Conceptual Design for a Severe Accident Heat Removal
System in a Pressurized Water Reactor Nuclear Power Plant
by
Daniel C. Flahive
An Engineering Project Submitted to the Graduate Faculty of Rensselaer Polytechnic
Institute in Fulfillment of the Requirements of the degree of
MASTER OF ENGINEERING IN MECHANICAL ENGINEERING
Approved:
_
Ernesto Gutierrez-Miravete, Project Adviser
Rensselaer Polytechnic Institute
Hartford, Connecticut
May, 2015
© Copyright 2015
by
Daniel C. Flahive
All Rights Reserved
ii
ABSTRACT
In response to the nuclear accident at the Fukushima Dai-ichi site in March of 201l,
nuclear power plants around the world have been asked to address the potential for an
event outside of the plants design basis. These scenarios include other external hazards
such as flooding, seismic, terrorist attack. In all the postulated events the site needs
coping strategies to protect the plant without the normally relied on safety equipment.
This report proposes a new system that can be put in place to remove heat from the
containment in a severe accident condition. In the event of fuel and reactor vessel
failure, the last line of defense is the containment building. This system can remove heat
from containment and maintain the pressure within the containment design limits.
This report does present a conceptual design for this system. The design would need to
be altered (i.e. pump size, heat exchanger size) but the general concept could be
implemented in an existing PWR plant.
iii
TABLE OF CONTENTS
ABSTRACT ..................................................................................................................... iii
TABLE OF CONTENTS ................................................................................................. iv
LIST OF TABLES ............................................................................................................ iv
LIST OF FIGURES ........................................................................................................... v
SYMBOLS/ABBREVIATIONS LIST ............................................................................. vi
1. INTRODUCTION ....................................................................................................... 1
2. BACKGROUND ......................................................................................................... 2
2.1
Fukushima Event ................................................................................................ 2
2.2
INPO Lessons Learned ...................................................................................... 3
2.3
PWR Postulated Event ....................................................................................... 5
2.4
Molten Core Concrete Interaction ...................................................................... 8
2.5
Flooding to Cool Corium ................................................................................. 10
2.6
Concern with Hydrogen ................................................................................... 16
2.7
System Design .................................................................................................. 17
3. METHODOLOGY .................................................................................................... 18
4. GROUNDRULES...................................................................................................... 19
5. EVENT PROGRESSION .......................................................................................... 20
6. CONCEPTUAL LAYOUT ....................................................................................... 22
7. EQUIPMENT ............................................................................................................ 25
7.1
Pumps ............................................................................................................... 27
7.2
Valves............................................................................................................... 27
7.3
Heat Exchanger ................................................................................................ 28
7.4
Piping ............................................................................................................... 28
8. FUNCTIONS ............................................................................................................. 30
8.1
System Standby/Containment Isolation ........................................................... 30
8.2
Containment Flooding...................................................................................... 31
iv
8.3
Containment Sump Cooling ............................................................................. 31
8.4
Electrical System .............................................................................................. 31
9. HEAT TRANSFER ................................................................................................... 33
10. HEAT EXCHANGER SIZING ................................................................................. 35
11. HYDRALIC MODEL ............................................................................................... 37
11.1 Flooding Case .................................................................................................. 38
11.2 Cooling Case .................................................................................................... 39
12. CONCLUSION.......................................................................................................... 41
13. REFERENCES .......................................................................................................... 42
APPENDIX A – FATHOM CODE FILES ..................................................................... 43
A.1
Flooding Case .................................................................................................. 43
A.2
Cooling Case .................................................................................................... 52
v
LIST OF TABLES
Table
Page
Table 1 Valve List
28
Table 2 Fluid Properties
33
Table 3 Heat Transfer Calculation
34
Table 4 Heat Exchanger Sizing
35
Table 5 Heat Exchanger Parameters
36
Table 7 Hydraulic Model Junctions
38
Table 8 Hydraulic Model Piping
38
Table 9 Cooling Case Performance
40
iv
LIST OF FIGURES
Figure
Page
Figure 1 Melted Corium in Concrete
9
Figure 2 Predicted Ablation from MCCI – Dry Cavity
12
Figure 3 Predicted Ablation from MCCI – Flooded Cavity
13
Figure 4 Reactor Vessel Location in Containment
22
Figure 5 Containment Sump/Pool Cooling System Configuration
24
Figure 6 Pipe and Instruments Diagram – System Standby
25
Figure 7 Pipe and Instruments Diagram – Containment Flood
26
Figure 8 Pipe and Instruments Diagram – Containment Cooling
26
Figure 9 Hydraulic Model
37
Figure 10 Flooding Case Model
39
Figure 11 Cooling Case Model
40
v
SYMBOLS/ABBREVIATIONS LIST
Symbol/
Meaning
Abbreviation
A
AC
Area
m²
Alternate Current
ANS
American Nuclear Society
BOP
Balance of Plant
CFR
Code of Federal Regulation
Cp
Heat Capacity
dP
Pressure Differential
EDG
HX
kJ/kg-K
Emergency Diesel Generator
Heat Exchanger
INPO
Institute of Nuclear Power Operation
LOCA
Loss of Coolant Accident
ṁ
Mass Flowrate
kg/s
MCCI
Molten Core Concrete Interaction
MWe
Megawatt Electric
MWt
Megawatt Thermal
NRC
Nuclear Regulatory Commission
PWR
Pressurized Water Reactor
q
RCS
SAMG
SG
Units
heat transfer
kW
Reactor Coolant System
Severe Accident Management Guidelines
Steam Generator
T
Temperature
°C, °F
U
Heat Transfer Coefficient
υ
Specific Volume
kW/m²-K
lbm/m³
vi
1. INTRODUCTION
In March 2011 the Fukushima Daiichi experienced a loss of all site backup AC power
following an earthquake and seismic induced tsunami at the site. The site is designed to
withstand expected external events however this event was beyond the design bases of
the plant. Both the earthquake and tsunami exceeded what the site was designed to.
Globally nuclear power plants are now assessing their capability to cope with events
beyond the design basis of the plant similar to the event at Fukushima Dai-ichi.
Regulators are requiring that sites evaluate their existing design basis and develop
coping strategies without relying on currently installed safety equipment.
In some extreme cases, regulators are asking sites to rely on no existing equipment or
piping outside of containment. This project will develop a system that can remove decay
heat from the reactor within containment. The only interaction would be a water supply
to remove the heat from containment (i.e. river or ocean). No existing pumps or heat
exchangers would be credited. The challenge of this design will be minimizing the
system design as space is typically limited.
The Fukushima nuclear accident was initiated by the Great East Japan Earthquake on
March 11, 2011 and the ensuing tsunami. There were 11 operating reactors and 3
shutdown reactors across 4 sites in the region affected by the tsunami. The operating
reactors (Units 1, 2 and 3) shutdown following the earthquake, as designed.
earthquake registered at 9.0 and exceeded the plant design based.
The
Investigations
following the event suggest that the plant survived the earthquake with no appreciable
damage and no loss of function to the plant safety systems (Reference 1).
1
2. BACKGROUND
2.1 Fukushima Event
The Fukushima Daiichi nuclear power plant included six boiling water reactors (BWRs).
On March 11th, 2011 Units 1, 2, and 3 were operating at full power and units 4, 5, and 6
were shutdown for refueling when a magnitude 9.0 earthquake occurred 112 miles (180
kilometers) off Japan’s east coast. The three operating units automatically scrammed on
seismic reactor protection system trips.
The earthquake caused damage to the offsite electric distribution to the site and lead to a
loss of offsite power event. The available emergency diesel generators automatically
started as designed and began to supply AC power to emergency systems at all units.
Three minutes after the earthquake, the Japan Meteorological Association issued a major
tsunami warning, indicating the potential for a tsunami at least 3 meters high. Workers
were notified of the warning, and operators were instructed to report to the control rooms
while non-essential personnel were evacuated to higher ground.
Forty-one minutes after the earthquake, the first of a series of seven tsunamis arrived at
the site. The maximum tsunami height impacting the site was estimated to be 14 to 15
meters, exceeding the site design basis tsunami height of 18.7 feet (6.1 meters) and site
grade levels at units 1, 2, 3 and 4, 32.8 feet (10 meters).
The tsunami flooded the emergency diesel generators and switchgear rooms for units 1
through 5. The seawater intake structure which serves as the ultimate heat sink for the
site was severely damaged and was rendered nonfunctional. All DC power was lost on
units 1, 2, and 4, while some DC power from batteries remained available on Unit 3
because some of those battery banks were not flooded. One air-cooled emergency diesel
generator continued to function and supplied electrical power to Unit 6, and later to Unit
5, to maintain cooling to the reactors and spent fuel pools.
The loss of AC power to units 1 through 4 meant the systems designed to remain decay
heat when the units shutdown (either planned for an outage (Units 4, 5 and 6) or in
response to an event (Units 1, 2 and 3).
2
With no core cooling to remove decay heat, core damage began on Unit 1 on the day of
the event. Steam-driven injection pumps were used to provide cooling water to the
reactors on units 2 and 3, but these pumps eventually stopped working. As a result of
inadequate core cooling, fuel damage also occurred in units 2 and 3. After debris caused
by the tsunami was removed, fire engines were moved into position and connected to
plant systems to restore water injection. Connection points had been installed previously
to support fire protection procedures, but the plant staff had difficulty locating them
initially because of the debris and because drawings had not been updated to show their
locations.
During the event, containment pressure remained high for an extended time, contributing
to hydrogen leakage from the primary containment vessel and inhibiting injection of
water to the reactors using low-pressure sources.
It is believed that hydrogen generated from the damaged fuel in the reactors accumulated
in the reactor buildings either during venting operations or from other leaks and ignited,
producing explosions in the Unit 1 and Unit 3 reactor buildings and significantly
complicating the response. The hydrogen generated in Unit 3 likely migrated into the
Unit 4 reactor building, resulting in a subsequent explosion and damage. The loss of
primary and secondary containment integrity resulted in ground-level releases of
radioactive material. Following the explosion in Unit 4 and the abnormal indications on
Unit 2 on the fourth day of the event, the site superintendent directed that all nonessential personnel temporarily evacuate for their safety, leaving approximately 70
people on site to manage the event.
2.2 INPO Lessons Learned
The Institute of Nuclear Power Operation (INPO) is an organization the sets industry
performance objectives and guidance to promote safety, reliability and promote
excellence in the operation of nuclear power plants in the US. The organization was
formed in response to the accident at Three Mile Island.
3
Tokyo Electric Power Company (TEPCO) requested INPO (along with the World
Association of Nuclear Operators (WANO) to conduct an independent review of the
accident at the Fukushima Daiichi Nuclear Power Station. The purpose of the review
was to identify and share operational and organizational lessons with the nuclear
industry.
From INPO Report, “Lessons Learned from the Nuclear Accident at the Fukushima
Daiichi Nuclear Power Station.” (Reference 2):
The following positive elements were critical to TEPCO’s response during the event:

