There are many parameters and calculations to consider when designing a nuclear power plant system. The main goal is to design a nuclear power plant system that generates a large amount of electricity, but does not exceed any safety limits. The power generation must be done in such a way that the temperatures in the core do not exceed material property limitations, and the critical heat flux in the hottest, most limiting channels are not exceeded. Additional things to consider are economic and environmental properties.
I. POWER REQUIREMENTS
The first task in designing a nuclear power plant system is to determine the net power output. This is usually done by the electric utility ordering the plant and depends on the network and the electrical grid growth requirements.7 For this project, the net power output desired is 800 MWe.
I.A. Power Limitations
The power generation in a plant must be done in such a manner that the temperatures anywhere in the core must not exceed material property limitations, and the critical heat flux in the hottest, most limiting channels cannot be exceeded.7 This is accomplished by adjusting the amount of fuel, reactor flow, operating conditions, and heat transfer surface area.7
II. SELECTION OF REACTOR TYPE
The second task in designing a nuclear power plant system is to select the reactor type. The selection of the reactor type is usually also chosen by the electric utility ordering the plant. Things to consider during selection are costs, desires to gain experience in a new field or type of power plant, and desires to use similar reactor type that the utility currently owns.7
II.A. Reactor Types
There are various types of reactors to choose from, for instance, pressurized-water reactors, boiling-water reactors, gas-cooled reactors, fast-breeder reactors, dual-purpose desalination-power reactors, etc.7
II.A.1. Westinghouse PWR
The reactor type chosen for this project is the Westinghouse PWR design. The Westinghouse PWR design consists of a primary, secondary, and tertiary loop. The primary loop contains the heat source, which comes from the energy resulting from the controlled fission reactions to be transformed into sensible heat in the coolant-moderator.1 The coolant is pumped to the steam generator, which then transfers the heat to the secondary loop, and the coolant returns back to the reactor vessel to be heated again. To maintain a system pressure high enough to prevent bulk boiling in the core, an electrically heated pressurizer is connected to the primary loop.1
The secondary loop is where the dry saturated steam produced in the steam generator flows to the turbine-generator and is expanded to convert the thermal energy into mechanical energy and, thus, electrical energy.1 The expanded steam then moves to the condenser where the latent heat of vaporization is transferred to the tertiary loop.
The tertiary loop is where the latent heat of vaporization is eliminated into the environment through the condenser cooling water. The specific location for rejected heat varies depending on the location of the site, but examples could be a river, lake, ocean, or cooling tower system.
III. STEAM CYCLE DESIGN
The third task in designing a nuclear power plant system is to design the steam cycle. The goal when designing the steam cycle is to conserve as much energy as possible, while still maintaining a large power output.
A thermodynamic cycle is a series of processes combined in such a way that the thermodynamic states at which working fluid exists are repeated periodically.2 The final state of one cycle is identical to the initial state of the next cycle.
III.A. Rankine Cycle
The Rankine cycle is the steam cycle that will be used for this project. The Rankine cycle can be simplified to four separate processes happening in four different components, more specifically, the pump, the boiler, the turbine, and the condenser. A simple schematic of a Rankine cycle can be seen below in Figure 1.
The first process in the Rankine cycle occurs in the pump. Isentropic compression occurs through the pump, causing work to be done on the pump, i.e., work in, without heat transfer.
The second process in the Rankine cycle occurs in the boiler. With constant pressure and varying temperature, heat is added to the system. Heat is first transferred to the subcooled liquid to raise the liquid to saturation temperature. The second heat transfer is then due to boiling at a constant pressure and saturation temperature until the working fluid is pure vapor.2
The third process in the Rankine cycle occurs in the turbine. Isentropic expansion occurs through the turbine, causing work to be done by the turbine, i.e. work out, without heat transfer.
The fourth process in the Rankine cycle occurs in the condenser. With constant pressure and temperature, heat is rejected in the condenser.
