ML17130A249
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CHAPTER 10 10.1-1 REV. 21, APRIL 2007 SECTION 10.0 AUXILIARY SYSTEMS
10.1
SUMMARY
DESCRIPTION
This section describes the reactor and plant auxiliary systems
that are required for operation, but which are not integral
portions of the reactor and power conversion equipment or their
safety systems.
CHAPTER 10 10.2-1 REV. 21, APRIL 2007 10.2 NEW FUEL STORAGE 10.2.1 Introduction
New fuel is stored in high density storage racks located in the
spent fuel pool.
10.2.2 Description
New fuel assemblies as received at the plant are stored in metal
boxes inside an outer metal container. Each outer container and
metal box holds two fuel bundles. After removing the box from the
outer container, the boxes are hoisted to the refueling floor and
the fuel bundles are receipt inspected and placed in the spent
fuel pool using the fuel-handling equipment. If a bundle fails
receipt inspection, it is tagged in accordance with procedure, placed back into the container, and left on the refueling floor. A
GE inspector checks the bundle. He either repairs the problem at
the site or returns the damaged bundle for a replacement. In the
event that the new fuel cannot be taken to the refueling floor
immediately, it can be off-loaded and stored in a roped-off area
with a security guard posted.
CHAPTER 10 10.3-1 REV. 26, APRIL 2017 10.3 SPENT FUEL STORAGE 10.3.1 Power Generation Objective
The power generation objective of the spent fuel storage
arrangement is to provide storage space for the spent fuel
assemblies which require shielding during storage and handling.
10.3.2 Power Generation Design Basis
- 1. Spent fuel storage racks for each reactor are designed to accommodate 3819 fuel assemblies.
- 2. Spent fuel storage racks are designed and arranged so that the fuel assemblies can be efficiently
handled during refueling operations.
- 3. Dry storage casks are designed so that spent fuel may be transferred from the wet storage racks into
the casks in an efficient manner
10.3.3 Safety Design Basis
- 1. All arrangements of fuel in the spent fuel storage racks are maintained in a subcritical configuration having a k eff 0.95 for all conditions.
- 2. Each spent fuel storage rack loaded with fuel and the pool structure are designed to withstand
seismic loading to minimize distortion of the spent
fuel storage arrangement or loss of spent fuel pool
level.
- 3. Fuel designated for dry cask storage will have a K eff < 0.95.
- 4. Dry storage casks are evaluated for design basis environmental events and are placed in
configurations which do not impact the structural
capabilities of the fuel pool, reactor building, or
cask transport path.
10.3.4 Description
The high density spent fuel storage racks provide storage at the bottom of the fuel pool for the spent fuel received
from the reactor vessel and new fuel for loading to the
reactor vessel. The fuel storage racks at the bottom of
the pool are covered with water (normally about 23 ft
CHAPTER 10 10.3-2 REV. 26, APRIL 2017 above the stored fuel) for radiation shielding. Sufficient shielding is provided by maintaining a minimum depth of
water at all times. The racks are freestanding, full-
length top entry and are designed to maintain the spent
fuel in a space geometry which precludes the possibility
of criticality (k eff will not exceed 0.95) under any conditions.
The high density spent fuel storage racks are of the "poison" type, utilizing a neutron absorbing material to
maintain a subcritical fuel array (see Figures 10.3.1, 10.3.2, and 10.3.3). The rack modules are rectilinear in
shape and are of nine different array sizes. The racks
are arranged in the spent fuel pool as shown in Figure
10.3.4 for Unit 2 and Figure 10.3.5 for Unit 3. A total
of 3819 storage locations are provided per pool. The
racks are capable of storing BWR fuel assemblies (with or
without their channels) with a maximum incore k-infinity of 1.270. Maintaining the incore k-infinity 1.270 will assure that the rack k-effective will be equal to or less than that determined in the safety design basis. TS
Amendment Nos. 175/178 demonstrated a k-infinity limit of
< 1.362 using an analysis of the criticality aspects of the storage of PBAPS fuel assemblies having a fuel
enrichment up to 4.5 weight percent of U-235. The
analysis methodology and results were described in the GE
report, GENE-512-92073, "Peach Bottom Atomic Power Station
Spent Fuel Storage K-infinity Conversion Analyses,"
November 1992. The method and the cross-section library
used consist of the GE MERIT computer code, using the
ENDF/B-IV cross section set, which is stated to have been
verified against extensive critical experiments. The
MERIT program is a three-dimensional Monte Carlo neutron
tracking code that calculates the system effective neutron
multiplication factor (K-effective) using a 190 group
cross-section library with the Haywood scattering kernel
of water. A spent fuel storage criticality validation is
performed for each reload to demonstrate that the reload
fuel assemblies meet incore K-infinity and rack K-
effective storage criticality requirements. Rack module
data is given in Table 10.3.2.
TS Amendment 287/290 approved use of neutron absorbing
inserts in the spent fuel pool (SFP) storage racks for the
purpose of criticality control in the SFPs. This amendment
modified TS 4.3, "Fuel storage" and added a new license
condition 2.C(14) to support the installation of NETCO-SNAP-IN neutron absorbing inserts into the individual cells of the existing PBAPS SFP storage racks. The
CHAPTER 10 10.3-3 REV. 26, APRIL 2017 installation of inserts addresses the degradation of the neutron absorbing material (Boraflex) previously installed
in the SFP racks. The rack inserts are manufactured by
NETCO using an aluminum and boron carbide composite
material produced by Rio Tinto Alcan, Inc.
Analyses of the safety considerations concerning the high density spent fuel storage racks are set forth in a
document entitled "Design Report of High Density Spent
Fuel Storage Racks for PECO Energy Company (PECO),
formerly Philadelphia Electric Company, Peach Bottom
Atomic Power Station Units 2 and 3, Revision 2," dated
July 21, 1986. This document describes the high density
spent fuel storage racks in detail and contains analyses
for seismic events, criticality concerns, structural
requirements, thermal and hydraulic requirements, and
postulated accident conditions associated with the high
density spent fuel racks. Additional analysis has been
performed that justifies application of the results of
this document for Spent Fuel Pool water temperatures as low as 40 F. The technical evaluation that allowed the use of GE-14 fuel in the Peach Bottom reactors is contained in ECR PB
99-02682. The acceptability of storing GE-14 fuel in the
Spent Fuel Pool is documented in "GE14 Spent Fuel Storage
Rack Analysis for Peach Bottom Atomic Power Station,"
Global Nuclear Fuel Document No. J11-03761-00-SFP, July
2000. GNF2 Fuel was evaluated in TS Amendment 287/290 for use of the NETCO-SNAP-IN inserts.
The dry storage cask consists of a fuel basket, a cask body, a protective cover, an overpressure system, penetrations with bolted and sealed covers for leak
detection and venting, closure bolts and locating pins.
The cask is designed to be lifted in a single failure
proof configuration.
Analysis of the safety considerations concerning the usage of the dry cask storage system is documented in the PBAPS
Independent Fuel Storage Safety Analysis Report (IFSSAR)
and PBAPS 10CFR 72.212 Report. These documents discuss in
detail the various design basis cask storage at PBAPS.
CHAPTER 10 10.3-4 REV. 26, APRIL 2017 10.3.4.1 High Density Spent Fuel Storage Racks The high density spent fuel racks are constructed of stainless steel materials and each rack module is composed
of cell assemblies, base plate, and base support assembly.
10.3.4.1.1 Cell Assembly
Each cell assembly is composed of (1) a full length enclosure constructed of 0.075 inch thick stainless steel, (2) sections of Bisco Boraflex which is neutron absorbing
material, and (3) wrapper plates constructed of 0.020 inch thick stainless steel. Additionally, NETCO-SNAP-IN inserts provide augmented neutron absorbing capability.
10.3.4.1.1.1 Cell Enclosure
The primary functions of the enclosure are to house fuel assemblies, to maintain the necessary separation between
assemblies for subcriticality and to provide structural
stiffness for the rack module. The inside square
dimension of the cell enclosure is 6.070 inches nominal
which accommodates either channeled or unchanneled fuel or
consolidated fuel assemblies. A partial plan view is
shown in Figure 10.3.6 and a partial elevation view is
shown in Figure 10.3.7.
10.3.4.1.1.2 Neutron Absorbing Material
The Bisco Boraflex manufactured by Brand Industrial Services provided the additional neutron absorbing media
required above that inherent in the rack structure
material. The Boraflex is fabricated to safety-related
nuclear criteria of 10CFR50, Appendix B, and it consists
of boron carbide particles as neutron absorbers held in
place by a nonmetallic binder. Boraflex contains an
initial minimum B 10 areal density of 0.021 gm/cm
- 2. It is a continuous sheet centered on the length of the active
fuel. Depending on the location of the cells in a rack
module, some cells have the Boraflex on all four sides, some of three sides and some on two sides. Cells with
four wrappers are located in the interior of the rack, cells with three wrappers are located on the periphery of
the rack, and cells with two (adjacent) wrappers are
located at the corners of the rack.
Since NETCO SNAP-IN rack inserts have been fully installed in the Peach Bottom Units 2 and 3 Spent Fuel
CHAPTER 10 10.3-5 REV. 26, APRIL 2017 Pool racks, Boraflex is no longer credited as a neutron absorbing material.
10.3.4.1.1.3 Wrapper Plate
The wrapper plates are attached to the outside of the cell enclosure by intermediate spot welding along the entire
length of the wrapper, forming the encapsulation of the
Boraflex. A water tight seal is not provided between the
wrappers and enclosures.
10.3.4.1.1.4 Neutron Absorbing Inserts The NETCO-SNAP-IN neutron absorbing inserts are manufactured by NETCO using an aluminum and boron
carbide composite material produced by Rio Tinto Alcan, Inc. The material contains 19% by volume of boron
carbide. The minimum certified areal density is 0.0105
grams/cm 2. An AA1100 aluminum alloy is used as a metal matrix to retain the boron carbide.
The inserts are designed to be an integral part of the
existing PBAPS spent fuel racks. The inserts are
nominally 0.075 inch-thick, are chevron shaped and have
a vertical length which is equal to the cell height of
the existing PBAPS spent fuel racks (169 inches). The
aluminum and boron carbide composite inserts function by
maintaining a greater than 90 degree bend angle when
formed, but are subsequently compressed to a 90 degree
bend angle when installed in the individual spent fuel
rack cells, which provides a bearing force against the
inside of the cell walls to retain the inserts in place.
10.3.4.1.1.4.1 Seismic and Structural Integrity A combination of analysis and testing has been used to
demonstrate acceptable structural and seismic
performance of the inserts.
The impact load of a fuel assembly on the neutron
absorbing inserts, generated by the horizontal
acceleration of a fuel assembly during a design-basis
seismic event, was determined to be 403 pounds per square inch (psi). Given that the NETCO-SNAP-IN insert yield stress is approximately 8000 psi, the deformation and subsequent failure of the insert due to seismically-
induced impact loads will not occur.
CHAPTER 10 10.3-6 REV. 26, APRIL 2017 To determine the stresses imparted on the cells of the existing PBAPS spent fuel racks by the inserts, the
limiting case involves the installation of the inserts
into the cells. The stresses imparted on the cell walls
during installation are not expected to exceed the
allowable stress. The additional stress that occurs
during installation of the inserts will have no effect
on the cell wall structural integrity and the stress
remains below the allowable value.
The increased load on the fuel racks from the inserts, which weigh approximately 18 pounds each, will be
insignificant and bounded by the existing design.
Analytical and confirmatory numerical analysis were used
to evaluate the stresses on the inserts during
installation. The stresses remained below the insert
material ultimate stress limit. Some instances were
identified of plastic deformation, particularly in the
wing and bend sections of the insert. However, sufficient elastic margin exists in the inserts, such
that adequate retention force is maintained between the
insert and cell walls.
Pre-installation testing demonstrated that adequate
retention force is maintained by the inserts, such that
they remain in place during normal (i.e., fuel handling)
and abnormal (i.e., design-basis seismic event) loading
conditions. In the unlikely event of warping or bowing
of an insert, any additional drag on a fuel assembly
will be recognized by a hoist load cell, which is
typically used during normal fuel handling activities.
Withdrawal testing showed that the inserts maintained a
static friction-based retention force well above the
established 200 pounds minimum removal criteria. During
a design-basis seismic event, the inserts must maintain
a retention force of 40.8 pounds to ensure that the
insert configuration remains unchanged, which is a 79.6%
reduction of the 200 pound minimum removal force
criteria. Over the 20-year expected life of the inserts, it is expected that the inserts will experience a stress
relaxation of approximately 50%.
CHAPTER 10 10.3-7 REV. 26, APRIL 2017 10.3.4.1.1.4.2 Fuel Handling Accidents The inserts have no effect on previously evaluated fuel handling accidents. This is based on the fact that the
installation of the inserts does not reduce the ability
of the cell wall or rack base plate to resist dynamic
impact loads resulting from a dropped fuel assembly, nor
does it affect whether a fuel assembly may become stuck
at the bottom of the existing racks.
An evaluation was performed to identify any previously
unanalyzed fuel handling accidents resulting from the
use of the tool used for installing and removing the
inserts. Due to similarities in geometry and the lower
weight of the insert tool and inserts compared to a fuel
assembly and grapple device used for normal fuel
handling activities, a postulated drop of the insert
tool and insert is bounded by previously analyzed fuel
handling accidents.
10.3.4.1.1.4.3 Criticality Analysis A SFP criticality analysis crediting the NETCO-SNAP-IN inserts was provided by Global Nuclear Fuel (GNF) report
NECD-33672P, Rev. 1, "Peach Bottom Atomic Power Station:
Fuel Storage Criticality Safety Analysis of Spent Fuel
Storage Racks with Rack Inserts." The analysis
determined a maximum k-effective of 0.92552 at a 95%
probability and 95% confidence level, and provides an
adequate reactivity margin to the regulatory k-effective
limit of 0.95.
Two computational methods were used by GNF in the
criticality analysis. GNF lattice design code TGBLA06
was used to calculate burned fuel compositions and the
in-core k-infinity values. The burned fuel compositions
were then used in MCNP-05P, the GNF proprietary version
of MCNP5, to obtain fuel storage rack k-effective
values. Tables 12, 13, and 14 of NEDC-33672P provide the biases
and uncertainties used to determine the maximum in-rack
k-effective. Biases are arithmetically added to the
calculated k-effective to account for conditions not
directly modeled in the base case analysis. Biases are
added for operational variables, abnormal or accident
conditions, and additional configurations. Uncertainty
components are statistically summed and then added to
the calculated k-effective. Uncertainties include
CHAPTER 10 10.3-8 REV. 26, APRIL 2017 manufacturing tolerances as well as computational uncertainties.
10.3.4.1.1.4.4 Abnormal or Accident Conditions The following abnormal accident conditions were
considered in the PBAPS SFP criticality analysis.
- 1. Missing NETCO-SNAP-IN insert, 2. Dropped fuel, 3. Damaged fue1, 4. No NETCO-SNAP-IN inserts on rack periphery, 5. Misplacement of a fuel assembly, 6. Lateral movement of a rack module,
- 7. Loss of SFP cooling, and
- 8. Inaccessible storage locations.
Analysis has determined the reactivity impact for the
above conditions and determined that they are either
bounded by other conditions or the corresponding
reactivity increase has been added to the calculated k-
effective.
10.3.4.1.1.4.5 Boraflex Credit for the Interim Period As part of a Peach Bottom license amendment (287/290, 5/21/16) for the NETCO SNAP-IN rack inserts, a license condition took credit for Boraflex during an interim period prior to installation of all inserts. Since NETCO SNAP-IN rack inserts have been fully installed in both Peach Bottom Unit 2 and Unit 3 Spent Fuel Pool racks, Boraflex is no longer credited as a neutron absorbing material.
GNF report 000N6365-R0, Revision 0, "Peach Bottom Atomic Power Station Units 2 and 3 Spent Fuel Pool Criticality Analysis Gap Sensitivity Study," provides a supplement to NEDC-33686P, Revision 1 that evaluates additional gap configurations. This analysis demonstrates that the results of NEDC-33686P, Revision 1 are bounding when compared to actual gap distributions in the pool.
10.3.4.1.2 Base Plate
The base plate is a 0.50 inch thick stainless steel plate with chamfered through holes centered at each storage
location which provides for a seating surface for the fuel
assemblies. These holes also provide passage for coolant
flow for each fuel assembly.
CHAPTER 10 10.3-9 REV. 26, APRIL 2017 10.3.4.1.3 Base Support Assembly
Each rack module is provided with base support assemblies which are located at the center of the four corner cells
within the module and at interior module locations to
distribute pool floor loading.
Each base support assembly is composed of a leveling block assembly, a leveling screw, and a support pad (see Figure
10.3.7). The top of the leveling block assembly is welded
to the bottom of the base plate. The leveling block
assembly is threaded at the bottom to accept the leveling
screw which sits in the support pad providing support for
the rack. The screw is remotely adjustable at rack
installation to obtain a level condition. The screw has
an adjustable range up to 1 inch. The leveling pad has a swivel joint to accommodate a maximum of 2 out-or-level condition of the pool liner.
The base support assemblies are welded to the bottom of the base plate at their appropriate support locations for
each rack (refer to Figure 10.3.7). The cell assemblies
are then positioned and welded to the top surface of the
base plate. The cell assemblies are positioned in a
checkerboard pattern with the space between four cell
assemblies forming a fifth storage locations. In addition
to being welded to the base plate, the vertical corners of
adjacent cells are welded to each other at two locations
along their length to form a integral structure. Along
the peripheral rows of the rack module, stainless steel
cover plates are welded between the cell assemblies to
enclose the non-cell locations. Some cover plates have
wrapper plates with Boraflex, identical to those used on
the cells, affixed to their inward side to satisfy
adjacent rack module criticality concerns. Each rack has
provisions for attachment of a lifting fixture for
installation and/or removal of the racks in the spent fuel
pool. The base plate has slotted holes at four locations
designed to accept lift rods which are inserted down
through the storage cells. The lift rods are connected at
the top of a lifting fixture.
The structure of the racks is designed to maintain the required spacing between stored fuel assemblies in the
event of impact of a fuel bundle dropped on the racks from
an elevation of 24 inches (maximum). For this case, the
integrity of the storage rack is not compromised and
damage to the racks is above the poison area; therefore,
CHAPTER 10 10.3-10 REV. 26, APRIL 2017 the criticality requirements are not violated. The structure of the racks is also analyzed for effects of the
impact of a fuel bundle dropped through an empty storage
cavity. The fuel drop accident analysis shows that the
structure absorbs the energy and the criticality
requirements are not violated. Analyses are also
conducted of the stresses on a storage rack due to maximum
uplift of the refueling crane on a fuel bundle which is
stuck. The evaluation of this case showed no permanent
deformation of the storage rack. The high density spent
fuel storage racks are seismic Category I equipment as
defined in NRC Regulatory Guide 1.13. These racks are
designed to withstand the effects of a maximum credible
earthquake and remain functional, in accordance with NRC
Regulatory Guide 1.29 and the Code of Federal Regulations, Title 10, Part 100.
The basic design criteria for the spent fuel storage rack are outlined by the NRC position paper. The NRC position
paper entitled "OT Position For Review and Acceptance of
Spent Fuel Storage and Handling Applications" dated April
14, 1978, as amended by the NRC letter dated January 18, 1979, offers two codes for deriving the allowable
stresses. The two codes are AISC Code or the ASME Code
III Subsection NF. The structural analysis herein is
based on the allowable stresses as outlined in ASME Code
III Subsection NF. The results of the seismic and
structural analyses are interrelated as the loads of the
seismic analysis are used in the structural analysis to
calculate stresses. The resulting margins of safety are
positive and satisfy the requirements of the ASME code.
The pool floor loads resulting from the seismic and
structural analyses also satisfy the requirements of the
PECO specification for the spent fuel storage racks. The
displacement results of the seismic analyses are used in
the lift-off stability calculation and show that factor of
safety against overturning is greater that the 1.5 minimum
requirement of the NRC position paper. The rack seismic
displacements are used in conjunction with thermal
displacements to show that there is no rack-to-rack, rack-
to-pool floor obstruction collision.
10.3.4.2 Spent Fuel Pool
The fuel pool together with the dryer-separator storage pool form a channel-shaped beam supported in the middle by
the biological concrete shield structure and at the outer
ends by the building walls.
CHAPTER 10 10.3-11 REV. 26, APRIL 2017 The pool floor carries a live load in addition to the water load. A system of large steel shapes is used to support the weight of the wet concrete only. Deep beam
action was checked and interactions of elements accounted
for. A finite element analysis was performed to check
temperature stresses in combination with other loads.
Hydrodynamic effects of water were also included in the
analysis.
Once the integrity of the system was ascertained, local stresses, embedments, connections, girder deflection, and
discontinuities were investigated.
The pool is lined with stainless steel and is designed to preclude inadvertent loss of water from the pool.
There are no connections to the fuel storage pool which could allow the fuel pool to be drained below the pool
gate between the reactor well and the fuel pool when the
pool gate is in place or below 10 feet above the top of
active fuel. Lines extending below this level are
equipped with syphon breaker holes to prevent inadvertent
pool drainage. Systems for maintaining water quality and
quantity are designed so that any maloperation or failure
of such systems will not cause fuel to be uncovered.
The fuel storage pool is designed to seismic Class I criteria and so that no single failure of structures or
equipment will cause the inability (1) to maintain
irradiated fuel submerged in water, (2) to reestablish
normal fuel pool water level, or (3) to safely remove
fuel. To prevent leakage, the pool is lined with
stainless steel. In addition to providing a high degree
of integrity, the lining is reinforced to withstand forces
that might occur when the transfer cask is moved in the
cask storage area.
Interconnected drainage paths are provided behind the liner. These paths are designed (1) to prevent pressure
buildup behind the liner plate, (2) to prevent the
uncontrolled loss of contaminated pool water to the
secondary containment, and (3) to provide expedient liner
leak detection and measurement.
Protection of the pool liner in the cask storage area for the normal cask lowering operation is provided by a 1-inch
thick steel wearing plate. This will prevent any damage
to the liner over plant life occasioned by normal fuel
cask handling. Additionally, interlocks are provided to
prevent the crane trolley, with a predetermined load on
CHAPTER 10 10.3-12 REV. 26, APRIL 2017 its main hook, from passing over the fuel pool. Strict administrative control is used for bypassing the
interlocks during cask handling operations.
The reactor building crane main hook and the lifting device associated with the cask are of a single failure
proof design such that a single failure will not result in
dropping the load. The available makeup water sources to
the spent fuel pool and associated flow rates are
presented in Table 10.3.1.
The spent fuel pool cask pit area restraining structure has been analyzed to withstand cask impacts due to
postulated design events. Cask placement in the center of
the cask pit area is controlled in accordance with station
procedures. This includes maintaining the appropriate
centering tolerance as well the angular placement of the
cask with respect to the trunnions and the restraining
structure.
10.3.4.3 Fuel Pool Level Alarms
Low water level alarms are provided locally and in the main control room in the event of water loss. The low
water level alarms are part of the fuel pool cooling
system. As a backup, flow alarms are provided in the
drain lines of the reactor vessel to drywell seal, drywell
to concrete seal, and fuel pool gate to detect leakage.
10.3.4.4 Dry Cask Storage
Dry Cask Storage of spent nuclear fuel has been evaluated for PBAPS. This program meets the requirements of 10CFR
72 and utilizes the General License issued under 10CFR 72.
10.3.4.4.1 Dry Cask Storage Rigging and Handling
The storage cask is handled by the Reactor Building Crane in a single failure proof configuration. Other cask
components such as the lid and basket hold down ring are
also handled in a single failure proof configuration
unless specifically evaluated otherwise in accordance with
UFSAR Section 10.A.11. All cask movement in the Reactor
Building will be consistent with NUREG-0612, over
designated safe load paths and either with the single
failure proof Reactor Building Crane or with the cask
transporter with the cask at the appropriate analyzed
heights. All cask movement in the Reactor Building will
be on designated safe load paths to ensure that cask drop
loads where applicable will not affect the safety of the
CHAPTER 10 10.3-13 REV. 26, APRIL 2017 PBAPS plant. The cask transporter is not single failure proof but has been evaluated for cask drops and found to
be acceptable. The cask is rigged in a single-failure-
proof configuration while in the spent fuel pool by
ensuring a procedurally controlled amount of water is
removed from the loaded cask.
10.3.4.4.2 Dry Cask Structural Considerations
A calculation was performed to consider the acceptability of a drop of a loaded cask from a cask transporter in the
Reactor Building onto the Elevation 135' floor. The
calculation demonstrates the structural adequacy of the
impact floor and the absence of effects on safety related
equipment beneath it at a lift height less than 2.5".
Superficial damage to the floor at el. 135' would be
repaired as appropriate. The lift height is procedurally
controlled.
The cask will be rigged in a single failure proof configuration while rigged from the reactor building crane
in the Reactor Building hatchway. The placement of the
cask in the hatchway will be such that credible cask
swinging in the hatchway during a seismic event will not
result in cask / plant impacts. Therefore, damage to the
plant is precluded.
The structural capacity of the Reactor Building floor at el. 234' has been evaluated for cask operations. These
analyses assumed a coincident maximum credible earthquake.
The designated cask laydown area is controlled by spent
fuel procedures.
The restraint structure (without the vertical guides) in the spent fuel cask pit area of the fuel pool was modeled
using finite element analysis techniques and evaluated for
structural adequacy under the effects of combined loadings
postulated for the structure consistent with the PBAPS
UFSAR Appendix C.
Cask impacts on the restraint structure due to postulated seismic (MCE) event were evaluated for the suspended cask (pendulum effect during hoisting) and for the free-
standing cask located on the wear plate (sliding and
tipping) and found acceptable.
Administrative controls and procedural requirements assure that the cask is appropriately centered within the
specified tolerance during the move-in/move-out and cask
placement operations.
CHAPTER 10 10.3-14 REV. 26, APRIL 2017 Passage of the cask transporter over the access road and haul path, with or without a cask, does not physically
affect systems, structures and components (SSCs) of the
PBAPS plants. The haul path and access road are well
defined by roadway markings to guide the cask transporter
driver. In the event the transporter fails or is
inadvertently driven off the designated haul path, off-
normal procedures exist to ensure that operations are
halted. Cask drops along the transport route were
evaluated and found acceptable.
10.3.4.4.3 Dry Cask Fuel Pool Operations
Cask insertion into the fuel pool will be controlled by procedures to ensure that the fuel pool level is
maintained. Fuel pool level will be procedurally
controlled by a combination of activities, including as
necessary, turning off the fuel pool cooling system, lowering and raising the fuel pool level control weir
gate, draining/filling the skimmer surge tanks, and
controlling the filling and draining of the casks as
appropriate. None of the involved activities can result
in inadvertent draining of the fuel pool.
Before each cask loading campaign a plan will be prepared that specifies the fuel assemblies to be moved from the
pool into the cask, and their specified assigned location
in the cask fuel basket. Selection of the correct fuel
assemblies from the fuel pool racks and correct placement
in the cask basket will be controlled by procedural
methods. Any mispositioning that could occur would be
detected by confirmatory monitoring. Normal site fuel
movement procedures will be used by appropriately
qualified personnel.
The elevation of the fuel bundles must be increased in order to provide clearance over the lip of the open dry
storage casks, which are taller than the cask for which
the spent fuel pool was originally designed.
The new normal-up setpoint needed to load an ISFSI cask will be used only for fuel cask loading and unloading
operations.
10.3.4.4.5 Dry Cask Storage Design Basis Events
The postulated events that could occur during cask operations are discussed in the IFSSAR and 10CFR 72.212
Report. The following is a summary of those events that
CHAPTER 10 10.3-15 REV. 26, APRIL 2017 could have potential impacts or interaction with PBAPS 2 and 3.
Fuel Bundle Drop
A design basis fuel bundle drop was evaluated and found to be bounded by existing accident analysis. Criticality, radiological releases and effects on the ISFSI cask and
fuel pool liner were evaluated and found acceptable.
Wrong Fuel Insertion
The TN-68 IFSSAR evaluates the impact on the cask if an incorrect fuel assembly is loaded. These events are
precluded due to administrative control and training of
personnel involved with loading the cask. Procedural
controls will ensure that only allowed fuel is selected
for loading into a cask. Additionally, verifications of
fuel being loaded will ensure that incorrectly loaded fuel
does not go undetected. The TN-68 IFSSAR concludes that
fuel with heat generation greater than allowed is not a
concern to the cask as long as the cask is submerged in
the fuel pool. Because there are multiple layers of fuel
verification prior to placing the lid on the cask, there
is no concern for incorrectly loaded fuel with a higher
than allowed heat generation rate. The TN-68 IFSSAR
evaluates the loading of a fuel assembly with higher than
allowed enrichment and determines the impact on the
criticality margin in the cask. For worst case conditions
and placement into the cask, a fuel assembly with an
initial enrichment of 5.0% was evaluated and determined to
not cause any criticality concerns. Loading verifications
prior to placing the lid on the cask would detect the
incorrectly loaded assembly and remove it prior to
continuing loading operations.
Cask Drop / Tip
Various scenarios involving the potential for cask drop/tip were evaluated. In no case, does the cask tip
over. For the drop scenarios, the cask remains within its
licensed design basis.
Natural Events
The cask and associated support equipment has been evaluated for various natural events. In all cases, the
cask was shown to not tip over. Cask drops are possible
when suspended on the transporter, however, these drops
have been evaluated as discussed earlier. Impacts of
CHAPTER 10 10.3-16 REV. 26, APRIL 2017 natural events have been considered in the analysis of the cask while in the Reactor Building and have been found to
not result in an uncontrolled lowering of the cask or
other damage to plant equipment due to cask impacts.
Fires and Explosions
The TN-68 cask is designed to withstand various fires and explosions. The cask is not combustible and therefore
poses no new significant fire or explosion threat to the
plant. The transporter has been evaluated and found
acceptable.
Cask Seal Leak
The worst case seal leakage has been demonstrated in the TN-68 SAR to be well below regulatory limits.
Cask Loading Operations Issues
During cask loading, various contingency actions may be required to perform in the event (for example) of the
inability to get the cask adequately drained and filled
with helium or the inability to meet the leak tightness
requirements. Procedures exist to direct actions in these
off-normal conditions. These conditions are governed by
TN-68 Tech Specs to ensure that cask parameters are not
exceeded. These actions may include returning the cask to
the spent fuel pool for unloading. Because there are
adequate controls on the cask in these off-normal
conditions, there is no impact to PBAPS 2 and 3
operations.
Cask Unloading Operations
If the cask is required to be unloaded or a seal repair is required, the cask is returned to the Reactor Building.
The cask transport route follows the same load path as for
transport to the ISFSI. A cavity gas sample shall be
obtained and analyzes and the cask depressurized to a
nominal atmospheric pressure. The cask is refilled with
water and the outlet line from the cask is piped below the
surface of the pool with a sparger attached at the
discharge end. Steam is quenched by the relatively cool
fuel pool water. Helium bubbles released from the cask
would rise to the surface to mix with the refueling floor
atmosphere and be dissipated by the HVAC system.
Particulates released from the cask would be scrubbed out
by the fuel pool water and filtered by the Spent Fuel Pool
Cooling and Cleanup System (SFPCCS). Except for Kr-85 and
CHAPTER 10 10.3-17 REV. 26, APRIL 2017 I-129, the constituents of the gas gap from leaking rods is scrubbed out by the fuel pool water and handled by the
SFPCCS. The dose due to noble gases from the unloading
operation, assuming 100% failed fuel in the cask, results
in a lower offsite dose than that from the refueling
accident analyzed in Section 14.6.4 of the PBAPS SAR.
The heat added to the pool water is well within the cooling capacity of the SFPCCS. It is capable of
receiving a full core offload directly from the reactor
and still keep the pool temperature at an acceptable
value. The heat from only 68 assemblies plus the latent
heat stored in the cask materials is insignificant
compared to that from a full core offload.
10.3.4.4.6 Dry Cask Storage Programs
As required by 10 CFR 72.212, the PBAPS radiation protection, emergency preparedness, security and training
programs were updated to incorporate dry cask storage.
10.3.4.4.7 Dry Cask Operations
Helium is used during dry cask operations. The exhaust of helium into the reactor building atmosphere is not a
concern to the plant since the helium is an inert gas and
readily dissipates. The exhaust will be vented to the Fuel
Floor area for further processing in the ventilation
system.
The cask is drained of water when it is raised to the fuel pool surface water level. Evaluations and procedural
controls have been developed to ensure that the cask is
not raised out of the pool such that the reactor building
crane or lift beam would exceed its single failure proof
rating.
The casks are dried to ensure that long term corrosion of the cask is minimized. The vacuum pumps exhaust will be
appropriately monitored by radiation protection personnel
and filtered as necessary. The vacuum pump discharge is
directed to the fuel floor area for further processing by
the refuel floor ventilation system. Cask over-
pressurization with helium is not a concern when using
procedurally controlled standard bottles and pressures.
This is due to the large volume of the cask compared to
the small volume of the helium bottle.
Leak testing is performed using calibrated leak testing equipment. This equipment is appropriately controlled and
CHAPTER 10 10.3-18 REV. 26, APRIL 2017 poses no significant risk to the plant. Leak testing ensures that the cask is properly sealed for transport and
storage outside of the reactor building.
The operations skid is designed to facilitate the evaluations required to drain, dry, inert, and test a
spent fuel cask. This equipment is not safety related.
Appropriate I&C measuring and test equipment will be
controlled in accordance with procedures and will be
within calibration frequency.
10.3.5 Safety Evaluation
The design of the spent fuel storage racks and dry storage casks provides for a subcritical effective multiplication
factor (k eff) for both normal and abnormal storage conditions. Under any condition the k eff is equal to or less than 0.95. The spent fuel pool concrete structure, as well as each spent fuel storage rack and fixture loaded
with fuel, are designed to seismic Class I criteria to
withstand the maximum credible earthquake.
The spent fuel pool is adequately protected from the effects of a turbine generated missile. The probability
of a turbine generated missile is small and is detailed in
Section 11.2. The fuel pool is protected against low
trajectory missiles by thick concrete walls between the
turbine and the pool as well as the thick concrete pool
walls. Once a high trajectory missile is generated, the
possibility of it landing in the pool is in the range of
10-4. Therefore, the combined risk to the fuel pool from a high trajectory turbine missile is insignificant.
The spent fuel pools are designed with substantial capability to withstand the effects of a tornado, including tornado-generated missiles. Discussion of this
capability is provided in Paragraph J.5.2.
Additional information is provided in Topical Reports APED-5696, "Tornado Protection for the Spent Fuel Storage
Pool" (General Electric, November, 1969) and "Tornado
Criteria for Nuclear Power Plants" (Bechtel Corporation, July, 1969).
The spent fuel storage pools are located in the reactor buildings which serve as secondary containment for the
reactors (subsection 5.3). Each reactor building is
designed to control leakage from the building and provides
filtration, through the standby gas treatment system, to
limit radioactive discharges in the event of an accident.
CHAPTER 10 10.3-19 REV. 26, APRIL 2017 Ventilation air from the spent fuel pool area is not normally filtered prior to exhaust to the atmosphere. The
standby gas treatment system is described in paragraph
5.3.3.
The consequences and assumptions used in evaluating a refueling accident are presented in paragraph 14.6.4. The
analysis provided in subsection 14.6.4 uses conservative
assumptions, similar to those provided in Regulatory Guide
1.25, to demonstrate that releases from a postulated
refueling accident result in doses which are well within
10CFR100 limits.
Provisions are made for level detection to ensure the fuel in the spent fuel storage is covered with sufficient water
for radiation shielding. Leakage detection
instrumentation is also provided to ensure an adequate
fuel pool water level is maintained. The design of the
spent fuel pool structure is such as to prevent
inadvertent draining of the pool.
The radiation levels are monitored in the refueling floor exhaust duct. Both low and high radiation signals are
alarmed in the control room and a high-high radiation
signal isolates the duct and initiates the standby gas
treatment system.
The high-density SFP storage racks utilize Boraflex as a
neutron absorber material for reactivity control. Due to
Boraflex degradation, PBAPS implemented an ongoing
Boraflex monitoring program, to include RACKLIFE
simulation of the rack degradation and blackness testing
using the BADGER B-10 areal density measurement system.
A SFP rack insert program has also been implemented that
will replace Boraflex. The effect of plant operation at
100% rated thermal power on Boraflex degradation is
accounted for by the Boraflex monitoring program until
installation of the SFP rack inserts is completed. A
reduction in the amount of Boraflex in the SFP racks
will reduce the criticality margin such that actions are
required to ensure that the Licensing Basis requirements
continue to be met. To ensure the SFP storage racks can
maintain criticality margin in accordance with the PBAPS
Technical Specification 4.3.1.1.b requirement of 5
percent (K eff 0.95), the peak in-core fuel bundle K inf is limited as follows:
1.A peak in-core fuel bundle K inf limit of 1.235 has been established and applies until all of the SFP
rack inserts are installed.
CHAPTER 10 10.3-20 REV. 26, APRIL 2017
- 2. A peak in-core fuel bundle K inf limit of 1.270 has been established and applies after all of the SFP rack inserts are installed.
The peak in-core K inf limit for the fuel bundles used in the representative equilibrium cycle core design is
1.2095, which is bounded by the K inf limits of 1.235 and 1.270. Therefore, these bundle K inf limits ensure the SFP criticality margin is maintained before and after
all of the SFP rack inserts are installed.
Dry cask storage casks were evaluated for various design basis events and normal conditions and found acceptable in
accordance with the IFSSAR and 10CFR 72.212 Report.
10.3.6 Inspection and Testing
Dry storage casks are appropriately inspected and tested to ensure design basis assumptions are met.
The in-service inspection program for the spent fuel storage racks involves periodic assessment of neutron
poison material performance.
10.3.6.1 Boraflex Inspection and Testing
This assessment may utilize jacketed Boraflex specimens contained in surveillance coupon assemblies hung on the
periphery of a rack module.
A computer based Boraflex performance model and direct measurement of the B-10 areal density of representative
in-service spent fuel storage rack panels may be used in
conjunction with or in replacement of coupon inspection.
10.3.6.2 Neutron Absorbing Inserts Surveillance Program The rack insert surveillance program is designed to
monitor the physical properties of the insert material by
performing periodic physical inspection and neutron
attenuation testing to confirm the ability of the material
to perform its intended function. 10.3.6.2.1 Fast Start Coupon Surveillance Program Exelon initiated a "Fast Start" coupon surveillance
program at LaSalle County Generating Station to provide
early performance data on the coupon exposure to maximum
temperature and gamma irradiation. The program consists of
CHAPTER 10 10.3-21 REV. 26, APRIL 2017 24 coupons suspended inside of a spent fuel storage rack cell and surrounded in all adjacent cells with freshly
discharged fuel. Two of the coupons will be removed
approximately every six months for testing, inspection and
comparison to their pre-installed condition. Initial
results showed essentially no change in the coupon
characteristics. Because the spent fuel pool chemistries
at PBAPS are similar to LaSalle, this program provides
information on initial material performance and is a basis
for confidence that early insert response to the SFP
environment is acceptable.
PBAPS will monitor LaSalle's program to identify any
unanticipated insert material performance issues including
review of their coupon test reports. Information obtained
will be used to evaluate the long-term and the full rack
insert surveillance programs at PBAPS and make any
necessary modifications.
10.3.6.2.2 Long-Term Coupon Surveillance Program
The long-term coupon surveillance program consists of a
specially designed monitoring tree to which a series of
surveillance coupons are attached. The monitoring tree, placed within the PBAPS spent fuel pools, will reside
there as long as the spent fuel storage racks with NETCO-SNAP-IN rack inserts continue to be used. Periodically, as described below, coupons will be removed and sent to a qualified laboratory for testing.
CHAPTER 10 10.3-22 REV. 26, APRIL 2017 Table 10.3.6.1 Long-Term Surveillance Coupons Coupon Type NumberObjective General 48 (See next Table)
Bend 24 Track effects along bend radii Galvanic (bi-metallic) 24 Trend galvanic corrosion with 304SS, Inconel 718 and Zircaloy coupons Specific coupons will be removed from the tree on a frequency
schedule in the following tables. The general coupons will be
subject to pre- and post-examination according to the following:
CHAPTER 10 10.3-23 REV. 26, APRIL 2017 Table 10.3.6.2 Long-Term Surveillance General Coupon Characterization
Pre-Characterization Post-Characterization Acceptance Criteria Visual (high
resolution
digital photo) X X Evidence of Visual
indications Dimension X X Min.
thickness:
0.005 inch less than
nominal thickness
Length Change: Any
change of +/-0.02
inch Width Change:
Any change of +/-
0.02 inch Thickness
Change: Any change
of +0.010 inch/-
0.004 inch Dry Weight X X Any change of +/- 5%
Density X X Any change of +/- 5% Areal Density X on select coupons X 0.0102 Boron-10 g/cm 2 minimum loading Weight Loss X Any change of +/- 5%
Corrosion Rate X < 0.05 mil/yr Microscopy X as required At the discretion of the test engineer Bend Coupon
Stress Relaxation X 50% stress reduction (to maintain 100 lbf
retention force)
The frequency for coupon inspection is shown in the following table.
CHAPTER 10 10.3-24 REV. 26, APRIL 2017 Table 10.3.6.3 Frequency for Coupon Inspection Coupon Type First Ten Years After 10 Years with Acceptable Performance General 2 coupons every 2 years 2 coupons every 4 years Bend 1 coupon every 2 years 1 coupon every 4 years Galvanic Couples 304
Stainless
Zircaloy Inconel 718 1 couple every 6 years
1 couple every 6 years
1 couple every 6 years 10.3.6.2.3 Full Rack Insert Surveillance Inspections Two rack inserts will be visually inspected by camera at the
frequency of the general coupon removal schedule described
above to visually monitor for physical deformities such as
bubbling, blistering, corrosion pitting, cracking, or
flaking. Special attention will be paid to development of
any edge or corner defects.
A region of high duty spent fuel storage rack cell locations
will be identified and will be monitored for fuel insertion
and removal events to ensure that their service bounds that
of the general population of storage locations. Once every
10 years, an insert will be removed from this region and will
be inspected for thickness along its length at several
locations and be compared with the as-built thickness
measurements of the removed insert to verify it has sustained
uniform wear over its service life. After the inspection, the
insert will not be reinstalled.
CHAPTER 10 10.3-25 REV. 21, APRIL 2007 TABLE 10.3.1 AVAILABLE MAKEUP WATER SOURCES
Rate Source Route (gpm)
Torus One RHR pump to fuel 10,000 pool (1)
Refueling water One refueling water pump 1,650 storage tank and/
to reactor well header, to or condensate fuel pool cooling pumps, to storage tank bypass filter, to fuel pool
Condensate storage High-pressure decontami-25 tank nation pump to fuel pool
Condensate storage Fuel pool makeup from con-60 tank densate transfer pump
Demineralized To demin. water supplies 150 water storage in service boxes
Total demineralized 1,885 water available
immediately
Total demineralized 11,885 water available
after 1 hr (2)
River water High-pressure service water 18,000 pumps via RHR cross-tie (2,3)
River water Fire waterhose stations 70 (2-3 in)
Total river water 70 available immediately
Total river water 18,070 available after 1 hr (3)
(1) Approximately 1 hr is required to install the removable spool before supply can be used.
(2) Can only be used if plant is shut down and RHR cooling is with RHR pumps A and/or C.
(3) Alternate to torus water using RHR pumps.
CHAPTER 10 10.3-26 REV. 21, APRIL 2007 TABLE 10.3.2 RACK MODULE DATA (PER UNIT)
QTY
ARRAY STORAGE LOCATIONS
RACK ASSY
DIMENSIONS (INCHES)
DRY WEIGHT (LBS)
PER RACK ASSY 1 9 x 14 126 54 x 89 x 180 10,000 2 10 x 14 280 64 x 89 x 180 11,200 1 11 x 14 Mod. 119 70 x 89 x 180 9,500 1 12 x 15 180 76 x 95 x 180 14,400 1 12 x 17 204 76 x 107 x 180 16,300 2 12 x 20 480 76 x 126 x 180 19,200 2 15 x 19 570 95 x 120 x 180 22,800 1 17 x 20 340 107 x 126 x 180 27,200 4 19 x 20 1,520 120 x 126 x 180 30,400 15 racks 3,819
Storage locations center-to-center spacing (inches) 6.28 Storage cell liner dimension (inches) 6.07
Intermediate storage location inner dimension (inches) 6.12
CHAPTER 10 10.4-1 REV. 26, APRIL 2017 10.4 TOOLS AND SERVICING EQUIPMENT 10.4.1 Introduction
All tools and servicing equipment necessary to meet the reactor
general servicing requirements are supplied for efficiency and
safe serviceability. The flow chart in Figure 10.4.1 defines in a
general way the steps that make up a routine refueling outage. The
heavy lines on the chart define the critical path in a normal
outage. Deviations to this path may be encountered under abnormal
circumstances. The following paragraphs describe the use of some
of the major tools and servicing equipment.
10.4.2 Fuel Servicing Equipment
Two fuel preparation machines located in each fuel storage pool
are used to remove and install channels to support inspection or
servicing of fuel assemblies. The fuel preparation machines are
also used for the placement of new fuel assemblies into the spent
fuel pool. These machines are designed to be removed from the
pool for servicing.
An equipment support railing is provided around the pool periphery
in order to tie off miscellaneous service equipment and for
personnel safety. Equipment lugs fabricated as part of the pool
liner are required for fixtures that might later be desired by
plant operating personnel. In addition, a curb with a plate of
thick stainless steel on top is provided around the entire
periphery of the refueling volume. Additional equipment may be
mounted by welding to, or drilling into, the plate. The curb may
be used as an additional support or tie-off area. Cable ways are
recessed into the floor around the pool periphery with openings to
pass cables into the pool from underneath this curbing.
The new fuel inspection stand is provided to restrain the fuel
assembly in a vertical position for inspection. The inspection
stand can hold two assemblies. The general purpose grapple is a
small, hand actuated tool used generally with fuel. The grapple
can be attached to the reactor building auxiliary hoist, the jib
crane, and the auxiliary hoists on the refueling platform. The
general purpose grapple is used to place new fuel in the
inspection stand and transfer it to the fuel pool.
A channel handling boom, with spring loaded takeup reel, is used
to assist the operator in supporting the weight after the channel
is removed from the fuel assembly. The boom is located between
the two fuel preparation machines. With the channel handling tool
attached to the reel, the channel may be conveniently moved
between fuel preparation machines.
CHAPTER 10 10.4-2 REV. 26, APRIL 2017 The complete channeling procedure is as follows. Using the refuel platform and the mast grapple, a spent fuel assembly is lifted
into the fuel preparation machine with the carriage lowered.
After raising the assembly to its high position, the channel is
unbolted from the fuel assembly using the channel bolt wrench
furnished for this purpose. This wrench is used to unscrew the
bolt and capture it. The channel handling tool is attached to the
channel handling boom and lowered to the channel. The tool is
attached to the channel triangular corner tabs by expanding two
fingers on the tool. The channel is then held, and the fuel
preparation machine carriage is lowered, causing the fuel bundle
to slide down out of the channel. The channel is then positioned
over the other fuel preparation machine, containing a new fuel
assembly, and the procedure is reversed. A channel storage rack
for accumulating channels is located on the wall between the fuel
preparation machines. A channel check gage may be mounted on the
wall adjacent to the fuel machines so the operator can check
channels. The channeled fuel is stored in the pool storage racks, ready for insertion in the reactor.
10.4.3 Servicing Aids
General and local area underwater lights are provided to
illuminate the internal region of the reactor vessel. Drop lights
are used for intense radial illumination where needed. These
lights are small enough in diameter to fit into fuel channels or
control blade guide tubes. A portable underwater television
camera and monitor are part of the plant optical aids. The
transmitted image can be viewed on the refueling platform. This
remote display assists in the inspection of the vessel internals, and general underwater surveillance in the reactor vessel and fuel
storage pool. General purpose, clear plastic viewing aids that
will float are used to break the water surface for better
visibility.
A portable underwater vacuum cleaner is provided to assist in
removing debris and miscellaneous objects from the pool floor or
the reactor vessel. The pump and the filter unit are completely
submersible for extended periods. Fuel pool tool accessories are
also provided to meet servicing requirements.
10.4.4 Reactor Vessel Servicing Equipment
Reactor vessel servicing equipment is supplied for safe handling
of the vessel head and its components, including nuts, studs, bushings, and seals.
The drywell head strongback is used for lifting the drywell head
and mirror insulation. Cruciform in shape, with four equally
spaced lifting points, the strongback is designed to keep the
CHAPTER 10 10.4-3 REV. 26, APRIL 2017 drywell head level during lifting and transport. Redundant rigging is used to connect the drywell head to the single-failure-
proof strongback, resulting in a single-failure-proof lift. An exception to the single-failure-proof requirements in ANSI N14.6-1978 is that two structural features added during modifications to upgrade this device were not impact-tested. (Ref. ECR 13-00378)
The Reactor Pressure Vessel (RPV) head strongback/carousel is used
for lifting the vessel head. The strongback/carousel is an
integrated piece of equipment consisting of a cruciform shaped
strongback, a circular monorail, and a circular storage tray.
The strongback is a box beam structure which has a hook box with
three pins in the center for engagement with the reactor building
crane main hoist hook. Each arm has a liftrod for engagement to
the four lift lugs on the RPV head. The monorail is mounted on
extensions of the strongback arms and four additional arms equally
spaced between the strongback arms. The monorail circle matches
the stud circle of the reactor vessel and it serves to suspend
stud tensioners and nut handling device.
The head strongback carousel service the following functions:
Lifting of Vessel Head - The strongback, when suspended from the reactor building crane main hook, transports the RPV head
plus the carousel with all its attachments between the
reactor vessel and storage on the pedestals. The strongback
and its connections to the RPV head are single-failure-proof.
One exception to the single-failure-proof requirements in
ANSI N14.6-1978 is that the hook and load pin material for
this device was not impact-tested as specified in ANSI N14.6-
1978. (Ref. ECR 13-00378)
Tensioning of Vessel Head Closure - The carousel, when supported on the RPV head on the vessel can carry up to eight
tensioners, its own weight, the strongback, storage of nuts, washers, thread protectors, and associated tools and
equipment. The stud tensioners are suspended equally spaced
from a monorail above the vessel stud circle. Each tensioner
has an air-operated hoist with individual controls.
The head holding pedestals are designed to support the vessel head
to permit seal replacement and seal surface cleaning and
inspection. The mating surface between vessel and pedestal is
selected to minimize the possibility of damaging the vessel head.
A reactor servicing platform permits the operator to work at a
level just above the reactor vessel flange, and permits servicing
access for the full core diameter. A service platform support is
provided, which rests on the vessel flange surface, and serves as
CHAPTER 10 10.4-4 REV. 26, APRIL 2017 both a track for the servicing platform and as a seal surface protector.
A separate seal surface protector made of aluminum is provided to
protect the sealing surface of the reactor pressure vessel flange
when the service platform or its track is not used.
A stud tensioner assembly is provided, and consists of four
tensioners transported by the reactor building crane main hoist or
the head strongback/carousel. The tensioners are controlled by a
hydraulic unit with pressure gages. Each tensioner contains:
- 1. An integral nut wrench for rotating the nut.
- 2. One stud elongation gage plus one elongation rod per stud to permit initial and periodic pressure/stretch
indication.
10.4.5 In-Vessel Servicing Equipment
The single or multiple instrument strongback is attached to the
reactor building crane auxiliary hoist and is used to lift
replacement in-core detectors from their shipping container. The
instrument handling tool is attached to the in-core detector by
the operators on the refueling platform. The single or multiple
strongback initially supports the in-core detector(s) as they are
lowered into the vessel, and the in-core detector is then
decoupled from the strongback. Final in-core detector insertion
is accomplished with the instrument handling tool. The instrument
handling tool is used for removing and installing fixed in-core
detectors, as well as for handling neutron sources and the WRNM
dry tubes.
10.4.5.1 Reactor Cavity Work Platform
The Reactor Cavity Work Platform (RCWP) is a tool used to allow in
vessel inspection/activities concurrent with fuel movement thereby
reducing refueling critical path time. The RCWP is a stainless
steel structure which consists of four (4) quadrants connected
using three sets of splice plates. The RCWP is stored in four (4)
specially designed sea land containers, and is brought to the
refueling floor and assembled prior to use.
The RCWP has an octagonal shaped structural framework with eight
radial legs which support four (4) personnel work baskets. The
eight support legs rest on the reactor building elevation 234'-0"
floor slab. The platform, when placed in the reactor cavity above
the open reactor pressure vessel (RPV), is slightly submerged into
the reactor cavity water. The bottom of the work baskets is
approximate elevation 231'-0" which provides approximately 7'-0"
CHAPTER 10 10.4-5 REV. 26, APRIL 2017 clearance to the underside of the refueling bridge. The RCWP legs have the capability to both extend/retract and rotate in order to
avoid obstructions such as the electrical pits and refuel bridge
gearbox, which are present on the operating deck. The RCWP has a
30-degree refueling opening in the direction of the fuel pool toallow for refuel bridge mast and fuel bundle movement whileperforming in vessel activities. Electrical power and station air
outlets are also provided in the RCWP work baskets.
10.4.6 Refueling Equipment
The refueling platform is used as the principal means of
transporting fuel assemblies back and forth between the reactor
well and the storage pool. The platform travels on rails
extending along each side of the reactor well and fuel pool. The
platform supports the fuel grapple and the frame-mounted and
monorail auxiliary hoists. The grapple is suspended from a
trolley system that can traverse the width of the platform.
Platform operations are controlled from either auxiliary hoist
control pendant or the fuel grapple controller consoles. The
platform contains a position-indicating system that indicates the
position of the fuel grapple over the core. The platform is
prevented from contacting the fuel pool and reactor walls by a
boundary zone interlock system.
One-half ton auxiliary hoists are mounted on both the reactor well
side of the refueling platform and on the platform trolley. These
hoists normally can be used with appropriate grapples to handle
control rods, in-core detectors, sources, and other internals of
the core. The auxiliary hoists can also serve as a means of
shuffling fuel elements and other equipment within the pool and
reactor.
A single operator is capable of controlling all the motions of
the platform required to handle the fuel assemblies during
refueling. Interlocks on both the grapple hoist and auxiliary
hoists prevent lifting a fuel assembly over the core with control rod withdrawn; interlocks also prevent withdrawal of a control rod
with a fuel assembly over the core attached to either the fuel
grapple or auxiliary hoists. Interlocks also block travel of the
refueling platform over the reactor in the startup mode. The
refueling interlocks are described and evaluated in subsection
7.6, "Refueling Interlocks."
A Service Pole Caddy platform is attached on the rear side of
either the Unit 2 or Unit 3 refueling platform at PBAPS. The
platform provides an auxiliary work station for unlatching and
latching the steam separator head bolts during refueling
activities. The platform can also be utilized for other
underwater servicing needs, such as jet pump beam bolt untorquing
CHAPTER 10 10.4-6 REV. 26, APRIL 2017 and steam line plug installation. The platform is provided with high torque service poles and a motorized hoist to handle the
poles.
10.4.7 Storage Equipment
In addition to the new and spent fuel storage racks, other storage
equipment is provided.
Defective fuel assemblies may be placed in special fuel cans which
would be stored in the defective fuel storage rack.
10.4.8 Under-Reactor-Vessel Servicing Equipment
The necessary equipment to remove CRD's during a refueling outage
is provided. An equipment handling platform with a rectangular
open center is provided. This platform is rotatable to provide
space under the vessel so the CRD can be lowered and removed.
A thermal sleeve installation tool (Figure 10.4.3) is used to
rotate the thermal sleeve (Figure 10.4.2) within the CRD housing.
Sleeve rotation permits disengagement of the guide tube. A rope
and pulley integral with the tool permits complete sleeve removal.
Miscellaneous wrenches are provided to install and remove the
neutron detectors. Flow through the drain tube pulls the fixed
in-core detector string into the in-core guide tube thus sealing
the opening in the in-core flange during in-core servicing. A
drain can be opened after in-core insertion to drain any residual
water. Correct seating of the in-core string is indicated when
drainage ceases.
10.4.9 Equipment Storage Pit
Large radioactive components, such as the steam dryer and steam
separator assembly are stored in the storage pit. The storage pit
is separated from the drywell by removable concrete blocks that
serve as a shield when the dryer and separator are stored and the
water level is lowered. Other large items, such as the pressure
vessel head and drywell head, are stored on the refueling floor.
To minimize worker exposure, a wet transfer of the dryer is
normally expected. To minimize operator exposure during a dry
transfer of the dryer assembly, the storage pit canal is deep
enough so that the top of the dryer can be kept at least 2 ft
below the operating floor level during transfer. The storage pit
is deep enough below the canal that, with the reactor well
drained, a minimum of 6 in of water shielding can be maintained
above the separator plenum dome.
CHAPTER 10 10.4-7 REV. 26, APRIL 2017 Special liner considerations account for the abrasion and high unit loadings that occur on areas where the dryer and separator
assemblies are placed.
The storage pit is lined with stainless steel for leaktightness
and corrosion resistance.
10.4.10 Reactor Building Crane
The reactor building crane for each unit is designed such that no
credible postulated failure of any crane component will result in
the dropping of the fuel cask; therefore, the consequences of this
accident are precluded.
The reactor building cranes have been evaluated using the criteria
of NUREG-0554 and NUREG-0612, Appendix C to establish the maximum
critical load (MCL) rating at which they can be considered single-
failure proof. The results of this evaluation resulted in a MCL
rating of 125 tons for the main hoist reactor building cranes.
Thus, for loads within this limit, a load drop is not credible.
The design of the main hoist is as follows:
A single hoist motor drives two separate shafts. The motor has
two centrifugally tripped limit switches, one outboard of each
hoist input pinion at each end of the motor shaft assembly. These
provide an automatic safety shutdown and protection from any
control or motor malfunction which might result in a runaway
condition of the load. Each motor driven shaft passes through a
150 percent capacity solenoid-actuated brake. A failure of either
the motor shaft, the connecting shafts, or the shaft couplings
singly would not result in a load drop as the brakes would be
effective in holding the load. On loss of power to the motor, both brakes engage. They can also be engaged by the operator.
Additionally, there is a 90 percent capacity eddy-current brake to
limit the rate of load lowering.
After the brake, each motor shaft enters its own gear reducer. If
a component of one gear case (gear teeth, shafts, bearings, or
structural component) should fail, the other gear reducer holds
the load with its brake with a safety factor of 5.
Each gear case is fitted on its output end with a pinion meshing
with the drum gear. A failure of a pinion, drum gear, pinion
shaft, or pinion bearing results in the load being carried by the
other similar set of parts on the other end of the drum. Again a
safety factor of 5 remains in the functioning parts. In each of
the main hoist gear cases, there is a mechanical load brake, with
cooling of the gear case oil, to offer additional safety in load
handling.
CHAPTER 10 10.4-8 REV. 26, APRIL 2017 In the event of failure of the drum shaft, drum bearing, or drum bearing bracket, the drum flange drops a fraction of an inch onto
machined structural seats so located that the drum is supported
and the remaining pinion and gear stay in mesh to restrain the
load. A safety factor of 5 still remains.
Two separate ropes are led from the drum, each being reeved
through a set of block sheaves, upper and lower, and back to an
equalizer bar, and are arranged for equal division of the load
between the two ropes. With both ropes functioning and equalized, the safety factor of the ropes is 7 on a static basis. If one rope
fails, the remaining rope supports the load with a residual safety
factor of 3.5 on a static basis. The equalizer bar is fitted with
double acting hydraulic cylinders and hydraulic accumulator to
minimize the shock when the entire load is transferred to one
rope. Therefore, load drop is precluded for a postulated single
rope failure. The equalizer bar is contained within structural
components so that if it breaks or if its pivot point breaks, the
parts are retained within the trolley and load drop is precluded.
To protect against overloading of the cables a load sensing system
consisting of tension type load cells supports the load sensing
sheave frame assembly. The load cells are supported by the load
cell support brackets attached to the trolley frame. To protect
against an unbalanced load, limit switches provide visual warning
indication to the crane operator. The limit switches, attached to
the trolley frame, are activated by movement in the equalizer bar
assembly.
All sheaves, both upper and block sheaves, are contained in heavy
structural casings which usually carry a negligible load. In the
event of a sheave pin failure, the sheaves rise to the top of the
block or drop to the base of the upper sheave housing and stop at
those points, and load drop is precluded. The block assembly
contains two 100 percent capacity hooks of "load carrying
devices." This redundancy in attachment to lifting assembly and
in load carrying capability are such that a single failure does
not cause load drop. Additional nondestructive testing (ultrasonic and magnaflux testing for the load block swivel and
the sheave shafts of the upper assembly) provides further
assurance that this crane is of a quality suitable for nuclear
services. Electrical circuits have been reviewed and it has been
determined that no single credible electrical component failure
causes the load to drop.
The auxiliary hoist is designed to satisfy the Single-Failure-
Proof Guidelines of Section 5.1.6 of NUREG-0612, and thus
eliminating the need to analyze the effects of drops of heavy
loads per the evaluation criteria of Section 5.1 of NUREG-0612.
CHAPTER 10 10.4-9 REV. 26, APRIL 2017 Protection of the pool liners in the cask storage area for the normal cask lowering operation is provided by a 1 in thick steel
wearing plate. This prevents any damage to the liner over plant
life occasioned by normal fuel cask handling.
The adequacy of the drywell head for a postulated drop of one of
the shield plugs was performed in a load drop analysis performed
in the Peach Bottom Calculation PS-0288, "Drywell Head Load Drop
Analysis."
