (CDS) in which changes in the amount of delayed neutrons is neglected, corresponding to taking dCi(t)/dt = 0 and Ci(t) = Ci,0 in the point kinetics equations. That means the point kinetics equations are as follows:

constant production of delayed neutrons

The above equation can be solved analytically and assuming that the reactivity is constant the solution is given as:

constant production of delayed neutrons 2

Point Kinetics Equations – Notes

Prompt neutron lifetime, l, is the average time from a prompt neutron emission to either its absorbtion (fission or radiative capture) or to its escape from the system. This parameter is defined in multiplying or also in nonmultiplying systems. In both systems the prompt neutron lifetimes depend strongly on:

  • material composition of the system
    • multiplying – nonmultiplying system
    • system with or without thermalization
    • isotopic composition of the system
  • geometric configuration of the system
    • homogeneous or heterogeneous system
    • shape of entire system
  • size of the system

In an infinite reactor (without escape) prompt neutron lifetime is the sum of the slowing down time and the diffusion time.

l=ts + td

In an infinite thermal reactor ts << td and therefore l ≈ td. The typical prompt neutron lifetime in thermal reactors is on the order of 10−4 second. Generally, the longer neutron lifetimes take place in systems in which the neutrons must be thermalized in order to be absorbed.

Systems in which most of the neutrons are absorbed at higher energies and the neutron thermalization is suppressed (e.g. in fast reactors), have much shorter prompt neutron lifetimes . The typical prompt neutron lifetime in fast reactors is on the order of 10−7 second.

Slowing Down and Diffision Times - Prompt Neutron Lifetime

Slowing Down and Diffusion Times for Thermal Neutrons in an Infinite Medium
Source: Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.

Prompt Neutron Lifetime - Reactor Types

Dependencies of asymptotic time period on the reactivity required for different reactor types with different prompt neutron lifetimes.

In multiplying systems, in which the absorption of a prompt fission neutron can initiate a fission reaction, l is equal to the average time between two generations of prompt neutrons (at keff=1). This time is known as the prompt neutron generation time.

Prompt Neutron Generation Time (or Mean Generation Time), Λ, is the average time from a prompt neutron emission to a capture that results only in fission. The prompt neutron generation time is designated as:

Λ = l/keff

In power reactors the prompt generation time changes with the fuel burnup. In LWRs increases with the fuel burnup. It is simple, fresh uranium fuel contains much fissile material (in case of uranium fuel about 4% of 235U). This causes significant excess of reactivity and this excess must be compensated via chemical shim (in case of PWRs) or via burnable absorbers.

Owing to these factors (high probability of absorption in fuel and high probability of absorption in moderator) the prompt neutron lives much shorter and prompt neutron lifetime is low. With fuel burnup the amount of fissile material as well as the absorption in moderator decreases and therefore the prompt neutron is able to “live”much longer.

The prompt neutron lifetime belongs to key neutron-physical characteristics of reactor core. Its value depends especially on the type of the moderator and on the energy of the neutrons causing fission. Its importance for nuclear reactor safety is well known for a long time.

The longer prompt neutron lifetimes can substantially improve kinetic response of reactor (the longer prompt neutron lifetime gives simply slower power increase). For example under RIA conditions (Reactivity-Initiated Accidents) reactors should withstand a jump-like insertion of relatively large (~1 $ or even more) positive reactivity and the PNL (prompt neutron lifetime) plays here the key role. Therefore the PNL should be verified in a reload safety evaluation process.

In some cases (especially in some fast reactors) reactor cores or can be modified in order to increase the PNL and in order to improve nuclear safety.

Delayed neutron fraction - yield

Source: Keepin, G. R., Physics of Nuclear Kinetics. Addison-Wesley, 1965.

Reactor-kinetic calculations with considering of such a number of initial conditions would be correct, but it also would be very complicated. Therefore G. R. Keepin and his co-workers suggested to group together the precursors based on their half-lives.Therefore delayed neutrons are traditionally represented by six delayed neutron groups, whose yields and decay constants (λ) are obtained from nonlinear least-squares fits to experimental measurements. This model has following disadvantages:

  • All constants for each group of precursors are empirical fits to the data.
  • They cannot be matched with decay constants of specific precursors.
Although this six group parameterization still satisfies the requirements of commercial organizations, a higher accuracy of the delayed neutron yields and a better energy resolution in the delayed neutron spectra is desired.