Fig. 1. This is a simple schematic of a Rankine cycle.2
III.A.1. Rankine Cycle Calculations
The equations for the work added by the pump, the heat transferred through the boiler, the work removed by the turbine, and the heat rejected through the condenser can be seen below.
pump |W ̇_P |=m ̇(h_2-h_1 ), s_1=s_2 (3.43)
boiler _2 Q_3=m ̇(h_3-h_2 ), P_2=P_3, h_3=h_g (3.40)
turbine W ̇_T=h_3-h_4, s_3=s_4, h_3=h_g (3.41)
condenser _4 Q ̇_1=m ̇|h_4-h_1 |, P_4=P_1, h_1=h_f,
T_4=T_1, (3.42)
The flow rate of the coolant also needs to be calculated here. The flow rate of the coolant affects the convective heat transfer coefficient. The equation to calculate the mass flow rate of the coolant can be seen below.
heat transferred by coolant=m ̇c_p (T_f2-T_f1 ) (ch.6)
IV. CHECKING EFFICIENCY
The fourth task in designing a nuclear power plant system is to use the steam cycle properties and derived equations from the to calculate the various efficiencies. The overall efficiency, efficiency of the pump, efficiency of the turbine, and actual efficiency can be calculated using the equations below. The estimated thermal efficiency should be around 33%. This is a conservative value.
efficiency η=(net work)/(gross heat added) (3.46)
η=(m ̇(h_3-h_4 )-m ̇|h_2-h_1 |)/(m ̇(h_g-h_2 ) ) (3.49)
η_(s,pump)=(work required by ideal isentropic pump)/(work required by actual pump) (3.50)
η_(s,pump)=(w_s/w_a )_pump, η_(s,pump)≤1 (3.50)
η_(s,turbine)=(actual work output by adiabatic turbine)/(theoretical work output by isentropic turbine) (3.51)
η_(s,turbine)=(w_s/w_a )_turbine, η_(s,turbine)≤1 (3.51)
η_(cyc,a)=(actual work (net))/(gross heat added) (3.52)
η_(cyc,a)≤η_(cyc,s) (3.53)
IV.A. Methods of Increasing Rankine Cycle Efficiency
The efficiency of the Rankine cycle can be increased, according to the theoretical Carnot cycle, by decreasing the temperature at which heat is rejected or by increasing the average temperature at which heat is added.2
IV.A.1. Lowering Heat Rejection Temperature
The net work of a Rankine cycle can be represented by the difference in the areas of heat added and heat rejected.2 If the heat rejection temperature is lowered, the net work area will be increased because of the area increase in the required heat addition. The net result of this process is a net gain in overall cycle efficiency.2
A disadvantage of lowering the heat rejection temperature is that it leads to an increase in the moisture content of the steam leaving the turbine.2 This results in a decrease in the turbine efficiency.
IV.A.2. Superheating
Superheating is one process that is used to increase the average temperature at which heat is added to a Rankine cycle. To do this, the steam is superheated above the saturation condition, which results in a higher steam temperature at the turbine limit.2 This can be done without increasing the maximum pressure in the cycle. The average temperature at which heat is added during the superheating process is higher than the average temperature for the preceding heat addition process, which results in an increase in the efficiency of the cycle.2 Another benefit to superheating is that the quality of the steam at the turbine exit is improved over that resulting from a cycle without superheat.2
IV.A.3. Increasing Maximum Pressure
Increasing the maximum pressure of the cycle is another process that is used to increase the average temperature at which head is added to a Rankine cycle. In general, a 100 psi increase in steam pressure will result in a 0.4% reduction in the amount of heat rejected.2
A disadvantage of using a high pressure cycle is the resulting high moisture content of the turbine exhaust steam. To reduce the moisture content, reheating is frequently used.2
IV.A.4. Reheating and Moisture Separation
In a cycle with reheat, the high moisture content steam is reheated after leaving the high-pressure turbine and before entering the low-pressure turbines.2 Reheating results in a gain in thermal efficiency because the turbines that use the dry, reheated steam operate at a higher efficiency than they would if the steam were not reheated. For each 50oF that the steam is reheated, the heat rejected will decrease by approximately 1.4%.2 A gain in overall efficiency is obtained by increasing the isentropic efficiency of the turbine with reheat.
IV.A.5. Regenerative Feedwater Heating
Feedwater heating is accomplished by utilizing a heat source, other than the fuel, within the cycle. This is known as regenerative heating. Regenerative heating increases the average temperature of the heat addition from external sources by using internal heat sources available in the cycle to preheat the feedwater. This process results in an increase in the thermal efficiency of the cycle. Waste heat rejection to plant cooling water can be reduce by up to 37% by using regenerative heating, depending on the amount and arrangement of feedwater heaters used.2 A T-s diagram of a Rankine cycle with superheat, reheat, and regenerative heating can be seen in Figure 2.
Fig. 2. This is a T-s diagram for a Rankine cycle with superheat, reheat, and regenerative heating.2
IV.B. Carnot Cycle
The Carnot cycle consists of a series of alternate, ideal, reversible, isothermal, and isentropic processes, which form a complete cycle through two specified temperature limits.2 The Carnot cycle can be used to calculate the maximum efficiency between two temperature limits, or vice versa. In this scenario, we will use the Carnot efficiency to calculate the temperature difference necessary to produce the desired power output in an ideal setting. The equation to do this can be seen below.
η_c=(T_H-T_L)/T_H =1-T_L/T_H (3.38)
IV.C. Fractions, Ratios, and Qualities
After checking the efficiencies, there are some ratios and qualities to be evaluated. The void fraction, slip ratio, equilibrium quality, real quality, and static quality can be calculated. The equations for these can be seen below, respectively.
α=V_g/((V_g+V_f ) ) (7.1)
s=u_g/u_f (7.15)
x_e=(h-h_f)/h_fg (7.16)
x=(Ï_g u_g A_g)/((Ï_g u_g A_g+Ï_f u_f A_f ) ) (7.17)
x_s=(Ï_g A_g ∆z)/((Ï_g A_g ∆z+Ï_f A_f ∆z) ) (7.19)
The void fraction represents the time averaged volumetric fraction of vapor in a two-phase control volume.6 It is assumed that the void fraction is a stationary random process, or a time-averaged deterministic quantity.6
The slip ratio is the gas phase velocity divided by the liquid phase velocity. The value of the slip ratio is 1 for homogeneous flow.
The equilibrium quality is typically thought of as the flow fraction of vapor, if thermodynamic equilibrium exists. The value of the equilibrium quality can be positive or negative, depending on the phase of the flow field.
The real quality, or flow quality, is the true flow fraction of vapor, which exists in a flow stream regardless of whether thermodynamic equilibrium exists or not.6 The flow quality is always between 0 and 1. If thermodynamic equilibrium exists, the equilibrium quality and flow quality are equal.
The static quality is defined as the mass fraction of vapor at a particular cross-section with unit thickness.6 The static quality is typically between 0 and 1, but it is not required.
V. REACTOR OUTPUT
The fifth task in designing a nuclear power plant system is to estimate the reactor output in MWt. This is done by using the desired electrical output and cycle efficiency. Since there is no theoretical limit to the flux generated in the core, there is no theoretical limit to the heat generated. However, the heat generated must be removed of the core will be destroyed.3 The reactor must be operated at such a power level that, with the available heat removal system, the temperature and heat flux anywhere in the core do not exceed specific safe limits. Therefore, the maximum power generation in a reactor core is limited by thermal considerations, rather than nuclear.3
The thermal efficiency of the system can be used to estimate the minimum reactor thermal power. The equation to do this can be seen below.
〖MW〗_t=(800 〖MW〗_e)/η (ch.11)
After the minimum reactor thermal power is calculated, an additional margin can be added for possible power uprates.7
VI. FUEL LATTICE
The sixth task in designing a nuclear power plant system is to design a basic fuel lattice, determined from nuclear, hydraulic, and heat-transfer considerations. These include, but are not limited to, fuel type, enrichment, cladding, fuel gap, fuel-rod diameter and pitch, fuel-element diameter, moderator-to-fuel ratio, cycle length, number of rods, number of assemblies, spacer grids, geometry, etc.7
VI.A. Fuel Type
The fuel type chosen for a nuclear power plant system affects the fissionable fuel density, the volumetric heat generation rate, the flux depression factor, and the conductivity integral.
The volumetric heat flux is dependent on the fuel type through the fissionable fuel density. The equations for the volumetric heat flux, fissionable fuel density, density of the fissionable fuel, and mass fraction of fuel in the fuel material can be seen below, in respective order.
q^”’=1.5477×〖10〗^(-8) G_f N_f σ_f Ï• [BTU/(hr∙〖ft〗^3 )] (4.8)
N_ff=Av/M_ff Ï_ff i (4.12)
Ï_ff=rÏ_f=rfÏ_fm (4.13)
f=(rM_ff+(1-r) M_Nf)/(rM_ff+(1-r) M_Nf+M_(O_2 ) ) (4.14)
VI.B. Enrichment
The fuel enrichment chosen for a nuclear power plant system also affects the fissionable fuel density and, thus, affects the volumetric heat generation. To have a design that reaches the required energy specified by the utility, the enrichment has to be sufficient enough to minimize or overcome any parasitic neutron capture.3
Fuel enrichment is limited by law. An enrichment of over 5% for a commercial reactor is prohibited. Therefore, a fuel enrichment of 4.95% of lower is typically used, to account for a 0.05% error.
VI.C. Cladding
One of the most important limiting factors in fuel element duty is the mechanical interaction of fuel and cladding. This fuel-cladding interaction produces cyclic stresses and strains in the cladding, and these in turn consume cladding fatigue life.4 Therefore, the reduction of fuel-cladding interaction is a principal goal of design. The reduction of fuel-cladding interaction can be achieved by increasing the gap between the fuel and the cladding, which will be explained in more detail in the proceeding section. Additionally, the equation for the temperature drop through the cladding can be seen below.
T_s-T_c=(q_s lnâ¡((R+c)/R))/(2Ï€Lk_c ) (5.31)
VI.D. Fuel Gap
The gap resistance between the fuel and the cladding is a function of the fuel and clad hot dimensions, the amount of burnup and fission gas release, the type, amount, and pressure of the prefill gas, the rod surface temperature, the clad material type and hardness, the fuel pellet and clad surface roughness, and the pellet cracking effects.4 Gap conductance is a result of the conduction across the gas film, the conduction between the fuel and clad at the points of contact, and the radiation from the pellet to the clad.
The gap heat transfer coefficient between the fuel and clad changes the slope of the fuel temperature. As the heat transfer of the fuel increases, the average fuel temperature increases, and the fuel thermally expands and makes contact with the cladding. The contact between the fuel and cladding increases the gap heat transfer coefficient and reduces the fuel surface temperature.4
Another consideration is the temperature drop through the fuel. The fuel pellet, clad, and coolant surfaces must all experience the same heat flow. With this information, the equations for the temperature drops through the fuel and the coolant can be seen below, respectively.
T_m-T_s=q_s/(4Ï€k_f L) (5.32)
T_c-T_f=q_s/h2Ï€(R+c)L (5.31)
VI.E. Fuel-Rod Diameter and Pitch
The fuel-rod diameter and pitch can be adjusted to obtain longer fuel cycles. The U-235-to-H ratio can be changed by changing the fuel-rod diameter or the fuel-rod pitch. If the pitch is increased, or the fuel diameter is decreased, the amount of moderator per rod increases. If the amount of moderator per rod increases, the effective cross-sections increase, and the fuel cycle can be lengthened.3
As the pitch-to-diameter ratio decreases, the circumferential variations in rod surface temperature and heat transfer coefficients increase. The fuel-rod pitch-to-diameter ratios are important when using various correlations. These correlations and their respective pitch-to-diameter ratios can be seen below.
Dittus-Boelter
Nu=0.023〖Re〗^0.8 〖Pr〗^0.4, P/D≤1.3 (6.12)
Markoczy
(〖Nu〗_∞ )_RB=ψ(〖Nu〗_∞ )_ct (6.15)
ψ=1+0.9120〖Re〗^(-0.1) 〖Pr〗^0.4 (1-2.0043e^(-B) ) (6.16)
for triangle B=(2√3)/π (P/D)^2-1 (6.19)
for 1≤P/D≤2.0
for square B=4/Ï€ (P/D)^2-1 (6.20)
for 1≤P/D≤1.8
Weisman
(〖Nu〗_∞ )_RB=ψ(〖Nu〗_∞ )_ct (6.15)
for triangle ψ=1.130(P/D)-0.2609 (6.21)
for 1.1≤P/D≤1.5
for square ψ=1.826(P/D)-1.043 (6.22)
for 1.1≤P/D≤1.3
El-Genk
forced laminar 〖Nu〗_fl=A〖Re〗^B 〖Pr〗^0.33 (6.23)
forced turbulent 〖Nu〗_ft=C〖Re〗^0.8 〖Pr〗^0.33 (6.24)
A=2.97-1.76(P/D) (6.25)
B=0.56(P/D)-0.3 (6.26)
C=0.028(P/D)-0.006 (6.27)
These correlations are used to calculate the Nusselt number, which can then be used to calculate the heat transfer coefficient. For many of these correlations, the Reynolds number and the Prandtl number need to be calculated. The equations for these dimensionless numbers, along with some of their parameters, can be seen below, respectively.
Re=(D_e VÏ)/μ (6.6)
Pr=ν/α (6.7)
ν=kinematic viscosity of fluid=μ/Ï (6.8)
α=thermal diffusivity of fluid=k/(Ïc_p ) (6.9)
D_(e,PWR)=4[P_assem^2-N_rods (πD_rods^2)/4-N_thimb (πD_thimb^2)/4]/(N_rods πD_rod+N_thimb πD_thimb ) (6.10)
VI.F. Moderator-to-Fuel Ratio
The moderator-to-fuel ratio is the ratio of moderator nuclei within the reactor to the number of fuel nuclei within the reactor.8 As the core temperature increases, the fuel volume, the fuel number density, and the moderator volume remain almost constant, but the moderator density number decreases due to thermal expansion.8
The moderator-to-fuel density affects the resonance escape probability and the thermal utilization factor, which then affects the infinite multiplication factor. At optimal value of moderator-to-fuel ratio, the infinite multiplication factor reaches its maximum value.8
VI.G. Number of Rods and Assemblies
The number of fuel assemblies and number of rods per assembly affects the core average linear power of the reactor. The larger the number of rods and assemblies, the smaller the core average linear power.
VI.H. Spacer Grids
Pressurized water reactors use spacer grids to maintain fuel rod spacing. The spacers can contribute up to 50% of the pressure drop in the core.5 The larger the number of spacer grids, the greater the pressure drop that will be contributed. A general method of predicting spacer grid pressure drop is given by Rehme and can be seen below.
∆P_grid=C_v ε^2 (ÏU_b^2)/(2g_c ) (6.46)
Spacer grids act to enhance the convective heat transfer by disrupting the momentum and thermal boundary layers on the fuel rod surface.5 Spacer grids also increase the turbulence in the flow, such that the convective heat transfer is larger.5
VII. CRITICAL HEAT FLUX
The seventh task in designing a nuclear power plant system is to estimate the critical heat flux, otherwise known as the burnout condition, for the specific assembly designed. When the wall superheat is raised, the flux reaches a maximum and a further increase of the wall temperature causes a decrease in the rate of heat flow. This point is known as the departure from nucleate boiling (DNB) or the critical heat flux (CHF).
Some of the parameters that affect the critical heat flux behavior are the flow memory affect, which is a function of quality, spacer grids, and surface roughness.6 In determining the operational limits for a PWR, a departure from nucleate boiling ratio of 1.3 is used to ensure that no critical heat flux will occur.6 The ratio for this can be seen below.
〖q”〗_(〖CHF〗_P )/q”(z) =1.3 (7.37)
VIII. PEAKING FACTORS
The eighth task in designing a nuclear power plant system is to choose an overall desired nuclear peaking factor. The utility usually prefer a large nuclear peaking factor to be mindful of licensing and safety limits. A very high peaking factor could lead to poor fuel utilization.7
Nuclear peaking factors represent the radial, axial, and local power distributions within the reactor core. The power distribution and peaking factors vary due to control rod locations and movement, plant operations, fuel assembly design, assembly burnup, assembly location within the core, and enrichment.3 The equations for various peaking factors can be seen below.
radial peaking factor F_R=F_xy=(peak radial power)/(average radial power) (ch.4)
axial peaking factor F_Z=(peak axial power)/(average axial power) (ch.4)
local peaking factor F_W=(peak flux in fuel element)/(average flux in fuel element) (ch.4)
nuclear peaking factors
F_R^N=(mean heat flux in hot channel)/(mean heat flux in average channel) (ch.4)
F_Z^N=maxâ¡ã€–heat flux in hot channel〗/(mean heat flux in hot channel) (ch.4)
F_Q^N=F_Z^N∙F_R^N=maxâ¡ã€–heat flux〗/(mean heat flux) (ch.4)
The average linear heat rate can be calculated using the plant power level, number of assemblies, number of fuel rods per assembly, and the length of the fuel rod. With the average linear heat rate, the peak linear heat rate can also be calculated. The equation for the average linear heat rate and peak linear heat rate can be seen below, respectively.
q Ì…^’=(core power(〖MW〗_t )(1000 kW/MW)(0.974))/(#of assem)(#fuel rods/assem)(ft/fuel rod) (ch.4)
VIII.A. Average Heat Flux
The average heat flux can be calculated using maximum design peaking factor, considering nuclear and engineering hot channel factors.7 This is done using estimates and the equations for this can be seen below.
use F_(Q,total,max)=2.86 (ch.11)
〖q”〗_avg=〖q”〗_(CHF,min)/F_(Q,total) (ch.11)
VIII.B. Heat Transfer Area
After the average heat flux is estimated, the necessary heat transfer area can be calculated. The equation to calculate this can be seen below.
A_H=Q_t/(q”) Ì… (ch.11)
VIII.C. Total Fuel Length and Mass
The total fuel length and fuel mass can be calculated from the specified fuel rod outside cladding diameter. The equation to calculate the total fuel length and total fuel mass can be seen below, respectively.
L=1/(Ï€D_0 ) A_H (ch.11)
M_f=(D_p^2 Ï)/(4D_0 ) A_H (ch.11)
IX. CORE SHAPE AND SIZE
The ninth task in designing a nuclear power plant system is to determine the overall core shape. The core shape will be an upright cylinder of a certain height-to-diameter ratio, in proportion with pressure drops, reactor pressure vessel diameter, etc.7 The objective here is to optimize the flow-heat transfer performance of the fuel.
Once the basic core lattice has been determined from nuclear, hydraulic, and heat-transfer considerations, the overall size of the core for a particular power output usually becomes a sole function of the thermal and mechanical design considerations, not nuclear.7
X. ADDITIONAL CONSIDERATIONS
Some other things to consider when designing a nuclear power plant system are listed below.
The design is an iterative process between nuclear, thermal-hydraulic, materials, structural, and economic concerns.7
The core behavior for anticipated transients, as well as over the fuel lifetime, need to be checked.
The consideration of low probability accidents, such as Loss of Coolant Accidents, Control Rod Ejection, Pump Locked Rotor, needs to be evaluated to determine if the safety systems are sufficient to diminish the transient at the allowable peaking factors.7
X. CONCLUSIONS
Overall, there are many things to consider when designing a nuclear power plant system. In short, desired power outputs and efficiencies are decided, which lead to various other parameters being calculated. All of these parameters are combined to make sure that the peaking factor and safety limit requirements are met. If all of the requirements and limitations are met, the reactor will be able to operate safely.
References
1. “Nuclear Power Reactors.†Elements of Nuclear Reactor Design, by L.E. Hochreiter, S. Ergun, G.E. Robinson, 2009, pp. 2.1-2.231.
2. “Thermodynamic Cycle for Power Plants.†Elements of Nuclear Reactor Design, by L.E. Hochreiter, S. Ergun, G.E. Robinson, 2009, pp. 3.1-3.62.
3. “Heat Generation in the Reactor Core and Structures.†Elements of Nuclear Reactor Design, by L.E. Hochreiter, S. Ergun, G.E. Robinson, 2009, pp. 4.1-4.44.
4. “Temperature Distribution in Fuel Rods, Plate Elements, and Structures.†Elements of Nuclear Reactor Design, by L.E. Hochreiter, S. Ergun, G.E. Robinson, 2009, pp. 5.1-5.73.
5. “Convective Heat Transfer and Single Phase Pressure Drop in the Reactor Core.†Elements of Nuclear Reactor Design, by L.E. Hochreiter, S. Ergun, G.E. Robinson, 2009, pp. 6.1-6.53.
6. “Two-Phase Flow Regimes, Boiling Heat Transfer, Critical Heat Flux, and Two-Phase Pressure Drop.†Elements of Nuclear Reactor Design, by L.E. Hochreiter, S. Ergun, G.E. Robinson, 2009, pp. 7.1-7.86.
7. “Preliminary LWR Core Thermal Design.†Elements of Nuclear Reactor Design, by L.E. Hochreiter, S. Ergun, G.E. Robinson, 2009, pp. 11.1-11.12.
8. “Moderator-to-Fuel Ratio.†Nuclear Power, www.nuclear-power/reactor-physics/reactor-dynamics/moderator-to-fuel-ratio/.