No vital equipment is located in compartments below the fuel pool
floor; therefore, no loss of function of vital equipment would
result from falling objects and flooding caused by a postulated
event.
Strict administrative control assures that the cask is not
unnecessarily lifted higher than required during maneuvering above
the refueling floor and also that the cask is not brought over the
reactor vessel or the fuel storage portion of the pool.
10.4.11 Heavy Loads Compliance The licensee has a defense-in-depth program to manage the handling
of heavy loads on site such that no credible load drop will
endanger the public safety and health. Loads that are either not
considered as 'heavy loads' (i.e., less than 1200 pounds) or have
been determined to not potentially impact irradiated fuel, the
reactor vessel, or safe shutdown equipment are not within the
scope of the 'heavy loads' program. Any heavy loads that have not
previously been evaluated will be evaluated prior to being lifted.
This evaluation would include ensuring at least one of the
following measures are in place for the lift:
Mechanical stops or electrical interlocks that prevent heavy
loads movement over irradiated fuel or safe shutdown equipment, Verification analysis that the consequences of a potential load
drop are within accepted bounds, or Use of a single-failure-proof handling system.
Lifts are conducted in accordance with good rigging practices and
in accordance with the licensee's approved procedures.
In July 1980, the NRC issued NUREG-0612, Control of Heavy Loads at
Nuclear Power Plants. This NUREG was issued to resolve NRC
Generic Technical Activity A-36. This generic issue involved
reviewing the adequacy of NRC requirements for controlling the
handling of heavy loads over or in proximity to spent fuel, the
reactor core, and safe shutdown equipment. NUREG-0612
recommendations were planned to be assessed by the NRC in two
phases (i.e., Phase I and Phase II). Phase I concerned itself
CHAPTER 10 10.4-10 REV. 26, APRIL 2017 with seven requirements that assured a defense-in-depth approach was taken in regards to handling heavy loads. Phase II of NUREG-
0612 concerned itself with plant specific analyses to ensure that
the potential for load drop was extremely small or if there was a
load drop no significant impact to spent fuel, the reactor vessel
or safe shutdown equipment would occur. The NRC issued a safety
evaluation report concerning Peach Bottom compliance to NUREG-0612
Phase I on 9/21/83. Peach Bottom submitted its intent concerning
compliance with Phase II. However, prior to NRC issuance of a
final SER for Peach Bottom, the NRC discontinued its review of
Phase II submittals. In Generic Letter 85-11, dated 6/28/85, the
NRC reported to the industry that due to their reviews of Phase I
activities and proposed Phase II activities, there did not warrant
a need to take further action on Phase II. The Peach Bottom
response to this generic letter on 2/11/86 stated that changes to
Phase II actions would be considered on a case-by-case basis.
Lifts are conducted in accordance with good rigging practices and
in accordance with the licensee's approved procedures.
10.4.11.1 NUREG-0612 Phase I Requirements All heavy loads that could be brought over or in proximity to
irradiated fuel, the reactor vessel or safe shutdown equipment are
handled in accordance with a defense in depth philosophy. The
following seven criteria ensure appropriate handling of heavy
loads is in place.
- 1. Safe Load Paths Safe Load Paths (SLP's) are established for the movement of heavy
loads to minimize the potential for heavy loads, if dropped, to
impact irradiated fuel in the reactor or spent fuel pool, or to
impact safe shutdown equipment required to be operable. These
load paths are controlled under approved design documents. Design
documents and procedures also control rigging exclusion zones and
height/weight restrictions. Administrative procedures require
that the cognizant supervisor review safe load paths prior to the
lift. Any deviations from designated SLP's must be approved by
engineering.
Concerning the Emergency Diesel Generator (EDG) Cranes, heavy
loads lifts may only be performed when the EDG is inoperable or
declared inoperable for purposes of heavy load lifts.
Load movements along the safe load paths are directed by a
qualified signalman.
- 2. Load Handling Procedures
CHAPTER 10 10.4-11 REV. 26, APRIL 2017 Load handling procedures are in place for the handling of heavy loads over or in proximity to reactor fuel or safe shutdown
equipment. A governing administrative procedure defines the
overall requirements to perform these lifting operations including
the NUREG-0612 Phase I requirements. Implementing procedure(s)
ensure that appropriate details of complicated lifts are defined.
As appropriate, the above procedures direct the identification of
required equipment, inspections and acceptance criteria required
before moving the load, the steps and proper sequence to be
followed in handling the load, definition and use of appropriate
safe load paths and require that Phase I requirements are met for
heavy load operations.
- 3. Crane Operator Training
For lifts performed within the scope of the heavy loads program, crane operators will be trained, qualified and conduct themselves
in accordance with Chapter 2-3 of ANSI B30.2-1976, 'Overhead and
Gantry Cranes'.
- 4. Special Lifting Devices
Special Lifting Devices used in areas where a load is carried over
or in proximity to the reactor vessel, spent fuel pool, or safe
shutdown equipment meet the requirements of ANSI N14.6-1978 with
the following exceptions:
a.Inspections shall be performed at least once per
operating cycle rather than annually.
b.Critical welds may be non-destructively examined in lieu
of the routine 150% load test
c.Dry fuel storage cask trunnions are waved from the
periodic load test or examination requirements.
d.The drywell strongback has two steel features that were
added during modifications that did not receive the
required materials testing (Charpy impact or drop weight
testing). (Ref. ECR 13-00378).
- e. The RPV head carousel hook and load pins did not receive the required materials testing (Charpy impact or drop
weight testing) (Ref. ECR 13-00378).
The stress design factor stated in section 3.2.1.1 of ANSI N14.6-
1978 is based on the combined maximum static and dynamic loads
that could be imparted on the handling device based on the
characteristics of the crane which will be used. Devices used for
CHAPTER 10 10.4-12 REV. 26, APRIL 2017 handling the spent fuel cask/lid are designed to ANSI N14.6 - 1986 which is equivalent to or exceeds the 1978 version.
- 5. Lifting Devices (not specifically designed)
Procedures are in place which require that all lifting devices not
specifically designed (i.e., slings) that are used to lift heavy
loads over or in proximity to the reactor vessel, spent fuel pool
or safe shutdown equipment meet the requirements of ANSI B30.9-
1971, 'Slings' or Twin-Path Extra TPXC Synthetic Round Slings
constructed with K-Spec fiber meeting the requirements of ASME
B30.9-2010, used in combination with engineered softeners and
abrasion protection devices as required by station procedures.
Additionally, a dynamic load factor of 25% of the dead load will
be used in selecting the sling.
- 6. Cranes (Inspection, testing and maintenance)
Procedures are in place which ensure that crane inspection, testing and maintenance is performed in accordance with ANSI B30.2
(1967 version), 'Overhead and Gantry Cranes'. Additionally, if
repairs of load sustaining members are made by welding, identification of materials shall be made and appropriate welding
procedures will be followed.
- 7. Crane Design
The Reactor Bldg, Turbine Bldg, Pump Structure, and EDG Cranes
meet the intent of the requirements of Chapter 2-1 of ANSI B30.2-
1976, 'Overhead and Gantry Cranes' and of CMAA-70, 'Specifications
for Electric Overhead Traveling Cranes'. Turbine Building Cranes
are upgraded to meet single failure proof requirements of NUREG-
0554 for the increased main and auxiliary hoist capacities of 115
Ton and 30 Ton respectively. Structural analyses are performed
using commercially available, NRC-approved computer programs.
Concrete anchor bolts are analyzed per American Concrete Institute (ACI) Code 349-01, approved by the NRC for this purpose.
10.4.11.2 NUREG-0612 Phase II Requirements
Heavy load lifts made by permanent station cranes and hoists, as
well as by mobile cranes and temporary rigging, are performed in a
manner to minimize the threat to irradiated fuel, the reactor
vessel or safe shutdown equipment. This necessitates that the
following three criteria of NUREG-0612, Phase II are met, except
for alternatives which may be approved on a case-by-case basis in
accordance with station procedures:
1.The lift is performed as single failure-proof equivalent (either using redundant rigging or increased safety factors)
or,
CHAPTER 10 10.4-13 REV. 26, APRIL 2017 2.The lifting system has electrical interlocks or mechanical stops such that loads could not be handled over or in
proximity to fuel, the reactor vessel, or safe shutdown
equipment or, 3.An evaluation is performed that ensures that a load drop
could not cause damage to fuel, the reactor vessel, or loss
of a safe shutdown function.
10.4.11.3 Safety Evaluation
Heavy load lifts are performed using a defense-in-depth program
such that no credible load drop will endanger the public safety
and health. The procedural controls that implement NUREG-0612
Phase I make the risk of a load drop very unlikely. In addition, single-failure-proof lifts are employed to further reduce the risk
of load drop to an acceptably low level. Where single-failure-
proof lifts are not used, the consequences of a postulated load
drop are evaluated, and must be demonstrated to be acceptable.
Resulting restrictions on load height, weight, lift configuration, and/or equipment required to be operable are procedurally
controlled.
CHAPTER 10 10.5-1 REV. 26, APRIL 2017 10.5 FUEL POOL COOLING AND CLEANUP SYSTEM 10.5.1 Power Generation Objective The power generation objectives of the fuel pool cooling and cleanup system are to provide fuel pool water temperature control
and to maintain fuel pool water clarity, purity, and level.
10.5.2 Power Generation Design Basis
- 1. The fuel pool cooling and cleanup system minimizes corrosion product buildup and controls water clarity
through filtration and demineralization.
- 2. The fuel pool cooling and cleanup system minimizes fission product concentrations which could be released
from the pool water to the reactor building environment.
- 3. The fuel pool cooling and cleanup system monitors fuel pool water level and maintains a water level above the
fuel sufficient to provide shielding for normal building
occupancy.
- 4. The fuel pool cooling and cleanup system limits the fuel pool water temperature during normal and refueling
operations.
10.5.3 Description The fuel pool cooling and cleanup system cools the fuel storage
pool by transferring decay heat through heat exchangers to the
service water system (Drawing M-363, Sheets 1 and 2). Water
purity and clarity in the storage pool, reactor well, and dryer-
separator storage pit are maintained by filtering and
demineralizing the pool water (Drawing M-364, Sheets 1 and 2).
See paragraph 10.3.4.2 for the description of the spent fuel pool.
Connections also exist to use the "B" filter-demineralizer to
process liquid radwaste, as shown on Drawing M-363, Sheet 1.
The system consists of three fuel pool cooling pumps, three heat
exchangers, filter-demineralizer(s), two skimmer surge tanks, and
associated piping, valves, and instrumentation. The three fuel
pool pumps are connected in parallel, as are the three heat
exchangers. The pumps and heat exchangers are located in the
reactor building below the bottom of the fuel pool.
The filter-demineralizers, which collect radioactive corrosion
products, are located in the radwaste building and are typically
arranged so that one filter-demineralizer is aligned to each
reactor unit and the third is a spare. Up to three filter-
CHAPTER 10 10.5-2 REV. 26, APRIL 2017 demineralizers may be aligned to one unit to support water clarity and water chemistry improvements, as required.
The pumps circulate the pool water in a closed loop, taking
suction from the skimmer surge tanks through the heat exchangers, circulating the water through the filter demineralizer, and
directing the processed spent fuel cooling water through the
system discharge lines located in the fuel pool and reactor well.
This return flow of spent fuel cooling water is discharged
downward from the discharge lines into the pool at an elevation
that is above the top of the storage racks. The cooled water
traverses the pool picking up heat and debris before starting a
new cycle by discharging over the skimmer weirs and scuppers into
the skimmer surge tanks. Makeup water for the system can be
transferred from the condensate storage tank to the skimmer surge
tanks. System and equipment parameters are listed in Table
10.5.1.
An evaluation of the fuel pool cooling system was performed for normal refueling of approximately 40% (320 bundles) of the core
every 24 months and a full core offload just before normal
refueling assuming all storage cells in the spent fuel pool are
filled. The evaluations assume that the offloaded fuel has
operated in the reactor at 3951 MW. For the normal refueling
offload of 40% of the core, the evaluation assumes a normal
complement of three fuel pool cooling trains (three fuel pool
cooling pumps and three fuel pool cooling heat exchangers) in
service as well as a single failure where only two fuel pool
cooling trains (two fuel pool cooling pumps and two fuel pool
cooling heat exchangers are in service). For the full core
offload case, no fuel pool cooling trains are assumed available
and fuel pool cooling is performed by the RHR system. The
evaluation also considered the time for the fuel pool to boil if
there is a loss of fuel pool cooling. See Table 10.5.2 for a
summary of the results of these cooling system evaluations.
When flooded up, the Fuel Pool Cooling system and the RHR fuel
pool assist mode can be used to remove decay heat from both the
spent fuel pool and the reactor vessel by cooling the spent fuel
pool. A cross-connection between the drain line from the skimmer
surge tank and the RHR system allows the RHR system to take a
suction from the fuel pool. This is called Fuel Pool Assist mode, when water is returned to the fuel pool and called Alternate Decay
Heat Removal (ADHR) mode when water is returned to the reactor
vessel through the normal shutdown cooling return line. In
addition, a Split Flow mode is available when the RHR discharge
flow is split between the fuel pool and the reactor vessel.
During ISFSI operation, it may be necessary to return a loaded dry
storage cask to one of the Spent Fuel pools for unloading. The
CHAPTER 10 10.5-3 REV. 26, APRIL 2017 heat introduced to the pool by the latent heat of the cask materials and the decay heat of the 68 contained assemblies is
less that the full core offload heat that the Fuel Pool Cooling
and Cleanup System has been analyzed for.
Since each refueling offload is cycle specific, then the
variations in the number of fuel assemblies discharged, the in-
core decay time, the fuel assembly transfer rate and the power
history can vary as long as analysis shows that the spent fuel pool bulk temperature will not exceed 150 F and localized boiling will not be expected to occur.
The system flow rate is larger than that required for two complete
water changes per day of the fuel pool, or one change per day of
the fuel pool, reactor well, and dryer-separator pit. The maximum
system flow rate is twice the flow rate needed to maintain water
quality.
For refueling operations, water to fill the reactor well and
dryer-separator storage pit is stored in the refueling water tank.
Water is transferred to the refueling area by two refueling water
pumps and/or via the CST and core spray system. During drainage, water can be pumped through one of the condensate filter-
demineralizer units before being returned to the storage tank.
When placing a dry fuel storage cask into the fuel pool, the water
level of the pool is managed by controlling the skimmer surge tank
level and the fuel pool level as needed. Procedures ensure that a
new cask is inspected and cleaned as necessary prior to placement
in the pool.
Fuel pool water is continuously recirculated. The circulation
patterns within the reactor well and fuel pool are established by
the placement of the diffusers in the reactor well and the
placement of skimmers and discharge lines in the fuel pool so as
to sweep particles dislodged during refueling operations away from
the work area and out of the pools.
Pool water clarity and purity are maintained by a combination of
filtration and ion exchange. The filter-demineralizer units are
located separately in shielded cells. The filter-demineralizer
maintains the Fuel Pool water quality to within the limits
specified in EPRI BWR Water Chemistry Guidelines for compatibility
with materials and equipment in the fuel pool that require
corrosion protection. Particulate matter is removed by the
filter-demineralizer unit in which finely divided, powdered ion-
exchange resin and fiber material serves as the filtering medium.
Alternately, a combination of powdered resin and cellulose may be
used as the disposable filter medium. The filter elements are a
CHAPTER 10 10.5-4 REV. 26, APRIL 2017 stainless steel mesh element mounted vertically in a tube sheet and replaceable as a unit. The filter vessel is constructed of
carbon steel and coated with a phenolic material. The resin is
replaced when the pressure drop across the filter is excessive or
when instrumentation indicates the ion exchange capacity is low.
Alarms, differential pressure indicators, and flow indicators
monitor the condition of the filter-demineralizers. Backwashing
and precoating operations are controlled from a local panel in the
radwaste building. The spent filter medium is removed from the
elements by backwashing with air and condensate, then flushed to
the waste sludge tank.
There are no connections to the fuel storage pool which could
allow the fuel pool to be drained below the pool gate between the
reactor well and the fuel pool when the pool gate is in place or
below 10 feet above the top of active fuel. Fuel pool cooling and
RHR discharge lines that extend below this level are equipped with
syphon breaker holes to prevent inadvertent pool drainage. A level
indicator, mounted at the valve rack, monitors reactor well water
level during refueling. Any significant leakage through the
refueling bellows assembly, drywell to reactor seal, or the fuel
pool gates is annunciated on the operating floor instrument racks
and in the main control room.
Instrumentation is provided for both automatic and manual
operation. The surge tanks have high and low level alarms and
pump trip switches. The pumps are controlled locally at the pump
or from a control panel near the filter-demineralizers. Pump low
suction pressure automatically turns off the pumps. A pump low
discharge pressure causes alarm annunciation in the main control
room and in the pump room. Also see paragraph 10.3.4.3 for spent
fuel pool instrumentation.
10.5.4 Inspection and Testing
No special equipment tests are required because at least one pump, heat exchanger, and filter-demineralizer are normally in
operation while fuel is stored in the pool.
Routine visual inspection of the system components, instrumentation, and trouble alarms is adequate to verify system
operability. Pool level indicators and associated alarms are
tested by simulating low water level to the sensors.
CHAPTER 10 10.5-5 REV. 25, APRIL 2015 TABLE 10.5.1 FUEL POOL COOLING AND CLEANUP SYSTEM
Design Core Thermal Power 4,030 MWt Total Pool, Well, and Pit Volume 111,400 cu ft Fuel Storage Pool Volume 53,350 cu ft System Design Flow 555 gpm Maximum Flow 1,665 gpm
Fuel Pool Cooling Water Pumps Quantity 3 Type
Horizontal, centrifugal
Design Flow/TDH (each) 580 gpm/250 ft Motor hp 60 hp
Fuel Pool Cooling Heat Exchangers Quantity 3 Heat Exchanger Capability
One exchanger in service
= 3.75 x 10 6 Btu/hr Two exchangers in service
= 7.50 x 10 6 Btu/hr Three exchangers in service
=11.25 x 10 6 Btu/hr Material Tube/Shell 304 SS/carbon steel Design Code ASME B&PV, Sec. VIII
CHAPTER 10 10.5-6 REV. 25, APRIL 2015 TABLE 10.5.1 (Continued)
Fuel Pool Filter-Demineralizers
Type Pressure precoat
Quantity 1 per unit, 1 common spare Design Filter Area 270 sq ft
Filter Capacity 550 gpm/unit
Pressure Drop 25 psi (dirty)
VIII
Holding Pump Flow 27 gpm Precoat Flow 450 gpm Flow Control Valve Pressure Drop 100 psi (max)
10 psi (min)
CHAPTER 10 10.5-7 REV. 25, APRIL 2015 TABLE 10.5.2
SUMMARY
OF COOLING SYSTEM ANALYSIS RESULTS
- 1) Heat Exchanger Capability
One exchanger in service = 3.75 x 10 6 Btu/hr Two exchangers in service = 7.50 x 10 6 Btu/hr Three exchangers in service = 11.25 x 10 6 Btu/hr
- 2) Maximum Pool Heat Load to insure exit temperature is below 150 F One exchanger in service = 8.66 x 10 6 Btu/hr Two exchangers in service = 17.33 x 10 6 Btu/hr Three exchangers in service = 26.0 x 10 6 Btu/hr
- 3) Normal Refueling a)Full Cooling Capability Equipment in service:
3 FPCCS Pumps (1665 gpm total SFP flow) 3 FPCCS Heat Exchangers (2400 gpm total service water flow, 90 o F service water temperature) Start of Offload (hours after shutdown): 80 Max. SFP Temperature: 140 o F Time to Boil from Max. Temperature: 11.4 hrs Makeup Flow Required at Boiling: 49 gpm Max Heat Load (MBTU/hr): 23.9 b)Single Failure Equipment in service:
2 FPCCS Pumps (1110 gpm total SFP flow) 2 FPCCS Heat Exchangers (1600 gpm total service water flow, 90 o F service water temperature) Start of Offload (hours after shutdown): 200 Max. SFP Temperature: 150 o F Time to Boil from Max. Temperature: 12.3 hrs Makeup Flow Required at Boiling: 40 gpm Max Heat Load (MBTU/hr): 19.5 TABLE 10.5.2 (continued)
CHAPTER 10 10.5-8 REV. 25, APRIL 2015
- 4) Full-Core Offload, Full Cooling Capability Equipment in service:
1 RHR Pump (5000 gpm total SFP flow) 1 RHR Heat Exchanger (4500 gpm total HPSW flow, 92 o F HPSW water temperature) Start of Offload (hours after shutdown): 150 Max. SFP Temperature: 140 o F Time to Boil from Max. Temperature: 6.0 hrs Makeup Flow Required at Boiling: 88 gpm Max Heat Load (MBTU/hr): 41.3
CHAPTER 10 10.6-1 REV. 21, APRIL 2007 10.6 SERVICE WATER SYSTEM 10.6.1 Power Generation Objective
The power generation objective of the service water system is to
supply water required for plant services.
10.6.2 Power Generation Design Basis
- 1. The service water system continuously supplies screened and chlorinated cooling water to the plant during normal
plant operation and shutdown periods.
- 2. System interconnections are provided to enable the emergency service water system to serve the reactor
building cooling water heat exchangers in the event of a
loss of off-site power. This design feature exists
although the heat sink, emergency service water (ESW),
for the reactor building closed cooling water (RBCCW)
system has been eliminated as a result of locking closed
the ESW-RBCCW cross-tie valves. Therefore, little, if
any, cooling would be provided to the service water
system loads during a loss of off-site power.
- 3. The service water system supplies cooling water to the core standby cooling equipment and space coolers during
normal plant operation and shutdown period only.
- 4. The system inhibits the release of radioactive material into the river.
10.6.3 Description
The service water system consists of three one-half capacity
service water pumps in the pump structure, three horizontal fuel
pool service water booster pumps in the reactor building, and
associated piping, valves, and instrumentation (Drawing M-314, Sheets 1 through 9).
The three service water pumps are vertical, turbine-type pumps, connected in parallel, taking suction from the pump structure, and
each delivering 14,000 gpm at a pump head of 155 ft. The pump
bearings are supplied from lube water pumps. Nominal system
pressure is 65 psig. The three fuel pool service water booster
pumps deliver service water to the fuel pool cooling heat
exchangers. These horizontal, centrifugal pumps are rated at 900
gpm at a pump head of 135 ft.
The safeguards equipment coolers and space air cooler are
automatically served by the emergency service water system.
CHAPTER 10 10.6-2 REV. 21, APRIL 2007 To inhibit leakage of radioactivity from potentially contaminated
systems (mechanical vacuum pump and fuel pool heat exchangers),
service water pressure is maintained higher than process fluid
pressure. A radiation monitor on the service water effluent
header from the reactor building cooling water heat exchangers
detects leakage of radioactive material from these exchangers. The
monitor indicates, records, and alarms in the main control room.
10.6.4 Inspection and Testing
The service water system components are proven operable by their
use during normal plant operations. Portions of the system
normally closed to flow can be tested to ensure their operability
and the integrity of the system.
CHAPTER 10 10.6-3 REV. 21, APRIL 2007 TABLE 10.6.1 SERVICE WATER SYSTEM DATA
Service Water Pumps
Quantity 3
Type Vertical, Turbine Type, Wet-Pit
Flow/Pump Head 14,000 gpm/155 ft
Bhp Rating 655 hp Speed 900 rpm
Motor Type Vertical, Induction Type
Voltage/Phase/Frequency 2,300 V/3 phase/60 Hz
Rated Horsepower 700 hp
Fuel Pool Service Water Booster Pumps
Quantity 3
Type Horizontal Centrifugal
Flow/Pump Head 900 gpm/135 ft
Bhp Rating 39 hp Speed 3,600 rpm
Motor Type Horizontal
Voltage/Phase/Frequency 460 V/3 phase/60 Hz
Rated Horsepower 40 hp
CHAPTER 10 10.7-1 REV. 26, APRIL 2017 10.7 HIGH PRESSURE SERVICE WATER SYSTEM 10.7.1 Safety Objective
The safety objective of the high pressure service water system is
to provide a reliable supply of cooling water for RHR under post-
accident conditions.
10.7.2 Safety Design Basis
- 1. The high pressure service water system is designed to seismic Class I criteria to withstand the maximum
credible earthquake without impairing system function.
- 2. The high pressure service water system is operable during flood conditions.
- 3. The high pressure service water system is designed with capacity and redundancy to supply cooling water to the
RHRS under post-accident conditions.
- 4. The high pressure service water system is operable during the loss of offsite power.
10.7.3 Power Generation Objective
The power generation objective of the high pressure service water
system is to supply cooling water to the RHRS for shutdown cooling
and for torus cooling.
10.7.4 Power Generation Design Basis
- 1. The high pressure service water system supplies a
reliable source of cooling water to the RHRS.
- 2. The high pressure service water system is designed for
remote-manual initiation.
- 3. The high pressure service water system inhibits leakage
of radioactive material from the RHRS to the environment.
- 4. The high pressure service water system provides an
additional source of water for post-accident containment flooding
by a cross tie between the high pressure service water system and
the RHRS.
10.7.5 Description Each high pressure service water system consists of four 4,500-gpm pumps installed in parallel in the pump structure (Drawing
CHAPTER 10 10.7-2 REV. 26, APRIL 2017 M-315, Sheets 1 through 4). Normal water supply to the suction of the pumps is from Conowingo Pond. When the high pressure service
water system is operated in conjunction with the emergency heat
sink (subsection 10.24, "Emergency Heat Sink"), the suction is
from the HPSW pump bay which is fed from emergency cooling tower
basin. The pump discharge is manifolded and provided with a
normally closed, motor-operated valve separating the four pumps
into groups of two. Two parallel headers run from the pump
structure to the reactor building. Each header delivers the
discharge from two pumps to two RHR heat exchangers also in
parallel. Under normal conditions, when the respective loop of
HPSW is in operation, the service water pressure on the discharge
side of the RHR heat exchanger is maintained positive with respect
to the RHRS side to inhibit leakage of radioactive material into
the environment. In the event of a design basis accident or
transient in which additional containment cooling capacity is required, a second HPSW pump can be aligned to a second RHR heat
exchanger by opening the cross-tie valve.
Under abnormal operating conditions RHRS pressure could exceed
high pressure service water system pressure. An RHR heat
exchanger leak under these abnormal conditions would result in
radioactive RHR water migrating into the high pressure service
water system and into the river. To limit the release of
radioactive water to the river from this potential release path, signals from the radiation monitors in the sample system which
samples the high pressure service water system upstream and
downstream of the RHR heat exchangers initiate an alarm in the
control room at a predetermined radiation level.
Flanged connection points are available on the high pressure
service water system, downstream of the RHR heat exchangers, to
allow for a temporary flow path of the RHR heat exchanger cooling
water in the event that the normal flow path becomes unavailable.
This alternative flow path is intended to be routed through
secondary containment. Therefore, this flow path may only be used
when secondary containment is not required.
An intertie is provided between units 2 and 3 high pressure
service water system to provide flexibility. A cross tie to the
RHRS provides the capability for primary containment flooding.
The high pressure service water system pumps are vertical
multistage turbine type. The pump mounting base is of watertight
construction to withstand the hydrostatic pressure at the design
flood condition. The pump design data is given in Table 10.7.1.
The high pressure service water system piping and valves are
designed as described in Appendix A.
CHAPTER 10 10.7-3 REV. 26, APRIL 2017 10.7.6 Safety Evaluation
The high pressure service water system pumps are installed in a
seismic Class I structure. The system meets seismic Class I
criteria and is protected against the design flood level.
Each pump is sized to accommodate the design heat removal capacity
of one RHRS heat exchanger. They have adequate head (1) to
maintain the high pressure service water system cooling water at a
higher pressure than the RHRS, thus inhibiting the release of
radioactive material to the environment, and (2) to permit
operation in conjunction with the emergency heat sink. Further, the pumps have both a normal and a standby power supply. In the
event of the loss of offsite power, the pumps are supplied from
the diesel generators and manually started as required.
Sufficient redundancy is provided in the number of pumps and power
supplies, and in the piping arrangement, so that no single system
component failure can prevent the system from supplying cooling
water to accommodate the normal shutdown mode and the containment
cooling mode. Therefore, core decay heat removal during the
shutdown periods, or containment cooling during the post-accident
condition, can be maintained.
10.7.7 Inspection and Testing
Pumps in the high pressure service water system are proven
operable by their use or testing during normal station operations.
Motor operated isolation valves can be tested to assure they are
capable of opening and closing by operating manual switches in the
control room and observing the position lights. Portions of the
high pressure service water system normally closed to flow can be
tested to ensure their operability and the integrity of the
system.
CHAPTER 10 10.7-4 REV. 23, APRIL 2011 TABLE 10.7.1 HIGH PRESSURE SERVICE WATER SYSTEM
EQUIPMENT DATA
High Pressure Service Water Pumps
Quantity 4 Per Unit
Type Vertical, Turbine Type
Flow/Head Design Point 4,500 gpm at 700 ft
Bhp at Rating
< 975 hp Speed 1,770 rpm Number of Stages 6
Pump Design:
Shut-Off Head
> 368 and < 445 psig
Material:
Bowl/Impeller Cast Carbon Steel or Moly Iron/Bronze or Cast Stainless Steel Discharged Head/Column Carbon Steel/Carbon Steel
Line Shaft Stainless Steel
Bearings Brass/Bronze/Rubber
Motor:
Type Vertical, Induction
Horsepower 1,000 hp Voltage/Phase/Frequency 4,160 V/3 Phase/60 Hz
CHAPTER 10 10.8-1 REV. 21, APRIL 2007 10.8 REACTOR BUILDING COOLING WATER SYSTEM 10.8.1 Power Generation Objective
The power generative objective of the reactor building cooling
water system is to provide cooling water to auxiliary plant
equipment associated with the nuclear steam supply system (NSSS).
10.8.2 Power Generation Design Basis
- 1. The reactor building cooling water system is designed to cool auxiliary plant equipment over the full range of
reactor power operation.
- 2. The reactor building cooling water system is designed to inhibit the release of radioactive material to the
environment.
10.8.3 Description
The reactor building cooling water system consists of two full-
capacity pumps, two full-capacity heat exchangers, one head tank, one chemical feed tank, and associated piping, valves, and
controls (Drawing M-316). The cooling water pumps and heat
exchangers are located in the reactor building auxiliary bay. The
head tank is located on the reactor building refueling floor. The
system equipment data is given in Table 10.8.1.
The system is a closed loop utilizing inhibited demineralized
water. The heat exchangers are designed with service (river)
water on the tube side and demineralized water on the shell side.
The reactor building cooling water system is designed for an
operating pressure of 140 psig.
The head tank, located at the highest point in the loop, accommodates system volume changes, maintains static suction
pressure on the pump, aids in detecting gross leaks in the reactor
building cooling water system, and provides for adding makeup
water. An automatic makeup control valve maintains water level in
the tank. The automatic function is not required and may be
valved out to monitor system inventory. High and low water levels
are alarmed in the main control room. An inhibitor is added as
necessary to the demineralized water by means of a chemical
addition tank to limit corrosion.
The reactor building cooling water system supply and return
headers penetrating the primary containment are each provided with
a motor-operated isolation valve outside the containment. These
isolation valves are manually controlled remotely from the main
control room.
CHAPTER 10 10.8-2 REV. 21, APRIL 2007 Electrical power for operating the reactor building cooling water
system pumps during failure of offsite power is supplied from the
standby power supply.
In the event of offsite power failure, the reactor building
cooling water system supply to the reactor water cleanup system
non-regenerative heat exchanger and pumps, instrument nitrogen
compressor skids, and various sample station coolers is isolated, and the water supply is maintained to the reactor recirculation
pump motor oil and mechanical seal water coolers and the reactor
building equipment drain sump cooler. In addition, water is
supplied to the drywell air cooling system and the drywell
equipment drain sump cooler, which are normally served by the
chilled water system, and to the CRD pump oil coolers and air
compressor jacket and after coolers, which are normally served by
the turbine building cooling water system.
The reactor building cooling water system can also supply water to
the fuel pool cooling heat exchangers, via removable spool pieces, in the event of loss of normal cooling water. The control and
instrumentation is designed for remote system startup from the
main control room.
These design features do exist although the heat sink, emergency
service water (ESW), for the reactor building closed cooling water (RBCCW) system has been eliminated as a result of locking closed
the ESW-RBCCW cross tie valves. These valves were locked closed because of the lack of required structural design of the piping, and due to the adverse hydraulic effects to safety related
components served by ESW. Therefore, the cooling effect of the
RBCCW system to any of the components described above will be
minimal.
A radiation monitor is provided at the cooling water return header
to indicate, record, and alarm leakage of radioactivity.
10.8.4 Inspection and Testing
Equipment in the reactor building cooling water system is proven
operable by use during normal plant operations. Motor operated
isolation valves can be tested to assure they are capable of
opening and closing by operating manual switches in the control
room and observing the position lights. Portions of the reactor
building cooling water system normally closed to flow can be
tested to ensure their operability and the integrity of the
system.
CHAPTER 10 10.8-3 REV. 21, APRIL 2007 TABLE 10.8.1 REACTOR BUILDING COOLING WATER SYSTEM
EQUIPMENT DATA
Reactor Building Cooling
Water System Pumps 2 (full-capacity)
Type Horizontal Centrifugal
Flow and Head 1,350 gpm at 140 ft
Bhp at Rating 65 hp Material:
Casing/Impeller/Shaft Cast Iron/Bronze/Stainless Steel Motor: Size 75 hp Voltage/Phase/Frequency 440 V/3 Phase/60 Hz
Speed 3,600 rpm
Reactor Building Cooling
Water System Heat Exchangers
Quantity 2 (full-capacity)
Type Horizontal, Shell and Tube
Heat Transfer Duty 25,500,000 Btu/hr
Shell Design:
Pressure/Temperature 150 psig/200 F Material Carbon steel
Flow Medium Inhibited Demineralized Water Tube design:
Pressure/Temperature 125 psig/200 F Material:
Tube Admiralty
Tube Sheet Carbon Steel
Tube Joint Rolled Flow Medium River Water
CHAPTER 10 10.9-1 REV. 21, APRIL 2007 10.9 EMERGENCY SERVICE WATER SYSTEM 10.9.1 Safety Objective
The safety objective of the emergency service water system is to
provide a reliable supply of cooling water to diesel generator
coolers, ECCS and RCIC compartment air coolers, Core Spray Pump
Motor Oil Coolers and other selected equipment during a loss of
offsite power or during a loss of normal station service water due
to the design flood condition or the loss of the Conowingo pond.
10.9.2 Safety Design Basis
- 1. The emergency service water system is designed to seismic Class I criteria.
- 2. The emergency service water system is operable during the design flood condition and loss of Conowingo pond.
- 3. The emergency service water system has sufficient capacity and redundancy so that no single active
component failure can prevent the system from achieving
its safety objective.
- 4. The emergency service water system is operable during the loss of offsite power.
10.9.3 Description
The emergency service water system is common to both Units 2 and
- 3. The system consists of two full-capacity pumps installed in
parallel in the seismic Class I portion of the pump structure, and
associated equipment coolers, valves, and controls (Drawing M-
315). Normal water supply to the suction of the emergency service
water system pumps is from Conowingo Pond. The pump discharge
piping consists of two headers with service loops to ensure water
supply to the diesel engine coolers. These two headers combine, forming a common header, to supply selected equipment coolers.
Valves in the supply headers provide loop isolation. A common
discharge header routes the system effluent normally to the pond.
The emergency service water system pumps are vertical, single-
stage, turbine type with an 8,000 gpm capacity developing a normal
average system pressure of 40 psig and a normal system flow of
approximately 4500 gpm. The pump mounting base is of watertight
construction to withstand hydrostatic pressure at the maximum
design flood condition. The pump design data is given in Table
10.9.1.
The emergency service water system is a standby system to provide
adequate cooling water supply to the emergency equipment coolers
CHAPTER 10 10.9-2 REV. 21, APRIL 2007 and compartment air coolers during a loss of offsite power or during a loss of normal station service water due to the design
flood condition or the loss of the Conowingo pond. During normal
plant operating conditions, the cooling water supply to the
equipment served by the emergency service water system, except the
diesel generator coolers, is normally provided from the
service water system. This allows testing of safeguards equipment
using service water without starting the emergency service water
system pumps.
Chemical injection and corrosion monitoring systems are installed
to mitigate corrosion damage to emergency service water system
piping.
The emergency service water system may also be operated in
conjunction with the emergency heat sink (subsection 10.24). This
configuration (closed loop) is the preferred system alignment
during the design flood condition and loss of Conowingo pond.
Both emergency service water pumps start after a 36 second time
delay whenever 4 kV power is available (following the loss of
offsite power or a diesel generator start). One of the emergency
service water system pumps is manually shut off if both pumps are
running and emergency service water system pressure is verified to
be adequate. All system supervisory instrumentation and controls
are located in the main control room.
The emergency service water system piping and valves are designed
as described in Appendix A.
10.9.4 Safety Evaluation
The emergency service water system pumps are installed in a
seismic Class I structure. The system meets seismic Class I
criteria, and the pumps are further protected against the design
flood level using watertight construction. The emergency service
water system is designed with redundant pumps and piping. Each
loop is powered from a separate division of both normal and
standby power. Therefore, the system is both redundant and single
failure proof and is operable in the event of a loss of offsite
power.
10.9.5 Inspection and Testing
The cooling of equipment served by the emergency service water
system, except the standby diesel generator coolers, is
functionally tested using the plant service water system. Pump
operation and diesel generator cooling capability is verified when
operability of the diesel generators is tested. Motor operated
CHAPTER 10 10.9-3 REV. 21, APRIL 2007 valves can be exercised to confirm operability. Emergency service water system operability is verified by flow and heat transfer
testing. Emergency service water system piping integrity is
verified by visual and ultrasonic inspection and corrosion
monitoring.
Emergency service water pump performance is verified in accordance
with ASME Code requirements. Cooling equipment minimum flows are
verified by magnetic or ultrasonic flow measurement devices.
The timer used to sequence the emergency service water pump during
a LOCA is tested (with offsite power available) in accordance with
surveillance test procedures. The test verifies the setting, operability, and functional performance of the relay, and provides
assurance that the automatic loading sequence is being maintained
and performs as required.
CHAPTER 10 10.9-4 REV. 21, APRIL 2007 TABLE 10.9.1 EMERGENCY SERVICE WATER SYSTEM EQUIPMENT DATA
Emergency Service Water System Pumps*
Quantity 2 (common for Units 2 and 3)
Type Vertical, Turbine Type
Flow/Head 8,000 gpm/96 ft
Bhp at Rating 237 hp Speed 1,170 rpm Number of Stages 1
Pump Design:
Shutoff Head 132 ft Maximum Working Pressure 200 psig
Material: Moly Iron
Bowl/Impeller Cast Iron/Bronze
Discharge Head/Column Carbon Steel/Carbon Steel
Line Shaft Stainless Steel
Bearings Rubber Motor Design:
Type Vertical Induction Type
Horsepower 250 hp Voltage/Phase/Frequency 4,160 V/3 Phase/60 Hz
- Emergency cooling water pump and motor are identical except
for the pump column length.
CHAPTER 10 10.10-1 REV. 21, APRIL 2007 10.10 TURBINE BUILDING COOLING WATER SYSTEM 10.10.1 Power Generation Objective
The power generation objective of the turbine building cooling
water system is to provide cooling water to auxiliary plant
equipment associated with the power conversion systems.
10.10.2 Power Generation Design Basis
The turbine building cooling water system is designed to cool non-
nuclear auxiliary plant equipment over the full range of plant
operation.
10.10.3 Description
The system consists of two full-capacity pumps, two full-capacity
heat exchangers (system design does allow use of both heat
exchangers if necessary due to high river temperatures or other
limiting operating conditions), one head tank, one chemical feed
tank, and associated piping, valves, and controls (Drawing M-316, Sheets 1 to 4). The cooling water pumps and heat exchangers are
located on the turbine building ground floor. The system design
data is given in Table 10.10.1.
The system is a closed loop utilizing inhibited demineralized
water. The heat exchangers are designed with service (river)
water on the tube side and demineralized water on the shell side.
The head tank, located at the highest point in the loop, accommodates system volume changes, maintains static suction
pressure on the pumps, aids in detecting gross leaks in the
turbine building cooling water system, and provides a means for
adding makeup water. An automatic makeup control valve maintains
water level in the tank. The automatic function is not required
and may be valved out to monitor system inventory. High and low
water levels are alarmed in the main control room. An inhibitor
is added as necessary to the demineralized water by means of a
chemical addition tank to limit corrosion.
In the event of offsite power failure, the turbine building
cooling water system is not operated. Under loss of offsite
power, the water supply to the instrument and service air
compressor skids, CRD pump lube oil coolers and the thrust bearing
housings is maintained from the reactor building cooling water
system. This design feature still exists although the heat sink, emergency service water (ESW), for the reactor building closed
cooling water (RBCCW) system has been eliminated as a result of
locking closed the ESW cross tie valves.
CHAPTER 10 10.10-2 REV. 21, APRIL 2007 Therefore, little, if any, cooling would be provided by the reactor building cooling water system during a loss of offsite
power.
10.10.4 Inspection and Testing
Equipment in the turbine building cooling water system is proven
operable by use during normal plant operations. Transfer valves
in the system can be tested to ensure that they are capable of
transferring the water supply of essential equipment from the
turbine building cooling water system to the reactor building
cooling water system on loss of offsite power. This design
feature still exists although the heat sink, emergency service
water (ESW), for the reactor building closed cooling water (RBCCW)
system has been eliminated as a result of locking closed the ESW
cross tie valves. Therefore, little, if any, cooling would be
provided by the reactor building cooling water system during a
loss of offsite power. System subsections normally closed to flow
can be tested to ensure their operability and system integrity.
CHAPTER 10 10.10-3 REV. 21, APRIL 2007 TABLE 10.10.1 TURBINE BUILDING COOLING WATER SYSTEM EQUIPMENT DATA
Turbine Building Cooling Water System Pumps
Quantity 2 (full-capacity)
Type Horizontal, Centrifugal
Flow and Head 525 gpm at 180 ft
Bhp at Rating 34 hp Material:
Casting/Impeller/Shaft Cast Iron/Bronze/Stainless
Steel Motor: Size 40 hp Voltage/Phase/Frequency 440 V/3 Phase/60 Hz
Speed 3,600 rpm
Turbine Building Cooling Water System Heat Exchangers
Quantity 2 (full-capacity)
Type Horizontal, Shell and Tube
Heat Transfer Duty 3,850,000 Btu/hr
Shell Design:
Pressure/Temperature 150 psig/200 F Material Carbon Steel
Flow Medium Inhibited Demineralized Water Tube Design:
Pressure/Temperature 125 psi/200 F
CHAPTER 10 10.10-4 REV. 21, APRIL 2007 TABLE 10.10.1 (Continued)
Material:
Tube Admiralty
Tube Sheet Carbon Steel
Tube Joint Rolled Flow Medium River Water
CHAPTER 10 10.11-1 REV. 21, APRIL 2007 10.11 CHILLED WATER SYSTEM 10.11.1 Power Generation Objective
The power generation objective of the chilled water system is to
provide cooling water to the auxiliary equipment inside the
10.11.2 Power Generation Design Basis
- 1. The chilled water system is designed to cool the auxiliary equipment inside the primary containment over
the full range of plant operation.
- 2. The chilled water system provides a reliable source of cooling water.
- 3. The chilled water system is provided with inter-ties with the reactor building cooling water system, which
serves the chilled water system during a loss of offsite
power. This design feature exists although the heat
sink, emergency service water (ESW), for the reactor
building closed cooling water (RBCCW) system has been
eliminated as a result of locking closed the ESW-RBCCW
cross tie valves. Therefore, little, if any, cooling
would be provided by the chilled water system during a
loss of offsite power.
10.11.3 Description
The chilled water system consists of three half-capacity, centrifugal refrigeration units, three half-capacity chilled water
pumps, an expansion tank, piping, valves, instrumentation, and
controls (Drawing M-327, Sheets 1 through 4). It is a closed-loop
system utilizing inhibited demineralized water. The pumps
circulate warm return water to the refrigeration unit chillers.
The chilled water is then piped to the drywell air coolers, the
recirculation pump motor coolers, and the drywell equipment sump
cooler. Two parallel supply headers and return headers penetrate
the primary containment. A motor operated isolation valve is
located outside the containment in each line. The inter-tie with
the reactor building cooling water system is made by motor
operated three-way valves. An automatic transfer from system to
system is made upon loss of offsite power. Chilled water system
shutdown requires a manual switchover. Chillers and pumps are
remotely controlled from the main control room. A standby start
feature is provided for each chilled water pump. Standby equipment
is provided to assure system reliability.
CHAPTER 10 10.11-2 REV. 21, APRIL 2007 10.11.4 Inspection and Testing
The chilled water system is proved operable by use during normal
plant operation. Portions of the system normally closed to flow
can be tested to ensure operability and integrity of the system.
CHAPTER 10 10.12-1 REV. 21, APRIL 2007 10.12 FIRE PROTECTION PROGRAM The Fire Protection Program (FPP) is described in a document
transmitted to the NRC on September 30, 1986 titled, "Fire
Protection Program, Peach Bottom Atomic Power Station, Units 2 and
3", and is hereby incorporated by reference into the UFSAR.
Chapter 1 of the FPP is an introduction.
Chapter 2 provides a general description of the fire detection and
suppression systems.
Chapter 3 presents an item-by-item comparison of the Peach Bottom
Atomic Power Station Units 2 and 3 fire protection program with
the guidelines set forth in Branch Technical Position APCSB 9.5-1, Appendix A, the requirements of Appendix R to 10CFR50, and the
requirements of the Fire Protection Safety Evaluation Report.
Chapter 4 provides a tabulation of the combustible loadings in
plant fire areas, describes fire barriers, and describes fire
detection and suppression systems in each area. The plant is
divided into 47 fire areas.
Chapter 5 provides an evaluation of the ability to safely shut the
plant down in the event of a fire in any one of the plant's 47
fire areas.
Chapter 6 addresses special topics.
Chapter 7 contains the fire protection requirements which have
been relocated from the Technical Specifications by Technical
Specifications Change Request 90-05, which was submitted to the
NRC on March 28, 1994. The relocation of these requirements was
in accordance with NRC Generic Letters (GL) 86-10, "Implementation
of Fire Protection Requirements," and GL 88-12, "Removal of Fire
Protection Requirements from Technical Specifications."
In addition to the above, administrative procedures, system
operating procedures, surveillance tests, and pre-fire strategy
plans have been established to implement the Fire Protection
Program.
CHAPTER 10 10.13-1 REV. 26, APRIL 2017 10.13 MAIN CONTROL ROOM AIR CONDITIONING 10.13.1 Power Generation Objective
The power generation objective of the main control room air
conditioning system is to provide a suitable environment for
continuous personnel occupancy and to ensure the operability of
control room equipment and instruments under normal and accident
conditions per 12.3.4.
10.13.2 Power Generation Design Basis
- 1. The system is designed to provide an environment with a controlled temperature. Humidity control is available
during periods of auxiliary boiler operation.
- 2. The system is capable of purging the main control room.
- 3. Redundant components are provided to ensure reliable system operation.
10.13.3 Safety Design Basis
- 1. The system is designed such that the control room is habitable even under the design basis accident
conditions.
- 2. The fresh air portion of the system is designed to be operable during the loss of offsite power by using the
standby power supplies.
- 3. The fresh air intake is filtered when main control room emergency ventilation is initiated to prevent iodine and
particulate contamination of the main control room air.
10.13.4 Description
The main control room air conditioning system consists of
ventilation air supply fans (normal), emergency air supply fans, air conditioning supply and return fans, filters, heating coils
and cooling coils, refrigerant water chillers, chilled water
pumps, dampers, duct work, instrumentation, and controls (Drawing
M-384).
Outside air is drawn through a filter by a fresh air supply fan
and is discharged to the inlet of the air conditioning supply fan
suction, and is then discharged to duct work leading to the main
control room and adjacent offices. This air is conditioned to
maintain a controlled temperature environment using heating and
CHAPTER 10 10.13-2 REV. 26, APRIL 2017 cooling coils. Humidity is conditioned during periods of auxiliary boiler operation. Normally, control room air is
recirculated by one of two return air fans. These fans take a suction from the north and south ends of the control room and
discharge to the air conditioning supply fan suction with filtered
outside air from the fresh air supply fans.
The control room heating coils are supplied from the auxiliary
steam supply. Cooling is provided by a chilled water system
consisting of two 100% chiller units, two 100% chilled water
pumps, and a piping system which also supplies chilled water to
the cable spreading room fan-coil supply unit and the health
physics and chemistry labs fan-coil supply unit. The fresh air
supply fans, both normal and emergency, are operable from the
standby power supply during the loss of offsite power. The
control room chiller and air conditioning supply and return fans
do not run following loss of offsite power. The instrumentation
and control for the main control room air conditioning system is
designed for automatic operation. One fresh air supply fan, one
air conditioning supply fan, and one return air fan are normally
in operation. Emergency cooling and ventilation systems for the
control room and other safety-related equipment rooms are
installed in seismic Class I structures and are provided with 100
percent redundancy. Monitoring and adjustment of the control room
emergency ventilation system air flow may be performed locally.
If an operating fan fails, the loss of duct pressure is sensed and
the standby fan starts automatically, the associated fan dampers
open, and an alarm sounds in the control room. Fans may also be
started manually.
A radiation monitoring system in the fresh air intake duct work
monitors the radioactivity level in the incoming outside air. This
system includes two flow switches that monitor air flow through
the fresh air intake duct work. If a high activity level or loss
of flow is detected, the operating normal fresh air supply fan
stops, one emergency air supply fan starts, and the air
conditioning supply and return air fans shut down. The air is
diverted through one of the two high efficiency and charcoal
filter trains automatically. The monitor also annunciates in the
control room. If a high-high activity level is detected, the
monitor will indicate in the control room.
The control room is capable of being purged with 100 percent
outside air. A once-through flow is established using the air
conditioning supply fans, with the return air fans discharging to
atmosphere at the radwaste building roof.
CHAPTER 10 10.13-3 REV. 26, APRIL 2017 10.13.4.1Control Room Habitability The primary design function of the Main Control Room (MCR) / Main
Control Room Emergency Ventilation (MCREV) System is to provide a
safe environment in which the operator can keep the nuclear
reactor and auxiliary systems under control during normal
operations and can safely shut down those systems during abnormal
situations to protect the health and safety of the public and
plant workers.
Technical Specifications 3.7.4 and its Bases are in place to
ensure that appropriate equipment is maintained operable and
inoperabilities are managed through compensatory actions and other
plant actions.
A Control Room Envelope (CRE) Habitability Program is required by
Technical Specifications 5.5.13. The program is established and
implemented to ensure that the CRE habitability is maintained such
that, with an operable MCREV system, CRE occupants can control the
reactor safely under normal conditions and maintain it in a safe
condition following a radiological event, hazardous chemical
release as applicable, or a smoke challenge. The program shall
ensure that adequate radiation protection is provided to permit
access and occupancy of the CRE under design basis accident (DBA)
conditions without personnel receiving radiation exposures in
excess of 5 rem total effective does equivalent (TEDE) for the
duration of the accident. The program includes elements required
by Technical Specification 5.5.13.
As a result of Technical Specification 3.7.4 and 5.5.13
requirements, PBAPS is committed to applicable portions of NRC Reg
Guide 1.197, NRC Reg Guide 1.196 as invoked by the Technical
Specifications or its Bases. PBAPS is committed to NRC Reg Guide
1.78 (6/74) and NRC Reg Guide 1.95 (2/75), as applicable, for
hazardous chemical assessments. The computer code HABIT is
utilized for hazardous chemical assessments, which was approved in
Revision 1 of Reg Guide 1.78. This is an exception to Revision 0, to which PBAPS remains committed. Additionally, Peach Bottom performs hazardous chemical assessments by probabilistic analysis in accordance with NUREG-0800, Standard Review Plan, Section 2.2.3.
10.13.5 Safety Evaluation
The fresh air portion of the main control room ventilation system
permits continuous occupancy of the main control room
under normal and accident conditions, including maximum credible
earthquake, contaminated outside air, and loss of offsite power.
The system has sufficient redundancy to maintain uninterrupted
main control room ventilation for personnel occupancy and
CHAPTER 10 10.13-4 REV. 26, APRIL 2017 instrument operability. Evaluation as to the expected dose rates under the design basis accident conditions is included in
paragraph 12.3.4.
10.13.6 Inspection and Testing
The main control room air conditioning system is proven operable
by its use during normal plant operation. Portions of the system
normally closed to flow can be tested to ensure operability and
integrity of the system.
CHAPTER 10 10.14-1 REV. 21, APRIL 2007 10.14 EMERGENCY VENTILATING SYSTEM 10.14.1 Safety Objective
The safety objective of the emergency ventilating systems is to
maintain suitable temperatures in the plant engineered safeguards
equipment rooms for equipment protection.
10.14.2 Safety Design Basis
- 1. The systems protect the safeguards equipment against overheating.
- 2. Selected systems shall be provided with redundant components for reliable operation.
- 3. The equipment is provided with alternate power supplies in the event of loss of offsite power.
- 4. The equipment is designed to seismic Class I criteria.
10.14.3 Description
The emergency ventilating systems include the following:
- 1. Emergency switchgear and battery rooms.
- 2. Standby diesel generator rooms.
- 3. Pump structure ventilation system (ESW/HPSW Compartment).
- 4. Pump rooms for the RHR, RCIC, HPCI, and core spray pumps.
The reactor building heating and ventilating system normally
supplies ventilation air to the RHR, RCIC, HPCI, and core spray
pump rooms (paragraph 5.3.2).
10.14.3.1 Emergency Switchgear and Battery Rooms
The system consists of a common air supply system and separate
exhaust systems for emergency switchgear and battery rooms (Drawing M-399). Outdoor air is filtered, conditioned by heating
coils when required, and discharged by one of the two supply fans
to the emergency switchgear and battery rooms of Units 2 and 3.
One of the two emergency switchgear room return air fans exhaust
air to atmosphere at the radwaste building roof or back to the suction of the supply fan as controlled by an air-operated damper.
One of the two battery room exhaust fans discharges exhaust air
from the battery rooms to atmosphere at the radwaste building
CHAPTER 10 10.14-2 REV. 21, APRIL 2007 roof. Loss of duct pressure automatically starts standby fans and sounds an alarm in the main control room.
The equipment is installed in a seismic Class I structure adjacent
to the main control room. The ventilation system is normally in
operation and continues to operate during accident conditions
including the loss of offsite power. All system controls are from
a local panel. Redundant fans are provided for reliable system
operation. A seismically supported, safety grade, pneumatic
supply has been provided to maintain the dampers open in accident
conditions.
10.14.3.2 Standby Diesel Generator Rooms
Each standby diesel generator room is provided with ventilation
air supply fans and an exhaust relief damper (Drawing M-385).
Combustion air for the diesel engine is taken from the room. The
ventilation systems are supplied with power from the diesels
during the loss of offsite power.
10.14.3.3 ESW/HPSW Compartments
The ESW/HPSW compartment housing the high pressure service water
pumps, emergency service water pumps, fire pumps, and service
water screen wash pumps is provided with a ventilation supply and
exhaust system in each of the two seismic Class I compartments.
The ventilation system is supplied with standby power during the
loss of offsite power. Redundant ventilation equipment is
furnished in each compartment for uninterrupted service. The pump
structure ventilation system for each HPSW subsystem is single
failure proof.
10.14.4 Safety Evaluation
The emergency equipment rooms are provided with cooling and
ventilating systems with sufficient redundancy to ensure proper
operation of equipment during normal and accident conditions. In
addition, equipment is designed and installed in accordance with
seismic Class I criteria and is supplied with normal and standby
power.
10.14.5 Inspection and Testing
The emergency ventilating systems are proved operable by use
during normal plant operation. The effectiveness of the energy
removal from the local environments can be evaluated by measuring
the compartment air temperatures where the equipment is located.
Portions of the systems normally closed to flow can be tested to
ensure operability and integrity of these systems.
CHAPTER 10 10.14-3 REV. 21, APRIL 2007 The instantaneous auxiliary relays used to sequence the diesel generator room vent supply fans and the residual heat removal
compartment fan coolers during a LOCA (with offsite power
available) will be tested in accordance with surveillance test
procedures. The test will verify the settings, operability, and
functional performance of the relays, and will provide assurance
that the automatic loading sequence is being maintained and will
perform as required.
CHAPTER 10 10.15-1 REV. 21, APRIL 2007 10.15 PLANT HEATING, VENTILATING, AND AIR CONDITIONING SYSTEMS 10.15.1 Power Generation Objective
The power generation objective of the plant heating, ventilating, and air conditioning systems is to control the plant air
temperatures and the flow of airborne radioactive contaminants to
ensure the operability of plant equipment and the accessibility
and habitability of plant buildings.
10.15.2 Power Generation Design Basis
The plant heating, ventilating, and air conditioning systems:
- 1. Provide appropriate temperature control for personnel comfort and equipment performance.
- 2. Provide sufficient filtered fresh air supply for personnel.
- 3. Provide air movement patterns from areas of lesser to areas of progressively greater contamination potential
prior to final exhaust.
- 4. Minimize the possibility of plant exhaust air recirculation into the plant air intake.
10.15.3 Description
10.15.3.1 General
The plant heating, ventilating, and air conditioning systems
provide heated or cooled air to main areas of the plant. Supply
air temperature is controlled by heating coils and cooling coils.
Generally, airflow is routed from areas of lesser to areas of
progressively greater contamination potential prior to final
exhaust. Also, the ventilation system has sufficient design
capacity to protect equipment from excessive temperatures.
The exhaust ventilation air from the turbine building and radwaste
building is discharged to atmosphere from the ventilation stack
above the reactor building roof. Exhaust from areas where
radioactive particulate may be present, such as equipment rooms, is not recirculated but is exhausted through high-efficiency
filters to atmosphere. Clean exhaust air from other plant areas
is not filtered prior to being released to atmosphere.
The reactor building heating and ventilating system is described
in paragraph 5.3.2. The main control room air conditioning system
is described in subsection 10.13, and the emergency heating and
ventilating systems are described in subsection 10.14.
CHAPTER 10 10.15-2 REV. 21, APRIL 2007 10.15.3.2 Turbine Building
The ventilation system supplies filtered and tempered outdoor air
to the operating floor, main condenser area, and equipment
compartments (Drawing M-387). The main condenser area is
maintained at a slight negative pressure to reduce exfiltration of
potential radioactive contaminants to the adjacent areas.
Ventilation air to the operating floor is recirculated or
exhausted as required to maintain space temperature. The exhaust
air from the operating floor and the main condenser area is
discharged to the atmosphere through the ventilation stack located
at the top of the reactor building. Air from potentially
contaminated equipment compartments is exhausted through high-
efficiency filters prior to release to the atmosphere at the
ventilation stack. Supplementary cooling in the main condenser
area and condensate pump room is provided by fan-coil units using
service water for cooling. Additionally, unit heaters are
provided in various areas for equipment freeze protection.
10.15.3.3 Radwaste Building
The ventilation system for the radwaste building maintains a
supply of filtered and tempered fresh air to all areas of the
radwaste building (Drawing M-389). Generally, air is distributed
from areas of lesser to areas of progressively higher
contamination.
Two exhaust systems are used: normal and equipment compartment
exhaust. The normal exhaust is unfiltered and is discharged to
atmosphere at the reactor building roof through the ventilation
stack. The equipment compartment exhaust air is filtered prior to
release to atmosphere from the ventilation stack. Air vented from
tanks containing radioactive liquids is exhausted through high-
efficiency filters prior to joining the equipment compartment
exhaust duct work.
10.15.3.4 Miscellaneous Rooms and Buildings
The cable spreading room, located beneath the main control room, is provided with its own supply and exhaust fans, filters, heating
and cooling coils, duct work, instrumentation, and controls.
Redundancy in the number of fans provides continued operation of
the system. These fans shut down on a loss of offsite power.
The computer room, located in the cable spreading room area, is
provided with self-contained air conditioning units, filters, and
controls to maintain constant temperature and humidity in the
room. These units operate from the standby power supply during a
loss of offsite power.
CHAPTER 10 10.15-3 REV. 21, APRIL 2007 The administration building, chemical laboratory rooms, shop and
warehouse building, water treatment building, and other structures
in the plant are provided with separate conventional heating, ventilating, and/or air conditioning system.
10.15.4 Inspection and Testing
The plant heating, ventilating, and air conditioning systems are
proved operable by their use during normal plant operation.
Portions of the systems normally closed to flow can be tested to
ensure operability and integrity of the systems.
CHAPTER 10 10.16-1 REV. 23, APRIL 2011 10.16 MAKEUP WATER TREATMENT SYSTEM 10.16.1 Power Generation Objective
The power generation objective of the plant makeup water treatment
system is to provide a supply of water suitable as makeup for the
plant and reactor systems and other water requirements.
10.16.2 Power Generation Design Basis
The makeup water treatment system is designed to provide reactor
quality water for pre-operational tests, startup, and normal power
operation.
10.16.3 Description
The makeup water treatment system is common to Units 2 and 3 (Drawings M-317 and M-319).
The makeup water treatment system receives river water from the
service water system. The system consists of a raw water
treatment system, a clarified water storage tank, a makeup
demineralizer system, a demineralized water tank, and associated
pumps, piping, and instrumentation.
The raw water treatment system consists of skid mounted equipment that is vendor supplied and operated. This equipment produces up to 400 gpm of clarified and filtered water with a nominal flow of 200 gpm for use in the makeup demineralizer system, domestic water, and other uses. This water is pumped to a 200,000 gallon clarified water storage tank. The clarified and filtered water is continuously monitored by vendor supplied turbidity and pH measuring devices which initiate an alarm on a vendor panel.
The makeup water demineralizer system consists of three feed pumps
taking suction on the clarified water storage tank and discharging
to vendor supplied ultra pure water equipment. The discharge from
the ultra pure water equipment goes to a 50,000-gal demineralized
water storage tank. The discharge to the storage tank is
monitored for quality by conductivity measuring devices which
initiate an alarm on a local panel. Also, silica content is
continuously monitored and recorded. The quality of water
discharged to the storage tank is controlled to maintain water
within the limits specified in EPRI BWR Water Chemistry
Guidelines.
The piping, tanks, and associated equipment of the demineralized
water treatment system are of corrosion-resistant metals which
prevent contamination of the makeup water with foreign material.
CHAPTER 10 10.16-2 REV. 23, APRIL 2011 10.16.4 Inspection and Testing The makeup water treatment system is an operational system in
daily use and as such does not require testing to ensure
operability. High demineralizer effluent conductivity
automatically initiates an alarm. Grab samples are tested in the
laboratory to check demineralizer performance and to ascertain
stored water quality.
CHAPTER 10 10.17-1 REV. 25, APRIL 2015 10.17 INSTRUMENT AIR, SERVICE AIR, AND INSTRUMENT NITROGEN SYSTEMS 10.17.1 Safety Objective
The safety objective of the instrument air, service air, and
instrument nitrogen systems is to provide a safety grade, pneumatic supply to support short-term and long-term operations of
safety equipment.
10.17.2 Safety Design Basis
- 1. The containment atmospheric control system containment purge and vent isolation valves are each provided with a
backup, safety grade, pneumatic (nitrogen) supply to the
valves' inflatable seals.
- 2. The containment isolation and flow control valves in the CAD vent lines are each provided with a separate, backup, safety grade, pneumatic (nitrogen) supply. The
control valves in the CADS supply are provided with a
safety grade supply of nitrogen from the CADS nitrogen
supply.
- 3. The ADS valves are provided with a separate short-term, safety grade, pneumatic supply and also a long-term, backup, safety grade, pneumatic supply of nitrogen. To
fulfill the requirements of Appendix R to 10CFR, Part 50
manual actions may be performed to connect a back-up
safety grade pnuematic nitrogen supply to enable remote
operation of safety relief valves. See FPP Table A-4
for actions required to credit this pneumatic supply.
- 4. A separate short-term, backup, safety grade, pneumatic supply is provided to each of the MSIVs.
- 5. The suppression chamber-to-secondary containment vacuum breaker air-operated valves are each provided with a
backup, safety grade, pneumatic supply.
- 6. The emergency switchgear and battery room dampers are provided with a backup, safety grade, pneumatic supply.
10.17.3 Power Generation Objective
The power generation objective of the instrument air and service
air systems is to supply suitable quality air at adequate pressure
for power plant operation.
CHAPTER 10 10.17-2 REV. 25, APRIL 2015 10.17.4 Power Generation Design Basis
- 1. The instrument air system supplies clean, dry, oil-free air, nominally at 100 psig, to station instrumentation
and controls.
- 2. The service air system supplies clean air, nominally at 100 psig, for station services.
- 3. Standby onsite power is provided to the backup air compressors, following a loss of offsite power, to
replenish compressed air storage as required. This
design feature exists although the heat sink, emergency
service water (ESW), for the reactor building closed
cooling water (RBCCW) system has been eliminated as a
result of locking closed the ESW-RBCCW cross tie valves.
Therefore, little, if any, cooling would be provided to
the backup air compressors during a loss of offsite
power.
- 4. Service air use is restricted during emergency conditions so that the instrument air supply shall not
be impaired.
- 5. Two separate systems are provided for each unit for the condensate filter demineralizer backwash operations.
- 6. Instrument nitrogen/instrument air is provided to the MSIV's for maintaining the valves open when operating
the steam cycle.
- 7. A separate air supply system is provided to selected radwaste equipment.
10.17.5 Description
The instrument air and service air systems (Drawing M-320) consist
of four air compressors per unit operating in parallel to supply
common discharge headers via individual air receiver tanks, piping, valves, and instrumentation. The instrument and service
air systems of Units 2 and 3 can be crosstied.
Two of the three larger compressors (650 SCFM) normally supply all
compressed air requirements for one reactor unit during normal
operation. The three larger compressors are fed from non-1E power
sources. During emergency conditions when neither station nor
offsite power are available, the smaller (419 SCFM) compressor, which is fed by a Class 1E power source, is designed to provide
desired operational flexibility. This design feature exists
although the heat sink, emergency service water (ESW), for the
CHAPTER 10 10.17-3 REV. 25, APRIL 2015 reactor building closed cooling water (RBCCW) system has been eliminated as a result of locking closed the ESW-RBCCW cross tie
valves. Therefore, little, if any, cooling would be provided to
the air compressors during a loss of offsite power.
The instrument air compressors normally operate (load and unload)
within a pressure range of approximately 10 psi. The service air
compressor, which can feed either of the instrument air headers
and the service air header, operates over approximately a 15 psi
range so that it will not assume control from the instrument air
compressors. In the unlikely event that header pressure decays to
97 psig, an air operated valve in the supply to the service air
header will close, thus utilizing the instrument and service air
compressors for instrument air only. The duty status of all three
compressors can be changed to allow for maintenance and
equalization of wear. During a loss of station or offsite power, only the 419 SCFM compressor (backup compressor), is fed by diesel
backed power. This design feature exists although the heat sink, emergency service water (ESW), for the reactor building closed
cooling water (RBCCW) system has been eliminated as a result of
locking closed the ESW-RBCCW cross tie valves. Therefore, little, if any, cooling would be provided to the air compressors during a
loss of offsite power. During a LOCA event the affected unit's
backup compressor will trip if running or will be prevented from
starting for 60 seconds. Identical backup compressors are
provided for each unit, and manual crossties are provided so that
either or both backup compressors can be utilized to supply either
unit. A single backup compressor is sufficient to supply
operational flexibility for one unit during shutdown following a
Each of the four compressors is of the 2-stage oil-free rotary
screw design. These compressors are water cooled and are complete
packaged units, incorporating an inter-cooler, after-cooler, oil
cooler, bleed-off cooler, and all controls and instrumentation in
a single sound attenuating enclosure. "Compressor trouble" alarms
actuated by any compressor trip functions are provided both in the
control room and locally on the compressors.
The lead compressors are rated at 691 ACFM at 125 psig maximum and
utilize a 150 HP motor with a 1.15 service factor. The backup
compressors are rated at 456 ACFM at 125 psig maximum and utilize
a 100 HP motor with a 1.15 service factor. All compressor motors
are capable of starting and accelerating at 75% of nominal
voltage.
The prefilters of the air dryer are the coalescing type and are
designed to remove effluent to 0.0013 ppmw. The dual tower
instrument air dryer package is rated at 900 scfm, is of the
heatless design and utilizes activated alumina desiccant for
absorption of moisture. The dryer is designed for a discharge dew
CHAPTER 10 10.17-4 REV. 25, APRIL 2015 point of -40 F. Each dryer incorporates a moisture sensing control which measures the actual moisture load present on the desiccant during each cycle. It then limits the number of
regeneration (purge) cycles to only those required to remove
moisture to maintain the required outlet dew point. The after
filters are designed to remove particulate to 0.9 micron absolute.
Also incorporated in the dryer skid package are a flow meter and a
dew point analyzer. The flow meter is provided with flow
recording and totalizing capabilities to enable continuous
monitoring of plant air usage. The dew point meter gives a
continuous readout of moisture level. There are local alarms on
each dryer skid to indicate high differential pressure across the
prefilters, high moisture content in the outlet air, and dryer
control malfunctions. There is one common alarm window per unit, located on C212L in the control room, to indicate an alarm
condition exists on either of the two dryer skids for that unit.
The discharge from the dryer package than passes to the plant
instrument air headers.
Breathing air stations are provided with air from the service air
system headers.
Since the air is supplied by non-lubricated compressors, the
instrument air system is supplied with clean, dry, oil-free air
for use by instrumentation and controls. This system supplies air
to the main steam isolation valves external to containment. The
main steam isolation valves are provided with accumulators for
reliable operation without compressor operation.
The control rod hydraulic control system air requirement is for
scram reset purposes only, and its demand is met by the CRDS
storage capacity. Other pneumatic-operated devices are also
designed for the fail-safe mode, and do not require a continuous
air supply under abnormal conditions.
The Condensate Filter Demineralizer backwash operation employs the
use of two separate air backwash systems. The primary system is a
high pressure air surge system which uses a non-lubricated, two
stage, water cooled compressor rated at 200 scfm at 200 psig.
Each unit has two separate receiver tanks (150 ft 3 each) which are cross-tied to allow either compressor to charge both units. In
addition each unit is provided with a backup low pressure air
scrub backwash system using low pressure centrifugal blowers rated
at 1400 scfm.
The containment atmospheric control system purge and vent valves
are supplied with separate safety grade pneumatic supplies to the
inflatable seals to maintain their leaktight condition. The
source of this pneumatic supply is from the Safety Grade
CHAPTER 10 10.17-5 REV. 25, APRIL 2015 Instrument Gas (SGIG) system. The SGIG supplies pressurized nitrogen gas from the CAD tank as a backup to normal instrument
air. The safety grade pneumatic supply is isolated from the
nonsafety grade portion of the air supply by spring-loaded, soft-
seat, check valves designed for zero leakage. The purge and vent
valves alarm upon opening or on loss of seal pressure.
The suppression chamber-to-secondary containment vacuum breaker
air-operated valves are each supplied with separate, safety grade, pneumatic supplies. There are two suppression chamber-to-
secondary containment vacuum breaker lines on each unit. Each
line is provided with a normally closed, fail open, air-operated
butterfly valve. Each of these valves is provided with a safety
grade pneumatic supply in order to maintain valve closure. One
valve on each unit is equipped with an inflatable seal which is
also supplied by the safety grade pneumatic supply. These valves
alarm upon opening or, for the valves equipped with the inflatable
seal, on loss of seal pressure. The source of this pneumatic
supply is from the Safety Grade Instrument Gas (SGIG) system. The
SGIG supplies pressurized nitrogen gas from the CAD tank as a
backup to normal instrument air. The safety grade pneumatic
supply is isolated from the nonsafety grade portion of the air
supply by spring-loaded, soft-seat, check valves designed for zero
leakage.
The safety grade supply to the CADS valves is described in
paragraph 5.2.3.9 of subsection 5.2.
A separate air supply system is provided to selected radwaste
equipment; the system contains two air compressors, associated
controls, a receiver, and separate piping to connect the air
supply to the equipment. This system eliminates the potential of
service air system contamination in other areas of the plant due
to backflow from radwaste equipment.
The ADS accumulators, which provide the short-term, backup, safety
grade supply, and their long-term, safety grade, pneumatic supply
are described in paragraph 4.4.5 of subsection 4.4.
An MSIV accumulator is located close to each isolation valve to
provide pneumatic pressure for valve closing in the event of
failure of the normal, non-safety grade, air supply system. The
accumulator volumes are designed for inboard and outboard
isolation valves when the normal pneumatic supply to the
accumulator has failed. The supply line to the accumulator is
large enough to make up pressure to the accumulator at a rate
faster than the rate the valve operation bleeds pressure from the
accumulator during valve opening and closing. The air supply
lines are provided with check valves to assure the integrity of
the accumulator air supplies.
CHAPTER 10 10.17-6 REV. 25, APRIL 2015 In order to eliminate the introduction of compressed air into the containment and to minimize the need for venting and discharge of
the primary containment gases to the environment, an instrument
nitrogen system is provided for pneumatic service to ensure the
oxygen concentration is maintained less than 5 percent inside the
drywell (Drawing M-333, Sheets 1 and 2).
Essentially, this system takes suction from the containment
nitrogen atmosphere and discharge to a receiver which will be the
source of supply for the required pneumatic services inside the
drywell. In this manner, no air will be added to the containment
atmosphere, but rather the containment nitrogen atmosphere will be
recycled, with any losses of nitrogen made up by the normal
inerting system.
The instrument nitrogen system lines are seismic Class I from the
containment penetrations to the second isolation valve, and have
automatic isolation valves which function as part of the primary
containment and reactor vessel isolation control system when
required.
Pneumatically operated devices located within the primary
containment are normally operated by the instrument nitrogen
system. A cross connection is provided between the instrument air
system and the instrument nitrogen system to service the
components in the primary containment should the instrument
nitrogen system be inoperable. Additionally, vital components, such as the main steam isolation valves and main steam relief
valves, are provided with accumulators for reliable operation
without compressor operation.
The emergency switchgear and battery room dampers are supplied
with safety grade pneumatic supplies to maintain the dampers open.
The source of the pneumatic supply is nitrogen cylinders. The
safety grade pneumatic supply is isolated from the non-safety
grade potion of the air supply by spring-loaded, soft seat, check
valves. The safety grade supply to the E.S.G.B.R. dampers is
described in paragraph 10.14.3.1.
10.17.6 Safety Evaluation
The safety grade, pneumatic supplies to the essential valves of
the CADS are provided so that the system can supply post-LOCA
nitrogen addition to the containment and can facilitate controlled
venting of containment. The safety evaluation for the CADS is
contained in paragraph 5.2.3.9.
The safety grade, pneumatic supply to the containment purge and
vent valves is provided to maintain pressure in the inflatable
CHAPTER 10 10.17-7 REV. 25, APRIL 2015 valve seats to assure leaktight conditions. The safety evaluation is contained in paragraphs 5.2.3.7 and 5.2.4.
The safety grade, pneumatic supply to each of the suppression
chamber-to-secondary containment vacuum breaker butterfly valve is
provided to maintain valve closure. It also provides the
pneumatic supply to the inflatable seal utilized in one of the
valves on each unit. The safety evaluation is contained in
paragraphs 5.2.3.6 and 5.2.4.
Each ADS valve is provided with a short-term, backup, safety
grade, pneumatic supply by means of its associated accumulator and
check valve to provide sufficient capacity to cycle the valve open
five times at atmosphere pressure, twice at 70% of containment
design pressure, or once at containment design pressure, all
within a 4-hour period.
A long-term, backup, safety grade, pneumatic supply has been
provided to the ADS valve accumulators inside the primary
containment to assure ADS valve operability for a period of 100
days following an accident.
A split ring header is installed inside the containment with three
ADS valves connected to one section of the split header and the
remaining two ADS valves connected to the other section of the
split header. The safety grade, pneumatic pressure is a series of
nitrogen cylinders located within the reactor building with a
connection provided outside the reactor building for the
installation of additional bottles, as required. Also, a long-
term, backup, safety grade pneumatic nitrogen supply has been
provided to SRVs RV2(3)-02-071E, H&J. This pneumatic supply is
provided to enable remote operation of the above valves for a
period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a Design Basis fire in Fire Areas
6S (Unit 2) and 13S (Unit 3) which has been postulated to render
the ADS valves available for only short-term operation. The
source of the pneumatic nitrogen supply is the Safety Grade
Instrument Gas (SGIG) system. The SGIG system is tied into the
6,000 gallon liquid nitrogen tank which supplies the Containment
Atmospheric Dilution (CAD) system. The CAD tank is located
outside of Fire Areas 6S (Unit 2) and 13S (Unit 3).
Spare primary containment penetrations, two for each unit, have
been modified to provide a permanent means of connection to each
section of the safety grade, pneumatic supply headers within each
drywell. Containment isolation has been provided for the
instrument gas supply lines into containment by the use of check
valves and other automatic valves outside the primary containment.
The outer, automatic valves are manually controlled from the
control room and automatically close on low differential pressure
CHAPTER 10 10.17-8 REV. 25, APRIL 2015 between pneumatic supply pressure and containment pressure or if gas flow becomes excessively high.
The MSIV accumulators are provided to supply a safety grade, backup, pneumatic supply to close the MSIV's by pneumatic pressure
following the loss of normal non-safety grade pneumatic supply.
The safety function of the accumulators is assured by a safety
grade check valve which isolates the accumulators and allows them
to perform their safety function.
The safety grade, pneumatic supply to the emergency switchgear and
battery room dampers is provided to assure continued operation of
the ventilation system.
10.17.7 Inspection and Testing
The instrument air and service air systems are proved operable by
their use during normal plant operation. Portions of the system
normally closed to flow can be tested to ensure operability and
integrity of the system.
The post-LOCA CADS is functionally tested in accordance with plant procedures. The atmospheric analyzing system is functionally tested in accordance with the Technical Requirements Manual.
Inspection and testing of the ADS is discussed in paragraph 4.4.8.
CHAPTER 10 10.18-1 REV. 21, APRIL 2007 10.18 DOMESTIC AND SANITARY WATER SYSTEM 10.18.1 Power Generation Objective
The power generation objective of the domestic and sanitary water
system is to provide the potable water supplies and sewage
treatment necessary for normal plant operations and shutdown
periods.
10.18.2 Power Generation Design Basis
- 1. Domestic water is chlorinated.
- 2. Sanitary system water (sewage) is treated prior to release.
10.18.3 Description
Domestic water is supplied from the clarified water system, discussed in subsection 10.16, "Makeup Water Treatment System."
The domestic water system consists of a 5,000-gal domestic water
storage tank, two domestic water pumps, a domestic water hydro-
pneumatic tank, hypo-chlorinator, and distribution piping (Drawing
M-317). Clarified and filtered water is chlorinated and stored in
the domestic water storage tank, then pumped to the hydro-
pneumatic tank, where it is pressurized for system distribution.
Water heating units are provided for domestic showers.
An onsite sewage treatment plant is provided to treat the normal
sewage prior to release. The facility has the capacity to handle
Units 2 and 3 and to handle the variable loading at the plant due
to population fluctuations between outage and non-outage periods.
The sewage treatment system is designed to provide an effluent
that meets the regulations of the Commonwealth of Pennsylvania.
10.18.4 Inspection and Testing
The domestic and sanitary water system is proved operable by its
use during normal plant operation. Portions of the system
normally closed to flow can be tested to ensure operability and
integrity of the system.
CHAPTER 10 10.19-1 REV. 25, APRIL 2015 10.19 PLANT EQUIPMENT AND FLOOR DRAINAGE SYSTEM 10.19.1 Power Generation Objective
The power generation objective of the plant equipment and floor
drainage system is to collect and remove waste liquids from their
points of origin to a suitable disposable area.
10.19.2 Power Generation Design Basis
- 1. Liquid waste drains are classified in accordance with radioactive contamination potentials and conductivity
levels and chemical content.
- 2. Potentially radioactive liquid wastes are collected separately from the non-radioactive wastes, in a
controlled and safe manner.
10.19.3 Description
The plant equipment and floor drainage system handles both
radioactive and potentially radioactive wastes. Radioactive
wastes are collected in the building sumps and transferred to the
radwaste building for treatment, sampling, and analysis prior to
disposal or reuse in the plant. Non-radioactive wastes are pumped or drained by gravity into the sewer system, storm drain system, or intake bay, and released.
10.19.3.1 Radioactive Equipment Drainage System
- 1. Reactor Building Drains
Reactor containment systems' equipment wastes are collected in two separate systems. The drywell
equipment drain sump system collects equipment drains
located in the primary containment. The reactor
building equipment drain sump system handles drainage
from equipment drains located in the secondary
containment. Equipment drains are collected in closed
piping and discharged to the equipment drain sump.
Pumps transfer these wastes to the radwaste system.
Containment is provided in transferring waste from the
sumps to the radwaste system by maintaining a minimum
water level in the sump, which seals the pump suction
lines. To prevent blowout of water seals, the drywell
equipment drain discharge line penetrating the primary
containment has two isolation valves which close upon a
high drywell pressure signal.
CHAPTER 10 10.19-2 REV. 25, APRIL 2015
- 2. Turbine Building
The turbine building radioactive equipment drainage system is collected in sumps located below the basement
level. Sump pumps transfer the liquid to the radwaste
system.
- 3. Radwaste Building
The radwaste building is provided with an equipment drain sump. Sump pumps transfer the liquid to the
radwaste system. Radioactive drainage within the
radwaste onsite storage facility is discussed in section
9.3.3.2.
- 4. Recombiner Building
The recombiner building is provided with an equipment drain sump. Sump pumps transfer the liquid to the
radwaste system.
10.19.3.2 Floor Drainage System
In general, floor drains from the primary containment, reactor
building, turbine building, recombiner building, and radwaste
building are collected in sumps located in the basement or lowest
level of the building. Sump pumps transfer the waste from the
sumps to the radwaste system.
10.19.3.3 Non-Radioactive Drainage System
Roof drains from the reactor buildings, turbine building, radwaste
building, and other buildings, and some floor drains in the
turbine building service areas are collected and typically discharged by gravity to the storm drain system. Floor and equipment drains associated with the Unit 2 and Unit 3 critical pump portions of the Circulating Water Pump Structure are collected in a common sump and pumped to the intake bay.
Rainwater from the roof drain is normally routed directly to the intake bay. Refer to UFSAR Section 12.2.10 for details regarding how this drainage system provides protection of critical equipment during external flood conditions.
10.19.3.4 Miscellaneous Drainage System
Non-radioactive chemical liquid wastes are collected, neutralized, and routed to the settling basin prior to release to the pond.
Oil drains and oil-contaminated liquid drains are collected in a
separate oil collection tank for offsite disposal.
CHAPTER 10 10.19-3 REV. 25, APRIL 2015 10.19.3.5 Torus Dewatering Facilities
A 1.2 million-gal capacity storage tank and associated valving and
piping is provided for dewatering the torus to support containment
suppression chamber inspections and/or modifications. The tank
provides temporary storage for the entire volume of the torus.
The Torus De-Watering System (TDWS) is only connected to the torus
for dewatering/polishing in plant modes 4 and 5. Removable spool
pieces with isolation valves are utilized to allow the TDWS pump
suction to be connected to the torus. The spool piece assemblies (with closed isolation valves) are fully qualified to allow the
torus to maintain a reliable source of water for ECCS operation
while operations with the potential to drain the vessel are in
progress.
Torus water transfer is via the torus dewatering pumps (Units 2
and 3) and sludge pump (Unit 3 only). The torus water may be sent
to either the Condensate Storage Tank or the Torus Dewatering Tank (TDT). Transfer is generally routed through a condensate filter
demineralizer to improve water quality prior to storage in the
TDT. Additionally, the TDWS may be used to filter and polish the
torus water by operating the system in closed loop from the torus
through a condensate filter demineralizer, and back to the torus.
10.19.4 Inspection and Testing
The plant equipment and floor drainage system is proved operable
by use during normal plant operation. Portions of the system
normally closed to flow can be tested to ensure operability and
integrity of the system.
CHAPTER 10 10.20-1 REV. 21, APRIL 2007 10.20 PROCESS SAMPLING SYSTEM 10.20.1 Power Generation Objective
The power generation objective of the process sampling system is
to monitor the operation of equipment and systems, and to provide
information for making operational decisions.
10.20.2 Power Generation Design Basis
- 1. The process sampling system is designed to obtain representative samples which can be used in the
radiochemical laboratory.
- 2. The process sampling system minimizes contamination and radiation effects at the sampling stations.
- 3. The process sampling system is designed to reduce decay and sample line plateout.
10.20.3 Description
Samples are taken at locations throughout the plant from the
process and auxiliary systems (Table 10.20.1). Sample points are
grouped as much as possible at normally accessible locations, and
drains are provided at these locations to limit the risk of
contamination. Lines are sized to ensure purging and sufficient
velocities to obtain representative samples. The samples are
analyzed and the resulting information is used to evaluate the
condition of the plant.
10.20.4 Inspection and Testing
The process sampling system is proved operable by its use during
normal plant operation. Grab samples are taken to verify the
proper operation of the continuous samplers. Portions of the
system normally closed to flow can be tested to ensure the
operability and integrity of the system.
CHAPTER 10 10.20-2 REV. 23, APRIL 2011 TABLE 10.20.1 PROCESS SAMPLING SYSTEM
Description Locations Purpose
Nuclear Steam Supply System
Reactor water Recirculation pump discharge Reactor water quality, crack arrest verification
Main steam Main steam line Carryover, moisture RHRS RHR HX outlet HX leakage
Cleanup Demineralizer
Filter-demineralizer Influent header Reactor water quality Filter-demineralizer Powdex unit effluent Filter condition Filter-demineralizer Precoat recycle line Element leakage Regenerative HX Return to reactor HX leakage Non-regenerative HX Cooling water outlet pipe HX leakage
Condensate System
Condensate Hotwell trays Tube leakage Condensate Condensate pump discharge Tube leakage, water quality Condensate demineralizer Influent header Condensate quality Condensate demineralizer Powdex unit effluent Filter condition Condensate demineralizer Effluent header Treated condensate quality Consensate demineralizer Precoat recycle line Element leakage
Feedwater Systems
Heater drains Heater No. 3 outlet Water quality Feedwater Heater No. 5 outlet Water quality Feedwater Reactor inlet Water quality Closed Cooling Water Systems
Turbine building cooling water Pump discharge Inhibitor concentration Reactor building cooling Pump discharge Inhibitor concentration
Circulating and Service Water Systems
Service water Pump discharge header Determine background Circulating water Condenser outlet Chlorine residual Circulating water Discharge canal Activity release
CHAPTER 10 10.20-3 REV. 23, APRIL 2011 TABLE 10.20.1 (Continued)
Description Locations Purpose
Circulating and Service Water Systems
Service water Pump discharge header Determine background Circulating water Condenser outlet Chlorine residual Circulating water Discharge canal Activity release
Liquid Radwaste System
Laundry drain tank Pump discharge Process data Laundry drain filter Filter effluent Water quality Floor drain collector tank Pump discharge Process data Floor drain filter Filter effluent Process data Floor drain demineralizer Demineralizer effluent Process data Floor drain sample tank Pump discharge Discharge suitability Waste collector tank Pump discharge Process data Waste surge tank Pump discharge Process data Waste collector filter Filter effluent Process data Waste demineralizer Demineralizer effluent Process data Waste sample tank Pump discharge Discharge suitability R/W fuel pool F/D precoat Precoat recycle line Element testing Fuel pool HX HX outlet Fuel pool quality, HX leakage Fuel pool filter demineralizer Powdex unit effluent Filter/demineralizer condition Chemical waste tank Pump discharge Process data Condensate phase separator decant Pump discharge Process data Cleanup phase separator decant Pump discharge Process data Centrifuge liquid effluent Liquid discharge pipe Process data
Makeup Water Treatment Systems
Raw water inlet GE M52 skid inlet Process data Filtered water outlet GE M52 outlet Process data Domestic water Pump discharge Chlorine residual Carbon filter Filter effluent Process data
CHAPTER 10 10.20-4 REV. 23, APRIL 2011 TABLE 10.20.1 (Continued)
Description Locations Purpose
Makeup Water Treatment Systems (Continued)
Cation unit Effluent Process data Anion unit Effluent Process data Mixed bed unit Effluent Water quality Dilute acid Header Process data Dilute caustic Header Process data Neutralizer tank Outlet pipe Process data Demineralized water storage tank Pump discharge Water quality Condensate transfer system Pump discharge Water quality Refueling water transfer system Pump discharge Water quality
Plant Off-Gas Systems
Air ejector discharge Header Activity; H 2 , 0 2 , and air in-leakage Off-gas stack sample Main stack Noble gas monitoring and particulate
and iodine samples to determine release rates Recombiner area monitoring Fan discharge from individual Activity equipment rooms, hydrogen analyzers, instrument racks, equipment sumps, and cooling water surge tank. Identifica-tion of specific source of leakage is obtainable.
Building ventilation exhaust Building ventilation stack Noble gas monitoring and particulate
and iodine samples to determine release rates Control room, radwaste, Fan discharge Activity recombiner ventilation
CHAPTER 10 10.21-1 REV. 21, APRIL 2007 10.21 COMMUNICATIONS SYSTEMS 10.21.1 Power Generation Objective
The power generation objective of internal and external
communications is to establish a combined system of loudspeakers
and telephones to provide convenient, effective operational
communications between various plant buildings and locations.
10.21.2 Power Generation Design Basis
- 1. Voice communication to points outside the station is provided by a dial telephone system.
- 2. Voice communication between various plant buildings and locations is provided.
10.21.3 Description
The following means of communication are provided in the plant:
- 1. A dial phone system with a self-contained power supply is provided for communicating with points outside the
station.
- 2. An intraplant communication system consisting of handsets and loudspeakers is provided for paging and
communications in all appropriate areas. The intraplant
system employs equipment that operates from the AC
instrument bus. Loudspeakers powered by individual
amplifiers are located throughout the station, with
muting facilities provided where required. Paging plus
two-party line channels are provided for simultaneous
operation.
- 3. A separate intraplant telephone system allows uninterrupted private communication for maintenance and
general use between the main control room, reactor
refueling area, and other selected plant areas. The
system is designed to ensure that no interference in
vital communication is caused by other plant
communication systems.
- 4. An evacuation alarm system is located in strategic points throughout the plant to warn personnel of
emergency conditions. Additional speakers are located, especially in high noise areas, to provide plant
evacuation signal.
CHAPTER 10 10.21-2 REV. 21, APRIL 2007 5. A distributed antenna system was installed throughout the plant to provide the capability of using low power walkie talkies as a supplementary means of communicating
within the plant. Because these walkie talkies are low
power, they will not interfere with plant
instrumentation.
- 6. A dedicated communications system, using the distributed antenna system is installed to allow plant operators to
communicate between the main control room and the Unit 2
and Unit 3 Alternative Control Stations. This system is
used for coordinating testing of the Alternative Control
Stations and in an Emergency, for a Safe Shutdown from
the Alternative Control Stations. See the Fire
Protection Program (FPP) for details.
10.21.4 Inspection and Testing
The communication systems are proved operable by use during normal
plant operation. Loss of offsite power operation can be tested to
ensure operability and integrity of the systems.
CHAPTER 10 10.22-1 REV. 21, APRIL 2007 10.22 STATION LIGHTING SYSTEM 10.22.1 Power Generation Objective
The power generation objective of the station lighting system is
to provide adequate normal and emergency indoor station lighting.
Power is supplied from a normal AC source, standby AC system, and
from the station battery system.
10.22.2 Power Generation Design Basis
- 1. Lighting intensities approximate levels recommended by the Illuminating Engineering Society.
- 2. Mercury vapor fixtures and mercury switches are not used inside the primary containment or directly above the
reactor on the refueling floor.
- 3. The main control room has a fluorescent lighted, glare-free, luminous ceiling to reduce glare and shadows at
the control boards.
- 4. Emergency lighting is provided in the control room, diesel generator rooms, emergency switchgear area, and
other points where lighting may be required under
abnormal conditions.
10.22.3 Description
The station lighting system is supplied from the station auxiliary
power system described in Section 8.0, "Electrical Power Systems."
Normal power is supplied from the unit auxiliary or the startup
transformers. The power supply for lighting areas required during
shutdown or abnormal conditions is automatically transferred to
the standby diesel generator system if the normal power supply is
lost.
The lighting distribution system has separate, dry-type lighting
transformers and circuit breaker type panel boards.
A separate emergency DC lighting system, energized from the
station batteries, is provided for safe exit lighting if all AC
power sources are lost.
Separate 8-hour, battery-powered lighting is provided to support
safe shutdown operations remote from the control room (Reference
Table A-4 of the PBAPS Fire Protection Program). Task lighting is
provided at the sites of the operations. The routes used to
access and egress the sites are also illuminated. See Fire
Protection Program (FPP) for details.
CHAPTER 10 10.22-2 REV. 21, APRIL 2007 10.22.4 Inspection and Testing
The station lighting system is proven operable during normal plant
operation. Loss of offsite power operation can be tested to
ensure operability and integrity of the system.
CHAPTER 10 10.23-1 REV. 23, APRIL 2011 10.23 PLANT AUXILIARY BOILERS 10.23.1 Power Generation Objective
The power generation objective of the plant auxiliary boiler
system is to supply necessary steam for plant uses.
10.23.2 Power Generation Design Basis
- 1. The plant auxiliary steam system operates independently from the nuclear steam system. Wherever the two systems
have interfaces, a positive means of separation is
provided.
- 2. The auxiliary steam system is designed to provide operational flexibility to accommodate the seasonal
steam demand.
10.23.3 Description
The plant auxiliary boilers are common to both Units 2 and 3. The
auxiliary steam system consists of two 40,000-lb/hr, oil-fired, water-tube package boilers and associated equipment and
instrumentation (Drawing M-324, Sheets 1, 1A, 2, 2A, 3, 3A, and 4). The boilers have a design and maximum allowable working pressure of 275 psig. The boilers are designed, fabricated, tested, and stamped in accordance with the ASME Boiler and
Pressure Vessel Code,Section I, and the rules and regulations of
the Commonwealth of Pennsylvania.
Process steam (200 psig system) and plant heating steam (50 psig
system) are distributed from the boiler steam outlet header.
Except for cold startup of a boiler or a startup subsequent to a
safety trip, the boiler control system is designed for unattended
operation.
The 200 psig process steam header provides the following process
steam during the plant startup until the nuclear steam pressure is
adequate:
- 1. Condensate deaeration in the hotwell during startup.
- 2. Turbine shaft seal steam during startup.
The condensate from this process is not returned to the deaerator
but is drained to the main condenser.
CHAPTER 10 10.23-2 REV. 23, APRIL 2011 10.23.4 Inspection and Testing
The auxiliary steam system is proven operable by its use during
normal plant operation. Portions of the system normally closed to
flow can be tested to ensure operability and integrity of the
system.
CHAPTER 10 10.24-1 REV. 25, APRIL 2015 10.24 EMERGENCY HEAT SINK 10.24.1 Power Generation Objective
The power generation objective of the emergency heat sink is to
provide an onsite heat removal capability so that the reactors of
Units 2 and 3 can be shut down in the event of the unavailability
of the normal heat sink.
10.24.2 Power Generation Design Basis
- 1. The emergency heat sink has a sufficient capacity for removing the sensible and decay heat from the reactors'
primary systems and auxiliary systems so that the
reactors can be shut down in the event of the
unavailability of the normal heat sink.
- 2. The emergency heat sink has the heat removal capacity to supply a source of cooling water to the emergency
service water system and the high pressure service water
system when required.
- 3. The emergency heat sink provides sufficient water storage capacity to permit emergency cooling tower operation until a makeup water supply can be
established.
- 4. The emergency heat sink can operate during a loss of offsite power and can withstand a seismic event.
10.24.3 Description
The emergency heat sink facility consists of a fireproof, multicell, mechanical, induced-draft cooling tower, constructed as
a seismic Class I structure, with an integral onsite 3.55 million-
gal water storage reservoir (Drawing C-2). The facility operates in conjunction with the high pressure service water pumps (subsection 10.7), the emergency service water pumps (subsection
10.9) at the pump structure, and the emergency service water
booster pumps (Drawing M-330). The equipment, valves, and piping
in the emergency heat sink system are designed in accordance with
seismic Class I criteria. Power requirements are supplied from
the standby power supply. Equipment data is shown in Table
10.24.1.
The high pressure service water pumps take suction from the pump
bays to supply water to the RHR heat exchangers. Sufficient head
is available to pump the return water directly to the emergency
cooling tower. Water supplied from the onsite reservoir flows by
gravity to the pump structure in two full-capacity lines, this
CHAPTER 10 10.24-2 REV. 25, APRIL 2015 flow being regulated by two motor operated flow control valves in series in each line. The emergency service water pumps, also
located in the pump structure, take suction from the pump bays and
supply water to standby diesel generator coolers and the CSCS's
pump room air coolers. The return water from the various coolers
is boosted in pressure by one of two full-capacity emergency
service water booster pumps and delivered to the emergency cooling
tower. After an extended period of time, the fuel pool cooling
system may be served by the emergency service water system by
adding cross connections.
Sluice gates in the pump structure isolate the high pressure
service water and emergency service water pump bays from Conowingo
Pond. These gates are manually closed prior to utilizing the
onsite reservoir.
The cooling tower is a mechanical, induced-draft type consisting
of three cells each capable of handling the heat transfer duty of
one RHR heat exchanger (one HPSW pump) plus the plant auxiliary
cooling requirement (one ESW pump). The tower heat transfer duty
is based on the heat duty occurring when the RHRS operating mode
is switched from the containment (torus) cooling mode to the
shutdown cooling mode at the time the reactor water temperature reaches 300 F. The emergency heat sink system supplies cooling water to two high
pressure service water pumps and one emergency service water pump
continuously until all the nuclear fuel can be shipped offsite.
Makeup water is supplied from an offsite source.
The installed capacity of 3.55 million gal of water stored in the
emergency cooling reservoir is adequate for 1 week of operation
without makeup. Makeup of water to the system will be initiated
as expeditiously as possible after shutdown of the reactors and
will be continued for an indefinite period as required.
After a two-reactor shutdown, assuming the turbine condensers are
not available as heat sinks, it is estimated that continued
operation of the RHRS in the shutdown cooling mode can cool the reactors to 212 F in approximately 12 hr and to 125 F in about 3 weeks. Based on controlled cooling tower operation at the rated flow condition, the total water consumed at the end of 7 days is
approximately 2.9 x 10 6 gal and a makeup rate of 250 gpm will be required after the first week. The turbine-condensers, if available, will be used as heat sinks for the removal of reactor
heat as long as effective. The minimum Conowingo Pond level required for effective operation of the main condenser circulating water pumps is approximately 93.8 feet (C.D.). Note: Water level
CHAPTER 10 10.24-3 REV. 25, APRIL 2015 in the pump bays will be lower due to level differential across the traveling screens.
The offsite makeup water supply to the plant will be made by truck
trailers, or temporary hose lines, and portable pumping equipment
will be available and used to withdraw water from waterholes in
the Susquehanna River or Rock Run Creek. Water transportation
into the plant will be initiated as soon as practicable after the
loss of the dam accident. The 250 gpm makeup rate required 1 week
after reactor shutdown represents about three 5,000 gal water
trucks per hour.
The feasibility of transporting a large quantity of water was
demonstrated during the 1965 drought period in York, Pennsylvania, when several million gallons were delivered by truck daily to the
potable water system of that city. Fuel oil is also delivered (two or three trucks per hour) to several generating stations on
the licensee's system.
10.24.4 Inspection and Testing
To assure that the emergency heat sink will function properly, the
tower and reservoir are inspected for integrity and reservoir
level. The high pressure service water, emergency service water, and emergency cooling water pumps are tested in conjunction with
their systems' testing. Portions of the system normally closed to
flow can be tested to ensure operability and integrity of the
system.
Flow measurement devices are provided in the emergency cooling
water system to facilitate testing of the emergency cooling water
pump and the emergency service water booster pumps in accordance
with AMSE Code requirements.
The timer used to sequence the emergency cooling water pump during
a LOCA will be tested in accordance with surveillance test
procedures. The test will verify the setting, operability, and
functional performance of the relay, and will provide assurance
that the automatic loading sequence will be maintained and will
perform as required.
CHAPTER 10 10.24-4 REV. 21, APRIL 2007 TABLE 10.24.1 EMERGENCY HEAT SINK EQUIPMENT DATA
Design Performance and Type
Type Induced Draft/
Counter Flow Design Wet Bulb Temperature 78.0 F Number of Towers/Number of Cells per Tower 1 / 3 Total Heat Load 357 x 10 6 Btu/hr Water Side High Pressure Flow Hot Water Flow 9,000 gpm Hot Water Temperature 160 F Cold Water Temperature 90 F Evaporation Loss at Rated Flow 7%
Low Pressure Flow Hot Water Flow 8,000 gpm Hot Water Temperature 100 F Cold Water Temperature 90 F Evaporation Loss at Rated Flow 1%
Total Water Concentration/Cell 3.69 gpm/ft 2 Water Load on Tower Base Area 187 gal/ft 2* Hot Water Overload Capability 50% (Approx.) Cold Water Temperature at Overload Flow 96 F
- Drift Water Loss at Rated Flow
<0.05% Retention Time through Tower 7.0 sec Air Flow Stack Height 20.0 ft Air Flow 8.01 x 10 6 lb/hr Draft Loss Inches 0.633 in H 2 0 Total Fan Power Demand, Bhp at Motor Coupling 185.0/cell
- Both systems at 50 percent overload.
CHAPTER 10 10.24-5 REV. 21, APRIL 2007 TABLE 10.24.1 (Continued)
Mechanical Equipment (per Cell)
Fans Number 1 RPM 116.7 rpm Blade Material Reinforced Fiberglass Epoxy
Tower and Cell Structure
Tower Height 57 ft 4 in Air Intake Height 10 ft 6 in Cell Dimension 48.0 ft x 48.0 ft Stack Height 20.0 ft
Emergency Cooling Water Pumps For specifications see Table 10.9.1
Emergency Service Water Booster Pumps
Quantity 2 (common to both units)
Pump Design Type Horizontal split Flow/Head 8,000 gpm/100 feet Bhp at Rating 230 Speed 1,170 rpm Number of Stages 1
Material Bowl Bronze Shaft AISI 303 Shaft Sleeve AISI 440c Wear Ring Bronze Impeller/Liner ASTM B143 Bronze
Motor Design Type Horizontal Induc-tion Type Horsepower 250 hp Voltage/Phase/Frequency 4,000 V/3 Phase/
60 Hz