It was recognised that the half-lives in six-group structure do not accurately reproduce the asymptotic die-away time constants associated with the three longest-lived dominant precursors: 87Br, 137I and 88Br.

This model may be insufficient especially in case of epithermal reactors, because virtually all delayed neutron activity measurements have been performed for fast or thermal-neutron-induced fission. In case of fast reactors, in which the nuclear fission of six fissionable isotopes of uranium and plutonium is important, the accuracy and energy resolution may play an important role.

In nuclear reactors the gamma radiation plays a significant role also in reactor kinetics and in a subcriticality control. Especially in nuclear reactors with D2O moderator (CANDU reactors) or with Be reflectors (some experimental reactors). Neutrons can be produced also in (γ, n) reactions and therefore they are usually referred to as photoneutrons.

A high energy photon (gamma ray) can under certain conditions eject a neutron from a nucleus. It occurs when its energy exceeds the binding energy of the neutron in the nucleus. Most nuclei have binding energies in excess of 6 MeV, which is above the energy of most gamma rays from fission. On the other hand there are few nuclei with sufficiently low binding energy to be of practical interest. These are: 2D, 9Be6Li, 7Li and 13C. As can be seen from the table the lowest threshold have 9Be with 1.666 MeV and 2D with 2.226 MeV.

Photoneutron sources

Nuclides with low photodisintegration
threshold energies.

In case of deuterium, neutrons can be produced by the interaction of gamma rays (with a minimum energy of 2.22 MeV) with deuterium:

Photoneutron - deuterium

Because gamma rays can be emitted by fission products with certain delays, and the process is very similar to that through which a “true” delayed neutron is emitted, photoneutrons are usually treated no differently than regular delayed neutrons in the kinetic calculations. Photoneutron precursors can be also grouped by their decay constant, similarly to “real” precursors. The table below shows the relative importance of source neutrons in CANDU reactors by showing the makeup of the full power flux.

Photoneutron balance

Despite the fact photoneutrons are of importance especially in CANDU reactors, deuterium nuclei are always present (~0.0156%also in the light water of LWRs. Moreover the capture of neutrons in the hydrogen nucleus of the water molecules in the moderator yields small amounts of D2O. This enhances the heavy water concentration. Therefore also in LWRs kinetic calculations, photoneutrons from D2O are treated as additional groups of delayed neutrons having characteristic decay constants λj and effective group fractions.

After a nuclear reactor has been operated at full power for some time there will be a considerable build-up of gamma rays from the fission products. This high gamma flux from short-lived fission products will decrease rapidly after shutdownIn the long term the photoneutron source decreases with the decay of long-lived fission products that produce delayed high-energy gamma rays and the photoneutron source drops slowly, decreasing a little each day. The longest-lived fission product with gamma ray energy above the threshold is 140Ba with a half-life of 12.75 days.

The amount of fission products present in the fuel elements depends on how long has been the reactor operated before shut-down and at which power level has been the reactor operated before shut-down. Photoneutrons are usually major source in a reactor and ensure sufficient neutron flux on source range detectors when reactor is subcritical in long term shutdown.

In comparison with fission neutrons, that make a self-sustaining chain reaction possibledelayed neutrons make reactor control possible and photoneutrons are of importance at low power operation.

See also: Delayed Neutron Fraction

See also: Effective Delayed Neutron Fraction – βeff

The delayed neutron fraction, β, is the fraction of delayed neutrons in the core at creation, that is, at high energies. But in case of thermal reactors the fission can be initiated mainly by thermal neutron. Thermal neutrons are of practical interest in study of thermal reactor behaviour. The effective delayed neutron fraction, usually referred to as βeff, is the same fraction at thermal energies.

The effective delayed neutron fraction reflects the ability of the reactor to thermalize and utilize each neutron produced. The β is not the same as the βeff due to the fact delayed neutrons do not have the same properties as prompt neutrons released directly from fission. In general, delayed neutrons have lower energies than prompt neutrons. Prompt neutrons have initial energy between 1 MeV and 10 MeV, with an average energy of 2 MeV. Delayed neutrons have initial energy between 0.3 and 0.9 MeV with an average energy of 0.4 MeV.

Therefore in thermal reactors a delayed neutron traverses a smaller energy range to become thermal and it is also less likely to be lost by leakage or by parasitic absorption than is the 2 MeV prompt neutron. On the other hand, delayed neutrons are also less likely to cause fast fission, because their average energy is less than the minimum required for fast fission to occur.

