NG-17-0111, Duane Arnold Energy Center, Revision 24 to Updated Final Safety Analysis Report, Chapter 9, Auxiliary Systems
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UFSAR/DAEC-1 T9.2-1 Revision 23 - 5/15 Table 9.2-1 EMERGENCY SERVICE WATER FLOW REQUIREMENTS a
Equipment River Water Temperature Flow for 3 ft/sec tube
velocity 95°F 90°F 85°F 80°F Diesel-generator 528 466 466 372 625 RHR and Core Spray room cooler 83 50 50 50 75* RCIC room cooler 17 12 12 8 20 HPCI room cooler 16 16 16 16 35 Control Building Chiller 199 132 100 82 250 RHR pump seal coolers (two) b 11 (5.5 ea) 8.8 (4.4 ea) 8 (4.0 ea) 6.6 (3.3 ea) N/A Core spray pump motor cooler c 3 3 3 3 1.2 d Heating and ventilation instrument air compressor 1 1 1 1 1 RHR service water pump motor coolers e 10 10 10 10 2.8 d Total flow 868 698.8 666 548.6 1010.0
- The required flow is greater than the 3 ft/sec flow for river temperatures over 85°F.
a Flow rates are given in gallons per minute for various river water temperatures.
b Flow rate required to maintain RHR pump seal inlet temperature 150ûF under DBA-LOCA Conditions (CAL-M10-010).
c Core Spray Pump - Two pumps arrangement with only one pump operating at a time. Each pump motor cooling coil requires a minimum of 3 gpm (ref. APED-E21-2723-007). Flow for 3 ft/sec is also for two pumps.
d The required flow is greater than 3 ft/sec for all river temperatures.
e RHRSW is a four pump arrangement with only two pumps operating at a time. Each motor cooling coil require a minimum flow of 5 gpm (ref. E025-019).
2011-019 2011-019 2008-022 2008-022 2014-004
UFSAR/DAEC-1 9.3-1 Revision 22 - 5/13 9.3 PROCESS AUXILIARIES 9.3.1 COMPRESSED AIR SYSTEMS
9.3.1.1 Design Bases
9.3.1.1.1 Power Generation Objective
The power generation objective of the compressed air systems is to provide a continuous supply of dry oil-free compressed air to plant instruments and general plant
services as required.
9.3.1.1.2 Power Generation Design Basis
The compressed air systems are designed to supply air at the required pressure and temperature. Instrument air is typically dried to a dewpoint of -40°F or lower at 100 psig.
9.3.1.2 Instrument and Service Air System
9.3.1.2.1 Description
Each compressor discharges through an integral aftercooler into a common discharge header and then into either or both of two air receivers. Instrument air then passes through an air dryer and a filter before entering the instrument air header that feeds the instrument components. There is a standby air dryer and filter which can be used for maintenance purposes. Service air is supplied directly from the receivers to the service air components. Instrument air is dried to a dewpoint of
-40°F at 100 psig. Refer to Figure 9.3-1. The service air system is automatically isolated from the instrument air system on low header pressure. Refer to Section 9.3.1.2.4.
The systems are so arranged that any one compressor can supply both instrument and service air requirements. The compressors are sized so that any one compressor has adequate capacity to supply the normal plant instrument and service air requirements, and any two are capable of satisfying peak air demand. The third compressor is provided as a spare. A fourth oil-free compressor located in the turbine building basement is used as a standby compressor.
Provisions within the air systems are made to mitigate the effects of system piping breaks. Should loss of air system header pressure occur, successive header isolations will result. Refer to Section 9.3.1.2.4. Also air accumulators or high pressure storage bottles have been provided locally for critical components of the Condensate and Feedwater system. This backup air system will allow the feedwater system to control reactor water level for a brief period after a loss of instrument air.
UFSAR/DAEC-1 9.3-2 Revision 22 - 5/13 All gaseous pneumatic systems in the primary containment use nitrogen as the working fluid in order to avoid oxygen (air) leakage into containment. Refer to Figure 6.2-44.. A dedicated nitrogen compressor located outside containment takes a suction from the inerted containment atmosphere and charges a nitrogen receiver located outside containment. The nitrogen receiver is then used to operate all pneumatic-operated valves within containment including, among others, the torus-drywell vacuum breakers, reactor pressure vessel head vent isolation valves, main steam safety/relief valves (SRVs) and main steam isolation valves (MSIVs). Safety-related accumulators are provided for the
SRVs and MSIVs. Refer to Sections 6.3.2.2, 7.3.1.1.1.7 and Figure 5.1-1 Sheet 1.
A safety-related air system is provided as a backup to the normal instrument air system for several critical safety-related components and systems. The safety-related air system is comprised of two separate Seismic Class I trains, each of which consists of a compressor, air receiver, and associated distribution piping, valves and service
connections. The distribution piping of the two safety-related air trains are not normally cross tied together and the compressor motors are supplied from different essential busses
such that no single failure will render both trains inoperable. The safety-related compressors are normally cooled with well water, but can be cooled with Emergency Service Water to ensure adequate post-accident cooling. Below is a list of components and systems whose operability is supported by the safety-related air system:
Component/System Supported Function Control Building Chiller System ventilation flow path and temperature control
Standby Filter Unit System flow control and the CBC System provides the ventilation flow path
Standby Gas Treatment System flow control and filter cooler bypass damper opening
Drywell Cooling Water valve closure Containment Isolation Valves
Containment Purge and Vent leak tightness (by pressurizing the Isolation Valves T-ring seal) when closed
Reactor Building to Torus Vacuum valve closure and leak tightness (by Breaker Butterfly Valves pressurizing the T-ring seal) when closed
9.3.1.2.2 Safety Evaluation UFSAR/DAEC-1 9.3-3 Revision 22 - 5/13 Instrument and service air systems are not safety-related. Any one of the four non-safety air compressors can satisfy the normal plant instrument and service air requirements; thus the four air compressors ensure a backup capability even in the unlikely event of a failure of three compressors.
The service air systems does not supply any safety-related equipment and total failure of the service air system, therefore, need not be considered in the Safety
Evaluation.
Although the normal instrument air system supplies some safety-related equipment, total failure of the system will not adversely affect the operation of the plant for the following reasons:
- 1. Testing was performed on the instrument air system as part of DAEC's response to NRC Generic Letter 88-14. See References 9 and 10. Results of the testing indicate, "The testing program developed to verify that air-operated safety-related components will perform their safety-related function upon loss of their normal air supply was successfully completed.
All safety-related components performed as designed."
- 2. The existence of the safety-related air system which can supply air to support the operation of safety-related equipment if the instrument air system becomes unavailable. Transition and separation from the instrument air system to the safety-related air system occurs automatically upon low safety-related air receiver pressure.
- 3. The pneumatically actuated components inside the drywell have a reliable source of compressed nitrogen to provide the required motive force, and therefore, do not rely upon either the safety-related air system nor the normal instrument air system.
Therefore, total failure of the normal instrument air system will not adversely impact the safe operation of the plant due to: the redundancies built into the design of the safety-related air-system, the redundancies built into the systems whose components are supplied safety-related air and the design of the components that are supplied air exclusively from the normal instrument air system that revert to their conservative
position on loss of air.
9.3.1.2.3 Testing and Inspection Requirements
The instrument and service air systems operate continuously and are observed and maintained during normal operations. An instrument air system blow down is performed periodically to remove any possible particulates from the system. Also, an instrument air quality test is performed periodically at various instrument air headers downstream of the air dryers. This test is performed to verify that the air quality (dew point, particulate
and UFSAR/DAEC-1 9.3-4 Revision 22 - 5/13 oil content) is consistent with manufacturer recommendations. Revisions to test frequencies are based on test results.
Routine testing is performed on the safety-related air compressors to ensure that they automatically start on low pressure in the associated safety-related air receiver.
Additionally, routine testing is performed on the safety-related air compressors to monitor the performance of the compressors and to monitor the integrity of the safety-
related air distribution piping and tubing. The routine test also verifies that the check valves which isolate the safety-related air system from the normal instrument air system
are working properly.
9.3.1.2.4 Instrumentation Requirements
The three air compressors that are normally in operation are controlled by a group
of three pressure switches, set for "load-no load" control over three overlapping pressure ranges. The lead compressor (as determined by the baseload selector) operates in the highest range. If the system demand exceeds the capacity of one air compressor, a second compressor will load and unload as required until a timed period of unloaded operation is exceeded. The second compressor will then return to standby (motor off) until the next call to start. The compressor starting sequence is controlled by the baseload selector and can be altered manually.
System isolation for the instrument and service air systems is accomplished by
pressure switches that provide the signals to close control valves. Interlocks are provided to automatically isolate the nonessential air lines from the air receivers on the detection
of low pressure in the air receivers. This provides a period for orderly shutdown or
repairs. On a decrease in air header pressure to a specific point or on loss of control power, the service air system will isolate. A further decrease in air header pressure to a
lower setpoint will isolate the nonessential Balance-of-Plant and Turbine Building instrument air headers. The Reactor Building instrument air header does not have an automatic isolation.
The safety-related compressors that are normally in standby are controlled by pressure switches monitoring the safety-related air receivers.
9.3.1.3 Breathing Air System
9.3.1.3.1 Description
The breathing air system is crosstied to the instrument air system and supplies grade 'D' air at a dew point of -40
°F or lower at a pressure of 100 psig. The system also consists of six-man stations located throughout the power block. Each station is designed to deliver 10 cfm to each breathing apparatus hookup, in accordance with NUREG-
0041.
UFSAR/DAEC-1 9.3-5 Revision 22 - 5/13 When necessary, breathing air for personnel use can be obtained from the instrument air mains or service air mains, both of which are oil free. When breathing air is supplied from the service air system, portable air filters are used to ensure the breathing air system air quality meets grade D air requirements as described in the Compressed Gas Association Air Specification G-7-1. The instrument air system meets the grade D requirements and, therefore, the use of portable filters is not necessary.
The breathing air system inside the drywell has been "abandoned in place" since it is supplied by service air and does not meet air quality requirements. This breathing air connection to the drywell has a removable spool piece inside the drywell, a blank flange
which is installed on the air supply line in the drywell and an isolation valve outside the containment. Procedures require the isolation valve to be verified closed and the blank
flange to be verified in place prior to plant startup.
9.3.1.3.2 Safety Evaluation
The breathing air system is not itself a safety-related system. Any interaction of the breathing air system with safety-related systems is kept to a minimum, specifically:
- 1. Pipe was routed such that a line break accident is inconsequential to plant safety.
- 2. Secondary containment was penetrated three times. The maximum possible inleakage from the atmosphere to secondary containment was
analyzed and was found to be acceptable. There was no penetration of primary containment.
9.3.1.3.3 Testing and Inspection Requirements
The breathing air system operates as required and is observed and maintained during normal operations. No special inspection or testing is required.
9.3.2 PROCESS SAMPLING SYSTEM
9.3.2.1 Design Bases
9.3.2.1.1 Power Generation Objectives
The power generation objectives of the process sampling systems are the
following:
- 1. Monitor the operation of plant equipment.
- 2. Provide information for making operational decisions with regard to effectiveness and proper performance.
UFSAR/DAEC-1 9.3-6 Revision 22 - 5/13
9.3.2.1.2 Power Generation Design Bases
The process sampling systems are designed to perform the following:
- 1. Obtain representative samples in forms that can be used in radiochemical laboratory analysis for the indication of changes in the constituents.
- 2. Minimize the contamination and radiation effects at the sampling stations.
- 3. Reduce decay and sample line plateout as much as possible.
9.3.2.2 System Description
Samples are taken from various streams and locations as indicated in Table 9.3-1.
Sample points are grouped at normally accessible locations, with drains to the contaminated waste system provided at these locations to limit the risk of contamination.
Lines are sized to ensure prompt purging and sufficient velocities to obtain representative samples when radioactivity measurements are made. Bottled grab samples are taken to
the laboratory for the appropriate analysis.
A postaccident sampling system has been installed to take postaccident reactor coolant and containment atmosphere samples. See Section 12.3.4 for a discussion of postaccident sampling.
The reactor recirculation system process sample line does have postaccident sample capabilities that could be used as a backup to the postaccident sampling system.
The isolation valves on the sample line have been provided with key-locked bypass switches to override a containment isolation signal to enable sampling with the containment isolated.
9.3.2.3 Safety Evaluation
The process sampling systems are not safety related.
9.3.2.4 Testing and Inspection Requirements
The process sampling systems are operational systems and as such require no
periodic testing to ensure operability.
9.3.2.5 Instrumentation Requirements UFSAR/DAEC-1 9.3-7 Revision 22 - 5/13 The process sampling systems are provided with continuous automatic, monitoring and alarm of undesirable conditions (except for the low-level radwaste processing and storage facility sample tank).
9.3.3 EQUIPMENT AND FLOOR DRAINAGE SYSTEM
9.3.3.1 Design Bases
9.3.3.1.1 Power Generation Objective
The power generation objectives of the equipment and floor drainage systems are to collect and remove all waste liquids from their points of origin and to route them to a suitable disposal area in a controlled and safe manner. Water from radioactive drains is collected for sampling and analysis before processing and disposal as described in Section 11.2. Drain line penetrations through containment barriers are designed to maintain containment during normal operations and design-basis accidents.
9.3.3.1.2 Power Generation Design Basis
Plant equipment and floor drainage systems will operate satisfactorily during normal plant operations and will retain their integrity following postulated accidents.
Nonradioactive drain systems are arranged to ensure that no infiltration of radioactive
wastes will occur.
9.3.3.2 Description
Plant equipment and floor drainage systems handle both radioactive and nonradioactive drains. Radioactive drains may contain potentially radioactive materials.
In general, radioactive drains are drained by gravity to a sump, the contents of which are pumped to the radwaste system for the determination of radioactivity before cleanup, reuse, or discharge. Nonradioactive drains are drained by gravity to a sump, the contents of which are pumped to the storm drain system that is entered at a point outside the building, or to the oil interceptor tank that serves the transformers.
Specific information regarding the number of drains, locations, capacities, and
types is provided in Figures 9.3-2 through 9.3-26.
9.3.3.2.1 Radioactive Equipment and Floor Drainage Systems
Reactor Building Drains
Reactor building radioactive equipment and floor drains are collected in two separate systems. One handles drainage from all equipment and floor drains located in the primary containment, and the other handles drainage from equipment and floor drains located in the secondary containment. The primary containment equipment and
floor UFSAR/DAEC-1 9.3-8 Revision 22 - 5/13 drain system begins with funnel drains at all items of equipment that require draining and floor drains located to facilitate the rapid and efficient removal of liquid waste from the
surface of the floor. Drainage collects in branch lines, and drains by gravity to the drywell equipment and floor drain sumps. Sump pumps transfer wastes from the sumps to the radwaste system.
The secondary containment equipment and floor drainage system begins with funnel drains at all items of equipment and floor drains located to facilitate the rapid and efficient removal of liquid waste from the surface of the floor, collects in branch lines, and drains by gravity to the reactor building equipment and floor drain sumps. Sump pumps transfer wastes from the sumps to the radwaste system.
Any leakage from the spent-fuel pool is channeled into one or more of the eleven 1-in.-diameter liner drain pipes that are provided to monitor leaks, as shown in Figure 9.3-21. Each pipe contains a manual gate valve that is normally closed. An operator will
periodically open each of these valves to check for leaks in the fuel pool liner. These drains are routed to the reactor building floor drain sump via a common trough.
Turbine Building Drains
The turbine building radioactive equipment and floor drainage system begins with funnel drains at all items of equipment that require draining and floor drains located to facilitate the rapid and efficient removal of liquid waste from the surface of the floor.
Drainage collects in branch lines, and drains by gravity into the turbine building equipment and floor drain sumps. Sump pumps transfer wastes from the sumps to the radwaste system.
Radwaste Building Drains
The radwaste building radioactive equipment and floor drainage system begins with funnel drains at all items of equipment and floor drains located to facilitate the rapid and efficient removal of liquid waste from th e surface of the floor. Drainage collects in branch lines, and drains by gravity to the radwaste building equipment and floor drain sumps. Sump pumps transfer wastes from the sumps to the radwaste system.
A conveyor floor drain sump receives drainage from the drum filling and storage area in the radwaste building, and sump pumps transfer wastes from this sump to the radwaste system.
Low-Level Radwaste Processing and Storage Facility (LLRPSF) Drains
The LLRPSF radioactive equipment and floor drains are collected in two separate systems. One system handles drainage from all equipment and floor drains located in the storage section of the LLRPSF, while the other system handles drainage from all equipment and floor drains located in the processing section of the LLRPSF.
UFSAR/DAEC-1 9.3-9 Revision 22 - 5/13 The storage area equipment and floor drainage system begins with funnel drains at all items of equipment that require draining. The floor drains are located to facilitate the rapid and efficient removal of liquid waste from the surface of the floor. Drainage collects in branch lines and drains by gravity to the storage area sump. Sump pumps transfer wastes from the sump to the radwaste system in the radwaste building.
The processing area equipment and floor drainage system begins with funnel drains at all items of equipment that require drainage. The floor drains are located to facilitate the rapid and efficient removal of liquid waste from the surface of the floor.
Drainage collects in branch lines and drains by gravity to either the processing area sump or the hydrolazing/decontamination sump. Sump pumps transfer wastes from the sumps to a sample tank for temporary holdup. The sample tank pump will transfer the liquid from the sample tank to the radwaste system in the radwaste building or to the environment.
Radiochemistry Laboratory Drains
The radiochemistry laboratory equipment and floor drainage system is divided into two categories that drain separately, by gravity, to two separate sumps. Each sump is provided with a duplex pump system that discharges the liquid waste from the sump to a header that drains by gravity to the radwaste system. One sump discharge, containing corrosive waste, is conveyed to the chemical waste tank; the other, containing detergents, is conveyed to the detergent drain tank.
9.3.3.2.2 Nonradioactive Water Drainage Systems
Drainage from building roofs is collected in branch lines, emptied to headers or main drain lines, and discharged to the storm drain system.
A separate drainage system is provided in the turbine building for nonradioactive equipment drains and floor drains. These are collected in branch lines that empty into a main drain line and drain, by gravity, to the nonradioactive wastewater sump. This waste is pumped into the storm drain system, the connection to which is outside the building.
There is a normally closed, manual isolation valve and alternate flow path for the diesel-generator room floor drains to the turbine building normal waste sump.
The manual isolation valves will prevent water backup into the diesel-generator rooms in case of leaking check valves during site flood conditions.
The turbine building pumps are each designed to deliver a flow rate of 50 gpm at 40-ft total discharge head, thus enabling them to remove water from the sump against the head created as a result of the maximum probable flood.
UFSAR/DAEC-1 9.3-10 Revision 22 - 5/13 9.3.3.3 Safety Evaluation Within the drainage system itself, the potential for inadvertent transfer of fluids from a contaminated volume to an uncontaminated volume does not exist as all areas of potential radioactive contamination are drained to the radioactive waste system sumps.
9.3.3.4 Tests and Inspection Requirements
Before being placed into service, the nonradioactive wastewater drainage system
proved leaktight when subjected to a hydrostatic test pressure equal to not less than a 10-
ft head of water.
9.3.3.5 Instrumentation Requirements
The drainage systems themselves do not require instrumentation for monitoring level, flow, temperature or radiation. All potentially radioactive drains are piped to the radwaste sump system which is shown in Figure 11.2-2. This figure shows the instrumentation that is used to monitor drainage into these sumps. LLRPSF drains are piped to sumps in the facility.
9.3.4 STANDBY LIQUID CONTROL SYSTEM
9.3.4.1 Design Bases
9.3.4.1.1 Safety Objective
The safety objective of the standby liquid control (SLC) system is to provide a backup method, independent of the control rods, to initiate and maintain the reactor subcritical as the nuclear system cools. Maintaining subcriticality thus ensures that the fuel barrier is not threatened by overheating in the improbable event that not enough of
the control rods can be inserted to counteract the positive reactivity effects of a colder moderator. The SLC system was modified in 1987 to meet the requirements of the final NRC rule on Anticipated Transients Without Scram (ATWS), given in 10CFR50.62 and
UFSAR/DAEC-1 9.3-11 Revision 22 - 5/13 9.3.4.1.2 Safety Design Bases
The SLC system meets the following safety design bases:
- 1. Backup capability for reactivity control is provided, independent of normal reactivity control provisions in the nuclear reactor, to be able to shut down the reactor if the normal control ever becomes inoperative.
- 2. The backup system has the capacity for controlling the reactivity difference between the steady-state rated operating condition of the
reactor with voids and the cold shutdown condition, including shutdown margin, to ensure complete shutdown from the most reactive condition at any time in core life.
- 3. The time required for the actuation and effectiveness of the backup control is consistent with the nuclear reactivity rate of change predicted between rated operating and cold shutdown conditions. A fast scram of
the reactor or operational control of fast reactivity transients is not specified to be accomplished by this system.
- 4. Means are provided by which the functional performance capability of the backup control system components can be verified periodically under conditions approaching actual use requirements. A substitute solution, rather than the actual neutron-absorber solution, can be injected into the reactor to test the operation of all components of the redundant control system.
- 5. The neutron absorber is dispersed within the reactor core in sufficient quantity to provide a reasonable margin for leakage or imperfect mixing.
- 6. The system is reliable to a degree consistent with its role as a special safety system; the possibility of unintentional or accidental shutdown of the reactor by this system shall be minimized.
9.3.4.2 Description
The SLC system is designed to inject a quantity of boron that produces a
concentration of Control capacity equivalent for DAEC to the requirements of the ATWS rule is achieved by running both SLC pumps simultaneously at their design minimum pumping rate of 26.2 gpm each,(References 1 through 3). described above will meet the ATWS shutdown requirements, as demonstrated in Section 15.3.1.
UFSAR/DAEC-1 9.3-12 Revision 22 - 5/13 The SLC system (see Figure 9.3-27) is manually initiated from the main control room to pump a boron neutron-absorber solution into the reactor if the operator believes
the reactor cannot be shut down or kept shut down with the control rods. However, the insertion of control rods is always expected to ensure prompt shutdown of the reactor
when required.
The SLC system is needed only in the highly improbable event that not enough control rods can be inserted in the reactor core to accomplish shutdown and cooldown in the normal manner.
The boron solution tank, the test water tank, the two positive-displacement pumps, the two explosive valves, and associated local valves and controls outside the primary containm ent. The liquid is piped into the reactor vessel and discharged near the bottom of the core shroud so that it mixes with the cooling
water rising through the core (see Section 5.3 and Section 3.9.5).
The boron absorbs thermal neutrons and thereby terminates the nuclear fission chain reaction in the uranium fuel.
A sparger, using compressed air, is provided in the tank for mixing. The storage tank piping to the suction of the SLC pumps is located on the side of the tank, rather than the bottom, to minimize the potential of line blockage from either foreign material or precipitation of the sodium pentaborate.
The low temperature alarm is required to be maintained at least 5 degrees above the saturation temperature.
The equipment containing the solution is installed in a room in which the air temperature is to be maintained within the range of 68 to 90
°F. In addition, a heater system maintains the solution temperature at 75 to 85
°F to prevent precipitation of the sodium pentaborate from the solution during storage.
Each positive-displacement pump is sized to inject the solution into the reactor at a minimum rate of 26.2 gpm. one pump is capable of injecting sufficient boron into the reactor to assure complete shutdown. The pump and system design pressure between the explosive valves and the pump discharge
is 1400 psig. The two relief valves are set at 1350 to 1400 psig. To prevent bypass flow from one pump in case of relief valve failure in the line from the other pump, a check valve is installed downstream of each relief valve line in the pump discharge pipe.
UFSAR/DAEC-1 9.3-13 Revision 22 - 5/13 The two explosive-actuated injection valves provide assurance of opening when needed and ensure that boron will not leak into the reactor even when the pumps are
being tested.
Each explosive valve is closed by a plug in the inlet chamber. The plug is circumscribed with a deep groove so the end will readily shear off when pushed with the
valve plunger. This opens the inlet hole through the plug. The sheared end is pushed out of the way in the chamber; it is shaped so that it will not block the ports after release.
The shearing plunger is actuated by an explosive charge with dual-ignition primers inserted in the side chamber of the valve. Ignition circuit continuity is monitored by a trickle current, and an alarm occurs in the control room if either circuit opens.
Indicator lights show which primary circuit opened. To service a valve after firing, a 6-in. length of pipe (spool piece) must be removed immediately upstream of the valve to gain access to the shear plug.
The SLC system is manually actuated by a two-position keylock switch which activates both injection pumps, opens both of the explosive valves, and isolates the reactor water cleanup system. A green light in the control room indicates that power is available to each pump motor contactor and that the contactor is open (pump not running). A pump will start even if the local switch at the pump is in the STOP position for test or maintenance. Pump discharge pressure and flow are indicated in the control room.
Although both pumps are started together, cross piping and check valves ensure that the boron solution will be injected even if only one pump runs and/or one explosive
valve opens.
Equipment drains and tank overflow are not piped to the waste system but to separate containers (such as 55-gal drums) that can be removed and disposed of independently to prevent any trace of boron from inadvertently reaching the reactor.
A high-point vent is located between the SLC explosive valves and check valve .
This vent line is required to allow for local leakage rate testing of valve in accordance with the Primary Containment Leakage Rate Testing Program.
9.3.4.3 Safety Evaluation
The SLC system is a special safety system not required for unit operation or to meet the single-failure criterion. The system is expected never to be needed for unit safety because of the large number of independent control rods available to shut down the
reactor.
UFSAR/DAEC-1 9.3-14 Revision 22 - 5/13 However, system reliability and the availability are enhanced by providing two sets of the components required to actuate the system. Pumps and explosive valves, are
provided in parallel in the design. Redundancy is not required for the tank heater or
heating cable.
The system is designed to bring the reactor from rated power to a cold shutdown at any time in core life. The reactivity compensation provided will reduce reactor power from rated to zero level and allow the cooling of the nuclear system to normal room temperature, with the control rods remaining withdrawn in the rated power pattern. It includes the reactivity gains that result from complete decay of the rated power xenon inventory. It also includes the positive reactivity effects from eliminating steam voids, changing water density from hot to cold, reduced Doppler effect in uranium, reducing neutron leakage from boiling to cold, and decreasing control rod worth as the moderator cools. The specified minimum final concentration of boron in the reactor core provides the capability of bringing the reactor, at any time in the cycle, to a subcritical condition.
to provide the specified shutdown margin, after operation of the SLC system, is 660 ppm (parts per million) (see Figure 9.3-30). The calculation of the minimum quantity of sodium pentaborate to be injected into the reactor is based on in the reactor coolant and the quantity of reactor coolant in the reactor vessel, recirculation loops, and the RHR system in the shutdown cooling mode at 70°F and reactor normal water level. The result is increased by 25% to allow for imperfect mixing, leakage, and volume in other small piping connected to the reactor.
will be achieved if the solution is prepared as defined in Section 9.3.4.2 and maintained above saturation temperature.
The cooldown of the nuclear system will require a minimum of several hours to remove the thermal energy stored in the reactor, cooling water, and associated equipment and to remove most of the radioactive decay heat. The controlled limit for the reactor vessel is 100
°F/hr, and normal operating temperature is approximately 550
°F. Usually, using the main condenser and various shutdown cooling systems to shut down the plant will require 10 to 24 hr before the reactor vessel is opened and much longer to reach room temperature (70
°F); this is the condition of maximum reactivity and, therefore, the condition that requires the maximum concentration of boron.
The SLC system equipment essential for the injection of the neutron-absorber solution into the reactor is designed as Seismic Category I for withstanding the specified earthquake loadings (see Section 3.2). Nonprocess equipment such as the test tank is designed as Nonseismic. The system piping and equipment are designed, installed, and tested in accordance with requirements stated in Sections 3.9 and 3.10.
UFSAR/DAEC-1 9.3-15 Revision 22 - 5/13 The SLC system is required to be operable in the event of a loss of offsite Power (LOOP); therefore, the pumps, heaters, valves, and controls are powered from the standby ac power supply or dc power in the absence of normal power. The pumps and valves are powered and controlled from separate buses and circuits so that a single failure in the power supplies will not prevent system operation.
The SLC system and pumps have sufficient pressure margin, up to the system relief valve setting of approximately 1400 psig, to ensure solution injection into the reactor above the normal pressure of approximately 1030 psig in the bottom of the reactor. The nuclear system relief and safety valves begin to relieve pressure at a nominal pressure of 1110 psig. Therefore, the SLC system positive-displacement pumps cannot overpressurize the nuclear system.
In addition, the SLC pumps need to be able to inject sufficient boron into the reactor to mitigate postulated ATWS events (Reference Section 15.3.1). A concern was raised in Information Notice 2001-13 (Reference 10) that insufficient margin was available between the required pump discharge pressure needed to ensure injection into the vessel during high pressure ATWS event and the SLCS relief valves opening settings to prevent the relief valves from opening and thereby divert sufficient boron injection from the reactor to cause unacceptable consequences. An evaluation was performed for the DAEC (Reference 11) that concluded that the potential for lifting the SLCS relief valves was minimal and that uninterrupted boron injection was assured during postulated ATWS events.
The following values show the bases for solution temperatures. The sodium pentaborate solution in the tank and pump suction line is required to be maintained 5
°F above its saturation temperature.
Concentration Solution Concentration
(%) Saturation Temperature
(°F) Minimum Required Solution Temperature
(°F) Minimum 53.5 58.5 Maximum 65 70 The heating system for the area of the building where the SLC system pumps and tanks B are located is designed to maintain the air temperature within the range of 68
° to 90°F. The electric heaters in the tank and the heat tracing in the pump suction line act as a backup to the area heating system. The tank heaters are designed to maintain the solution within the range of 75
° to 85°F and the heat tracing is designed to maintain the suction piping within a range of 75
° to 95°F.
UFSAR/DAEC-1 9.3-16 Revision 22 - 5/13 9.3.4.4 Tests and Inspection Requirements Operational testing of the SLC system is performed in at least two parts to avoid inadvertently injecting boron into the reactor. With the valves to and from the solution tank closed and the three valves to and from the test tank opened, demineralized water in the test tank can be recirculated by locally starting either pump.
The injection portion of the system can be functionally tested by valving the injection lines to the test tank and actuating the system from the control room. Both injection valves open on actuation. System operation is indicated in the control room. After functional tests, the injection valves and explosive charges must be replaced and all the valves returned to their normal positions as indicated in Figure 9.3-27.
By closing a local locked-open valve to the reactor, leakage through the injection valves can be detected at a test connection in the line between the containment isolation check valves. Position indicator lights in the control room indicate that the local valve is closed for tests or open and ready for operation. Leakage from the reactor through the first check valve can be detected by opening the same test connection when the reactor is
pressurized.
The test tank contains demineralized water for approximately 8 minutes of pump operation. Demineralized water from the makeup is available for refilling or flushing the system.
Should the boron solution ever be injected into the reactor, either intentionally or inadvertently, then after making certain that the normal reactivity controls will keep the reactor subcritical, the boron is removed from the reactor coolant system by flushing for gross dilution followed by operating the reactor cleanup system. There is practically no
effect on reactor operations when the boron concentration has been reduced below approximately 50 ppm.
The concentration of the sodium pentaborate in the solution tank is determined periodically by chemical analysis.
9.3.4.5 Instrumentation Requirements
Instrumentation consisting of solution temperature indication and control, tank level, and heater system status is provided locally at the SLC tank.
A temperature indicator in the pump inlet at the pressure point permit daily recording of the temperature of the liquid control solution in the piping between the SLC tank and the pump inlet. A temperature indicator monitors line temperature to ensure that the sodium pentaborate solution does not fall below about 75
°F. This could happen only if both the room air heating units and the line trace heaters failed in very cold
weather.
UFSAR/DAEC-1 9.3-17 Revision 22 - 5/13 High or low temperature, or high or low liquid level, causes an alarm in the control room. Tank level is also monitored in the control room. SLC pump discharge header pressure and flow indication are provided locally and in the control room.
9.3.5 HYDROGEN WATER CHEMISTRY SYSTEM
9.3.5.1 Design Bases
9.3.5.1.1 Power Generation Objectives
The power generation objectives of the hydrogen water chemistry system (HWCS) are to inject hydrogen into the feedwater to control intergranular stress
corrosion
cracking of austenitic stainless steel piping and components, and to provide the hydrogen supply for main generation cooling.
9.3.5.1.2 Power Generation Design Bases
The HWCS is designed to meet the following design bases:
- 1. It will supply hydrogen for feedwater injection at a rate of 0 to 5 SCFM to each feed pump suction.
- 2. It will supply hydrogen for main generator purge and makeup (continuous fill) requirements.
- 3. It will supply the offgas system with air or oxygen to ensure a stoichiometric mixture for recombination of hydrogen and oxygen.
- 4. It will inject oxygen into the suction lines of running condensate pumps in order to keep the oxygen level in the condensate and feedwater systems high enough to minimize general corrosion.
- 5. It will automatically isolate the hydrogen and oxygen injection systems in the event of system failures.
9.3.5.2 Description
Intergranular stress corrosion cracking in the reactor coolant system is controlled
by injecting hydrogen into the feedwater. The hydrogen in the feedwater enters the reactor where it removes (by combining with) dissolved oxygen, which is radiolytically produced. The composition of non-condensible gas which leaves the reactor via the
condenser and the offgas is changed. Flows of both oxygen and hydrogen are decreased, and the ratio of oxygen to hydrogen is also decreased. In order to prevent combustible mixtures of hydrogen in the offgas stream, air or oxygen is injected.
UFSAR/DAEC-1 9.3-18 Revision 22 - 5/13 Reduction of oxygen concentration to less than approximately 30 ppb could result in an increase in the corrosion rate for carbon steel piping in the feedwater and condensate systems. Oxygen is injected into the condensate pump suction lines to preclude this problem.
Hydrogen is stored in vendor-supplied hydrogen tube trailers, and in six high
pressure hydrogen storage vessels which serve as backup when the tube trailer supply is not available. A tube trailer has a capacity of approximately 125,000 cubic feet, which is sufficient for at least a ten-day injection supply. Hydrogen from the tube trailer can be used for feedwater injection, main generator fill and makeup, and recharging the
hydrogen storage tanks.
The hydrogen supply line is 1/2-inch stainless steel pipe, coated and wrapped for underground installation.
Hydrogen for injection is supplied to the suction of each feedwater pump through one of two redundant flow control valves. The hydrogen flow rate is programmed to increase in proportion to reactor power when in "Cascade mode". Hydrogen flow can also be adjusted manually if necessary using the "Automatic" or "Manual" controller modes of operation for the system. An isolation valve is provided to terminate the hydrogen injection if the feedwater pump is tripped or on a HWCS isolation signal.
9.3.5.3 Instrumentation Requirements
The following instrumentation and controls are associated with the HWCS:
- Concentrations of dissolved oxygen and hydrogen in the recirculation water are indicated and recorded.
- Residual offgas oxygen concentration is indicated and recorded.
2011-016 2011-016 UFSAR/DAEC-1 9.3-19 Revision 22 - 5/13
- Local atmospheric hydrogen monitors at six locations around the plant alarm on high concentration, and initiate HWCS isolation on a high-high concentration.
- Storage tank temperatures are indicated locally.
- Emergency shutdown buttons are provided to immediately isolate the hydrogen system. The oxygen system will be shut down after a 12 minute time delay. Air addition, if in service, will continue to operate to ensure that all excess hydrogen has had time to reach the offgas system.
The effect of hydrogen injection is monitored by instruments which measure electrochemical corrosion potential and crack growth rate. Sensors located outside the drywell are exposed to reactor coolant from a recirculation loop sample line. To verify that they were exposed to conditions which are representative of those in the primary system, additional sensors were placed in the recirculation piping and in-core via LPRM assemblies. These additional sensors are no longer in use. See Section 7.6.1.6.3.
9.3.6 ZINC INJECTION (GEZIP) SYSTEM
9.3.6.1 Description
GE Nuclear Energy has developed a system to inject zinc into the BWR primary system called GEZIP (General Electric Zinc Injection Passivation). The GEZIP process maintains trace quantities of ionic zinc in the reactor water for the purpose of reducing radiation buildup by maintaining/reducing CO-60 buildup on primary system surfaces.
The GEZIP system, SUS 63.01, consists of a zinc addition skid that is designed to inject trace amounts of Depleted Zinc Oxide (DZO) into the feedwater during normal plant operation. The system consists of a simple recirculation loop off of the feedwater system. The zinc solution is obtained by passing a stream of feedwater from the feedwater pumps' discharge header tap by the feedwater regulating valves. This feedwater then goes through a dissolution vessel containing pelletized DZO next to the turbine lube oil conditioner. The feedwater dissolves the pellets as it passes through the zinc vessel carrying the dissolved DZO into the feedwater pumps' suction header located on the condenser bay mezzanine. Manual valves are used UFSAR/DAEC-1 9.3-20 Revision 22 - 5/13 to control feedwater flow to the reactor. Instrumentation associated with the skid includes a calibrated flow meter, a differential pressure indicator, and a temperature
indicator.
9.3.7 NOBLE METAL CHEMICAL ADDITION
9.3.7.1 Description
Noble Metal Chemical Addition (NMCA) is a process used to inject noble metals into the reactor coolant to enhance the effectiveness and efficiency of Hydrogen Water Chemistry (HWC) in mitigating Intergranular Stress Corrosion Cracking (IGSCC) in Boiling Water Reactor (BWR) vessel internals. In addition, use of NMCA allows lowering injection rates of HWC which in turn reduces radiation exposure to plant personnel. NMCA may be performed using Classic NobleChem TM (injection during hot standby conditions) or On-Line NobleChem (injection during power operation
conditions) application specifications and procedures.
NMCA treatments have been applied into the DAEC's reactor coolant in an effort to mitigate IGSCC in the reactor vessel internals. The reactor water limits in the Technical Requirements Manual were changed to support classic application of the
NMCA (References 4 through 9).
UFSAR/DAEC-1 9.3-21 Revision 22 - 5/13 REFERENCES FOR SECTION 9.3
- 1. General Electric Company, Anticipated Transient Without Scram (ATWS) Response to NRC Rule 10CFR50.62, GE/NEDE-31096-P, December 1985.
- 2. General Electric Company, Assessment of ATWS Compliance Alternatives, GE/NEDC-30921, March 1985.
- 3. General Electric Company, Duane Arnold ATWS Assessment, GE/NEDC-30859-1, March 1985.
- 4. IES Utilities Inc., RTS-290, Reactor Water Conductivity Limit Change for Noble Metal Chemical Addition, NG-96-1297, July 5, 1996.
- 5. IES Utilities Inc., Noble Metal Chemical Addition 10CFR50.59 Safety Evaluation, SE 96-07, August 1996.
- 6. IES Utilities Inc., 10CFR50.59 Safety Evaluation for Noble Metal Pretreated Fuel Pins, SE 96-09, August 1996.
- 7. IES Utilities Inc., Noble Metal Chemical Addition Monitoring Equipment, Engineering Change Package 1573, October 3, 1996.
- 8. IES Utilities Inc., TRMCR-004, October 20, 1999.
- 9. IES Utilities Inc., 10CFR50.59 Safety Evaluation for 2 nd Noble Metal Chemical Addition, SE 99-046, August 1999.
- 10. NRC Information Notice 2001-13, "Inadequate Standby Liquid Control System Relief Valve Margin," August 10, 2001.
- 11. NG-01-0909, "Response to Request for Additional Information (RAI) to Technical Specification Change Request TSCR-042, Extended Power
Uprate," August 16, 2001.
UFSAR/DAEC-1 T9.3-1 Revision 12 - 10/95 Table 9.3-1 Sheet 1 of 4 LOCATION OF SAMPLING POINTS Description Locations Purpose Reactor Reciruclation System
Reactor primary coolant water Reactor vessel outlet Monitor reactor when cleanup is isolated a Reactor Water Cleanup System and Fuel Pool Cooling and Cleanup system
Filter-demineralizer influent Filter inlet pipe Demineralizer efficiency a
Filter-demineralizer effluent Filter outlet pipe Filter efficiency a Nuclear Steam Supply System
Main steam Main steam line Carryover steam quality, H 2 and 0 2 , Monitor corrosion and radioactivity Suppression pool Suppression pool reciruclation
pipe Monitor corrosion and radioactivity Standby liquid control system Reciruclation pipe Borate concentration
Reactor shutdown cooling system RHR system header Check corrosion inhibitor concentration Condensate System
Condensate Condensate pump discharge Condensate quality and tube leaks a
Condensate demineralizer effluent Demineralizer system outlet Treated condensate quality b
a Conductivity b Conductivity and dissolved oxygen.
UFSAR/DAEC-1 T9.3-2 Revision 17 - 10/03 Table 9.3-1 Sheet 2 of 4 LOCATION OF SAMPLING POINTS Description Locations Purpose Reactor Feedwater System
Reactor feedwater system After last heater of each train Water analyses c
Reactor feedwater system After first heater of each train Water analyses c Reactor Building Cooling Water System
Cooling water sample Outlet of each major heat exchanger Determine location of heat exchanger leaks Cooling water sample Pump discharge Check corrosion inhibitor concentration Main Condenser Circulating Water System
Influent Inlet to cooling water heat exchanger Determine radioactivity d Effluent Cooling water blowdown canal Radioactivity d Liquid Radwaste System
Radwaste surge tanks Radwaste surge pumps discharge Process data Waste collector tank Pump discharge Process data
Floor drain collector tank Pump discharge Process data c Conductivity and total iron.
d Conductivity and pH.
UFSAR/DAEC-1 T9.3-3 Revision 12 - 10/95 Table 9.3-1 Sheet 3 of 4 LOCATION OF SAMPLING POINTS Description Locations Purpose Liquid Radwaste System (Continued)
Detergent drain tanks Pump discharge Process data
Waste sample tanks Pump discharge Process data
Fuel pool filter-demineralizer
influent Filter inlet Fuel pool quality a Floor drain sample tank Pump discharge Process data
Chemical waste tank Pump discharge Process data
Chemical waste sample tank Pump discharge Process data
Waste collector filter Filter outlet Filter efficiency
Waste collector demineralizer Dimneralizer outlet Demineralizer efficiency a
Floor drain filter Filter outlet Filter efficiency
Floor drain demineralizer Demineralizer outlet Demineralizer efficiency a
Low-Level Radwaste Processing and Storage Facility sample tank Sample tank inlet Process Data Makeup Demineralizer System
Cation effluent Outlet pipe Demineralizer efficiency
Anion effluent Outlet of each unit Demineralizer efficiency a, d Condensate storage tank Demineralizer water transfer pump discharge Water quality a
a Conductivity.
d Conductivity and pH.
UFSAR/DAEC-1 T9.3-4 Revision 12 - 10/95 Table 9.3-1 Sheet 4 of 4 LOCATION OF SAMPLING POINTS Description Locations Purpose Offgas System
Air ejector sample After air ejectors Radioactivity, H 2 , 0 2 , and air leakage Offgas filter samples Inlet and outlet Process data
Stack sample Main stack Radioactivity
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UFSAR/DAEC-1 T9.4-1 Revision 23 - 5/15 Table 9.4-1 AREA TEMPERATURE REQUIREMENTS
Area Maximum Temperature for Normal Equipment Operation a (°F) Calculated Maximum Room Temperature
(°F) Control Room
104 76 Electrical and switch-gear rooms 104 102 Battery rooms 104 102 RHR and core spray rooms 104 104 HPCI room 104 104 RCIC room 104 104 Diesel-generator room Generator 140 104 Diesel 140 138
a Operation at temperatures higher than those listed will not affect safe system operation but may reduce the expected operating life of the equipment.
2015-003 2015-003 UFSAR/DAEC-1 T9.4-2 Revision 23 - 5/15 Table 9.4-2 Sheet 1 of 3 PERIODIC TESTING - SAFETY-RELATED VENTILATION SYSTEMS Test Item/System Test Description
Standby filter units -
control building 1. Isolation signal will be simulated to start the selected train. Performance of the fan will be checked
against its required flow characteristics for flow
through the HEPA filters and charcoal adsorbers.
Differential-pressure reading across each filter will
be recorded at the set air flow rate through the train.
- 2. It will be demonstrated that the electric coil functions properly and its output meets the required value.
- 3. The second train will be tested by similar actions.
- 4. Automatic initiation shall be demonstrated by verifying that upon receipt of a high radiation test signal at the air intake radiation monitors, the system automatically switches to the isolation mode.
Ventilation systems -
emergency diesel-generator rooms 1. Supply fan will turned on and it will be demonstrated that all the dampers are operating properly to provide
proper ventilation.
- 2. Performance of the fan will be checked against its preoperational balancing characteristics.
- 3. The second system will be tested by similar actions.
Ventilation system -
battery room 1. Exhaust fan will be turned on and its performance checked against its preoperational balancing
characteristics . 2. It will be demonstrated that the control system functions properly to meet the accident operating modes as designed.
- 3. Each fan will be tested by similar actions.
UFSAR/DAEC-1 T9.4-3 Revision 23 - 5/15 Table 9.4-2 Sheet 2 of 3 PERIODIC TESTING - SAFETY-RELATED VENTILATION SYSTEMS Test Item/System Test Description
Room cooling units -
RHR and core spray rooms, RCIC room, HPIC room
- 1. Room cooling unit will be turned on and performance of the fan checked against its
preoperational balancing characteristics. 2. It will be demonstrated that the cooling effect of the cooling coil meets the cooling requirement.
- 3. It will be demonstrated that the control system functions properly; if the room temperature rises to the setpoint then the ventilation system will turn on.
- 4. The second equipment will be tested by similar actions.
HVAC systems -
control room 1. The operating system will be checked to ensure that all dampers are operating properly to meet the set control room pressure.
- 2. Performance of the fans will be checked against the preoperational balancing characteristics.
- 3. Differential-pressure reading across the filter will be recorded at the set air flow rate.
- 4. It will be demonstrated that the cooling coil functions properly to meet the cooling requirements of the accident modes.
- 5. It will be demonstrated that the control system functions properly. The room temperature setting shall govern the system operation.
- 6. The second system will be tested by similar actions.
UFSAR/DAEC-1 T9.4-4 Revision 23 - 5/15 Table 9.4-2 Sheet 3 of 3 PERIODIC TESTING - SAFETY-RELATED VENTILATION SYSTEMS Test Item/System Test Description
Isolation dampers -
reactor building 1. Isolation signal to the selected damper will be cut off and the damper operation observed to ensure that it is
closed. 2. The closure time of the damper will be observed not to exceed 10 sec.
a 3. It will be demonstrated that the pilot lights are operating properly during the operation of the damper. 4. After test, the isolation signal will be replaced to the normal connection.
- 5. Similar tests will be done to each damper.
Isolation valves Refer to Technical Specifications.
a The dampers in the exhaust line from the refueling floor will be observed to close within 5 sec, the time required for isolation in this instance as described in Section 6.2.3.
UFSAR/DAEC-1 T9.4-5 Revision 23 - 5/15 Table 9.4-3 Summary Table of New Activated Carbon Physical Properties (Derived from ANSI/ASME N509-1980 Table 5-1)
Test Test Method Acceptance Value Performance Requirements Methyl Iodide, 30
°C, 95% RH (1) ASTM D3803-1989 <5.0% penetration Methyl Iodide, 30
°C, 95% RH (2)
ASTM D3803-1989
<0.5% penetration for a 6 bed Physical Properties Particle Size Distribution
ASTM D2862 using 8 x 16 U.S. Mesh Retained on #6 Sieve: 0.1% maximum Retained on #8 Sieve: 5.0% maximum Through #8, on #12 Sieve: 60% maximum Through #12, on #16 Sieve: 40% minimum Through #16 Sieve: 5.0% maximum Through #18 Sieve: 1.0% maximum Ball Pan Hardness ASTM D3802 92 minimum
C C 4, Activity (on base) ASTM D3467 60 minimum Apparent Density ASTM D2854 0.38 g/cm3 minimum Ash Content (on base) ASTM D2866 state value Ignition Temperature ASTM D3466 330°C minimum Moisture Content ASTM D2867 state value pH of Water Extract ASTM D3838 state value
Notes: (1) SFU Conditions (2) SBGT Conditions 2013-009