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Thermal Hydraulics Flashcards

1
Q

Which one of the following is an example of significant radiative heat transfer?

A

Heat transfer from the fuel cladding to the reactor coolant through a stable vapor layer

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2
Q

Refer to the drawing of a pool boiling curve (see figure below). In which region of the curve does the
most efficient form of heat transfer occur?

A

Region II

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3
Q

Refer to the drawing of a pool boiling curve (see figure below).
Which region of the curve contains the operating point at which the hottest locations of a reactor
normally operate to transfer heat from the fuel cladding to the coolant at 100 percent power?

A

Region II

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4
Q

Why does nucleate boiling improve heat transfer in a reactor core?

A

Heat is removed from the fuel rod as both sensible heat and latent heat of vaporization, and the
motion of the steam bubbles causes rapid mixing of the coolant

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5
Q

Convection heat transfer improves when nucleate boiling begins on the surface of a fuel rod because:

A

the motion of the steam bubbles causes rapid mixing of the coolant.

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6
Q

How does the convective heat transfer coefficient vary from the bottom to the top of a fuel assembly if
reactor coolant enters the fuel assembly as subcooled water and exits as superheated steam?

A

Increases, then decreases.

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7
Q

Nucleate boiling affects heat transfer from a fuel rod primarily by…

A

improving the convective heat transfer from the fuel rod to the coolant.

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8
Q

Subcooled water enters the bottom of an operating reactor core. As the water flows upward past the
fuel assemblies, steam bubbles form on the surface of a few fuel rods and are swept away.
If the coolant at the surface of the affected fuel rods had remained subcooled, average fuel temperature
in the affected fuel rods would have been __________ because single-phase convection is a
__________ efficient method of heat transfer than boiling.

A

higher; less

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9
Q

Case 1: Subcooled reactor coolant enters the bottom of a fuel assembly in a reactor operating at
power. As the coolant flows upward through the fuel assembly, the water heats up and exits the fuel
assembly still subcooled.
Case 2: Same as above except that reactor pressure is decreased such that the coolant begins to boil
halfway up the fuel assembly, which results in a saturated steam-water mixture exiting the fuel
assembly.
Assume that departure from nucleate boiling is avoided in both cases and that power level does not
change. As compared to Case 1, the average fuel temperature for Case 2 will be __________ because
boiling is a __________ efficient method of heat transfer.

A

lower; more

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10
Q

Subcooled reactor coolant enters the bottom of a fuel assembly and exits the top of the fuel assembly
as a saturated steam-water mixture. How does the convective heat transfer coefficient change as the
coolant travels upward through the fuel assembly?

A

Increases only

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11
Q

Subcooled water enters a fuel assembly in a reactor operating at power. As the water flows upward
through the fuel assembly, the water begins to boil and exits the fuel assembly as a saturated
steam-water mixture.
If fuel assembly power is unchanged and system pressure is increased such that all of the water
remains subcooled, the average fuel temperature in the fuel assembly would be __________ because
boiling is a __________ efficient method of heat transfer

A

higher; more

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12
Q

Initially, subcooled water is flowing into a fuel assembly with subcooled water exiting the fuel
assembly several degrees hotter than when it entered. No boiling is occurring in the fuel assembly.
Assume that fuel assembly thermal power and water flow rate remain the same.
System pressure is decreased, causing some of the water in contact with the fuel rods to boil during
transit through the fuel assembly, but the water exiting the fuel assembly remains subcooled.
Compared to the initial conditions, the average fuel temperature in the fuel assembly will be
__________; and the temperature of the water exiting the fuel assembly will be __________.

A

lower; higher

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13
Q

Subcooled nucleate boiling is occurring along a heated surface. If the heat flux is increased slightly,
what will be the effect on the differential temperature (ΔT) between the heated surface and the fluid?
(Assume subcooled nucleate boiling is still occurring.)

A

Small increase in ΔT as vapor bubbles form and collapse.

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14
Q

Which one of the following characteristics will enhance steam bubble formation in water adjacent to a
heated surface?

A

The presence of gases dissolved in the water.

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15
Q

What type of boiling is described as follows?
The bulk temperature of the liquid is below saturation, but the temperature of the heat transfer surface
is above saturation. Vapor bubbles form at the heat transfer surface, but condense in the bulk liquid
so that no net generation of vapor is obtained.

A

Subcooled nucleate boiling

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16
Q

Which one of the following is a characteristic of subcooled nucleate boiling but not saturated nucleate
boiling?

A

TBulk Coolant is less than TSat

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17
Q

Which one of the following is a characteristic of saturated nucleate boiling but not subcooled nucleate
boiling?

A

TBulk Coolant equals TSat

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18
Q

Which one of the following describes a reason for the increased heat transfer rate that occurs when
nucleate boiling begins on the surface of a fuel rod?

A

The motion of the steam bubbles causes rapid mixing of the coolant.

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19
Q

Which one of the following modes of heat transfer is characterized by steam bubbles moving away
from a heated surface and collapsing in the bulk fluid?

A

Subcooled nucleate boiling

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20
Q

Which one of the following characteristics will enhance steam bubble formation in the coolant
adjacent to a fuel rod?

A

Surface scratches or cavities in the fuel cladding.

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21
Q

A nuclear power plant is currently shut down after several months of operation at 100 percent power.
The shutdown cooling system is in operation, maintaining an average reactor coolant temperature of
280°F. A pressure control malfunction causes reactor coolant pressure to slowly and continuously
decrease from 100 psia while reactor coolant temperature remains constant.
Which one of the following describes the location where nucleate boiling will first occur?

A

At a scratch on the surface of a fuel rod near the top of a fuel assembly.

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22
Q

If departure from nucleate boiling occurs on the surface of a fuel rod, the surface temperature of the
fuel rod will…

A

increase rapidly.

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23
Q

Which one of the following describes the heat transfer from a fuel rod experiencing departure from
nucleate boiling? (Note: ΔT refers to the difference between the fuel rod surface temperature and
the bulk coolant saturation temperature.)

A

Steam bubbles begin to blanket the fuel rod surface, causing a rapid increase in the ΔT for a given
heat flux

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24
Q

Departure from nucleate boiling should not be allowed to occur in the core because…

A

as steam bubbles begin to blanket the fuel rod, its temperature rises sharply.

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25
Q

Which one of the following is indicated by a rapid increase in the temperature difference between the
fuel cladding and the bulk coolant?

A

Departure from nucleate boiling is occurring

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26
Q

Which one of the following reactor coolant system parameters has the least effect on margin to
departure from nucleate boiling?

A

Pressurizer level

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27
Q

An adequate subcooling margin during a loss of coolant accident is the most direct indication that
__________ is being maintained.

A

core cooling

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28
Q

Which one of the following parameter changes will reduce the departure from nucleate boiling ratio?

A

Increasing reactor coolant temperature.

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29
Q

Which one of the following will increase the departure from nucleate boiling ratio?

A

Increasing pressurizer pressure.

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30
Q

A nuclear power plant is operating with the following initial conditions:
• Reactor power is 45 percent in the middle of a fuel cycle.
• Axial and radial power distributions are peaked in the center of the core.
Assuming reactor power level does not change, which one of the following will increase the
steady-state departure from nucleate boiling ratio?

A

Core xenon-135 builds up in proportion to the axial and radial power distribution with automatic
rod control.

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31
Q

A nuclear power plant is operating with the following initial steady-state conditions:
• Reactor power is 45 percent in the middle of a fuel cycle.
• Axial and radial power distributions are peaked in the center of the core.
Which one of the following will decrease the steady-state departure from nucleate boiling ratio?

A

The operator decreases reactor coolant boron concentration by 5 ppm with no control rod motion.

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32
Q

A nuclear power plant is operating with the following initial conditions:
• Reactor power is 55 percent in the middle of a fuel cycle.
• Axial and radial power distributions are peaked in the center of the core.
Which one of the following will decrease the steady-state departure from nucleate boiling ratio?

A

Core xenon-135 depletes in proportion to the axial and radial power distribution with no control
rod motion

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33
Q

A nuclear power plant is operating with the following initial conditions:
• Reactor power is 45 percent in the middle of a fuel cycle.
• Axial and radial power distributions are peaked in the center of the core.
Which one of the following will decrease the steady-state departure from nucleate boiling ratio?

A

A pressurizer malfunction decreases reactor coolant system pressure by 20 psig

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34
Q

A reactor is shutdown with all control rods inserted. The reactor coolant system (RCS) is at normal
operating temperature and pressure. Which one of the following will decrease the departure from
nucleate boiling ratio for the reactor? (Assume the reactor remains shutdown.)

A

Reducing RCS flow rate by 3 percent.

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35
Q

A nuclear power plant is operating with the following conditions:
• Reactor power is 55 percent in the middle of a fuel cycle.
• Axial and radial power distributions are peaked in the center of the core.
Which one of the following will increase the steady-state departure from nucleate boiling ratio?

A

A reactor trip occurs and one control rod remains fully withdrawn from the core.

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36
Q

A nuclear power plant is operating with the following initial conditions:
• Reactor power is 45 percent in the middle of a fuel cycle.
• Axial and radial power distributions are peaked in the center of the core.
Which one of the following will increase the steady-state departure from nucleate boiling ratio?

A

A reactor trip occurs and one control rod remains fully withdrawn from the core.

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37
Q

A reactor is shut down at normal operating temperature and pressure with all control rods inserted.
Which one of the following will decrease the departure from nucleate boiling ratio for this reactor?
(Assume the reactor remains shutdown.)

A

Decreasing reactor coolant pressure by 10 psig.

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38
Q

Which one of the following parameter changes would move a reactor farther away from the critical
heat flux?

A

Decrease reactor power.

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39
Q

How does the critical heat flux vary from the bottom to the top of a typical fuel assembly during
normal 100 percent power operation?

A

Decreases continuously.

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40
Q

The heat flux that causes departure from nucleate boiling is the…

A

critical heat flux.

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41
Q

The critical heat flux is the heat transfer rate per unit __________ of fuel rod that will initially cause
__________.

A

area; departure from nucleate boiling

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42
Q

How does critical heat flux (CHF) vary with core height during normal full power operation?

A

CHF decreases from the bottom to the top of the core.

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43
Q

A reactor is operating at steady-state 75 percent power. Which one of the following parameter
changes will cause the core to operate closer to the critical heat flux? (Assume reactor power does not
change unless stated.)

A

Decrease reactor coolant flow rate by 5 percent.

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44
Q

Which one of the following is most likely to result in fuel cladding damage?

A

Operating at a power level that exceeds the critical heat flux.

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45
Q

A small increase in differential temperature at the fuel cladding-to-coolant interface causes increased
steam blanketing and a reduction in heat flux. This describes which type of boiling?

A

Partial film boiling

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46
Q

Refer to the drawing of a pool boiling curve (see figure below).
Choose the region of the curve where transition boiling is the primary heat transfer process.

A

Region III

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47
Q

Refer to the drawing of a pool boiling curve (see figure below).
Which one of the points shown marks the onset of transition boiling?

A

B

Region 2

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48
Q

Which one of the following describes the heat transfer conditions in a fuel assembly that is
experiencing transition boiling?

A

Alternate wetting and drying of the fuel rod surface

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49
Q

Which one of the following describes the conditions in a fuel assembly that is experiencing transition
boiling?

A

Alternate wetting and drying of the fuel rod surface.

50
Q

Refer to the drawing of a pool boiling curve (see figure below).
Which one of the following describes the heat transfer conditions in a fuel assembly that is
experiencing region III heat transfer?

A

Alternate wetting and drying of the fuel rod surface.

51
Q

Refer to the drawing of a pool-boiling curve (see figure below).
With heat flux continuously increasing, the point at which the critical heat flux is reached (point B),
marks the beginning of…

A

partial film boiling

52
Q

Refer to the drawing of a pool boiling curve (see figure below).
Which one of the following regions represents the most unstable mode of heat transfer?

A

Region III

53
Q

Film boiling heat transfer is…

A

heat transfer through a vapor blanket that covers the fuel cladding

54
Q

Reactor power is increased sufficiently to cause steam blanketing of several fuel rods. This condition
is being caused by…

A

departure from nucleate boiling.

55
Q

If the fission rate in a reactor core steadily increases, the mode of heat transfer that occurs immediately
after the critical heat flux is reached is called…

A

transition boiling.

56
Q

Refer to the drawing of a pool boiling curve (see figure below).
Which one of the points shown marks the smallest ΔT at which stable film boiling can exist?

A

Region 3

57
Q

Refer to the drawing of a pool boiling curve (see figure below).
Which one of the following describes the conditions in a fuel assembly that is experiencing region IV
heat transfer?

A

Complete steam blanketing of the fuel rod surface.

58
Q

During a loss of coolant accident, some fuel rods may experience stable film boiling. Which one of
the following types of heat transfer from the fuel cladding will increase significantly when stable film
boiling begins?

A

Radiation

59
Q

The departure from nucleate boiling (DNB) ratio is defined as the…

A

critical heat flux divided by the actual heat flux.

60
Q

In the definition of departure from nucleate boiling ratio, the term “actual heat flux” refers to the…

A

heat transfer rate per unit area at any point along the fuel rod.

61
Q

A reactor is operating at steady-state 100 percent power near the end of a fuel cycle with all control
rods fully withdrawn. At what axial location in a typical fuel assembly will the maximum departure
from nucleate boiling ratio occur?

A

At the bottom of the fuel assembly.

62
Q

If a reactor is operating with the departure from nucleate boiling ratio at its limit, which one of the
following is indicated?

A

A small fraction of the fuel rods may be experiencing critical heat flux

63
Q

Core heat transfer rate is maximized by the presence of…

A

turbulent flow with nucleate boiling.

64
Q

The heat transfer coefficient for the core will be directly increased if: (Assume bulk coolant
subcooling.)

A

nucleate boiling occurs in the coolant.

65
Q

Increasing the coolant flow rate through a reactor core affects the heat transfer rate from the fuel,
because a higher coolant flow rate results in a __________ laminar film thickness and a __________
coolant temperature adjacent to the fuel.

A

smaller; lower

66
Q

Which one of the following will minimize core heat transfer?

A

Laminar flow with no nucleate boiling.

67
Q

A nuclear power plant is operating at 100 percent power. The reactor coolant subcooling margin will
be directly reduced by…

A

increasing reactor coolant temperature.

68
Q

The difference between the actual temperature and the saturation temperature of a liquid is the…

A

subcooling margin.

69
Q

Which one of the following must be present to assure adequate core cooling following a small loss of
coolant accident?

A

Subcooling margin greater than zero.

70
Q

Which one of the following will increase the reactor coolant system (RCS) subcooling margin with the
reactor operating at full power?

A

Decreased RCS hot leg temperature.

71
Q

A 60°F/hour reactor coolant system (RCS) cooldown and depressurization with natural circulation is
in progress. After one hour, the RCS subcooling margin will be minimum in the…

A

reactor vessel head

72
Q

A reactor coolant system (RCS) cooldown and depressurization is in progress on natural circulation
following a loss of offsite power. The following conditions currently exist:
RCS Tcold = 520°F, decreasing
RCS Thot = 538°F, decreasing
Pressurizer pressure = 2,000 psia, decreasing
If the cooldown rate is being maintained at 50°F/hr, which one of the following locations is most likely
to experience sustained steam voiding?

A

Reactor vessel head

73
Q

Which one of the following is most likely to result in steam bubble formation in the reactor vessel head
while maintaining a 60°F subcooling margin in the hottest reactor coolant system (RCS) hot leg?

A

Performing a 50°F/hr RCS cooldown with natural circulation.

74
Q

Which one of the following is most likely to result in steam bubble formation in a reactor vessel head
while maintaining a 40°F subcooling margin in the hottest RCS hot leg?

A

Performing a 50°F/hr RCS cooldown with natural circulation.

75
Q

A nuclear power plant maintains the reactor coolant system (RCS) cold leg temperature (Tcold) at
557°F from 0 percent to 100 percent power. At 100 percent power, the reactor differential
temperature (Thot - Tcold) is 60°F.
If this plant also maintains RCS pressure constant at 2,235 psig, which one of the following is the
approximate RCS subcooling margin at 50 percent power?

A

66°F

76
Q

Assume that a 30°F subcooling margin is maintained in the reactor coolant system (RCS) hot legs
during each of the following cooldown operations for a shutdown reactor. Which one of the
following will maintain the greatest subcooling margin in the reactor vessel head?

A

Performing a 25°F/hr RCS cooldown with all reactor coolant pumps running

77
Q

Refer to the drawing of a fuel rod and adjacent coolant flow channel (see figure below).
With a nuclear power plant operating at steady-state 100 percent reactor power at the beginning of a
fuel cycle, which one of the following has the greater temperature difference?

A

Fuel pellet centerline-to-pellet surface

78
Q

During a plant cooldown and depressurization with forced circulation, reactor coolant system (RCS)
loop flow indications and reactor coolant pump (RCP) motor current indications become erratic.
These abnormal indications are most likely caused by…

A

RCP cavitation

79
Q

Single-phase coolant flow resistance in a reactor core is directly proportional to the square of coolant
__________; and inversely proportional to __________

A

velocity; coolant channel cross-sectional area

80
Q

Refer to the drawing of a section of pipe that contains flowing subcooled water (see figure below).
Given:
• Pressure at P1 is 24 psig.
• Pressure at P2 is 16 psig.
• Pressure change due to change in velocity is 2 psig.
• Pressure change due to change in elevation is 10 psig.
The pressure decrease due to friction head loss between P1 and P2 is __________; and the direction of
flow is from __________.

A

4 psig; right to left

81
Q

Refer to the drawing of a section of pipe that contains flowing subcooled water (see figure below).
Given:
• Pressure at P1 is 26 psig.
• Pressure at P2 is 34 psig.
• Pressure change due to change in velocity is 2 psig.
• Pressure change due to change in elevation is 8 psig.
The pressure decrease due to friction head loss between P1 and P2 is __________; and the direction of
flow is from __________.

A

2 psig; left to right

82
Q

Refer to the drawing of a section of pipe that contains flowing subcooled water. (See figure below).
Given:
• Pressure at P1 is 30 psig.
• Pressure at P2 is 32 psig.
• Pressure change due to change in velocity is 2 psig.
• Pressure change due to change in elevation is 2 psig.
The pressure decrease due to friction head loss between P1 and P2 is __________; and the direction of
flow is from __________.

A

2 psig; right to left

83
Q

Refer to the drawing of a section of pipe that contains flowing subcooled water (see figure below).
Given:
• Pressure at P1 is 34 psig.
• Pressure at P2 is 20 psig.
• Pressure change due to change in velocity is 2 psig.
• Pressure change due to change in elevation is 8 psig.
The pressure decrease due to friction head loss between P1 and P2 is __________; and the direction of
flow is from __________.

A

4 psig; right to left

84
Q

Refer to the drawing of a section of pipe that contains flowing subcooled water (see figure below).
Given:
• The pressure at P1 is 20 psig.
• The pressure at P2 is 20 psig.
• The pressure change caused by the change in velocity is 2 psig.
• The pressure change caused by the change in elevation is 8 psig.
The pressure decrease due to friction head loss between P1 and P2 is __________; and the direction of
flow is from __________.

A

6 psig; right to left

85
Q

A reactor is producing 3,400 MW of thermal output with a reactor vessel differential temperature (ΔT)
of 60°F and a reactor vessel mass flow rate of 1.4 x 108 lbm/hr. If core ΔT is 63.6°F, what is core
bypass mass flow rate? (Assume bypass flow ΔT equals 0°F.)

A

7.92 x 106 lbm/hr

86
Q

A reactor is producing 3,400 MW of thermal output with a reactor vessel differential temperature (ΔT)
of 60°F and a reactor vessel mass flow rate of 1.0 x 108 lbm/hr. If core ΔT is 63.6°F, what is core
bypass mass flow rate? (Assume bypass flow ΔT equals 0°F.)

A

5.66 x 106 lbm/hr

87
Q

A reactor is producing 3,400 MW of thermal output with a reactor vessel differential temperature (ΔT)
of 60°F and a reactor vessel mass flow rate of 1.1 x 108 lbm/hr. If core ΔT is 63.6°F, what is core
bypass mass flow rate? (Assume bypass flow ΔT equals 0°F.)

A

6.23 x 106 lbm/hr

88
Q

Adequate core bypass flow is needed to…

A

equalize the temperatures between the reactor vessel and the reactor vessel upper head

89
Q

Which one of the following describes a function of core bypass flow?

A

Provides cooling to various reactor vessel internal components.

90
Q

Which one of the following is a function of core bypass flow?

A

Provides mixing of coolant in the reactor vessel head.

91
Q

Maximizing the elevation difference between the core thermal center and the steam generator thermal
center and minimizing flow restrictions in the reactor coolant system (RCS) piping are features of
nuclear power plant designs that…

A

ensure RCS natural circulation flow can be established.

92
Q

Which one of the following must exist for natural circulation flow to occur?

A

The heat sink must be located higher than the heat source.

93
Q

The driving head for natural circulation flow through the core is developed by differences in
__________ between the hot leg and the cold leg

A

water density

94
Q

If the steam generator thermal centers were at the same elevation as the reactor core thermal center,
natural circulation flow in the reactor coolant system would…

A

not occur.

95
Q

A reactor is shut down with natural circulation core cooling. Decay heat generation is equivalent to
1.0 percent of rated thermal power. Stable natural circulation mass flow rate is 1,000 gpm.
When decay heat generation decreases to 0.5 percent of rated thermal power, stable natural circulation
flow rate will be approximately…

A

794 gpm

96
Q

A reactor is shut down with natural circulation core cooling. Decay heat generation is equivalent to
1.0 percent of rated thermal power. Core differential temperature (ΔT) has stabilized at 16°F.
When decay heat generation decreases to 0.5 percent of rated thermal power, core ΔT will be
approximately…

A

10°F

97
Q

Sustained natural circulation requires that the heat sink is __________ in elevation than the heat
source and that there is a __________ difference between the heat sink and the heat source.

A

higher; temperature

98
Q

Which one of the following conditions must occur to sustain natural convection in a fluid system?

A

A density change in the fluid.

99
Q

A reactor is shut down with natural circulation core cooling. Decay heat generation is equivalent to
1.0 percent of rated thermal power. Core differential temperature (ΔT) has stabilized at 16°F.
When decay heat generation decreases to 0.333 percent of rated thermal power, core ΔT will be
approximately…

A

8°F.

100
Q

A reactor is shut down with natural circulation core cooling. Decay heat generation is equivalent to
1.0 percent of rated thermal power. Core differential temperature (ΔT) has stabilized at 13°F.
When decay heat generation decreases to 0.5 percent of rated thermal power, core ΔT will be
approximately…

A

8°F.

101
Q

A reactor is shut down with natural circulation core cooling. Decay heat generation is equivalent to
1.0 percent of rated thermal power. Stable natural circulation flow rate is 800 gpm.
When decay heat generation decreases to 0.5 percent of rated thermal power, stable natural circulation
flow rate will be approximately…

A

635 gpm

102
Q

Sustained natural circulation requires that the heat source is __________ in elevation than the heat
sink; and that there is a __________ difference between the heat source and the heat sink.

A

lower; temperature

103
Q

A nuclear power plant was operating at steady-state 100 percent power when a loss of offsite power
occurred, resulting in a reactor trip and a loss of forced reactor coolant circulation. Thirty minutes
later, reactor coolant system (RCS) hot leg temperature is greater than cold leg temperature and steam
generator (SG) levels are stable.
Which one of the following combinations of parameter trends, observed 30 minutes after the trip,
indicates that natural circulation is occurring? (CET = core exit thermocouple)

A

RCS Hot Leg RCS Cold Leg SG RCSCET
Temperature Temperature Pressures Subcooling
Decreasing Stable Stable Increasing

104
Q

A nuclear power plant was operating at steady-state 100 percent power when a loss of offsite power
occurred, resulting in a reactor trip and a loss of forced reactor coolant circulation. Two hours later,
reactor coolant system (RCS) hot leg temperature is greater than cold leg temperature and steam
generator (SG) levels are stable.
Which one of the following combinations of parameter trends, observed two hours after the trip,
indicates that natural circulation is not occurring? (CET = core exit thermocouples)

A

RCS Hot Leg RCS Cold Leg SG RCSCET
Temperature Temperature Pressures Subcooling
Stable Stable Decreasing Decreasing

105
Q

A reactor had been operating at 100 percent power for 3 months when a loss of offsite power occurred,
causing a reactor trip and a loss of forced reactor coolant flow. If forced reactor coolant flow is not
restored, which one of the following describes the relationship between reactor coolant hot leg and
cold leg temperatures one hour after the reactor trip?

A

Hot leg temperature will be greater than cold leg temperature because natural circulation cooling
flow occurs in the same direction as forced reactor coolant flow.

106
Q

A reactor is shut down at normal operating temperature and pressure with all reactor coolant pumps
stopped. Stable natural circulation cooling is in progress with a minimum of 50°F subcooling.
Which one of the following, if increased, will not affect natural circulation flow rate?

A

Reactor coolant pressure

107
Q

Fully-developed natural circulation flow rate will be greatest when…

A

all reactor coolant pumps stop at the same time as the reactor trip.

108
Q

Natural circulation flow can be enhanced by…

A

increasing the temperature difference between the heat source and the heat sink.

109
Q

Which one of the following will enhance natural circulation flow in the reactor coolant system?

A

Steam generator level is increased.

110
Q

A nuclear power plant was operating at a constant power level for the last two weeks when a loss of
offsite power occurred, which caused a reactor trip and a loss of forced reactor coolant flow. Natural
circulation reactor coolant flow developed and stabilized 30 minutes after the trip.
Which one of the following combinations of initial reactor power and post-trip steam generator
pressure will result in the greatest stable natural circulation flow rate 30 minutes after the trip?

A

Initial Post-trip Steam
Reactor Power Generator Pressure
100 percent 1,000 psia

111
Q

A nuclear power plant was operating at a constant power level for the last two weeks when a loss of
offsite power occurred, which caused a reactor trip and a loss of forced reactor coolant flow. Natural
circulation reactor coolant flow developed and stabilized 30 minutes after the trip.
Which one of the following combinations of initial reactor power and post-trip steam generator
pressure will result in the smallest stable natural circulation flow rate 30 minutes after the trip?

A

Initial Post-trip Steam
Reactor Power Generator Pressure
25 percent 1,100 psia

112
Q

A nuclear power plant was operating at steady-state 100 percent power when a loss of offsite power
occurred, which caused a reactor trip and a complete loss of forced reactor coolant flow. Natural
circulation reactor coolant flow developed and stabilized approximately 30 minutes after the trip.
Which one of the following combinations of reactor power history and post-trip steam generator
pressure will result in the greatest stable natural circulation flow rate?

A

Days At Post-trip Steam
Full Power Generator Pressure
100 percent 1,000 psia

113
Q

A few minutes ago, a nuclear power plant experienced a loss of offsite power that caused a reactor trip
and a loss of all reactor coolant pumps. Natural circulation flow is currently developing in the reactor
coolant system (RCS).
Which one of the following operator actions will promote the development of natural circulation in the
RCS?

A

Establish and maintain steam generator water level high in the normal operating range.

114
Q

A nuclear power plant was operating at steady-state 100 percent power when a sustained loss of offsite
power occurred, which caused a reactor trip and a complete loss of forced reactor coolant flow.
Which one of the following combinations of reactor power history and post-trip steam generator
pressure will result in the smallest stable natural circulation flow rate?

A

Days At Post-trip Steam
100 Percent Power Generator Pressure
10 1, 100 psia

115
Q

During the reflux boiling method of core cooling, steam from the reactor core is condensed in the
__________ side of a steam generator and flows back into the core via the __________. (Assume the
steam generators contain U-tubes.)

A

hot leg; hot leg

116
Q

Which one of the following describes the method of core heat removal during reflux core cooling
following a loss of coolant accident?

A

Convection with natural circulation coolant flow.

117
Q

A nuclear power plant is experiencing natural circulation core cooling following a loss of coolant
accident. Which one of the following, when it first occurs, marks the beginning of reflux core
cooling? (Assume the steam generators contain U-tubes.)

A

Steam condensation in the hot legs is unable to pass completely through the steam generators to
enter the cold legs.

118
Q

A reactor coolant system natural circulation cooldown is in progress with steam release from the steam
generator (SG) atmospheric steam relief valves (operated in manual control). If high point voiding
interrupts natural circulation, which one of the following will occur? (Assume feedwater flow rate,
SG relief valve position, and core decay heat level are constant.)

A

SG level will increase and SG pressure will decrease.

119
Q

A reactor coolant system natural circulation cooldown is in progress with steam release from the steam
generator (SG) atmospheric steam relief valves (operated in manual control). Assume feedwater
flow rate, SG relief valve position, and core decay heat level are constant.
If high point voiding interrupts natural circulation, SG levels will gradually __________; and core exit
thermocouple indications will gradually __________

A

increase; increase

120
Q

A reactor coolant system (RCS) cooldown on natural circulation is in progress. The cooldown rate is
being controlled by releasing steam from the steam generator (SG) atmospheric relief valves in
Manual control.
If voids interrupt the RCS natural circulation flow, which one of the following will occur? (Assume
feedwater flow rate, SG relief valve positions, and decay heat level are constant.)

A

SG pressure will decrease and core exit thermocouple (CET) temperatures will increase.

121
Q

A reactor coolant system natural circulation cooldown is in progress with steam release from the steam
generator (SG) atmospheric steam relief valves (operated in manual control). Assume feedwater
flow rate, SG relief valve position, and core decay heat level remain constant.
If high point voiding interrupts natural circulation, SG steam flow rate will __________ and core exit
thermocouple temperatures will __________.

A

decrease; increase

Thermo (10 decks)

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