These two effects (lower fast fission factor and higher fast non-leakage probability for delayed neutrons) tend to counteract each other and forms a term called the importance factor (I). The importance factor relates the average delayed neutron fraction to the effective delayed neutron fraction. As a result, the effective delayed neutron fraction is the product of the average delayed neutron fraction and the importance factor.

βeff = β . I

The delayed and prompt neutrons have a difference in their effectiveness in producing a subsequent fission event. Since the energy distribution of the delayed neutrons differs also from group to group, the different groups of delayed neutrons will also have a different effectiveness. Moreover, a nuclear reactor contains a mixture of fissionable isotopes. Therefore, in some cases, the importance factor is insufficient and an importance function must be defined.

For example:

In a small thermal reactor with highly enriched fuel, the increase in fast non-leakage probability will dominate the decrease in the fast fission factor, and the importance factor will be greater than one.

In a large thermal reactor with low enriched fuel, the decrease in the fast fission factor will dominate the increase in the fast non-leakage probability and the importance factor will be less than one (about 0.97 for a commercial PWR).

In large fast reactors, the decrease in the fast fission factor will also dominate the increase in the fast non-leakage probability and the βeff is less than β by about 10%.

Table of main kinetic parameters.

Table of main kinetic parameters.

Mean Generation Time with Delayed Neutrons, ld, is the weighted average of the prompt generation times and a delayed neutron generation time. The delayed neutron generation time, τ, is the weighted average of mean precursor lifetimes of the six groups (or more groups) of delayed neutron precursors.

It must be noted, the true lifetime of delayed neutrons (the slowing down time and the diffusion time) is very short compared with the mean lifetime of their precursors (ts + td << τi). Therefore τi is also equal to the mean lifetime of a neutron from the ith group, that is, τi = li and the equation for mean generation time with delayed neutrons is the following:

ld = (1 – β).lp + ∑li . βi => ld = (1 – β).lp + ∑τi . βi

where

  • (1 – β) is the fraction of all neutrons emitted as prompt neutrons
  • lp is the prompt neutron lifetime
  • τis the mean precursor lifetime, the inverse value of the decay constant τi = 1/λi
  • The weighted delayed generation time is given by τ = ∑τi . βi / β = 13.05 s
  • Therefore the weighted decay constant λ = 1 / τ ≈ 0.08 s-1

The number, 0.08 s-1, is relatively high and have a dominating effect of reactor time response, although delayed neutrons are a small fraction of all neutrons in the core. This is best illustrated by calculating a weighted mean generation time with delayed neutrons:

ld = (1 – β).lp + ∑τi . βi = (1 – 0.0065). 2 x 10-5 + 0.085 = 0.00001987 + 0.085 ≈ 0.085

In short, the mean generation time with delayed neutrons is about ~0.1 s, rather than ~10-5 as in section Prompt Neutron Lifetime, where the delayed neutrons were omitted.

Despite the fact the number of delayed neutrons per fission neutron is quite small (typically below 1%) and thus does not contribute significantly to the power generation, they play a crucial role in the reactor control and are essential from the point of view of reactor kinetics and reactor safety. Their presence completely changes the dynamic time response of a reactor to some reactivity change, making it controllable by control systems such as the control rods.

Delayed neutrons allow to operate a reactor in a prompt subcritical, delayed critical condition. All power reactors are designed to operate in a delayed critical conditions and are provided with safety systems to prevent them from ever achieving prompt criticality.

For typical PWRs, the prompt criticality occurs after positive reactivity insertion of βeff(i.e. keff ≈ 1.006 or ρ = +600 pcm). In power reactors such a reactivity insertion is practically impossible to insert (in case of normal and abnormal operation), especially when a reactor is in power operation mode and a reactivity insertion causes a heating of a reactor core. Due to the presence of reactivity feedbacks the positive reactivity insertion is counterbalanced by the negative reactivity from moderator and fuel temperature coefficients. The presence of delayed neutrons is of importance also from this point of view, because they provide time also to reactivity feedbacks to react on undesirable reactivity insertion.

Press the “clear and run” button and try to increase the power of the reactor.

Compare the response of the reactor with the case of Infinite Multiplying System Without Source and without Delayed Neutrons (or set the β = 0).

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.

See above: