Design of Hardened Containment Vent Systems for Decay
Heat Removal and Severe Accident Conditions
by
Matthew James Fallacara
An Engineering Project Submitted to the Graduate
Faculty of Rensselaer Polytechnic Institute
in Partial Fulfillment of the
Requirements for the degree of
MASTER OF ENGINEERING IN MECHANICAL ENGINEERING
Approved:
_________________________________________
Ernesto Gutierrez-Miravete, Project Advisor
Rensselaer Polytechnic Institute
Hartford, Connecticut
August, 2013
i
CONTENTS
CONTENTS ...................................................................................................................... ii
LIST OF TABLES ............................................................................................................ iv
LIST OF FIGURES ........................................................................................................... v
LIST OF ACRONYMS .................................................................................................... vi
ACKNOWLEDGMENT ................................................................................................. vii
ABSTRACT ................................................................................................................... viii
1. Introduction and Background ...................................................................................... 1
1.1
Review of the Events at Fukushima Dai-ichi ..................................................... 1
1.2
Boiling Water Reactors ...................................................................................... 2
1.3
Current HCVS Licensing Requirements ............................................................ 6
1.4
Potential Future HCVS and FCVS Licensing Requirements ............................. 9
1.5
HCVS/FCVS Installed in Foreign Countries ..................................................... 9
2. Theory and Methodology .......................................................................................... 11
2.1
HCVS Design Requirements for EA-13-109 ................................................... 11
2.2
HCVS Wetwell Sizing for Decay Heat Steam/Energy Release ....................... 14
2.2.1
Mass Flow Rate Requirements ............................................................ 17
2.2.2
Required HCVS Pipe Sizing ................................................................ 19
2.3
Severe Accident Conditions in Containment ................................................... 21
2.4
Drywell Venting ............................................................................................... 27
2.5
Effects of Filtering ........................................................................................... 28
2.5.1
Decontamination Factors for Suppression Pools ................................. 29
2.5.2
Engineered Filter Designs .................................................................... 30
3. Results and Discussion .............................................................................................. 33
3.1
Design of HCVS for Decay Heat Removal and Severe Accidents .................. 33
3.1.1
Non-Severe Accident Calculations ...................................................... 35
3.1.2
Severe Accident Calculations .............................................................. 37
ii
3.2
Discussion of Filter Benefits ............................................................................ 41
3.3
Additional Considerations for Severe Accidents ............................................. 43
4. Conclusion ................................................................................................................. 46
5. References.................................................................................................................. 47
6. Appendix.................................................................................................................... 50
6.1
Mass Flow Rate Requirements Spreadsheet – Non-Severe Accidents ............ 50
6.2
Required HCVS Pipe Sizing Calculation Spreadsheet-Non-Severe Accidents 52
6.3
Mass Flow Rate Requirements Spreadsheet – Severe Accidents .................... 54
6.4
Required HCVS Pipe Sizing Calculation Spreadsheet – Severe Accidents .... 55
6.5
Properties of Hydrogen Vapor – Excerpted from [26] .................................... 57
6.6
Properties of Hydrogen – Interpolation Table ................................................. 58
6.7
Specific Volume Chart - Hydrogen.................................................................. 59
6.8
Enthalpy Chart - Hydrogen .............................................................................. 60
iii
LIST OF TABLES
Table 1: Decay Heat Power Following Shutdown [4] ..................................................... 15
Table 2: Hydrogen Flammability Limits [17] ................................................................. 24
Table 3: Mass Flow Rate Requirements .......................................................................... 35
iv
LIST OF FIGURES
Figure 1: GE Mark I Containment, Modified from [3] ..................................................... 2
Figure 2: Simplified Diagram of a BWR [11] ................................................................... 3
Figure 3: GE Mark II Containment, Modified from [2] .................................................... 5
Figure 4: Mark I Containment with Wetwell and Drywell HCVS [8] ............................ 16
Figure 5: Volume % Hydrogen vs. % Metal-Water Reaction Following a LOCA [17] . 26
Figure 6: IMI Nuclear (CCI) Filter Technology [2 - Enclosure 4] .................................. 31
Figure 7: Simplified HCVS/FCVS Design [2 - Enclosure 4] .......................................... 33
Figure 8: Suppression Pool Bypass for Mark II [19] ....................................................... 44
v
LIST OF ACRONYMS
AC – Alternating Current
BWR – Boiling Water Reactor
BWROG – BWR Owners’ Group
CNS – Cooper Nuclear Station
CS – Core Spray
DC – Direct Current
DF – Decontamination Factor
ECCS – Emergency Core Cooling Systems
EDG – Emergency Diesel Generator
EPRI – Electric Power Research Institute
EPU – Extended Power Uprate
FCVS – Filtered Containment Vent System
GE – General Electric
GL – Generic Letter
HCVS – Hardened Containment Vent System
HPCI – High Pressure Coolant Injection
IPE – Individual Plant Examination
ISG – Interim Staff Guidance
LOCA – Loss of Coolant Accident
LPCI – Low Pressure Coolant Injection
LWL – Low Water Level
MWt – Megawatts Thermal
NRC – Nuclear Regulatory Commission
PCPL – Primary Containment Pressure Limit
PR – Pressure Ratio
PWR – Pressurized Water Reactors
RCIC – Reactor Core Isolation Cooling
RHR – Residual Heat Removal
SBO – Station Blackout
SRV – Safety/Relief Valve
TMI-2 – Three Mile Island Unit 2 Nuclear Plant
vi
ACKNOWLEDGMENT
I would like to thank Dr. Ernesto Gutierrez-Miravete for his guidance and help in
completing this project and the Rensselaer professors and staff for assisting me complete
this degree. I would also like to thank Zachry Nuclear Engineering, Inc. for their support
and financial assistance in helping me obtain my Master of Engineering. In addition, I
would like to thank Tom Driscoll for his willingness to spend a weekend reading and
commenting on this paper. Finally, I would like to thank Andie and my family for their
assistance and encouragement.
vii
ABSTRACT
Hardened Containment Vent Systems (HCVS) for Boiling Water Reactor (BWR) Mark I
and II containments are designed and sized to prevent potential containment failure due
to overpressurization during accident conditions. This paper calculates the energy from
both decay heat following reactor shutdown and zirconium/steam reactions in terms of a
mass flow rate for steam and gas at the containment design pressure limit. In addition,
the allowable mass flow rates for different sized piping systems with a range of total
resistance coefficients are calculated. This paper then compares the results of the two
different mass flow rate calculations and determines required HCVS design sizing.
HCVS for both wetwell and drywell venting conditions are evaluated. Finally, this paper
provides discussion on the use of wet and dry filters in conjunction with wetwell venting
through suppression pools.
viii
1. Introduction and Background
1.1 Review of the Events at Fukushima Dai-ichi
On March 11, 2011, a magnitude 9.0 earthquake struck off the coast of Japan.
This earthquake resulted in a tsunami, approximately 45 feet tall, striking the Japanese
coast approximately 40 minutes following the earthquake. The earthquake and
subsequent tsunami caused extensive damage across northeastern Japan, including to the
Fukushima Dai-ichi nuclear power plant. At the time of the earthquake, Fukushima
Nuclear Units 1, 2, and 3 were in operation and Units 4, 5, and 6 were shutdown for
routine refueling and maintenance outages. Following the earthquake, the operating units
automatically shut down and offsite power was lost for all units. The emergency diesel
generators (EDG) at all of the units started, as designed, to provide alternating current
(AC) power to critical systems to cool and shutdown the cores. However, the tsunami
damaged the EDGs at Units 1-5 such that all AC power was lost, and the units entered a
condition known as Station Blackout (SBO). Note that one of the EDGs for Unit 6 was
still operational following the tsunami. In addition to the loss of AC power, direct
current (DC) power for all of the units was lost early on following the
earthquake/tsunami [1].
Due to the loss of all power, the cores and containments at Units 1, 2, and 3
could not be cooled. This led to the rise of pressure and temperature inside the primary
containment, which includes the drywell, wetwell, and interconnecting vent piping, for
the applicable units (refer to Figure 1). These units were equipped with hardened
containment vents from the wetwell, which were designed to open and relieve high
pressure/temperature in containment to prevent containment failure. Operators at Units
1, 2, and 3 considered operating each unit’s hardened containment vent; however, the
systems were not able to be operated remotely, due to the loss of all power, or manually,
due to temperatures and radioactivity levels near the systems. This lead to the pressure
and temperature inside the three units’ containments exceeding their design limits, which
lead to core damage, high radiation levels, hydrogen production, and containment
failure. The containment failure allowed for the uncontrolled release of radiation to the
environment. In addition, containment failure led to the release of hydrogen to other
1
buildings, where it was able to create flammable mixtures and cause detonations and
additional damage to the units [1].
Figure 1: GE Mark I Containment, Modified from [3]
1.2 Boiling Water Reactors
There are two basic types of nuclear reactors that are utilized in the United States
to produce electricity: BWRs and Pressurized Water Reactors (PWR). This paper will
focus only on BWRs; see the NRC website for information on PWRs [11]. Figure 2
below shows a simplified BWR. In a BWR, the reactor vessel houses the fuel rods,
which is located inside of a structure called containment. The fuel rods produce a
2
substantial amount of power in the form of heat through a process known as nuclear
fission. Fission is the process of splitting atoms. The fuel rods consist of pellets
contained in metal tubes, which consist of an approximate mix of 3.5% Uranium-235
and 96.5% Uranium-238. Uranium-235 is considered a fissile material, in that it is
capable of sustaining a chain reaction. Neutrons are utilized to split or fission a
Uranium-235 atom, and in the process, additional neutrons are released in conjunction
with the daughter products of the atom and heat/energy. These neutrons then find other
Urainum-235 atoms and continue the fission process, which is self-sustaining [5].
The reactor vessel for a BWR is filled with a water-steam mixture. Water (also
called coolant) enters the reactor vessel and is directed to the bottom of the fuel rods, or
reactor core. The coolant flows up through the reactor core and is heated to form steam
(thereby “cooling” the core). The steam exits the top of the reactor through main steam
piping and flows out of containment and into a turbine. The turbine converts the thermal
energy of the steam into mechanical energy, through the spinning of the turbine blades.
The turbine then turns the electrical generator and creates electricity. The steam that has
passed through the turbine is condensed back into water in the condenser, and finally is
pumped back into the reactor [5].
Figure 2: Simplified Diagram of a BWR [11]
3
In a nuclear power plant, containment is extremely important for protecting the
health and safety of the public, since it provides the last barrier against the uncontrolled
release of radiation. For BWRs, containment has three functions: 1) condense steam and
contain fission products released during a loss of coolant accident (LOCA) so that the
radiation release rates do not exceed the government limits, 2) provide a heat sink for
certain safety related equipment, and 3) provide a source of water for cooling systems
[15]. The major components of BWR primary containment are the drywell, suppression
pool/wetwell, and the interconnecting vent pipe system (see Figure 1). Secondary
containment consists of the reactor building located around primary containment;
however, it does not have the same functions as primary containment. This paper refers
to primary containment as containment and secondary containment as the reactor
building. The drywell houses the reactor vessel, piping, recirculation pumps and other
equipment. Its purpose is to contain steam released during a LOCA and direct it to the
wetwell, along with preventing any radioactive material from leaking out. The wetwell
consists of a free volume that is approximately half-filled with water and located lower
than the drywell. In a Mark I containment, the wetwell (also called the torus) is toroidal
or donut shaped, with the bottom of the drywell located above the center. In a Mark II
containment, the wetwell is cylindrical, located directly below the drywell (see Figure 3
below). The wetwell’s purpose is to condense steam from a LOCA, along with
preventing any radioactive material from leaking out. The interconnecting vent system
provides the path that connects the drywell and the wetwell. This allows for steam and
water to pass to the suppression pool for condensing and allows non-condensable gases
to be released to the wetwell gas space. In a Mark I, eight large vent pipes connect the
drywell and wetwell. Inside the wetwell, the pipes exhaust into a vent header. This
header extends circumferentially around the torus. Downcomer pipes extend from the
bottom of this header into the pool. In a Mark II containment, downcomer pipes connect
the drywell and wetwell [15]. In addition, there is piping that branches off the main
steam piping in containment and is routed to the wetwell. For conditions of high
pressures in the reactor vessel, the safety relief valves (SRV) in this branch piping will
open and allow steam to pass directly to the wetwell and suppression pool for pressure
relief.
4
Figure 3: GE Mark II Containment, Modified from [2]
BWRs have many safety systems that are utilized to remove heat either when the
reactor is shutdown or during accident conditions. During normal plant operation, heat is
removed from the reactor by the generation of steam, which exits the reactor vessel, and
is utilized to create electricity. When the reactor is shutdown, the fission process stops;
however, the fuel along with the metal in the reactor vessel is still extremely hot.
Therefore, the residual heat removal (RHR) system is utilized to remove this heat, called
decay heat. The water in the reactor vessel is circulated through RHR heat exchangers,
which cool the water. The reactor core isolation cooling (RCIC) system provides
makeup water to the reactor vessel for core cooling during abnormal or accident
5
conditions, when the main steam piping is isolated or when the normal supply of water
to the reactor vessel is lost. Water, from both dedicated tanks and the suppression pool,
is available to be pumped into the reactor vessel from RCIC. Turbine driven pumps are
utilized to pump the water into the reactor vessel and the pumps are powered by steam
supplied from the main steam lines. This steam exhausts to the wetwell. Finally, the
emergency core cooling systems (ECCS) provide core cooling under LOCA conditions.
There are multiple systems that are part of the ECCS: high pressure coolant injection
(HPCI), low pressure coolant injection (LPCI), and core spray (CS). HPCI supplies
makeup water to the reactor vessel under small and intermediate sized breaks or LOCAs,
where the pressure in the reactor vessel remains high. Similar to RCIC, HPCI utilizes
turbine driven pumps. LPCI utilizes many of the same components as RHR and provides
cooling water from the suppression pool to the reactor vessel under large break LOCAs
where the pressure in the reactor vessel quickly decreases. In addition to injection into
the reactor vessel, LPCI provides spray into containment (both drywell and wetwell
sections), which reduces pressure and temperature inside containment. CS pumps water
from the suppression pool and sprays it onto the top of the reactor core to cool the fuel
[5].
All six of the Fukushima reactors are BWRs, and here in the United States,
approximately one-third of all operating reactors are BWRs. Fukushima Dai-ichi Units
1, 2, 3, 4, and 5 use the General Electric (GE) Mark I containment design, which is the
same design as 23 of the nuclear power plants in the United States. This design has the
smallest containment volume of all operating plants in the United States. In addition,
Fukushima Unit 6, along with eight of the plants in the U.S., is a GE Mark II
containment design, which is slightly larger in containment volume than the Mark I
design.
1.3 Current HCVS Licensing Requirements
Following the accident at the Three Mile Island Unit 2 Nuclear plant (TMI-2) in
1979 in Pennsylvania, the NRC and the nuclear industry evaluated many changes to
nuclear plants across the country. One topic that was evaluated was how to reduce the
vulnerability of BWR Mark I containments to severe accident challenges (i.e., accidents
6
with core damage). Many of the areas that were investigated were allowed to be
evaluated by the individual plants under the Individual Plant Examination (IPE)
program; however, one area that the NRC felt should be required by all Mark I units was
the installation of hardened wetwell vents. The NRC only singled out Mark I nuclear
plants for installation of hardened containment vent because they have the smallest
containment free volume. The free volume is the space that would be pressurized during
a LOCA or severe accident. The average containment free volume for a Mark I reactor is
between 250,000 and 300,000 ft3 [2 – Enclosure 2]. A Mark II design has an average
containment free volume slightly larger than a Mark I (up to 25%) [16]. However, all of
the other types of reactors operated in the United States have containment free volumes
that are greater than 1,000,000 ft3. The BWR Mark III reactor averages a containment
free volume of approximately 1,500,000 ft3, and the large dry PWR containments (which
is a majority of containments in the United States), have containment free volumes of
approximately 2,000,000 ft3 [2 – Enclosure 2]). The type of reactor does not determine
the power level of the reactor. BWR Mark I power levels range from some of the
smallest (slightly less than 2000 MWt) to some of the largest (3500 to 4000 MWt).
Therefore, since the power levels are comparable between the different reactor types, an
accident at a Mark I reactor could pressurize containment quicker and at a higher level
than the other reactor types, which could create a greater chance of containment failure
due to overpressurization.
Although not issued as an Order, the NRC prepared Generic Letter (GL) 89-16
[6], which provided instructions along with an example of an already installed hardened
wetwell vent to all U.S. nuclear plants with Mark I containments. The major reason for
the request for installing a hardened wetwell vent was to avoid exceeding the primary
containment pressure limit (PCPL) during severe accidents. Prior to GL 89-16, all of the
plants had procedures that instructed them to vent containment if the PCPL limit was to
be reached; however, it would be through non-hardened/duct vents that would not be
able to handle the high pressures, potentially fail, and cause an uncontrolled release of
radiation. Following the issuance of this GL, all facilities with Mark I containments
installed a hardened wetwell vent. The NRC found all of the installed designs to be
acceptable and did not issue an Order with detailed design requirements. The NRC did
7
not request that owners with Mark II containments provide feedback to GL 89-16, nor
did they recommend installation of hardened wetwell vents due to their larger volume
containments [6]. The NRC only requested that owners evaluate the use of hardened
vents as part of their IPE program. As part of this program, three of the eight Mark II
reactors installed a hardened vent [16].
As evaluated following the events at Fukushima, the installed hardened vents
could have been quite beneficial in mitigating the consequences of the event if the vents
were able to be operated. The NRC issued Order EA-12-050 [1] in March of 2012 along
with an Interim Staff Guidance (ISG) on the Order (JLD-ISG-2012-02 [7]) in August of
2012 requiring all facilities with Mark I and Mark II containments in the United States to
install new HCVS or justify the installation of their existing hardened vents, to meet the
new design requirements. The purpose of these design requirements is to allow for the
removal of decay heat, preventing core damage, and maintaining control of containment
pressure following an event that causes a loss of heat removal systems, such as SBO.
Following the issuance of EA-12-050, the NRC continued to evaluate HCVS and
how to maintain containment integrity if the vents are required to be used during severe
accidents (i.e., accidents with damage to the core). SECY-12-0157 [2] was issued by the
NRC Staff, which evaluated additional requirements for venting. The options listed for
consideration were as follows: 1) maintain status quo as documented in EA-12-050 and
not issue any additional requirements, 2) require HCVS be capable of operating under
severe accident conditions, 3) install engineered filters as part of HCVS, and/or 4)
develop severe accident confinement strategies (i.e., venting in conjunction with the use
of other heat removal systems such as RCIC or ECCS). The five NRC Commissioners
determined that Option 2 would be a requirement at the current time and additional
options would require further evaluation (see Section 1.4) [28]. Order EA-13-109 was
issued in June of 2013 and superseded EA-12-050. All of the original requirements from
EA-12-050 are included in EA-13-109 along with additional requirements to ensure that
venting functions are available during severe accident scenarios. A severe accident is
classified as an accident involving extensive core damage, including accidents involving
a breach of the reactor core by the molten core debris. During these accidents, expected
8
conditions in containment include elevated pressures, temperatures, radiation levels, and
combustible gas concentrations (including hydrogen).
1.4 Potential Future HCVS and FCVS Licensing Requirements
As discussed above, Option 3 (engineered filters) and Option 4 (confinement
strategies) from SECY-12-0157 are not required to be addressed at the current time by
BWR Mark I and II nuclear plants. The major difference between the existing
requirements in EA-13-109 and Option 3 from SECY-12-0157 is the installation of an
engineered filter that is a capable of reducing the release of radioactive materials passing
through the HCVS by certain standards. This is called a filtered containment vent system
or FCVS. Different types of radioactive releases, mainly aerosols and iodine, would be
required to be reduced by certain decontamination factors (DF). A number of companies,
including Areva and Westinghouse, have filtering systems that would be capable of
being installed to meet additional requirements issued by the NRC.
Option 4 (confinement strategies) is different from the rest of the options since it
would involve establishing technical acceptance criteria and not just installation of new
equipment. Options 2 and 3 would have defined attributes (venting capability to not
exceed maximum containment pressure and DF). Confinement strategies involve
extensive interactions with all stakeholders (plants) and different strategies would most
likely be employed by different plants depending on their best solution. This solution
involves utilizing different heat removal systems (RCIC and ECCS) at different times
following an accident and certain actions may be different for different types of
accidents. Since very few plants could be considered the same, and based on the
availability of public information on specific plants, this option is only discussed in this
paper in general terms and is not evaluated.
1.5 HCVS/FCVS Installed in Foreign Countries
As part of the preparation of SECY-12-0157, the NRC researched containment
venting strategies for nuclear plants in other countries. It was found that many countries
in Europe, including Sweden, Finland, Germany, France, Switzerland, and the
Netherlands, installed FCVS following the accidents at TMI-2 and Chernobyl (in 1986
9
in Ukraine). These countries determined that based on operating experience from severe
core damage, FCVS were necessary to increase defense in depth in regards to
minimizing accident fission product releases, along with the management of the
production of hydrogen. These countries deemed the installation of engineered filters as
essential safety enhancements. The NRC and the nuclear industry in the U.S. studied
filtered vents, but they determined the installation of filtered vents were not considered
essential safety enhancements; therefore, not required to be installed. Vents installed
following issuance of GL 89-16 were intended to help prevent core damage from
occurring and not mitigating core damage if it occurred as part of a severe accident.
Following Fukushima, many countries have determined that FCVS will be
required to be installed in order to mitigate the events of severe accidents and limit the
potential release of radiation following a severe accident. These countries include Japan,
Taiwan, Belgium, Romania, South Korea, and Canada. Note that one Canadian plant
already installed a FCVS in 2007 prior to Fukushima. Other countries, including the
United States, Mexico, China and others are still evaluating what types of venting
strategies will be required to be installed, including potential for FCVS. As part of the
evaluation process, the NRC has performed cost-benefit analyses based on the cost of
filters, probability for a severe accident, and cleanup cost [2 – Enclosure 2]. Many
foreign countries did not perform cost-benefit analyses on FCVS on which strategies to
implement.
10
2. Theory and Methodology
2.1 HCVS Design Requirements for EA-13-109
The current licensing requirements for HCVS for BWR Mark I and II nuclear
plants in the United States are listed in Order EA-13-109. As discussed above, Order
EA-12-050 was originally issued and required plants to install HCVS to prevent core
damage when heat removal capability was lost due to conditions such as extended loss
of electric power. However, this Order was superseded by EA-13-109 so that the HCVS
will not only function during these conditions, but also during severe accidents (i.e.,
when core damage has occurred). All of the original requirements listed in EA-12-050
are reflected in EA-13-109, in addition to the new requirements for severe accidents.
This paper will first examine the requirements of the design of HCVS.
Mark I and II reactors are obligated to comply with the listed requirements in
Attachment 2 to EA-13-109. These requirements are broken down into two phases. In
Phase 1, licensees are required to install a HCVS from the wetwell that can provide
venting capability during severe accidents. In Phase 2, licensees are required to either
install a HCVS to provide the same venting capabilities during severe accident
conditions from the drywell or develop and implement a strategy that makes it unlikely
that a licensee would need to vent from the drywell during a severe accident.
The first grouping of functional requirements for Phase 1 for HCVS (Section 1.1
of Attachment 2 to EA-13-109) discuss the operational requirements of the new HCVS.
These focus on the actions of nuclear plant operators. The system is required to be
accessible and functional for conditions during times when there is a severe accident and
potential loss of electric power. If a nuclear plant operator is required to perform certain
actions in order to make the system function, minimal actions should be necessary to
lower the risk of failure. Operators need to have the ability to operate the system during
these accident conditions. Venting from a remote location, such as the Control Room, is
preferable and allows an Operator to be more responsive, since this is where Operators
perform the majority of actions for operating nuclear plants. If the valves will be
operated manually, the potential environmental conditions, including radiation and
temperature, need to be evaluated to ensure that Operators can access the areas. At
11
Fukushima, the operators could not utilize their hardened containment vents. After the
plants lost all AC and DC power, alternate sources of power were not available to
operate the valves remotely in the system. In addition, they could not operate the valves
manually due to radiation and temperature [1].
The second grouping of functional requirements for Phase 1 (Section 1.2 of
Attachment 2 to EA-13-109) focus on the design features of the HCVS. The first major
design feature is that HCVS shall have the capacity to vent steam/energy equivalent to
one percent of the licensed thermal power. This vent must be able to restore and
maintain containment pressure below the design pressure limit of containment. The
discharge of the vent needs to be such that it is released above main plant structures and
minimizes any potential cross flow of vented fluids for sites that have multiple units.
These two objectives are in place so that venting does not cause adverse effects
elsewhere onsite. Venting above the main plant structures will limit the potential impact
on personnel at the site. At Fukushima, there was an explosion in Unit 4 and hydrogen
leaking from Unit 3 to Unit 4 is believed to have caused this explosion [7]. The HCVS
shall be powered so that it can be operated from a remote location and it shall also have
the ability to be operated manually. As discussed above for Operator actions, it is much
preferred if the HCVS can be operated from a remote location, since it will keep the
Operators out of more hazardous conditions. However, having two means of operation
will provide defense in depth so that if for some reason the valve cannot be remotely
operated, another option is available. The manual operator may require shielding to be
installed so personnel can access the required locations. Also similar to above,
instrumentation shall be installed such that not only is the status of the vent system
known (opened or closed), but also potential radiation that is released from the vent is
known. In addition, the system shall be designed such that is can withstand the effects of
severe accidents. The system shall be able to cope with high temperature, pressure, and
radiation while venting steam, hydrogen and other non-condensable gases that may be
generated during accidents where the core is damaged. For combustible gases, such as
hydrogen, the flammability limit shall not be reached while venting or the system shall
be able to withstand the dynamic loading resulting from detonation or deflagration.
12
Finally, the HCVS shall be inspected, tested, and maintained such that it will be able to
function if required to during accident conditions.
The HCVS is required to meet two quality standards for Phase 1. The piping for
the system will be routed from the wetwell to the outside atmosphere. All piping that
penetrates containment is required to meet special conditions, such that it cannot fail and
cause a breach/uncontrolled opening of containment. This requires that there are two
isolation barriers installed as close to the containment penetration as possible, so that
containment can be isolated. There are two barriers in case one barrier fails, and the
barriers are subject to stringent inspection and testing criteria. The HCVS will be no
different from any other containment penetration, and the system up to and including the
second containment isolation barrier needs to meet these requirements. In addition, all of
the HCVS components are required to be designed such that they can withstand a
seismic event (i.e., earthquake) and still be functional. At Fukushima, the earthquake
struck the plant first. This did not cause the SBO condition, because the EDGs (which all
nuclear plants in the United States have) started as designed and provided the units with
AC power. However, the tsunami struck the plant following the earthquake and the put
the units into the SBO condition. Therefore, the HCVS has to be able to withstand
seismic conditions.
For Phase 1, there are two programmatic requirements. Each plant will be
required to develop, implement, and maintain procedures that are necessary for the
operation of the HCVS. In addition, operators will be required to be trained on how to
operate the system through all situations. These will not be discussed further in this
paper.
For Phase 2, if a HCVS is installed from the drywell, the system will be required
to meet all of the conditions listed for the wetwell HCVS. In addition, this vent path
must be capable of venting the drywell atmosphere, which may be different from the
wetwell atmosphere. If it is determined that a drywell vent is not required, it will then
need to be proven why. Any instrumentation and equipment that will be required to
accomplish this task will need to be installed and procedures developed, along with
training.
13
2.2 HCVS Wetwell Sizing for Decay Heat Steam/Energy Release
One of the major requirements for HCVS is that they are capable of relieving the
energy inside containment in order to prevent or to mitigate the consequences of a severe
accident. Requirement 1.2.1 of EA-13-109 states that HCVS for both Mark I and II
containments shall have the capability to vent steam/energy equivalent to one percent of
licensed/rated thermal power (unless a lower value is justified by analyses) and be able
to maintain containment pressure below the primary containment design pressure. This
was an original requirement for EA-12-050 and JLD-ISG-2012-02 provided additional
guidance on meeting this Requirement. Note, an ISG has not yet been prepared for EA13-109, and one is expected by October 31, 2013 for Phase 1 requirements and another
by April 30, 2015 for Phase 2 requirements. The HCVS shall be capable of keeping the
containment pressure below both the PCPL and primary containment design pressure.
These two limits may be different and the vent will be required to be capable of limiting
containment pressure so neither limit is exceeded. A limit of one percent of the rated
thermal power was chosen based on studies that have shown that suppression pools are
capable of absorbing the decay heat generated during the first three hours following
shutdown. Decay heat continues to decrease well under one percent as time passes.
Table D.1 of ANSI/ANS-5.1-2005 [4] presents the approximate decay heat power
relative to the operating power following the shutdown of a reactor (partial shown below
in Table 1). At 1.00 x 104 seconds (2.78 hours) following reactor shutdown, the decay
heat power is approximately 1.04%.
14
Table 1: Decay Heat Power Following Shutdown [4]
Time After
Shutdown
(Seconds)
Time After
Shutdown
(Hours)
%
Reactor
Power
1.00 x 100
1.00 x 101
1.00 x 102
1.00 x 103
1.00 x 104
1.00 x 105
1.00 x 106
1.00 x 107
1.00 x 108
1.00 x 109
2.78 x 10-4
2.78 x 10-3
2.78 x 10-2
2.78 x 10-1
2.78 x 100
2.78 x 101
2.78 x 102
2.78 x 103
2.78 x 104
2.78 x 105
7.669%
5.442%
3.426%
2.077%
1.044%
0.5838%
0.2633%
0.09436%
0.03891%
0.01105%
The ISG states that Licensees may use a value lower than one percent if analyses
can justify that primary containment and the suppression pool can handle it. In addition,
the location where venting will occur needs to be considered (drywell vs. wetwell).
Finally, the pressure difference between the drywell and the wetwell needs to be taken
into account.
As part of this section, a generic HCVS for the wetwell shall be designed. This
design will need to perform two calculations: 1) determine the mass flow rate required to
be vented in order to remove a minimum of one percent of the rated thermal power from
a nominal sized reactor and 2) determine the size of the pipe to remove the required
mass flow rate. It is assumed that the suppression pool is capable of absorbing the decay
heat following the shutdown of the reactor until the decay heat power level reaches one
percent.
For the purposes of this paper, design information from the Cooper Nuclear
Station (CNS) in Brownville, NE will be utilized to size a HCVS. CNS is a BWR Mark I
nuclear plant, making it very similar to the Fukushima units that sustained extreme
damage. CNS installed a hardened vent from its wetwell in the early 1990’s following
the issuance of GL 89-16. All of the information that will be utilized is taken from
publically available records. Some of the information is taken from the response of
Cooper to EA-12-050 and the NRC [9] (Note: responses for how plants will meet the
15
requirements of EA-13-109 will not be prepared until after the ISGs are issued). The
calculations will be compared to results from this response for the wetwell. The
calculations in this paper do not constitute a calculation or design basis for CNS, but are
used to demonstrate the approximate sizing of a HCVS for a nuclear plant the size of
CNS. Figure 4 below provides a depiction of a Mark I containment with the HCVS
installed coming off the top of the wetwell. A drywell HCVS is shown for reference for
Section 2.4.
Figure 4: Mark I Containment with Wetwell and Drywell HCVS [8]
16
2.2.1
Mass Flow Rate Requirements
The current power level for CNS is 2419 MWt [11]; however, CNS is preparing
to increase the power level of the reactor as part of an Extended Power Uprate (EPU) to
approximately 2915 MWt [9]. This value is equal to the thermal power generated by the
reactor during one hour. One percent is equivalent to 29.15 MWt, which is equal to 9.95
x 107 BTU using a conversion factor of 1 MWt = 3.412 x 106 BTU. This is the required
power that the HCVS will be required to vent out of containment.
As discussed in the JLD-ISG-2012-02, the containment vent shall be sized to be
rated for the lower of the PCPL and primary containment design pressure. For CNS, the
primary containment design pressure (56 psig) is less than the PCPL (62 psig), and
therefore will be utilized [9].
Steam/gases will be vented to the wetwell through two possible routes: 1) from
the reactor vessel/main steam piping, through the SRVs, and into the bottom of the torus
or 2) through the interconnecting piping of the drywell and torus, into the header, and
through the downcomers. The first scenario occurs as long as there is no LOCA or
breach of the reactor vessel and the second scenario occurs if a LOCA or breach of the
reactor vessel occurs. Both the downcomers and SRV discharge piping are lower in the
suppression pool than the inlet to the HCVS, which is designed to be at the top of the
torus. The pressure of the vented flow that enters the HCVS will be reduced due to the
static head of the water in the torus. The dimensions of torus and water levels are as
follows (reference dimensions and heights taken from [10]):

Minor diameter of the torus is 28.75 ft.

Low Water Level (LWL) is 12.583 ft above the torus bottom.

Bottom of downcomers is 9.583 ft above the torus bottom.

The assumed maximum water level in torus is 28 ft above the bottom of the
torus.

The SRV discharge piping is located at the bottom of the wetwell.
An assumed maximum water level is set at 28 ft, which is just below the top of the torus
and thus the HCVS entrance. This is necessary because if the torus was full of water, the
vent would not be able to properly vent steam and remove pressure (at which point
venting would only be able to be performed through a drywell vent – see Section 2.4).
17
For this calculation, three different scenarios will be considered. The first
scenario (Low) is when water in the torus is at LWL and venting is through the
downcomers. The static head of water would be the minimum (3 ft) when steam/energy
from the drywell enters the wetwell and the pressure at the downcomer exit is equal to
the drywell pressure. The second scenario (High) is if the water level is at the assumed
maximum, venting is again through the downcomers, and the pressure at the downcomer
exit is equal to drywell pressure, which will give a static head of 18.417 ft. Finally a
third scenario (Max) will consider if the water level in the torus is at the maximum and
the steam/gas enters the wetwell at the bottom through the SRV discharge piping. In this
scenario, the full height of the water (28 ft) reduces the pressure. Note the scenario of the
steam/gas entering at the bottom through the SRV discharge piping, while the wetwell is
at LWL is not evaluated; however, this scenario would develop a head between the Low
and High cases (12.583 ft). The reduction of the pressure that will enter the HCVS
entrance due to the static head of water in the torus is calculated as follows:
(1)
Where:
Pv = Pressure at HCVS entrance, psia
PD = Pressure of Drywell, psia
LT = Static Water Height in the Torus, ft
DW = Density of water at saturated conditions and pressure of Drywell,
lb/ft3
Mass flow will be relieved from containment using the following energy balance
relationship:
(2)
Where:
Q’ = 1% of rated thermal power, Btu
Q’V = rate of heat removal that is vented, Btu
Q’I = rate of heat injected into containment, Btu
w = flow rate (vented and injected), lb/hr
h = enthalpy at specific pressure/temperature (vented and injected), Btu/lb
18
For this paper, two different scenarios will be evaluated: 1) during venting, no flow from
any safety systems (RCIC and ECCS) is injected into containment and 2) during venting,
water from safety systems is injected into the drywell and/or reactor vessel. The
following assumptions are made as part of the mass flow calculations:

Vented steam is at saturated vapor conditions with a quality of 1.0 at the vent
entrance, not superheated.

No heat loss from the drywell or wetwell.

For Scenario 2 only, during venting, the volume of fluid in primary containment
(drywell and wetwell) remains constant. This means that WV = WI.

For Scenario 2 only, the injection fluid is water at 100°F. At this temperature, h I
= 68.05 Btu/lb. This is an assumed temperature based on potential water flow
from safety systems into containment. This water could either be from dedicated
tanks located outside of the reactor building, or it could be water pumped from
the suppression pool, cooled through heat exchangers and returned to primary
containment.
For Scenario 1, Q’I goes to zero and WV is solved based on the enthalpy of the steam
exiting through the HCVS at the exit pressure. For Scenario 2, the enthalpy of water
injected into containment is subtracted from the enthalpy of the steam exiting through
the HCVS, and WV is solved for.
2.2.2
Required HCVS Pipe Sizing
Once the mass flow rate is determined, the required pipe size for the HCVS can
be calculated. As part of this scenario, the pipe size will be calculated based upon a
range of total resistance coefficients (K) for the system. When steam is vented from the
HCVS, there will be a number of pressure losses from the exit of the wetwell until the
discharge. Through the HCVS, the pressure will drop as the steam passes through
valves, elbows, tees, and potentially, different sized pipes. Since the specific design of
the flowpath for the existing CNS HCVS pipe is unavailable, the pipe sizes will be based
upon a total resistance coefficient. In addition, as discussed in [9], the current CNS vent
is a 10-inch pipe. As part of this calculation, three different pipe sizes will be considered:
8-inch, 10-inch and 12-inch. Although CNS has a 10-inch line, 8-inch will be considered
19
since there are BWRs that installed this size piping (see Enclosure 1 to GL 89-16 for
discussion of the Pilgrim Nuclear Power Station hardened vent). A 12-inch pipe will be
evaluated in case the 10-inch pipe is not sufficient. All pipes will be designed as
standard schedule or thickness and carbon/stainless steel. Per ASTM A524 [12], piping
of this size is subject to hydrostatic pressures of 1200 psi and higher, so the maximum
pressure that will be vented through the HCVS will not be a concern.
The pipe sizing calculation will be calculated using the Modified Darcy Formula
from Crane Technical Paper No. 410 [13]:
(3) [13]
Where:
Y is the net expansion factor for compressible fluid
v is the specific volume of steam, ft3/lb
K is the total resistance coefficient for the HCVS system
d is the internal pipe diameter, inch
ΔP is the pressure difference, psig
The first objective is to determine if the flow through the HCVS will reach sonic
velocities when the steam exits the HCVS. Per A-22 of [13], the specific heat ratio for
water is approximately 1.3. The pressure ratio (PR) is required to be calculated and is
equal to the pressure drop in the vent piping, ΔP (entrance pressure minus atmospheric
pressure) divided by the absolute pressure (in psia) of the pressure at the entrance to the
vent, P’1.
A chart is then created using the information from the chart on page A-22 of [13]
for k = 1.3. Additional nominal total resistance coefficients are added (12, 14, 16, and
18) and the ΔP/P’1 and Y are interpolated. Then, K and Y are interpolated for the
calculated PRs. For these specific PRs for that size pipe, when the K value is greater
than the interpolated value, the fluid that exits the HCVS will be sub-sonic. When the K
value is less than the interpolated value, the fluid that exits the HCVS will be at sonic
velocity. If the steam will be exiting at sonic velocity, ΔP will be required to be
recalculated and multiplied by the factor ΔP/P’1 at that specific K value (which reduces
the ΔP). Once all of the parameters are known, the Modified Darcy Formula can be used
20
to calculate the flow capacity of each of the different sized pipes for different total
resistance coefficients.
Note the calculations for Mark II containments would follow the same
methodology and assumptions used for Mark I containments. The only major difference
would be specific information for the suppression pool depth and injection/discharge
elevations.
2.3 Severe Accident Conditions in Containment
One of the biggest concerns in a nuclear plant that occurs during severe accident
conditions is the production of hydrogen. Hydrogen has multiple concerns if generated
in containment. The first concern is it will cause an increase in pressure and energy in
containment. Up to 20% of the pressure in containment during severe accidents can be
the result of the production of hydrogen and other non-condensable gases [19].
Secondly, there is a big risk of potential combustion and detonations from hydrogen
reactions that can cause damage to containment, along with the equipment/components
inside, and potentially cause the uncontrolled release of radiation. At Fukushima, it is
believed that Units 1, 2, 3, and 4 all experienced detonations due to hydrogen [1 and 16].
Hydrogen Production
During severe accidents, the major source of hydrogen production is generated
from the reaction of steam and the fuel rod metal (zirconium) that contains the fuel
pellets. When the plant is normally operating, the core is surrounded by water (Note the
transition from water to steam occurs above the fuel rods in the upper part of the reactor
vessel). In the event of a LOCA, the main purpose of the safety systems is to pump water
back into the reactor to keep the reactor core covered and cool. In the event of an
accident in which the safety systems cannot provide water to the core (similar to what
happened at Fukushima when all electric power was lost), the temperature of the water
in the reactor vessel will increase due to the decay heat of the fuel. Some of this steam
will be discharged through the SRVs to the suppression pool. While this is happening,
the water level in the reactor vessel will lower until the core becomes uncovered. When
this occurs, steam will react with the zirconium. In order to produce hydrogen from this
21
reaction, the zirconium temperature is required to be greater than 1832°F, which occurs
when the core is partially or fully uncovered. The chemical formula for the
steam/zirconium reaction is as follows from NUREG/CR-2726 [17]:
(4) [17]
The reaction is exothermic, which means that the energy that is released creates a chain
reaction, similar to the fission process. The creation of energy increases the temperature
of the zirconium in that area, which then causes the reaction to keep occurring.
In addition to the steam/zirconium reaction, there are other reactions in
containment that can produce hydrogen. Similar to zirconium, the steel that is in the
reactor, along with the reactor walls itself can produce hydrogen. For this to occur, the
temperature is required to be higher (~2200°F). This is much less likely than for
zirconium since the steel in the vessel does not contain fuel pellets, which can create
substantial amounts of heat and energy. Like the zirconium reaction, the process is
exothermic, but to a much smaller degree (~277.6 Btu/lbm produced or 10% the amount
of heat/energy released for zirconium/steam) [17].
Also, during both normal and accident conditions, hydrogen can be produced
through what is known as the radiolysis of water. The radiation in the reactor can cause
water molecules to break down into many different forms, but among these are hydrogen
and oxygen. Note that this happens at a much slower rate than the zirconium and steel
reactions with steam [17].
In addition, hydrogen is produced during potential molten core-concrete
interactions. If the fuel melts during a severe accident, penetrates the reactor vessel, and
falls to the drywell floor, it will react with the concrete. Steam and carbon dioxide will
be released from the concrete and will react with the metallic constituents of the melted
core to produce hydrogen and carbon monoxide. Other gases such as methane and
hydrocarbons can also be generated in a process called hydrogenation. Similar to
hydrogen, the other non-condensable gases that are produced will cause pressure
increases inside of containment [17].
Finally, three other sources of materials in containment can react with water and
generate hydrogen: zinc-based paint, aluminum, and galvanized material. When these
22
materials corrode due to the presence of water, hydrogen can be produced. These
reactions generate less hydrogen than the other reactions, and are slower than radiolysis
of water [17].
The calculations discussed above (Mass Flow Rate Requirements and Required
HCVS Pipe Sizing) will be performed again to account for the energy that could be
generated for the production of hydrogen. As discussed in [27], the energy from metalwater reaction can represent several times the energy from decay heat. Therefore, this
will most likely require larger piping systems (~18-inch diameter). This calculation will
only focus on the energy produced during the zirconium steam reaction (in addition to
the decay heat energy) since the energy from zirconium/steam reaction is ten times
higher than for the steel/steam reaction (2765 Btu/lb vs. 277.6 Btu/lb). In addition, the
steel/steam reaction will not produce nearly the amount of hydrogen or energy since the
temperatures of most of the steel in containment will not be capable of reaching 2200°F.
This calculation conservatively assumes that all of the available zirconium reacts,
creating approximately 4400 lbm of hydrogen along with the associated energy [17]. In
addition, the methodology above for venting decay heat only through the HCVS only
considered the properties of steam when venting; however, for this scenario, hydrogen
will also have a significant presence and its properties also need to be considered.
Hydrogen Combustion/Flammability/Deflagration/Detonation
Once hydrogen is produced in containment, the next concern is combustion. The
chemical formula for the combustion of hydrogen is :
(5) [17]
A great deal of energy is produced when hydrogen and oxygen are allowed to combust.
From this energy come both high pressure, which can cause damage to containment
along with equipment inside, and high temperature [17]. In order for substantial
combustion of hydrogen to take place, two conditions are required: 1) the mixture must
be flammable and 2) ignition source must be present. The flammability limits are
defined as a combustible mixture with a limiting molar fraction of fuel that will cause a
23
flame to propagate indefinitely [18]. Table 2 below shows the required concentrations of
hydrogen in air in order for hydrogen combustion to occur.
Table 2: Hydrogen Flammability Limits [17]
Hydrogen Concentration in Air
0% - 4%
4% - 14%
14% - 59%
59% - 75%
75% - 100%
Possible Reaction
Noncombustible
Combustible
Combustible (Possibly Detonable)
Combustible
Noncombustible
The 4% value is termed the lower flammability limit, in that this is approximately the
minimum hydrogen concentration that is required for combustion. The 75% is termed
the upper flammability limit. In addition to the hydrogen concentration needing to meet
these requirements, the oxygen concentration also is required to be approximately 5% by
volume or greater to enable the reaction.
Note that in an atmosphere that is altered by the presence of another gas that is
not combustible (such as steam, nitrogen, carbon dioxide, etc.), the lower flammability
limit will increase and the higher flammability limit will decrease. If a significant
volume of this third gas exists, there comes a point when the atmosphere is considered
inert. A mixture is considered inert if a flame is not able to propagate through it. During
a LOCA, the amount of steam produced from the break in the reactor vessel/main piping
could cause an atmosphere to be considered inert. However, this would be a temporary
condition, since following a LOCA, the steam will condense and the atmosphere would
no longer be inert. In addition, plants can purposefully add a third gas to inert an
atmosphere (see below on inerting containment) [17].
Once a flammable mixture is present, an ignition source is required for
combustion. Hydrogen/air mixtures only require a small input of energy in order to
spark. Sources such as electrical equipment or small static electrical charges are
sufficient. The closer the mixture is to one of the flammability limits, the more energy is
required for ignition. Similar to how a third gas can shrink the flammability region, a
third gas also increases the amount of energy required for ignition [17].
24
Combustion waves are classified as either deflagrations or detonations.
Deflagrations are waves/flames that travel at subsonic speeds relative to unburned gas
and then propagate through thermal conduction. Deflagrations that occur in containment
will create loads that are quasi-static on the containment walls and equipment. These
types of combustions cause gradual rise in pressure and temperature. In addition, all of
the hydrogen may not be burned in these types of combustions [18].
A hydrogen jet that is injected through a pipe (such as a HCVS) with steam could
create a turbulent diffusion flame. In a diffusion flame, the burning rate is controlled by
the rate of mixing oxygen and fuel (hydrogen). A mixture that has reached its
flammability limit can be ignited through either an outside ignition source or
spontaneous ignition temperature. The spontaneous ignition temperature is in the range
of 959-1076°F. These can be quite useful since the mixture of fuel and oxygen can be
more controlled, and deliberate ignition schemes can be used to initiate combustion [17].
Detonations are combustion waves that travel at supersonic speeds relative to the
unburned gas in front of it (can be as high as 6600 ft/sec). The compression of the
unburned gas by the shock waves raises the gas temperature high enough to initiate rapid
combustion. The detonation wave can create pressures that can be as much as 15 times
the initial pressure; however, when shock waves reflect off rigid walls, they can be
almost 30 times the original pressure, and cause catastrophic damage to containment
walls, pipes, and equipment [17].
Inerting of Mark I and II Containments
As discussed in Section 1.3 above, Mark I and II containments have the smallest
free volumes among the operating reactors in the United States. Figure 5 below shows
the production of hydrogen in containment based on the percent of metal-water reactions
that takes place.
25
Figure 5: Volume % Hydrogen vs. % Metal-Water Reaction Following a LOCA [17]
As the figure demonstrates, approximately 1-2% metal-water reaction will
produce enough hydrogen to create a flammable environment. A Mark III would require
approximately 10% metal-water reaction, while a dry PWR takes up to approximately
40% metal-water reaction to create a flammable concentration. Therefore, all Mark I and
II BWRs in the United States are inerted during operation. These containments are filled
with nitrogen prior to operating to decrease the oxygen concentration by volume below
4-5%. This essentially guarantees that deflagration and detonation will not take place.
Nitrogen is used in this application since it is relatively cheap for the size of
containment, it does not cause corrosion, and it can be stored as a pressurized liquid.
26
However, even with inerted containments, oxygen concentrations have to be monitored.
Leakage of air from the outside, leakage from compressed air systems, and radiolysis of
water can introduce oxygen into containment. Due to the inerting of containment, the
energy that could be produced from the hydrogen/oxygen reaction (5.2 x 104 Btu/lb for
every 2 lb of hydrogen that reacts) will not be considered as part of the pipe sizing
calculation for an HCVS. It is assumed that significant quantities of oxygen will not be
either generated either through radiolysis during the accident sequence or through inleakage.
2.4 Drywell Venting
The requirement of adding a HCVS that will relieve pressure from the drywell, in
addition to the wetwell, was included as part of Phase 2 of EA-13-109. During venting
situations, there are concerns that the wetwell vent may not be available at certain times
or effective at reducing pressure or the concentrations of hydrogen and other noncondensable gases in containment. Injection of water from safety systems to cool the
core (with the fuel either in the reactor vessel or relocated to the containment floor) can
cause the wetwell to overfill. This would prevent the wetwell HCVS from performing its
function.
The concern with the use of a drywell vent and the reason why wetwell venting is
the much-preferred path is because of the reduction in radioactivity. In order to vent out
of the HCVS from the wetwell, the steam/gas mixture must pass through the pool in the
wetwell. The water is able to scrub out some of the radioactive materials and hold them
in the water before the steam/gas is vented. The effectiveness of reducing the release
fraction for radioactive materials from containment is defined as the decontamination
factor (DF) (see Section 2.5.1 for additional discussion of DF for suppression pools). A
drywell vent will not have the added benefit of reducing the radioactivity in the steam
prior to release, and drywell venting would release significantly more radioactivity than
wetwell venting.
The preferred location for the drywell vent would be near the top of the drywell.
This location is expected to have higher temperatures and concentrations of hydrogen,
which would make the vent more beneficial. In addition, the reactor building (secondary
27
containment) surrounds the upper part of the drywell. Venting from here will reduce the
potential leakage from the drywell to the reactor building [2]. This was an issue at
Fukushima in which Units 1, 3, and 4 had detonations in their reactor buildings.
This paper will perform the same calculations for Mass Flow Rate Requirements
and Required HCVS Pipe Sizing for both decay heat only and decay heat +
zirconium/steam energy to calculate the maximum flow rate requirement for discharging
steam/energy from a drywell HCVS. This calculation is not as complex as for the
wetwell vent since the calculations do not consider the pressure reduction due to the
static head of the suppression pool. These calculations consider many of the same
assumptions, including the steam/gas inside the drywell at saturated conditions.
2.5 Effects of Filtering
Following the issuance of SECY-12-0157, the five NRC Commissioners voted and
approved that HCVS shall be capable of operation under severe accident conditions.
However, Options 3 (FCVS) and 4 (severe accident confinement strategies) were not
approved and additional research was requested to be performed before the approval of
additional rulemaking [28]. As part of Enclosure 7 to SECY-12-0157, proposed Draft
Orders were prepared to show that if approved, what the rulemaking would consist of for
Option 3, installation of filtered containment vents. These requirements were very
similar to the requirements for HCVS capable of operating during severe accidents
(Option 2), with the additional requirement that the containment vent system would be
required to include an engineered filter. This filter would need to be capable of reducing
the release of radioactive material by an amount that is reasonably achievable. The
specified decontamination factors are approximately 1000 for aerosols and 100 for
iodine using filters that are currently available. The filters are required to be sized for
severe accident conditions and all of the parameters that these would entail (e.g.,
radioactive and non-radioactive aerosols, containment pressure and temperature, flow
rates, gas composition, and decay heat). This filter shall be capable of passive operation
with no operator action for 24 hours following initiation of venting.
During accident conditions, the reactor core releases radiation in many different
forms and elements; however, the major forms that are of concern are aerosol particulate
28
(containing cesium (CS) and iodine (I)), gaseous forms of elemental and organic iodine,
and noble gases [20]. One of the main functions of containment is to contain these
radioactive particles and gases and prevent their release. However, HCVS will release
radioactive materials if venting is required to prevent the overpressurization of
containment and reduce the concentrations of hydrogen and other non-condensable
gases. In order to curtail the release of these materials and reduce the potential adverse
effects on the population and environment, engineered filters can be installed as part of
the HCVS (creating a FCVS). Foreign nuclear plants have installed filtered vents, as
discussed in Section 1.5. Although nuclear plants in the United States have not installed
filters as part of hardened containment vents, they have taken advantage of filtering
through the suppression pool to reduce the release of radioactivity if venting is required.
Note that engineered filters and suppression pool filtering methods are only effective for
reducing aerosol and gaseous iodine releases; they do not have any effect on the release
of radioactive noble gases [19].
2.5.1
Decontamination Factors for Suppression Pools
As part of GL 89-16, hardened containment vents were originally proposed to be
installed off the wetwell to take advantage of the natural filtering through water.
Steam/gases that are vented either from the reactor vessel/piping through the SRVs or
from the drywell through the interconnecting piping to the suppression pool, will be
filtered as they pass through the pool before they reach the wetwell atmosphere and are
vented out. A DF quantifies the efficiency of venting, and it is calculated based on the
aerosol mass concentration that initially enters the pool divided by the concentration that
leaves the pool [22]. NUREG-1150 [23] first calculated decontamination factors for
suppression pools. Prior to breach of the reactor vessel, the median DF for aerosols for
suppression pools was found to be between 60 and 80. After the breach of the reactor
vessel, these values were reduced to 7-10 (median).
Further research into DF for suppression pools has shown that it is inappropriate
to characterize a DF as just one value for all accident conditions. The particle size
distribution will vary for different accident scenarios and the efficiency of the pool is
highly dependent on this particle size. Particles can range from <0.1 μm to greater than
29
10 μm and are removed by different mechanisms. The smaller particles can be removed
by diffusion, intermediate particles removed by interception with bubbles in the pool,
and larger particles through gravitational settling [2 - Enclosure 5]. In addition there are
a number of other factors that affect the DF. A larger DF will be achieved by
maximizing the pool height, increasing the mass flow rate into the pool, keeping the pool
subcooled (i.e., temperature below saturated conditions), and having steam included in
the gas stream that is inlet to the pool [20 and 21].
As part of the MELCOR analysis in Enclosure 5a to SECY-12-0157, the NRC
defines a DF for suppression pools between 100 and 300. This is a conservative
estimation and is based on pool depth, pool temperatures, and other factors as discussed
above.
2.5.2
Engineered Filter Designs
There are two different major types of engineered filters that either have been
installed or are available for installation today: wet filters and dry filters. All filters are
designed to be in standby and ready for use when required.
Wet Filters
Wet filters use the same generic principle as suppression pools by using water as
the fundamental medium for filtering. The two Beznau nuclear power plants in
Switzerland chose to install the Sulzer-EWI system in the early 90’s. This system
consists of a water tank with nozzles and baffles. In this arrangement, steam/gas is
vented out of containment through a pipe, through the containment isolation barriers and
into the filter. The steam/gas is discharged through nozzles into the bottom of the tank
and directed upwards where the gas expands. As the gas travels, it is subjected to
numerous changes in direction through baffles. The gas then exits through the top of the
filter unit. The reduction in the radioactivity of aerosols is due to breaking the gas/steam
mixture into small bubbles to increase their surface area, and a DF of greater than 1000
is achieved. The breakdown and many changes of direction causes even energy
dissipation in the filter. The filter capacity is large such that there is no risk of flow
obstruction. The unit is capable of passing 0.5-1% of the steam/energy from the reactor
30
at the design pressure of containment, and up to double that with minimal efficiency
reductions. Chemical conditioning of the water with chemicals such as sodium carbonate
(Na2CO3), sodium hydroxide (NaOH), and sodium thiosulfate (Na2S2O3) convert volatile
iodine to non-volatile iodine [24].
Many other existing engineered filters (e.g., FILTRA-MVSS by ABB, sliding
pressure venting system at Gosgen in Switzerland) along with newer versions (e.g.,
Westinghouse’s FILTRA-MVSS, AREVA FCVS, IMI) are very similar in design to the
Sulzer-EWI. Figure 6 below shows the IMI filter design. All designs consist of tanks of
water located outside of containment. The steam/energy is discharged through nozzles or
venturis at the bottom of the tank, and passed through impingement/baffle plates to
break down the gases. This creates changes in direction, which enables for a long dwell
time and maximum diffusion capture of aerosol products. Some of the older designs use
gravel beds at the top of the filter, while newer designs use metal-fiber beds to remove
water droplets and additional micro-aerosol products before exiting. These designs use
chemical conditioning to capture elemental iodine. The designs of these filters allow for
venting the required flow rates at maximum containment pressures. The systems are
passive so that no actions are required for a minimum of 24 hours following the start of
venting activities. The piping upstream and downstream of the filter is inerted to
eliminate hydrogen concerns. The DF for these filters has been found to be greater than
1000 for aerosols and greater than 100 for elemental iodine capture [2 – Enclosure 4, 22,
and 24].
Figure 6: IMI Nuclear (CCI) Filter Technology [2 - Enclosure 4]
31
Dry Filters
Dry filters are different from wet filters in that they utilize no water. Older
designs in Sweden (FILTRA) and France vent through either large beds of gravel and/or
sand. As the steam and gas pass through these materials, the filters hold up and capture
the radioactive particles, and condense the steam. Per [24], for FILTRA, the DF for
aerosols and iodine is said to be greater than a 1000; however, during testing the DF was
found to be between 2.3 and 5 [22]. For the sand beds, aerosols were tested to have a DF
of >6000 [22]. The DF for elemental and organic iodine is approximately 10 [2 –
Enclosure 4]. Neither of these designs is currently marketed for new installations.
The dry filter systems that are marketed (in Germany by FZK and around the
world by Westinghouse) is a metal-fiber filter. In the German design, the steam/gas is
passed through pads of stainless steel fibers that get continuously smaller and have a DF
for aerosols >60,000. In the Westinghouse design, the steam/gas is passed through metal
fibers with silver zeolite beds and have DFs ~10,000 for aerosols and 10-100 for
elemental and organic iodine. In these designs, the venting strategy requires delays of
more than a day after shutdown such that the heat loading on the filter is less and a
longer time for particulate to settle in containment. Based on these limitations, wet filters
appear to be a better design for venting BWRs, which may require early venting times [2
– Enclosure 4, 22, and 25].
32
3. Results and Discussion
3.1 Design of HCVS for Decay Heat Removal and Severe Accidents
Figure 7 shows a simplified diagram of a potential FCVS design. In this design,
there are two hardened vents connected by a common header that discharges to a wet
filter: one vent from the wetwell/torus and one vent near the top of the drywell. Without
the filter installed on the common discharge, Cooper and other nuclear power plants in
the United States with Mark I and II containments will install similar arrangements to
comply with EA-13-109.
Drywell Vent
Wet Filter
Wetwell Vent
Figure 7: Simplified HCVS/FCVS Design [2 - Enclosure 4]
Phase 1 of this Order requires installation of the wetwell vent, and this design would
meet the requirements listed in Attachment 2 of the Order as follows:

The sizing and capacity of the HCVS would meet the 1% reactor thermal power at
maximum containment pressure or PCPL if it were designed per Section 3.1.1
below.

The pipe has two isolation barriers to provide containment isolation. These valves
are required to be operated both remotely (with designated power sources, such as
batteries), and manually. Shielding would need to be provided along with manual
valve operators to allow for personnel to open these valves when required.

The system would need to be designed and supported such that it can withstand the
effects of a seismic event.
33

Instrumentation would be provided such that the positions of the valves, radiation
levels in the piping, and other conditions such as pressure temperature are known.

Discharge of the vent would need to be above the elevation of all main plant
structures along with minimizing cross flow to other units for multi-unit sites.

Finally, for severe accidents, the system would be required to cope with high
temperature, pressure, radiation, hydrogen, and other non-condensable gases. The
vent shall ensure the flammability limit in containment or the venting system during
venting, or the system shall be designed to withstand the dynamic loading from
deflagration and detonation. Section 3.1.2 presents discussion of these conditions,
including pipe size calculations.
The installation of the drywell vent as shown above would meet the requirements of
Phase 2 of Attachment 2 to EA-13-109. Many of the requirements discussed above
would need to be designed the same as the wetwell vent. The same sections that present
calculations for HCVS wetwell pipe sizing also present calculations of drywell pipe
sizing. The location of the drywell vent is shown being near the top of containment,
which will allow for venting of the higher temperature gasses in containment, including
hydrogen. The figure shows two pathways with double isolation. One pathway is
considered completely passive (no operator action required to vent), with both valves
open, but a rupture disc installed. A rupture disc is a non-reclosing pressure relief disc
that is a one-time use component. This would only allow for venting when the pressure
in containment exceeds the set pressure of the rupture disc (i.e., containment design
pressure). The second pathway is setup the same as the wetwell path with two normally
closed isolation valves. This double pathway arrangement could be installed on the
wetwell HCVS as well. The double arrangement eliminates the operator action to vent
containment when pressure reaches its limit, and operator action would only be required
if early venting was desired or if the pathway needed to be closed following venting.
The installation of an engineered filter is not required as part as EA-13-109,
however, it may be required as part of future Orders. Section 3.2 discusses the benefits
of filters along with the deficiencies.
34
3.1.1
Non-Severe Accident Calculations
Mass Flow Rate Requirements
Appendix 6.1 presents the full spreadsheets and results for mass flow rate
requirements; however, Table 3 shows the summarized results below:
Table 3: Mass Flow Rate Requirements
WV (lb/hr)
Wetwell - Torus Water Level
Low
High
Max
Drywell
-
No Injection/Makeup
Flow Into Containment
8.43 x 104
8.44 x 104
8.45 x 104
8.42x 104
Injection/Makeup Flow
Into Containment
8.94 x 104
8.96 x 104
8.97 x 104
8.94 x 104
In this table, the torus water levels are as follows: Low is if the static head of the
water in the torus is 3 ft, High is if the static head is 18.417 ft, and Max is if the static
head is at 28 ft. The fourth column calculates the drywell HCVS required flow rates.
As presented in the table, the most conservative scenario for the wetwell HCVS
is when the pressure is reduced by the largest static head in the torus (Max) and when it
is assumed that makeup flow is available and is injected when the wetwell is vented. The
static head has a small effect on flow requirements comparatively (~0.2% increase in
lb/hr from the Low case to the Max case). The availability of makeup/injection of water
into containment has a much bigger effect (comparatively) on the maximum mass flow
rate out of the HCVS (~6% increase). The temperature of the assumed makeup/injection
drives this difference. The required drywell flow is slightly less the Low case, since
there is no pressure reduction taken due to static head.
If the makeup temperature was assumed at 130°F (enthalpy of 98.0 Btu/lb) rather
than 100°F (enthalpy of 68.1 Btu/lb), the required flow for Max conditions would
increase to 9.21 x 104 lb/hr. This would be an approximate 3% increase above water at
100°F. Knowing the maximum temperature of makeup fluid injected by the safety
systems into containment is a critical parameter for obtaining conservative results.
The mass flow rate calculations for the wetwell HCVS are slightly less than the
mass flow rate provided by CNS to the NRC (90,853 lb/hr) [9]; however, the results are
very similar. Note that the flow rate in the CNS submittal was calculated based on a
35
containment pressure of 41.3 psig rather than the maximum containment design pressure
of 56 psig (it is stated that was to allow for use of RCIC, which requires venting at a
lower pressure – see additional discussion in Section 3.3). This reduction in pressure
would slightly increase the flow rate. In addition, some of the design parameters that
CNS utilized to calculate their flow rates, such as water level in the torus and
temperature of makeup/injection water, are not documented in their response and remain
assumptions for the calculations in this paper.
Required HCVS Pipe Sizing
The required HCVS pipe sizing calculations utilize the results from above
assuming there is makeup/injection flow (at 100°F) into the primary containment when
venting. The wetwell pipe sizing considers all three static head conditions (Low, High,
and Max). The drywell pipe sizing does not consider any static head conditions.
As discussed in Section 2.2.2, it is important to calculate the critical value of K
(system resistance coefficient) based on the pressure ratio (PR) for each case. The
critical values of K are as follows: 1) Low case, K = 12.282, 2) High case, K = 9.573, 3)
Max case, K = 8.285, and 4) Drywell case, K = 12.806. These values determine if the
flow out of the pipe will be at sub-sonic or sonic velocity. If the K value is greater than
the calculated critical value, the flow will be sub-sonic, while a K value less than the
critical value will lead to sonic flows and the differential pressure will be required to be
reduced. The Modified Darcy Formula is then calculated to determine the maximum
mass flow rates allowable for a certain pipe size with a range of resistance coefficients.
These are then compared with the mass flow rates calculated above.
Appendix 6.2 presents the full spreadsheets and results for HCVS pipe sizing.
Four tables present the flow capacity for three different pipe sizes (8-inch, 10-inch, and
12-inch NPS) at a range of resistances. The pipe size is based on “Standard” pipe sizes
(STD) from Table B-18 of [13]. The results are summarized as follows:

For the “Low” condition and the drywell pipe vent, the 8-inch pipe will be
acceptable if K ≤ 6. The 10-inch and the 12-inch pipe will be acceptable for all K
values up to 20. Note that flow capacities for higher K values may be acceptable
36
and can be calculated if required, but a maximum of 20 will allow for a system
with a significant number of losses.

For the “High” condition, the 8-inch pipe will be acceptable if K ≤ 4. The 10inch will be acceptable if K ≤ 16. The 12-inch pipe for this condition will be
acceptable for all K values up to 20.

For the “Max” condition, the 8-inch pipe will be acceptable if K ≤ 4. The 10-inch
will be acceptable if K ≤ 14. The 12-inch pipe for this condition will be
acceptable for all K values up to 20.
The CNS response to the NRC states the 10-inch wetwell vent is capable of
venting the specified flow of 90,853 lb/hr. Therefore, the total system resistance would
be expected to be less than or equal to 14 (Max case), depending on the pressure of the
steam/gas that enters the vent and assuming max water level in the torus.
The results of this section provide design criteria to consider for installation of a
venting line. Note that this same philosophy can be applied for Mark II containment with
changes for the physical design of the suppression pool. A specific path with valves and
other piping components would need to be designed based on the specific plant
parameters and interferences. Once a specific design and path is decided upon, the total
system loss coefficient can be calculated. Based on this calculation the results from the
tables can be utilized to determine what size pipe would be necessary meet Requirement
1.2.1 of EA-13-109.
3.1.2
Severe Accident Calculations
Mass Flow Rate Requirements
Appendix 6.3 presents the results of the mass flow rate calculation for severe
accident conditions. For this calculation, the “Max” condition for a wetwell vent was the
only case utilized, since it envelops the other cases for wetwell (Low and High) and
drywell vents as presented above. In order to produce 4400 lbm of hydrogen, 100,000
lbm of zirconium is required [17], and this will create 2.77 x 108 Btu. This energy is 2.78
times the amount of energy from 1% decay heat (9.95 x 107 Btu).
In the calculations above that only considered 1% decay heat energy, the
calculations utilized the properties of steam (enthalpy and specific volume), which
37
would be the dominant gas that is vented. For severe accident conditions, there will be
other non-condensable gases, including hydrogen, which will be generated and will
affect the flow rate that is required to be vented. The mixture of gases in containment
will depend on numerous factors and this paper cannot fully evaluate different gas
mixtures. However, this paper evaluates the effect on venting if hydrogen was dominant
in terms of venting and compare this to steam.
Appendix 6.5 presents an excerpt from [26] that provides vapor properties of
hydrogen. A venting pressure of 70.7 psia is equal to 4.81 atm, so the values will need to
be interpolated between the 2 atm and 5 atm columns. In addition, at 70.7 psia with
saturated conditions in containment (which has been assumed for these calculations for
steam), the temperature would be 303.6°F. The hydrogen table only presents data up to
300K (80.6°F); therefore, the properties of hydrogen need to be extrapolated. To
calculate the enthalpy and specific volume of hydrogen at 70.7 psia and 303.6°F, the last
six values in the table were utilized to create charts, and linear trend line equations were
found using Excel. Linear trend lines were found to be acceptable since the R2 values
were 1 and 0.999 for specific volume and enthalpy, respectively. See Appendices 6.6,
6.7, and 6.8 for the interpolation table and the charts with trend lines.
The calculated enthalpy of 2564 Btu/lb was utilized to calculate the maximum
flow rate of 1.51 x 105 lb/hr (based on hydrogen properties). The maximum calculated
flow rate for steam properties is 3.39 x105 lb/hr, which is more than twice as much as for
hydrogen. Both are significantly higher than the 8.97 x 104 lb/hr calculated for this case
only considering decay heat and based on steam properties.
Required HCVS Pipe Sizing
The calculated mass flow rates for severe accident conditions are utilized to
calculate the required pipe size and Appendix 6.4 presents these results. The required
pipe size for steam conditions was calculated the same as above, except that the higher
calculated flow rate of 3.39 x 105 lb/hr was utilized. For this energy, a 12-inch STD NPS
pipe will not be adequate for venting. An 18-inch pipe system would be acceptable if K
≤ 8.285 (critical K value). A 20-inch pipe allows for K ≤ 12, which would be a
38
reasonable total resistance for a pipe system. For 22-inch and 24-inch systems, flow
resistance coefficients up to 20 (and possibly higher) would be acceptable.
To evaluate the required pipe size conditions based on hydrogen properties, the
calculated specific volume of 60.64 ft3/lb at 70.7 psia and 303.6°F is utilized. The results
show that even though the required mass flow rate based on hydrogen conditions is less
than half of steam conditions, the required pipe sizes are more restrictive based on
hydrogen properties. An 18-inch pipe system would be acceptable if K ≤ 4. A 20-inch
pipe system would be acceptable if K ≤ 6, while a 22-inch pipe system would be
acceptable for K ≤ 10. Finally, a 24-inch pipe system would be acceptable for K ≤ 16.
These conditions require either systems with flow resistance coefficients that are less
than for steam, or larger sized pipes.
These calculations considered maximum energy conditions, which was very
conservative. As part of the NRC study in Enclosure 5a of SECY-12-0157, the in-vessel
hydrogen produced during severe accident conditions (zirconium/steam reaction) was
between 500 and 750 kg-mol, or 1000 to 1500 kg, which is equivalent to approximately
2000 to 3000 lb. This is less than the 4400 lb of hydrogen considered to be produced for
these calculations. More detailed analyses could determine the maximum quantity of
hydrogen produced from this reaction for specific reactors, which could allow for
consideration of lower energy production and therefore, smaller piping systems.
The calculations only evaluated saturated conditions at maximum containment
design pressure for venting (pressure at 70.7 psia, temperature = 303.6°F). Due to the
rapid production of hydrogen and increase in energy in containment, it is possible that
the temperature could be higher at the time of venting. Calculations were performed for
both mass flow rate and required pipe size for a temperature of 500°F to evaluate the
effect of higher temperatures in containment for hydrogen conditions, and the results are
shown in Appendix 6.3 and 6.4. At a higher temperature, the mass flow rate is reduced
to 1.19 x 105 lb/hr. This has a minor effect on the pipe size calculations. There was no
change to the allowable size based on flow coefficients up to 18” pipe; however, for
larger pipe systems, the allowable flow resistance coefficient is increased. Therefore, a
lower temperature is more conservative when sizing a pipe system.
39
Note that if a pressure above the containment design pressure is expected at the
time of venting (due to potential delays/difficulties in opening the system), the system
may not be able to handle the energy based on hydrogen properties. In the calculations
that considered steam properties, a higher differential pressure between the inlet to the
HCVS and the outside atmosphere allowed for either a smaller pipe system or a system
with a higher total resistance coefficient. However, this will not be the case based on
hydrogen properties. The mass flow rate is based upon the enthalpy, and increasing
pressure causes a negligible increase in the enthalpy. The mass flow rate will be nearly
constant for increasing pressure. However, there is a much larger effect on specific
volume (decreases) with increasing pressure. The Modified Darcy Formula is inversely
proportional to the square root of the specific volume, so a smaller specific volume will
decrease the amount of flow through a pipe with a certain flow resistance. Therefore, a
larger pipe system or a system with less flow resistance may be required if it is expected
that the pressure in containment could exceed containment design pressure prior to
venting.
Additional Severe Accident Discussion
The calculations above demonstrate the magnitude of the piping system that
would be required to vent the energy that could be in containment during severe
accidents. Since Mark I and II containments are inerted, this paper is not concerned with
deflagration and detonation in containment prior to venting; however, this could be an
issue during and following venting. As the steam/gas is discharged from containment
through the vent line, oxygen will mix and create a flammable mixture either in the
piping system or at the HCVS discharge. At this point, the system would need to be able
to withstand potential detonation. The system would also be required to be capable of
handling a diffusion flame from a hydrogen jet. Following venting, if the vent is not
closed, air (i.e., oxygen) could flow back into containment and create potential
flammable conditions. One suggestion is that a check valve or back draft damper be
installed in the line near the discharge. Either would allow flow out of the HCVS, but
would not allow for flow of air back into containment. The HCVS could be inerted up to
either of these devices, and would not allow mixing until the steam/gas discharge. This
40
would limit the section of piping that would need to be evaluated for deflagration and
detonation.
Another suggestion is the use of igniters in the HCVS. During venting, it is
possible to know the percentages of hydrogen and oxygen and determine when the gas
mixture would be considered flammable. At the time when the gas mixture becomes
flammable but not yet detonable, igniters could be used to burn the hydrogen
(deflagration). This would increase the pressure/temperature in the system, but at a much
slower rate than if the mixture were to detonate. Instrumentation to detect the volumetric
percentages of hydrogen and oxygen are currently utilized in containments and could
also be used as part of HCVS.
3.2 Discussion of Filter Benefits
The NRC performed studies on the consequences of severe accidents based on
different mitigation strategies [2 – Enclosure 5B]. These strategies included both
wetwell and drywell venting, with and without core and containment spray. Two of the
cases specifically evaluated the effects of wetwell venting (Case 3) and drywell venting
(Case 12) without any spray. In these two cases, both filtered and unfiltered cases were
evaluated. As would be expected, filters had a beneficial impact on the dose to the
population, contaminated area, and economic cost of recovery. The effects were
evaluated radially up to 50 miles from the site. For the wetwell venting case, a filter
(with a DF of 10) reduced the dose to the population and the cancer fatality rate by
approximately 60%, while the contaminated area and economic cost of recovery was
reduced by approximately 85%. The drywell venting case evaluates filters with DFs of
1000 and 5000. Both filters reduced the dose to the population and the cancer fatality
rate by approximately 95%, while the contaminated area and economic cost of recovery
was reduced by approximately 99%. Note a filter with a DF of 5000 did not provide
significantly better results than the filter with a DF of 1000. The consequences on the
area and population when comparing wetwell and drywell venting with filters
demonstrated that the drywell venting with filters were more efficient than unfiltered
wetwell venting through the suppression pool (population dose, cancer fatality rate and
contaminated area were approximately double, while the economic cost was more than
41
four times as high for unfiltered wetwell venting). However, filtered wetwell venting
produced the lowest consequences (20% lower for population dose and cancer fatality
rate, 70% lower for contaminated area and 30% lower for economic cost when compared
to drywell filtering). This study did not evaluate any cases using both wetwell and
drywell venting, with and without filters. This would be useful since it is expected that
the suppression pool may completely fill due to injection of water by the safety systems
and make wetwell venting not possible later on in the accident.
Another benefit of filter installation is that the piping from containment to the
filter can be inerted. This would prevent meeting and exceeding the lower flammability
limit in the system. In addition, the filter would also prevent the backflow of air/oxygen
into containment.
One of the unknowns with adding filters to wetwell vents is how much of a
benefit it achieves. Filter performance is very dependent on the particle size distribution,
and suppression pool filtering will greatly affect the particle size distribution that passes
through and reaches the filter. The NRC study utilizes an assumption that the DF would
be reduced from 1000 or greater (nominal value) down to 10 (following the suppression
pool); however, further research and testing should be performed to evaluate the particle
sizes that would pass through the suppression pool and how efficient a filter would be
further reducing the radioactivity [19]. However, note that as saturation conditions in the
suppression pool are reached, the particle size distribution that will pass to the filter will
shift back towards the distribution that would be seen without the suppression pool, thus
making the filters more beneficial.
Another concern is that many of the filters are designed to accept a flow rate up
to the 1% decay heat energy. It is unclear if the filters would be able to accept the higher
flow rates associated with severe accident conditions, which could be up to an equivalent
of 4% of the decay heat.
Finally, one of the big concerns for plants is the cost-benefit analysis. Installation
of filters is expected to cost at least $15 million, with some plants believing that the cost
could range up to $45 million. When evaluating the chances of a severe accident
occurring at a unit and the difference in the current cost for cleanup (with and without a
filter) versus the cost of the filter (materials, installation, and maintenance), the NRC
42
study demonstrates that filter installation is not cost beneficial. There is discussion in
this study that the assumed cleanup cost is too low and when doubled, filter installation
becomes cost beneficial if filter costs are equal to or less than $15.353 million. Although
the cost-benefit analysis does not favor filter installation, the foreign countries that either
have installed filters or will require filter installation going forward have not considered
cost-benefit analyses. Instead, they have only focused on the benefits of installed filters
from a defense-in-depth perspective and considered filters to be essential safety
enhancements to protect against potential severe accidents [2 – Enclosure 2].
3.3 Additional Considerations for Severe Accidents
The following discussion topics are not currently included in the requirements of
EA-13-109; however, they could have a large effect on HCVS that are capable of
handling severe accident conditions.
Suppression Pool Bypass
For Mark II containments, there is a concern regarding what is called
“suppression pool bypass”. During severe accidents with breach of the reactor vessel, the
barrier between the drywell and the wetwell for a Mark II containment could be
breached and the effective DF through a wetwell HCVS filter is greatly reduced. Unlike
the Mark I, where the wetwell is a torus located outside of the drywell and connected
only by piping, the Mark II drywell is located directly above the wetwell. During a
severe accident where the core melt breaches the reactor vessel and falls to the floor of
the drywell, there is the potential that the floor or some of the piping that passes from the
drywell to the wetwell could fail. This would cause the atmosphere in the wetwell to be
the same as the drywell. Figure 8 below shows this condition, where containment is
equipped with both a wetwell and a drywell HCVS. This is not a major concern if a filter
is installed at the end of the vent piping (FCVS); however, it is a major concern if a filter
is not installed. The DF for the suppression pool would be made obsolete following this
failure since the wetwell atmosphere would not be “filtered” through the pool. EA-13109 discusses this issue; however, no requirements are provided currently for utilities.
This loss of filtering capability is an issue that will be resolved as part of the NRC
43
rulemaking addressing broader severe accident management and filtering strategies. As
part of EA-13-109, the NRC states that utilities with Mark II containments may install
engineered filters to address this issue.
Figure 8: Suppression Pool Bypass for Mark II [19]
Confinement Strategies
Both the NRC [2 – Enclosure 5a] and EPRI [19] performed studies on the
consequences of severe accidents based on different mitigation strategies. These
strategies included both wetwell and drywell venting, with and without core and
containment spray, and containment flooding. One of the “key insights” that was seen in
both studies is that venting alone (wetwell or drywell, with or without a filter) is not
enough to mitigate the worst-case scenario for severe accidents: breach of containment,
which allows for uncontrolled release of radioactivity and contamination. Other
44
strategies, such as core or containment spray, or containment flooding, are necessary in
conjunction with venting to prevent breach of containment. The injection of water will
prevent either breach of the reactor vessel or containment failure, if the reactor vessel
has been breached and the core has relocated to the drywell floor. Venting will only
prevent containment failure due to potential overpressurization in containment.
Therefore, it is necessary that confinement strategies (Option 4 discussed in [2]) are
further analyzed to determine what safety systems will be required to be available in
order to cope with severe accidents, in addition to venting, with or without filters.
Following the events of Fukushima, the NRC and the nuclear industry are evaluating
strategies to ensure that safety systems are available to allow for injection of water into
containment for accident scenarios. These strategies will be required to be evaluated in
conjunction with containment venting during severe accidents.
Venting Pressure/Cycling
This paper assumes that all venting occurs at containment design pressure of 56
psig/70.7 psia. However, the BWR Owners’ Group (BWROG) is proposing venting
earlier (at 25 psig) rather than waiting for the containment design limit. This would be
necessary to allow for use of RCIC along with other safety systems [2 – Enclosure 4].
EPRI on the other hand proposes only venting near the containment design pressure and
cycling of the vents. In EPRI’s study, containment would first be vented when the
pressure limit is reached until the pressure in containment was reduced to approximately
40 psig, and then the vent would be shut until the pressure increased again to the
pressure limit. The vent would be cycled opened and closed for the remainder of the
accident. The reasoning is that the higher pressure maximizes the scrubbing performed
by the suppression pool for the following reasons: 1) increased timing before venting, 2)
increased velocity of the gas injected into the suppression pool (similar theory venturis
in wet filters), and 3) decreased chances of reaching saturation conditions in the wetwell
(with reduced DF) due to greater subcooling [19]. Both proposals have their benefits, but
further research is required to determine which strategy is superior.
45
4. Conclusion
As demonstrated in this paper, venting through HCVS is necessary in order to
help mitigate the consequences of severe accident conditions at nuclear power plants
with Mark I and II containments. The calculations performed present pipe sizes and flow
resistance coefficients that would be required in order to design HCVS to relieve
steam/gas in containment based on both decay heat only and decay heat + energy
generated from severe accidents. Much larger piping systems are required to mitigate
severe accident conditions. These calculations rely on some conservative assumptions,
but other assumptions regarding timing of venting and atmosphere in containment during
venting need to be further analyzed to verify that the results are bounding.
Review of literature and studies on the effectiveness of venting through
engineered filters demonstrate that they are beneficial in reducing the effects on
population and land contamination. However, the effectiveness of filters in conjunction
with suppression pool “filtering” for wetwell HCVS is unknown. In addition, filter
installation is very expensive and current cost-benefit analyses do not demonstrate they
are beneficial in this regard.
Finally, although vents/filters are beneficial in preventing overpressurization of
containment, they alone do not prevent the worst-case scenario from occurring following
a severe accident: containment failure leading to the uncontrolled release of radiation. If
containment fails, filters would prevent the uncontrolled release of radiation.
Confinement strategies must be evaluated further so they can be used in conjunction
with venting to prevent the uncontrolled release of radiation and contamination and
mitigate the consequences of severe accidents.
46
5. References
[1] EA-12-050, March 12, 2012, “Issuance of Order to Modify Licenses with Regard to
Reliable
Hardened
Containment
Vents”,
NRC
ADAMS
Accession
Number
ML12054A694
[2] SECY-12-0157, November 26, 2012, “Consideration of Additional Requirements for
Containment Venting Systems for Boiling Water Reactors with Mark I and Mark II
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[3] Mark I Containment Report, March 19, 2011, http://files.gereports.com/wpcontent/uploads/2011/10/NEI-Mark-1-White-Paper.pdf
[4] ANSI/ANS-5.1-2005, American Nuclear Society, Decay Heat Power in Light Water
Reactors
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[6] Generic Letter 89-16, Installation of a Hardened Wetwell Vent, September 1, 1989,
NRC ADAMS Accession Number ML13017A234
[7] JLD-ISG-2012-02, Compliance with Order EA-12-050, Reliable Hardened
Containment Vents, Interim Staff Guidance, Rev. 0, NRC ADAMS Accession Number
ML12229A475
[8]
BWR
Mark
I
Containment
Figure,
Modified
from
http://upload.wikimedia.org/wikipedia/commons/a/a0/BWR_Mark_I_Containment%2C
_diagram.png
[9] Overall Integrated Plan in Response to March 12-2012, Commission Order
Modifying Licensees with Regard to Reliable Hardened Containment Vents (Order EA12-050), Cooper Nuclear Station, Docket No. 50-298, DPR-46, February 28, 2013, NRC
ADAMS Accession Number ML13070A094
[10] GE NE-T23-00786-00-09, Evaluation of Steam Ingestion in the ECCS Suction
Strainers for Cooper Nuclear Station, Non-Proprietary Report, NLS2001079 Enclosure
3, NRC ADAMS Accession Number ML012640451
[11] Nuclear Regulatory Commission Website, http://www.nrc.gov/
[12] ASTM A524-96 (Reapproved 2012), Standard Specification for Seamless Carbon
Steel Pipe for Atmospheric and Lower Temperatures
47
[13] Crane Technical Paper No. 410, Flow of Fluids Through Valves, Fittings and Pipe,
1980
[14] EA-13-109, June 6, 2013, “Order Modifying Licenses with Regard to Reliable
Hardened Containment Vents Capable of Operation Under Severe Accident Conditions”,
NRC ADAMS Accession Number ML13130A067
[15] Chapter 6.5, Primary Containments, General Electric Advanced Technology
Manual, NRC ADAMS Accession Number ML11263A359
[16] SECY-11-0093, July, 12, 2011, “Recommendations for Enhancing Reactor Safety
in the 21st Century; The Near-Term Task Force Review of Insights from the Fukushima
Dai-ichi Accident,” NRC ADAMS Accession Number ML111861807
[17] Camp, A. L., et al., NUREG/CR-2726, “Light Water Reactor Hydrogen Manual”,
Sandia National Laboratories, August 1983, NRC ADAMS Accession Number
ML071620344
[18] Sherman, M. P., “Hydrogen Combustion in Nuclear Plant Accidents and Associated
Containment Loads,” Nuclear Engineering and Design, Volume 82, Issue 1, October 1,
1984, Pages 13-24
[19] EPRI Technical Report, Investigation of Strategies for Mitigating Radiological
Releases in Severe Accidents, BWR Mark I and Mark II Studies, September 2012,
http://www.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=00000000000102
6539
[20] Wassel, A.T., et al., “Analysis of Radionuclide Retention in Water Pool,” Nuclear
Engineering and Design, Volume 90, Issue 1, November 3, 1985, Pages 87-104
[21] Dehbi, A., et al., “Aerosol Retention in Low-Subcooling Pools Under Realistic
Accident Conditions,” Nuclear Engineering and Design, Volume 203, Issues 2-3,
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[22] NEA/CSNI/R(2009)5, “State-of-the-Art Report on Nuclear Aerosols,” Nuclear
Energy Agency, Committee on the Safety of Nuclear Installations, December 17, 2009,
http://www.oecd-nea.org/nsd/docs/2009/csni-r2009-5.pdf
[23] NUREG-1150, “Severe Accident Risks: An Assessment for Five U.S. Nuclear
Power Plants,” U.S. Nuclear Regulatory Commission, http://www.nrc.gov/readingrm/doc-collections/nuregs/staff/sr1150/
48
[24] Rust, H., et al., “Pressure Release of Containments During Severe Accidents in
Switzerland,” Nuclear Engineering and Design, Volume 157, Issue 3, August 1995,
Pages 337-352
[25] Schlueter, R. O., Schmitz, R. P., “Filtered Vented Containments,” Nuclear
Engineering and Design, Volume 120, Issue 1, June 1, 1990, Pages 93-103
[26] Brookhaven National Laboratory, Selected Cryogenic Data Notebook, Section 3
Properties
of
Hydrogen,
Revised
August
1980,
http://www.bnl.gov/magnets/staff/gupta/cryogenic-data-handbook/Section3.pdf
[27] Dallman, R. J., Galyean, W. J., Wagner, K. C., “Containment Venting as an
Accident Management Strategy for BWRs with Mark I Containments,” Nuclear
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[28] VR-SECY-12-0157, Commission Voting Record for SECY-12-0157, March 19,
2013, NRC ADAMS Accession Number ML13078A012
49
6. Appendix
6.1 Mass Flow Rate Requirements Spreadsheet – Non-Severe
Accidents
50
Mass Flow Rate Requirements Spreadsheet Continued
51
6.2 Required HCVS Pipe Sizing Calculation Spreadsheet-Non-Severe
Accidents
52
6.2 continued
53
6.3 Mass Flow Rate Requirements Spreadsheet – Severe Accidents
54
6.4 Required HCVS Pipe Sizing Calculation Spreadsheet – Severe
Accidents
55
6.4 cont
56
6.5 Properties of Hydrogen Vapor – Excerpted from [26]
57
6.6 Properties of Hydrogen – Interpolation Table
58
6.7 Specific Volume Chart - Hydrogen
59
6.8 Enthalpy Chart - Hydrogen
60