The seismically isolated emergency response centers at the Fukushima Daiichi
and Daini nuclear power stations filled a vital need in protecting emergency
response personnel and ensuring access to the site could be maintained during
the accident.

Emergency response personnel took innovative and resourceful actions to
reestablish critical safety functions and plant monitoring capability. Actions to
restore power and heat removal capability at the Fukushima Daini Nuclear
Power Station were particularly noteworthy.

The response of TEPCO employees during and following the event reflected high
levels of professionalism, courage, dedication, and personal ownership.
The following are considered the most significant operational lessons from the event:

When periodic reviews or new information indicates the potential for conditions
that could significantly reduce safety margins or exceed current design
assumptions, a timely, formal, and comprehensive assessment of the potential for
substantial consequences should be conducted. An independent, cross-functional
safety review with a plant walkdown should be considered to fully understand the
nuclear safety implications. If the consequences could include the potential for
common-mode failures of important safety systems, compensatory actions or
countermeasures must be established without delay.

Emergency and accident response strategies and implementing actions must give
highest priority to maintaining core cooling. Emergency response centers must
maintain continuous awareness of the status of core cooling; changes to the
method of core cooling must be made deliberately and with a clear strategy to
establish an alternate cooling method; and, when there is reason to question the
quality or validity of core cooling information, deliberate actions must be taken
immediately to ensure a method of cooling is established.

Plans must address the immediate emergency response needs for human
resources, equipment, and facilities in the first few hours of an event, as well as
the need for a long-duration response capability. In addition, plans should
4
address how to engage the domestic and international nuclear industry to obtain
needed support and assistance during an event.

Training and periodic drills must be sufficiently challenging and realistic to
prepare operating crews and emergency response personnel to cope with and
respond to situations that may occur during a multi-unit nuclear accident,
including a nuclear accident resulting from a natural disaster.

Because the specific sequence of initiation events for beyond-design-basis events
is unknown, emergency response strategies must be robust and provide multiple
methods to establish and maintain critical safety functions using a defense-indepth approach.

Optimum accident management strategies and associated implementing
procedures (such as emergency operating procedures and accident management
guidelines) should be developed through communications, engagement, and
exchange of information among nuclear power plant operating organizations
and reactor vendors. Decisions to deviate from these strategies and procedures
should be made only after rigorous technical and independent safety reviews that
consider the basis of the original standard and potential unintended
consequences.

Emergency response strategies for extreme external events should consider the
traumatic human impact of such events on individual responders and leaders and
provide for appropriate training, assistance, and contingency plans.

Nuclear operating organizations should consider the safety culture implications
of the Fukushima Daiichi event, focusing on strengthening the application of
safety culture principles associated with questioning attitude, decision-making,
the special and unique aspects of the nuclear technology, and organizational
learning.
The system designed in this report will focus on addressing the second bullet of the
operational lessons from the event, designing a system focused on the highest priority of
maintaining core cooling. More specifically establishing an alternate cooling method if
the accident degrades to a severe accident with reactor vessel failure.
2.3 PWR Postulated Event
The Fukushima nuclear power plant units were BWR design while the focus of the
system designed within this report are for a PWR design. The PWR design still is
susceptible to external events and could have led to a severe accident if exposed to
similar environment conditions that occurred at Fukushima on March 11, 2011.
5
There is a lot of uncertainty of the progression of a severe accident and no perfect event
timeline could be developed that would encompass all PWR designs. The description he
is a general progression of events for a PWR if the site were to experience a loss of all
AC power (off-site and onsite emergency power) similar to happened at Fukushima.
When the event initiates the reactor is expected to scram as normally designed. This is
an automatic response by the plant due to a number of different events (e.g. seismic
event, loss of offsite power). In a pressurized water reactor (PWR) control rods will
remain above the nuclear fuel during power operation. They are held in place by
powered drive mechanisms designed to release on a variety of signals, operator action or
in the case of a loss of power. The control rod drive mechanisms are powered to hold
the rods in place and on a loss of power release. On a loss of power the control rods rely
only on gravity to insert them into the core and shutdown the reactor.
They are
composed of a material (typically boron) that will quickly add negative reactivity to the
reactor. At this point the reactor is shutdown however the principle decay heat from the
reactor remains and needs to be managed.
These control rods rapidly introduce negative reactivity and the critical reaction is
shutdown in a fraction of a second. In the two most significant accidents at light water
reactors (Fukushima & Three Mile Island) the reactors effectively shutdown as designed.
The accident eventually developed because the residual decay heat could not be
managed.
The PWR control rods drop into the core halting the chain reaction. Within a few
seconds of scramming the reactor decay heat of the core is less than 1% of the normal
operation heat load. 1% of the normal operating heat load still represents a significant
amount of heat. In this report, considering a nominal reactor size, 1% is 30 MWt. After
the initial drop in reactor power, decay heat in the reactor drops at a much lower rate.
Without backup AC power onsite the auxiliary feedwater system will refill the SGs.
The auxiliary feedwater system typically consists of two pumps, one AC motor driven
and the other steam driven. Given the loss of site AC power, the motor driven pump
would be unavailable. The steam turbine driven pump is supplied steam from the steam
6
generators to drive the pump. The initial heat removal mechanism for the reactor is to
boil water in the steam generators and release the steam to the atmosphere. This steam
in separated from the reactor coolant and therefore does not release radiation to the
atmosphere, only steam. A portion of the steam produced by the steam generator is
diverted to the steam driven auxiliary feedwater pump.
Controls on this system are battery powered in most plants so in the event of a loss of
offsite power and the emergency diesel generators (EDS) the system can still operate.
The pump runs on a steam supply and controls for this system rely on backup batteries.
This system will temporarily remove heat from the RCS. Most plants have sufficient
battery power for several hours before needing to be recharged by the EDGs onsite. In
the case of Fukushima the all except one of the site EDG were flooded by the tsunami.
Assuming that SG cooling is lost, the water in the SGs would continue to boil and its
level would decrease. Eventually the SG will boil dry and the steam generators will
cease to remove heat from the RCS. Without SG heat removal temperature and pressure
will increase in the RCS. The pressure relief valves for the reactor will lift and water
volume in the RCS will boil away.
The water volume in the reactor will eventually drop and the fuel rods will become
uncovered. Without water over the fuel it will over heat and begin to melt. Once the
reactor is completely dry and the fuel has melted in the bottom of the reactor the fuel
could melt through the bottom of the reactor and present the ex-vessel core severe
accident the system proposed in this document is designed to address.
The molten core will mix with the metal mass of the fuel assemblies and reactor internal
structures and form a molten substance referred to as corium.
The corium could
continue to melt through the reactor vessel and breach the reactor vessel, falling to the
floor of the reactor vessel cavity.
When the core is ex-vessel many plant have strategies to flood containment to cover the
melted core. Flooding the core provides several benefits:

Cooling of the molten core
7

Mitigate interaction with the concrete

Provide some amount of radiative scrubbing
The system proposed in this document will be capable of flooding the containment if
necessary. Some plants may accomplish this function with current installed systems and
procedures. If not, the first function of the system will be to flood the containment
vessel and cover the molten core. The external pump will be used to deliver water into
containment. The same piping used to deliver to the heat exchanger will be used and a
motor operated valve inside containment will be used.
2.4 Molten Core Concrete Interaction
In the most severe of severe accidents, the nuclear fuel in the reactor would melt and the
molten core or corium can melt through the reactor vessel and fall to the reactor cavity
floor. On the concrete floor of containment the core will begin to melt through the
concrete floor. This process is known as Molten Core Concrete Interaction (MCCI).
Within the concrete floor of containment is a steel liner which is the containment
pressure boundary. If sufficient concrete melt occurs and the steel liner is met by the
corium, the containment boundary would be breached.
The core of a nuclear reactor consists of uranium dioxide in the fuel, zirconium in the
fuel rod cladding and carbon steel and stainless steel in other structures. At high
temperatures zirconium is oxidized by water vapor, so the main constituents of core
melt, or corium, are UO2, ZrO2, Zr, Fe, Cr and Ni. The melting point of the pure oxides
is around 2700°C, while the metals melt at 1350-1900 °C. Mixtures of different species
do not have single melting points. Instead, they change from solid to liquid over a range
of temperatures, between so-called solidus and liquidus temperatures. The density of
corium is around 6000-7000 kg/m3. (Reference 3)
8
Figure 1 Melted Corium in Concrete (Reference 3)
The molten core, concrete interaction (MCCI) is illustrated in Figure 1. The solid
concrete and the molten corium pool may be separated by a thin layer of partly molten
concrete. The concrete melt will travel upwards because it is less dense than the
overlying core melt. The melting core will produce gas bubbles to rise around and
through the corium.
The bubbles will be highly radioactive, non-condensable gasses.
Mitigation of MCCI will not only protect the integrity of the containment boundary, it
will reduce the release of these gases from the concrete and improve the coping
capability of the plant with the event.
The oxides in corium and concrete are miscible with each other, but the metallic species
are immiscible with the oxides. Because the metals are lighter than the corium oxides, a
metallic layer can form on the surface of the corium pool. When concrete oxides are
added to the melt, its density decreases eventually below the density of the metals. After
this, the metallic layer may relocate to the bottom of the pool. On the other hand, intense
stirring of the pool by the rising gas bubbles may cause the metals and the oxides to be
mixed with each other. (Reference 5)
The melting of the concrete is limited by the heat transfer from the corium to the
concrete. The thermal conductivity of the concrete is very low and therefore a lot of heat
9
goes into heating and melting the concrete. While melting of concrete is not desirable in
a severe accident, the process does remove substantial heat from the corium.
Upon vessel failure, there may be jet of corium leak through the failure region of the
vessel. Some initial erosion of the concrete may take place in the region of contact on
the concrete floor. This initial erosion will likely last only a short time period before is
crusts over preventing further ablation of the concrete. Significant concrete melt would
not be likely until the pool of corium has collected on the concrete floor.
However, the rapid initial impingement heat transfer may initiate a sustained concrete
spalling. Spalling means that cracks form in the concrete and break off into pieces
instead of simply melting the concrete. The cracks could for from rapid evaporation of
water within the concrete. Pieces of the cracked concrete could float to the surface of
the corium. These pieces could melt more rapidly and further spalling could occur in the
concrete below the pieces that had cracked and broken free of the floor.
This process
could continue and become self-sustaining more rapidly ablating the concrete.
Heat is also transferred from the ex-vessel corium pool to the fluid above the material
(air or water in a flooded cavity) by thermal radiation and convection. Without action to
flood the reactor vessel cavity, usually there is air above the pool surface. Cooling by air
is limited compared to the presence of water.
2.5 Flooding to Cool Corium
The system designed within the report will utilize the cooling of water over the pool of
corium. Once the core is flooded, natural circulation will carry hot water surrounding
the corium up, transferring heat to the entire water volume of the flooded containment.
To aid in the function of this system, the piping of the system designed within this report
can support discharging cooled water to various location in the flooded lower
containment. The necessity for this feature and the specific pipe routing would be
addressed on a plant by plant basis.
10
The rate of concrete ablation is limited by the heat transfer from the melt to concrete. If
sufficient cooling can be provided by water in lower containment ablation of the
concrete could be halted. If concrete ablation can be halted and continuous heat removal
from containment can be sustained, containment integrity should be maintained.
The impact of flooding the reactor cavity on MCCI has been analyzed and results have
shown its effectiveness. Even in the cases where there is no flooding of the corium in
the reactor cavity, the ablation of the concrete is expected to halt prior to breaching the
containment vessel.
Figure 2 shows the concrete ablation approximately 160 cm in depth into the concrete.
This result is dependent on the corium mass and the geometry of the reactor vessel
cavity but uses the limiting (smaller) diameter cavity of the cases analyzed in
Reference 5. Therefore while less ablation of the concrete would be preferred, it is not
expected the corium would melt through the concrete floor of lower containment and
breach the containment vessel liner.
The ablation of the concrete will introduce gases into containment that could be just as
significant to the breaching of containment as containment pressure increases and a
potential combustion of the gases could challenge containment integrity as well.
The melting and oxidation of fuel cladding and other structural materials in the reactor
will produce hydrogen. Following the vessel breach, MCCI will generate hydrogen and
carbon monoxide be released into the containment atmosphere.
If sufficient
concentrations of these gases collect in containment, relative to the proportion of steam
and oxygen are present the containment atmosphere will become flammable. If a weak
ignition source exists, such as a spark from electrical equipment, combustion of these
flammable gases will begin.
Depending on the concentration of flammable gases, highspeed flames could occur and
possible lead to an explosion. The dynamic loads from an explosion could induce loads
on the containment structure leading to failure and therefore a breach of the containment.
11
The secondary containment failure at Fukushima Dai-ichi Units 1, 3, and 4 were due to a
distinct hydrogen ignition events.
Mitigation strategies have been developed by nuclear power plants to control the
concentration of such gases by intentionally igniting the gasses if the concentration of
gases is flammable but not explosive. In some PWR designs, igniters are permanently
installed inside the containment vessel and powered to provide an ignition source. This
actions mitigates the risk of such gases reaching explosive levels.
To limit the MCCI and volume of combustible gases in containment, the reactor cavity
flooding effectiveness is shown in Figure 3. The system designed within this report will
be capable of flooding the reactor cavity within the one day specified in the worst case
for Figure 3 or sooner.
The required flooding time would be dependent on the
availability of the portable pump and site operators. To provide the most effective
coping capability for the system equipment could be staged at the first signal there could
be in issue. The action to flood the reactor vessel cavity could be initiated at the first
sign of fuel damage. If these two steps are taken, the reactor vessel core should be
flooded prior to a reactor vessel breach. In that case the MCCI concrete ablation could
even improve on the results in Figure 3 for the flooding after 10 seconds case.
Figure 2 Predicted Ablation from MCCI – Dry Cavity (Reference 5)
12
Figure 3 Predicted Ablation from MCCI – Flooded Cavity (Reference 5)
Flooding the ex-vessel corium provides several benefits to mitigate MCCI.
First,
flooding will provides a cooling mechanism. A pool in lower containment will absorb
the decay heat from the corium up to the point where boiling occurs. Once the pool
begins to boil it will continue to remove heat from the corium however the quantity of
available water will decrease. Boiling will also increase the pressure in containment so
this would not be a sustainable heat removal mechanism. Removing the heat from the
pool is the only mechanism to continuously remove the heat from the corium. The
system designed within this document will be capable of this function.
Flooding the ex-vessel corium also limits the release of gases and radiation from the fuel
and concrete interaction. The corium interaction with concrete will produce gasses as
previously discussed in this report. If the gasses are generated beneath a sufficient level
of water in the reactor cavity, rather than being released into the containment
atmosphere, they will be absorbed by the water. If these gases, which are flammable and
13
in a large concentration, explosive, are retained within the water volume, the potential
for an explosive atmosphere in containment will not develop.
2.5.1
Uncertainty in Response to Flooding the Reactor Cavity
There is potentially and significant phenomenological uncertainty for the action of
flooding the reactor cavity. The amount and rate of steam production when the corium
beaches the reactor vessel and initially interacts with the water outside of the vessel is
difficult to predict. The timing of when water is added to the reactor vessel cavity, when
reactor vessel failure occurs, and the rate at which the corium exits the reactor vessel
will influence the steam generation. The following three conditions can:

Molten core debris is discharged into water

Water is sprayed on the upper surface of the molten pool

Water is added to solidified debris
The rate of steam generation will directly influence the containment pressure response.
Rapid steam generator would be much greater if the reactor vessel failure occurs after
the reactor cavity has been flooded than is it is added following vessel failure and corium
discharge onto the concrete floor.
The containment’s pressure response on the
containment shell integrity would depend on the containment conditions (i.e. pressure,
temperature) at the time of the steam generation. Even in the case where the corium
discharges into the already flooded reactor vessel cavity, the pressurization rate is much
less than the containment design-basis, a large break loss of coolant accident (LOCA).
In a large break LOCA, the RCS is assumed to have a large RCS diameter pipe
(diameter 30 inches to 42 inches) break at normal operating temperature and pressure.
In this accident a greater amount of steam is released into containment than the
conditions for the accident being considered in this report.
Provided the containment pressure can be maintained near normal operating conditions
prior to the reactor vessel failure, the containment pressure response would not threaten
the containment. However, if the containment pressure was elevated (due to hydrogen
burns, MCCI, and so on), or the containment heat removal systems were not operable, or
14
the pressure suppression function is not available, the pressurization due to steam
generation might challenge containment integrity. In this case it would be important that
the system designed within in the report would be available by this time to begin
removing heat from containment.
Another concern in a nuclear accident after the reactor has shutdown is for the core to
return to re-criticality. The design of the fuel is specifically configured to provide a core
geometry that can go critical in the right operational conditions. The amount of fuel and
spacing of the fuel in the fuel assemblies are needed in order for a critical reaction to
occur. Once the core debris has deformed (melted) and relocated to the bottom of the
reactor vessel and even more so, relocation external to the reactor vessel, there is an
extremely low probability of re-criticality occurring.
The response and recovery of the corium in this event is also an uncertainty. Once the
core has been badly damaged the geometry of the fuel would be significantly different
than the fuel pin arrangement of the designed fuel assembly. The change in geometry
will likely be less favorable to cooling than the pre-damaged fuel assembly. In the case
of the Three Mile Island accident, after losing level and uncovering the fuel inside the
reactor, the core was re-flooded approximately 200 minutes into the accident. Despite
be re-flooded portions of the molten core did not solidify immediately and remained in a
molten state and drop in the lower plenum of the reactor vessel approximately 30
minutes after re-flooding of the core. This example from Three Mile Island occurred
with the core remaining in the reactor vessel and core debris was consolidated in a
limited volume. In an ex-vessel event, which is the basis for the system designed in this
report, the core debris will have move area to disperse in the reactor vessel cavity and
corium geometry would likely improve from a coolability standpoint.
2.5.2
External Vessel Cooling without Reactor Vessel Failure
If the action is taken to flood the reactor vessel cavity prior to reactor vessel failure, the
vessel will be surrounded by water which could provide some cooling before the vessel
fails. The amount of cooling would vary between different plant designs dependent on
the configuration of insulation on the reactor vessel. If there is sufficient cooling it is
15
possible that the reactor vessel integrity could be preserved the though damaged, the
core would remain within the reactor vessel.
While the core remains in its design location even when fuel damage begins, cooling
through the reactor vessel from flooding will be limited. The fuel will be separated from
the reactor vessel wall and the only effective way to cool the core in this location is to
provide water directly to the RCS. As the event progresses, the molten core will fall to
the bottom of the reactor vessel. The core material, once relocated to the lower plenum
region, flooding the reactor cavity can removed a significant fraction if not all of the
decay heat. In contact with the metal of the reactor vessel, conduction through the vessel
can remove heat from the molten core.
The system designed in this report could function by removing heat while the molten
core remains in the reactor vessel. In this case, instead of directly removing heat from
the corium in the reactor vessel cavity, the heat conducted from the reactor vessel would
heat the flooded lower containment and the system would function otherwise as
designed.
2.6 Concern with Hydrogen
The build-up of hydrogen in containment is known concern during a severe accident.
Hydrogen is generated first, in the melting of the fuel, and second, in an ex-vessel severe
accident, through MCCI.
In a PWR design hydrogen would be released into the
containment atmosphere. The light gases such as hydrogen have a tendency to stratify
and collect at local high points such as ventilation ducts or under ceilings at various
levels in containment. The stratification and local buildup of hydrogen and other gases
could reach flammable or explosive concentrations. In addition to hydrogen, carbon
monoxide gas is generated during MCCI.
Unlike hydrogen, carbon monoxide has
approximately the same density as steam and is less susceptible to stratification.
The system designed in this report will not specifically address the hydrogen, carbon
monoxide or other combustible gasses generated in containment during a severe
16
accident. Nuclear power plants have existing systems (e.g. igniters) to address the
buildup of hydrogen inside containment. This system will however provide two benefits
related to hydrogen in containment. First, the mitigation of MCCI will reduce the total
volume of gases generated. Second, the flooded level in containment will retain some of
the gases generated during MCCI.
2.7 System Design
The system designed within the report will used the flooding method to cool the exvessel corium. The behavior of corium has been studied and analyzed to determine how
it will behave during a severe accident however with limited opportunities for testing
there is still uncertainly in the results. The common approach that the industry favors,
despite the uncertainty is to flood the cavity to provide some cooling of the molten, exvessel corium. The system designed within this report provides that function of flooding
lower containment. Once flooded, the system can cool the flooded pool to remove the
decay heat from the corium.
17
3. METHODOLOGY
This project develops a feasible design for an in-containment decay heat removal
system. This system will focus on the pressurized water reactor (PWR) designed plants.
Ground rules to bound the event and heat removal requirements will be determined.
Heat transfer calculation and a hydraulic model will be used in the design process.
The ground rules for designing this system will assume conditions about the affected
units. A nominal reactor size will be selected. The design would need to be adjusted if
implemented at a plant to the specific plant power but the general concept could be used.
The plant will shutdown normally and initial decay heat removal will be performed
through the installed SGs. While the Fukushima plant does not have SGs however this
response is consistent with the Fukushima event following the earthquake but prior to
the tsunami. The design will focus on addressing the most severe accident where fuel
failure has occurred and breached the reactor vessel.
A conceptual design has been created consisting of two pumps and a heat exchanger. To
prove the feasibility of the design the heat transfer and hydraulic performance of the
system will be proven sufficient to remove the required decay heat from the reactor.
18
4. GROUNDRULES
The concepts of the design within this project could be used as a defense in depth for
most pressurized water reactor designs with some alterations. To bound the scope of the
design the follow assumptions will be used.
Initial heat removal through SG boiloff – Pressurized water reactor designs transfer heat
to a steam generator during normal operation. In an accident these SGs are full or water
that can be boiled-off and steam released to the atmosphere for initial heat removal in the
event. Because of this initial heat removal it is assumed that heat in the first hour of the
event is removed from the SGs. This assumption is consistent with what happened
during the Fukushima event. Cooling for all units at Fukushima lasted at least the first
hour after the earthquake.
Sized for nominal reactor size 1 MWe, 3MWt (assuming 33% thermal efficiency). In
the US operating PWR designs range between 502 MWe and 1336 MWe
(www.nrc.gov). The 1 MWe was chose to prove the concept of the design. If this
design is implemented in a nuclear power plant it would need to be modified to fit the
specific unit. The heat removal this system will be designed for will be based on 3MWt
and the 1979 ANS decay heat approximations.
Focus on pressurized water reactor plant design – This design will focus on a severe
accident heat removal. Installed plant systems are designed to remove heat from the
nuclear fuel provided the core remains in a controlled geometry. This design will
provide a heat removal capability without relying on the installed existing systems.
The purpose of this design is to support a severe accident where there has been damage
to the nuclear fuel. In this case it is assumed that not only the fuel in the reactor has
melted but also melted through the bottom of the reactor. This is considered and exvessel core damage event. At this point the radioactive material has breach the first two
of the three boundaries (the fuel cladding and the reactor vessel). The only remaining
barrier is the containment. This systems objective is to remove heat and prevent over
pressure of the containment.
19
5. EVENT PROGRESSION
The system designed within the report will not focus on the reason for the loss of core
cooling but simply how the event progresses following the loss of core cooling.
When the event initiates the reactor is expected to scram. This is an automatic response
by the plant due to a number of different events (e.g. seismic event, loss of offsite
power). The PWR control rods drop into the core halting the chain reaction. Within a
couple seconds of scramming the reactor decay heat of the core is less than 1% of the
normal operation heat load.
Without backup AC power onsite a steam turbine driven pump (steam provided from the
decay heat in the reactor and refill the SGs. They system in most plant designs does not
require AC power. The pump runs on a steam supply and controls for this system rely
on backup batteries. This system will temporarily remove heat from the RCS. Most
plant have sufficient battery power for several hours before needing to be recharged by
the emergency diesel generators EDG onsite. In the case of Fukushima the all except
one of the site EDG were flooded by the tsunami.
Assuming that SG cooling is lost, the water in the SGs would continue to boil and its
level would decrease. Eventually the SG will boil dry and the steam generators will
cease to remove heat from the RCS. Without SG heat removal temperature and pressure
will increase in the RCS. The pressure relief valves for the reactor will lift and water
volume in the RCS will boil away.
The water volume in the reactor will eventually drop and the fuel rods will become
uncovered. Without water over the fuel it will over heat and begin to melt. Once the
reactor is completely dry and the fuel had melted in the bottom of the reactor the fuel
could melt through the bottom of the reactor and present the ex-vessel core severe
accident the system proposed in this document is designed to address.
When the core is ex-vessel many plant have strategies to flood containment to cover the
melted core. Flooding the core provides several benefits:
20

Cooling of the molten core

Mitigate interaction with the concrete

Provide some amount of radiative scrubbing
The system proposed in this document will be capable of flooding the containment if
necessary. Some plants may accomplish this function with current installed systems and
procedures. If not, the first function of the system will be to flood the containment
vessel and cover the molten core. The external pump will be used to deliver water into
containment. The same piping used to deliver to the heat exchanger will be used and a
motor operated valve inside containment will be used.
21
6. CONCEPTUAL LAYOUT
Figure 1 shows a typical PWR containment. In this design, the reactor vessel is located
in the lower containment. In the event of a break in the reactor coolant system (RCS),
reactor coolant, water, will spill into a sump in the bottom of containment. Plant safety
systems will deliver additional water to the RCS to cool the core which will spill into
containment and flood the sump and lower containment.
The design proposed in this report will take advantage of the lower containment. In this
postulated accident in-place safety systems would have been unable to maintain core
cooling (i.e. loss of emergency power). The core will have melted and broken through
the reactor vessel wall. Figure 2 shows the postulated location of the core after this
accident and a flood level of containment.
Figure 4 Reactor Vessel Location in Containment (Image Reference)
Flooding the lower containment in this case allows the fuel to remain under water.
Many plants have this action as part of their Severe Accident Management Guidelines
(SAMG). This system design could be adapted to include this function if necessary.
22
The fuel geometry will no longer support a critical reaction. Therefore the heat removal
required will be limited to the decay heat.
In this configuration the heat from the core will be transferred to the water. Steam
bubbles will form some will collapse within the water, heating up the water. Others will
transfer to the air volume of containment, filling containment with steam.
Some heat in the water will be conducted through the containment steel liner and
concrete to the earth and external air. Most of the heat however will generate steam in
containment. If the heat cannot be removed, steam will continue to increase the pressure
in containment and eventually above the design pressure.
This design will use a two pump and a heat exchanger to remove heat from the water in
containment. Details on pump type (electric/diesel) and will be discussed further in the
equipment sizing section (Section 6)This approach will provide a continuous heat
removal capability long after the accident until steps to cleanup and decommission the
plant can be taken.
23
Core
Figure 5 Containment Sump/Pool Cooling System Configuration
24
7. EQUIPMENT
Figures 4, 5 and 6 show the components needed for the system proposed in this report.
The general concept could be customized to meet the needs of other plants. The main
components general specifications will be provided here, pumps, HX, valves and piping.
Other components such as instrumentation would need to consider plant specific
requirements.
Figure 6 Pipe and Instruments Diagram – System Standby
25
Figure 7 Pipe and Instruments Diagram – Containment Flood
Figure 8 Pipe and Instruments Diagram – Containment Cooling
26
7.1 Pumps
This system will require two pumps. The first will be a portable diesel driven pump.
The pump will be staged near a large volume water supply. It is typical for nuclear
power plants to be located near a large body of water (i.e. river, ocean) which is used for
balance of plant (BOP) heat removal during normal operation. Large hoses or temporary
piping would be used to connect the pump discharge to the plant.
The second pump will be an installed electric pump. An external generator will provide
power to the pump inside containment. This pump will circulate water in the flooded
containment sump through the heat exchanger.
Both pumps will require the same flowrate, 130 kg/s. This flowrate was chosen to
provide sufficient heat removal capacity with the heat transfer capability available given
the heat exchanger design and temperature difference between two loops of the system.
7.2 Valves
Valves that will need to be included in the design to support function of the system and
containment isolation (more details of this requirement discussed in Section 8.1). The
function of containment isolation is required as part of the design basis for nuclear
power plant. One design goal of this system is to not impact the design basis of the plant
(Section 3). The other valves in the system will support function of the system.
Table 1 provides a list of valves shown in Figure 6 and valve types. The specific
functions these valves support will be further explained in Section 7.0
27
Table 1 Valve List
No.
V1
V2
V3
V4
V5
V6
V7
V8
V9
Type
Check Valve
Gate Valve
Check Valve
Check Valve
Motor Operated Gate Valve
Check Valve
Motor Operated Gate Valve
Gate Valve
Check Valve
7.3 Heat Exchanger
The heat exchanger will be a plate design. This design has been chosen due to its large
heat transfer area relative to the overall size of the heat exchanger unit. Plate heat
exchangers consist of a series of plates stacked together channeling different fluids
between alternating plates. The plates are often corrugated with a chevron pattern to
provide flow distribution and to promote turbulence. The plates are typically 0.5 mm to
1.2 mm thick, and the gap between the plates is typically from 2 to 5 mm. Gaskets seal
the plate edges to prevent the cold and hot fluids from mixing, while also preventing
leakage to the environment.
Plates are made from malleable corrosion-resistant
materials, such as stainless steel and titanium, in a wide range of sizes. For the hot fluid
channel, gaskets seal the cold fluid port; for the cold fluid channel, gaskets seal the hot
fluid port (Reference 6).
Section 10 provides details on sizing of the heat exchanger.
7.4 Piping
The pipe material is stainless steel. Stainless steel is commonly used in nuclear power
plants due to its corrosion resistant properties. The water used in the reactor contains
boron to control reactivity however it is highly corrosive. On the cooling side of the
28
system the water used can be from various sources such as lakes, rivers or oceans.
Because of the uncertainty in water quality the corrosion resistance of stainless steel is
used.
Two pipe sizes will be used in this design. In choosing the piping size, the smallest
piping is the most economical for a system design, therefore the smallest size, large
enough to carry the required flowrate will be used. Both loops have relatively the same
flowrate and therefore the same pipe size will be used. At 2500 gpm (the design
flowrate plus some margin) 8 inch piping is the smallest piping that flow velocities and
pressure losses are provided for (Reference 7).
On the suction side of the two pumps minimizing piping losses is particularly important
to maintain Net Positive Suction Head (NPSH) for the pumps. The next standard pipe
size, 10 inch nominal piping is therefore used.
29
8. FUNCTIONS
8.1 System Standby/Containment Isolation
A requirement in nuclear power plants is to maintain containment integrity to prevent the
release of radioactive material. This is even more so important in the event of a severe
accident. 10 CRF 50 Appendix A (Reference 8) provides these requirements.
From Reference 8:
Criterion 56—Primary containment isolation. Each line that connects directly to the
containment atmosphere and penetrates primary reactor containment shall be provided
with containment isolation valves as follows, unless it can be demonstrated that the
containment isolation provisions for a specific class of lines, such as instrument lines,
are acceptable on some other defined basis:
(1) One locked closed isolation valve inside and one locked closed isolation valve
outside containment; or
(2) One automatic isolation valve inside and one locked closed isolation valve
outside containment; or
(3) One locked closed isolation valve inside and one automatic isolation valve
outside containment. A simple check valve may not be used as the automatic
isolation valve outside containment; or
(4) One automatic isolation valve inside and one automatic isolation valve outside
containment. A simple check valve may not be used as the automatic isolation
valve outside containment.
Isolation valves outside containment shall be located as close to the containment as
practical and upon loss of actuating power, automatic isolation valves shall be designed
to take the position that provides greater safety.
This system will required water going in and out of containment and providing electrical
power to equipment inside containment.
There will be two containment piping penetrations. The first will be the inlet water
supply. On the outside of containment a manual valve will be located and on the inside
of containment a check valve will be location (Appendix A). This configuration will
meet the requirement or 10 CFR 50 Appendix A, Criterion 56.
The second piping penetration will be the HX outlet. An automatic valve will be located
inside containment and manual valve outside containment (Appendix A)
30
8.2 Containment Flooding
The first function is the containment flooding capability.
Some plants have this
capability in their design already. If not, this feature could be implemented to flood the
containment. The external diesel powered pump can flood water into containment.
To initiate that the containment flood function the portable diesel driven pump will need
to be connected to the system. Valve V2 and V5 should be open and V7 and V8 should
be closed. Closing V7 or V8 will block the outlet flow path from containment. During
the initial flooding stage, water will be directed into the containment building sump to
flood containment and flood the ex-vessel corium. Flooding of containment will cool
the ex-vessel containment and help retain the gasses released during MCCI as discussed
in Sections 2.4 and 2.5.
Once the containment is flooded the heat exchanger function can be employed to provide
containment sump cooling as described in Section 7.3.
8.3 Containment Sump Cooling
Once the containment is flooded the heat exchanger function can be employed to cool
the flooded containment. This is the main function of the system. The in-containment,
electric pump will circulated water through the heat exchanger. The external diesel
driven pump will circulate water from the site ultimate heat sink (i.e. ocean, river)
through the other side of the heat exchanger to remove heat from containment.
8.4 Electrical System
To avoid impacting the existing plant electrical systems this system will be completely
independent. Electrical power to the in containment components will be provided from
a portable generator outside of the containment. These components include the in
containment pump and two motor operated valves located inside containment.
31
A portable generator will be located outside of containment and connect to dedicated
electrical penetrations to power the pumps and valves in containment. The dedicated
electrical penetrations are pressure maintaining electrical conductors through the
containment boundary.
Providing dedicated electrical penetrations will ensure the
existing systems in the plant are not impacted by the installation of this system.
32
9. HEAT TRANSFER
Two components are evaluated for the heat transfer. First the two cooling loops must
have sufficient flow and temperature differential to carry the required heat. Second, the
heat exchanger needs to be sufficiently sized to transfer the required heat.
The total heat transfer balance can be determined using the following inputs:

Cooling loop temperature in and out

Containment water in and out

Flow rate of each loop
The cooling loop assumes 80°F water temperature. This is a nominal high temperature.
Most large volumes of water (e.g. lake, ocean) will remain less than 80°F. A lower
temperature will provide additional cooling capacity for the system. A high water
temperature would require some modification to the system design.
The containment water temperature will rise due to the decay heat of the damaged
nuclear core. The pool will rise in temperature to boiling and begin to produce steam in
containment. The steam produced will begin to pressurize containment. A typical
design pressure for containment is 45 psig or 59.7 psia. For this design it’s assumed the
pressure increases to 50 psia. This increases the temperature of the water to 280°F,
saturated temperature at 50 psia.
The fluid properties for each loop are shown in Table 2.
Table 2 Fluid Properties
Loop
Tin
(°F)
Tin
(°C)
Tout
(°F)
Tout
(°C)
Cooling
Containment
80
280
26.56
137.7
180
180
82.11
82.11
Pressure
(psia)
14.7
50
υ
ΔT
(°C)
(ft³/lbm)
55.5556
-55.556
0.016071
0.017264
Cp
(kJ/kg-K)
4.178
4.178
The coolant will need to transport the assumed 3 MW of heat (Section 3). Table 3
provides estimated temperatures in each loop and flowrates to demonstrate the capability
to transfer the required heat.
33
𝑞 = 𝑚̇𝐶𝑝 ∆𝑇
[1]
Table 3 Heat Transfer Calculation
Loop
Cooling
Containment
Tin
Tin
(°F) (°C)
80 26.56
280 137.7
Tout
Tout
(°F)
(°C)
180 82.11
180 82.11
34
ΔT
Flow
Cp
(°C) (kg/s) (kJ/kg-K)
55.6
130
4.178
-55.6
130
4.178
q
(kW)
30175
-30175
10.HEAT EXCHANGER SIZING
Using the flowrates and temperatures calculated above this section will size the heat
exchanger needed to remove the required heat. A plate heat exchanger is expected to be
used. This design provides a lot of surface area in a relatively compact design. This is
an advantage as the system would have to fit into an existing containment. Most plants
have limited free space available.
𝑞 = 𝑈𝐴∆𝑇𝑙𝑚
[2]
The heat transfer coefficient selected was a typical value for plate heat exchanger
(Reference 9). In the design phase identifying the specific heat transfer coefficient
cannot be done without selecting a specific design and knowing the specific flow and
temperature conditions in the system. Therefore the average of the range provided (1000
to 4000 W/m²-K) was assumed.
Table 4 Heat Exchanger Sizing
Tin
(°F)
Cooling
Containment
Tin
(°C)
80 26.56
280 137.7
Tout
(°F)
Tout
(°C)
180
181
Pres.
(psia)
82.1
82.7
ΔTlm
(°C)
A
(m2)
(kW/m²-K)
14.7 55.8
50 -55.8
216
216
2.5
2.5
U
Q
(kW)
30150
-30150
Pressure drop in a plate heat exchanger consists of three components: (1) pressure drop
associated with the inlet and outlet manifolds and ports, (2) pressure drop within the core
(plate passages), and (3) pressure drop due to the elevation change. For the purpose of
this report the pressure drop due the elevation change is ignored because both loops
draw suction and discharge to the same location. The pressure drop on one fluid side in
a plate heat exchanger is given by (Reference 9):
∆𝑝 =
1.5𝐺𝑝 2 𝑁𝑝
2𝑔𝑐 𝜌𝑖
4𝑓𝐿𝐺 2
+ 2𝑔
𝑐 𝐷𝑒
1
1
1
𝐺2
𝑖
𝑐
(𝜌) + (𝜌 − 𝜌 ) 𝑔 ±
𝑜
𝑚
Where: Gp=ṁ/(π/4)Dp2 is the fluid mass velocity in the port,
Dp is the port/manifold diameter,
35
𝜌𝑚 𝑔𝐿
𝑔𝑐
[3]
Np is the number of passes on the given fluid side, and
De is the equivalent diameter of flow passages (usually twice the plate spacing).
Note that the third term on the right-hand side of the equality sign of Equation 13.35 is for the momentum
effect which is generally negligible for liquids.
Therefore:
∆𝑝 =
1.5𝐺𝑝 2 𝑁𝑝
2𝑔𝑐 𝜌𝑖
4𝑓𝐿𝐺 2
+ 2𝑔
𝑐 𝐷𝑒
1
(𝜌)
[4]
𝑚
Without selecting a specific heat exchanger for the conceptual design, nominal
parameters are used to estimate the heat exchanger pressure drop for both loops in the
system.
Table 5 provides the heat exchanger parameters.
Table 6 provides the
calculated pressure drop using Equation 4.
Table 5 Heat Exchanger Parameters
Parameter
Inlet/Outlet Port
Diameter (Dp)
Heat Exchanger
Passes (Np)
Plate Length
Plate Width
Number of Plates
Plate Spacing
De
Value
.203 m
Notes
Port diameter is the same size at the inlet piping, 8 inches
(Section 7.4
1
2m
0.75 m
144
.002 m
.004 m
Required Surface Area/Plate Area
Minimum Standard Gap (Section 7.3)
Twice the spacing between the heat exchanger plates.
Table 6 Heat Exchanger Pressure Drop Calculation
Loop
Cooling
Containment
Gp
(kg/s∙m2)
814.6
814.6
Dp
(m)
653.4 0.203
653.4 0.203
G
(kg/s∙m2)
Np
De
(m) (m)
1 0.004
1 0.004
f
0.1
0.1
ρᵢ
ρₒ
(kg/m³)
(kg/m³)
996.7
927.8
970.3
970.4
dP
(Pa)
983.5 34073
949.1 35333
ρm
(kg/m³)
To ensure the pumps are large enough to meet the required flowrates a bounding, high
pressure drop for the plate heat exchanger will be assumed in the hydraulic model
developed in Section 11. The model will use a second order quadratic resistance curve
based on the pressure drop of 10 psid at 2000 gpm.
36
dP
(psid)
4.942
5.125
11.HYDRALIC MODEL
This design will require two loops, each loop containing a pump and a common heat
exchanger. The first loop will be short to circulate water in the containment through the
heat exchanger and discharging back to containment.
Because the loop is short
hydraulic resistance will primarily be the resistance through the heat exchanger.
The second loop will use an external water source and deliver cool water to the outside
of the heat exchanger. Most power plants are located near a large water source. This is
necessary for heat balance during the normal operation of the plant. This design will
utilize that water source (i.e. lake, river, ocean). A portable pump can be used to draw
from this source. It’s likely that a long length of hose or pipe could be needed. 1000 ft
of hose will be assumed. After exiting the plant 100 ft of host will be assumed to deliver
the heated water to a dedicated location. This water could be drained back into the
original water source but is not necessary.
Figure 9 Hydraulic Model
37
Table 7 Hydraulic Model Junctions
Junc.
Description
J1
J2
J3
J4
J5
J6
J7
J8
J9
Containment Circulation Pump
Heat Exchanger Containment Side
Containment Water
Cooling Water Supply
Cooling Water Pump
Heat Exchanger Cooling Side
Cooling Water Discharge
Tee
Tee
In the hydraulic model, piping includes inputs such as length, size and additional losses.
Additional losses include valves and entrance and exit losses. Both loops will include
entrance losses on the suction piping and exit losses on the discharges. The relative on
the Table 8 provides the details on the piping in the model based on the In the piping,
most of the piping is assumed to be 8 inches nominal diameter. This selection
Table 8 Hydraulic Model Piping
Junc.
Dia (in)
Length (ft)
Valves/Losses
P1
P2
P3
P4
P5
P6
P7
P8
P9
10
8
8
10
8
8
8
8
8
10
10
50
20
1000
1000
20
10
10
1 Entrance Loss
1 Check Valve
1 Check Valve, 1 Exit Loss
1 Entrance Loss
1 Gate Valve, 2 Check Valves
2 Gate Valves, 1 Exit Loss
1 Gate Valve, 1 Check Valves
None
None
11.1 Flooding Case
The first action for the system will be to flood the lower containment. This system
configuration will include one pump operating and delivering water to the containment
sump. The system alignment for this condition is shown in Figure 7.
38
Starting with the base model, the pump J1 is turned off and closed to reverse flow. The
check valve in pipe P2 would prevent any reverse flow through the pump. Pipe P6 is
also closed to simulate the closed valves V7 and V8 in Figure 7.
Figure 10 Flooding Case Model
The output from this case shows that the system can be successfully aligned to flood
containment and the reactor cavity at greater than 2000 gpm. The output file for this
case is presented in Appendix A.1.
11.2 Cooling Case
After the lower containment is flooded the system will be realigned to its cooling
function. Figure 8 shows the valve alignment for this function. From the flooding
function alignment, Valves V7 and V8 are opened, Valves V5 is closed and the pump
inside containment is started.
39
Starting with the base model, pumps J1 and J5 remain on. Pipe P7 is closed to simulate
the closing of Valve V5. This closes the flowpath for the external pump (J5) from
delivering to containment. Figure 11 shows the model aligned for the cooling function.
Figure 11 Cooling Case Model
The output from this case shows that the system can be successfully aligned to provide
the required cooling as described in this report. Table 9 summarizes the flowrates and
heat removal in the Cooling Case. The complete output file for this case is presented in
Appendix A.2.
Table 9 Cooling Case Performance
Mass Flow (kg/s)
Flow (gpm)
Inlet Temperature (°F)
Outlet Temperature (°F)
Heat Transfer (BTU/s)
Heat Transfer (kW)
Loop
Containment
Cooling
130
130
2216
2062
280
80
188
173.2
-28586
28586
-30158
30158
40
12. CONCLUSION
This report provides a conceptual design for a severe accident decay heat removal
system in a PWR nuclear power plant. In the most severe of severe accidents where the
reactor vessel fails and failed fuel falls to the reactor cavity floor, this system can help
the plant cope with the accident.
The designed system consisting of two pump and a plate heat exchanger will first flood
the reactor vessel cavity and the assumed ex-vessel corium to provide the initial cooling.
Once the lower containment is flooded, the system will circulate the flooded pool in
lower containment through a heat exchanger and remove heat from containment. The
cases run within this report show the system can remove the required heat for the
nominal conditions presented in this report.
The system designed in the report does nominal inputs such as core power level. If this
design is implemented in a PWR plant, the design will likely need to be altered to the
specific plant design it is being implemented for.
41
13.REFERENCES
1. World
Nuclear
Association,
Fukushima
Accident,
http://www.worldnuclear.org/info/Safety-and-Security/Safety-of-Plants/Fukushima-Accident/
2. INPO 11-005 Addendum August 2012, Lessons Learned from the Nuclear Accident
at the Fukushima Daiichi Nuclear Power Station.
3. Sevón, T., (2005) Molten Core . Concrete Interactions in Nuclear Accidents. Theory
and Design of an Experimental Facility.
4. Severe Accident Management Guidance Technical Basis Report, Volume 1:
Candidate High-Level Actions and Their Effects. EPRI, Palo Alto, CA: 2012.
5. Zhong, H., (2011) A Study on the Coolability of Ex-vessel Corium by Late Top Water
Flooding.
6. Serth, Robert W. Lestina, Thomas G. Process Heat Transfer Principles, Application
and Rules of Thumb, 2nd Edition, 2014.
7. Flow of Fluids through Valves, Fittings, and Pipe, CRANE Technical Paper No.
410, 1988.
8. 10 CFR 50 Appendix A, General Design Criteria for Nuclear Power Plants,
http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-appa.html
9. Incorpera, DeWirr, Bergman, Lavine, Fundamentals of Heat and Mass Transfer, 5th
Edition; 2002
10. Kreith, Frank Goswami, D. Yogi. (2007), Handbook of Energy Efficiency and
Renewable Energy. Taylor & Francis.
42
APPENDIX A – FATHOM CODE FILES
A.1
Flooding Case
A.1.1 Flooding Case Input
43
44
45
46
47
48
A.1.2 Flooding Case Output
49
50
51
A.2
Cooling Case
A.2.1 Cooling Case Input
52
53
54
55
56
57
A.2.2 Cooling Case Output
58
59
60