RS-17-126, Quad Cities Nuclear Power Station, Units 1 & 2, Revision 14 to Updated Final Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems
Text
QUAD CITIES - UFSAR Revision 5, June 1999 5.1-1 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS
5.1
SUMMARY
DESCRIPTION
The equipment and evaluations presented in th is chapter are applicable to either unit.
[5.1-1]
The reactor coolant system includes those syst ems and components which contain or transport reactor coolant, in the form of water or steam, to and from the reactor pressure vessel (RPV).
These systems form a major portion of the reactor coolant pressure boundary (RCPB).
Chapter 5 of this report provides informatio n regarding the reactor coolant system and pressure-containing appendages out to and in cluding the outermost isolation valve in the main steam and feedwater piping.
[5.1-2]
The RCPB includes all those pressure-containing components such as the RPV, piping, pumps, and valves, which are:
A. Part of the reactor coolant system (RCS), or
B. Connected to the RCS up to and including any and all of the following:
- 1. The outermost containment isolatio n valve in system piping which penetrates the primary containment;
- 2. The second of the two valves normally closed during normal reactor operation in system piping which does not pene trate the primary containment; and
- 3. The RCS safety relief valve (SRV), relief valves (RVs), and safety valves.
The topics of the RCS and connected systems that are discussed in this chapter include RCPB integrity, the RPV and appurtenances, and the major RCPB allied subsystems. Diagram of Nuclear Boiler and Reactor Recircu lation Piping is shown in FSAR Figure 5.1-1. Table 5.1-1 includes information such as overall dimensio ns, design pressures and temperatures, power ratings, and design codes for the major compon ents of the RCS. The total water and steam volume of the reactor vessel and recirculation system is approximately 15,679 cubic feet at 68 o F. Additional parameters of the RCS are summarized in Tables 5.4-1 through 5.4-6.
[5.1-3]
5.1.1 Reactor Coolant Pressure Boundary Integrity
Section 5.2 addresses the integrity of the RCPB. This section includes discussions of
overpressurization protection, RCPB materials, inservice inspection and testing of the RCPB, and RCPB leakage detection.
To protect against overpressure, relief valves ar e provided that can di scharge steam from the RCS to the suppression pool. The automatic de pressurization system (ADS) also acts to automatically depressurize the RCS in the event of a (small break) loss-of-coolant QUAD CITIES - UFSAR Revision 9, October 2007 5.1-2 accident (LOCA) in which the high pressure coolant injection (HPCI) system fails to maintain sufficient reactor vesse l water level. Section 6.3 pr ovides more details regarding the ADS and HPCI systems. Depressurization of the RCS allows the low-pressure core cooling systems to function.
5.1.2 Reactor Pressure Vessel and Appurtenances
The RPV and appurtenances are described in Secti on 5.3. The major safety consideration for the RPV is the ability to function as a radioactive material barrier. Various combinations of loading were considered in the vessel design. The vessel meets the requirements of applicable codes and criteria.
The possibility of brittle fracture was considered, and suitable design and operatio nal limits have been established to avoid conditions where brittle fractures are possible.
Figures 5.3-4a through 4c of Section 5.3 provide more information regarding the pressure-temperature limits. Refer to Reference 1 for detailed design information on the RPV, the purchase specifications for the RPV, RPV manufacturers data, and the seismic analysis of the RPV.
5.1.3 Reactor Coolant System Subsystems
Section 5.4 deals with subsystems that are closely allied to the RCPB. These include the
reactor recirculation system, the hydrogen wa ter chemistry (HWC) system, the main steam line flow restrictors, the reactor core isolatio n cooling (RCIC) system, the residual heat removal (RHR) system, and the reactor water cl eanup (RWCU) system. A brief description of these subsystems is provided in the following paragraphs.
The reactor recirculation system (refer to Section 5.4.1) prov ides coolant flow through the core. Adjustment of the core coolant flow ra te changes reactor power output, thus providing a means of following plant load demand withou t adjusting control rods. The recirculation system is designed so that no fuel damage will result during operat ional transients caused by reasonably expected single operator e rrors or equipment malfunctions. The arrangement of the recirculation system rout ing plus the appropriate placement of pipe break restraints is such that a piping failu re cannot compromise containment integrity.
The HWC system (refer to Section 5.4.3) is us ed to inject hydrogen into the reactor coolant to limit the dissolved oxygen concentration. Su ppression of dissolved oxygen, coupled with high purity reactor coolant, reduces the suscep tibility of reactor piping and materials to intergranular stress corrosion cracking.
The main steam line flow restrictors (refer to Section 5.4.4) are venturi-type flow devices that are welded into each steam line between the RPV and the first main steam line
isolation valve (MSIV). The restrictors are de signed to limit the loss of coolant resulting from a main steam line break outside the prim ary containment so that reactor vessel water level remains above the top of the core during the time required for the MSIVs to close.
QUAD CITIES - UFSAR Revision 6, October 2001 5.1-3 Two isolation valves are installed on each main steam line, one inside and the other outside the primary containment. The MSIVs are discussed in Section 6.2.
The RCIC system (refer to Section 5.4.6) pr ovides makeup water to the core during a reactor shutdown when the reactor become s isolated from the main condenser and feedwater flow is not available. The system is started automatically upon receipt of a low reactor water signal, or is manually started by the operator. Water is pumped to the core from the contaminated condensate storage t ank by a turbine-driven pump using reactor steam. For 10 CFR 50 Appendix R considerat ions, the electric-driven safe shutdown makeup pump (SSMP) provides backup to the RCIC system of either Unit 1 or Unit 2. For details on the SSMP system refer to Section 5.4.6.5.
The major equipment of the RHR system (refer to Section 5.4.7) includes four main system pumps, two heat exchangers, and four RHR se rvice water pumps. The RHR system can be used to remove heat under a variety of situat ions. In one of its three modes of operation (shutdown cooling), the RHR system removes decay heat during normal shutdown and reactor servicing. A second mode of RHR sy stem operation (containment cooling) removes heat from the primary containment followi ng a loss-of-coolant accident. The third operational mode of the RHR system is low-pre ssure coolant injection (LPCI). Low pressure coolant injection is used during a postulated loss-of-coolant accident. LPCI operation is described in Section 6.3.2. Other features of the RHR system include: supplementing the fuel pool cooling system; draining the cond enser to the suppression chamber by taking water from a condensate pump; transferring wa ter from the RPV to the main condenser or to the suction of the condensate pumps; tr ansferring water from the suppression chamber to the radwaste system, or via the radwaste system to the main condenser; and delivering and returning reactor water to the fuel pool system demineralizer for cleanup.
The RWCU system (refer to Section 5.4.8) recirculates a portion of the reactor coolant through a filter-demineralizer to remove so luble and insoluble impurities. The RWCU system maintains RCS coolant inventory by re turning the same quantity of water that was extracted.
Section 5.4 also discusses: main steam line and feedwater piping (refer to Section 5.4.9), valves (refer to Section 5.4.12), and the safety and relief valves (refer to Section 5.4.13).
5.1.4 Piping and Instrumentation Diagrams
The piping and instrumentation diagrams (P&IDs) applicable to the RCS and connected systems are identified in Table 5.1-2. This t able is organized according to the drawing topic first and then the applicable unit.
[5.1-4]
5.1.5 General Arrangement
The general arrangement drawings for the re actor coolant system and connected systems are shown on M-6, M-8, and M-9. These drawin gs are applicable to both Unit 1 and Unit 2.
QUAD CITIES - UFSAR Revision 9, October 2007 5.1-4 5.1.6 References
- 1. Quad Cities Reactor Pre ssure Vessel Design Report.
(Sheet 1 of 4)
Revision 6, October 2001 QUAD CITIES - UFSAR Table 5.1-1 REACTOR COOLANT SYSTEM DATA Reactor Vessel
Internal height 68 ft 7-5/8 in Internal diameter 251 in Design pressure and temperature 1250 psig at 575°F
Maximum heatup rate and normal cooldown rate 100°F within a 1-hour period Base metal material SA-302 Grade B Top Head thickness 4 in. (minimum)
Shell thickness 6-1/8 in (minimum)
Bottom Head thickness 6-1/8 in. (minimum)
Design lifetime 40 years Base metal initial NDT (assumed) 40°F maximum Cladding material Weld deposited ER-308 electrode Cladding thickness 1/8 in (minimum)
Design code ASME
^*^ Section III Class A, 1965
Recirculation Loops Number 2 Material Stainless steel Design Pressure and temperature Suction 1175 psig at 565°F Discharge 1325 psig at 580°F Design Code ASME Section I, 1965 ASME Section I, 1968
USAS B 31.1, 1967
- ASME Boiler and Pressure Vessel Code.
QUAD CITIES - UFSAR Table 5.1-1 (Continued)
REACTOR COOLANT SYSTEM DATA (Sheet 2 of 4)
Recirculation Pumps Number 2 Type Vertical, centrifugal, single stage
Power rating 6000 hp Speed 1800 rpm Flow rate 45,000 gal/min Design pressure and temp. 1450 psig at 575°F Developed head 570 ft Design code ASME Section III Class C, 1965 ASME Section III Class C, 1968
Recirculation Valves Number 8 Type Motor operated gate Design code ASME Section I, 1965 ASME Section I, 1968
USAS B 31.1, 1967
Jet Pumps Number 20 Material Stainless steel
Overall height (top of nozzle to diffuser discharge) 18 ft 7 in Diffuser diameter 20-3/4 in QUAD CITIES - UFSAR Table 5.1-1 (Continued)
REACTOR COOLANT SYSTEM DATA (Sheet 3 of 4)
Revision 8, October 2005 Main Steam Lines Number 4 Diameter 20 in Material Carbon Steel Design Code ASME Section I, 1965 ASME Sections I and III, 1968
USAS B 31.1, 1967 Electromatic Relief Valves Number Capacity (each)
Pressure setting (analytical limit)
Design Code 4
558,000 lbm/hr each at 1120 psig
<1115 psig (2)
< 1135 psig (2)
USAS B 31.1 1967 ASME Section III Code Class 1, 1980 Edition, Winter 1980 Addenda without Code Stamp ASME Section III Code Class 1, 1995 Edition, with 1996 Addenda without Code Stamp Safety Valves Number 8 Capacity (each) 644,543 lbm/hr each at 1240 psig Pressure Setting 1240 psig (2) 1250 psig (2)
1260 psig (4) Design Code ASME Section III, 1965 ASME Section III, 1968
USAS B 31.1, 1967 QUAD CITIES - UFSAR Table 5.1-1 (Continued)
REACTOR COOLANT SYSTEM DATA (Sheet 4 of 4)
Revision 10, October 2009 Safety/Relief Valve (Target Rock)
Number 1 Capacity (each) 598,000 lbm/hr at 1080 psig
Pressure Setting (safety function setpoint)
Pressure Setting (analytical limit for relief function) 1135 psig
<1135 psig Design Code ASME Section III, 1965 ASME Section III, 1968
USAS B 31.1, 1967 Reactor Core Isolation Cooling System TURBINE Steam Pressure Inlet 150 - 1120 psia Exhaust 25 psia Power 80 - 500 hp Steam Flow 6,000 - 16,500 lb/hr
Pump Number 1 Type 5 - stage, horizontal, centrifugal
Discharge Developed Head - over a reactor pressure range of 1135 psia - 165 psia 2800 ft at 1135 psia -
525 ft at 165 psia
Flow 400 gal/min NPSH 20 ft
Control Power 125 Vdc 120 Vac (Sheet 1 of 1) Revision 13, October 2015 QUAD CITIES - UFSAR Table 5.1-2 APPLICABLE REACTOR COOLANT SYSTEM P&IDs Topic Unit Drawing Main steam piping 1 M-13 Main steam piping 2M-60 Reactor feed piping 1M-15 Reactor feed piping 2 M-62 Pressure suppression piping1M-34 Pressure suppression piping2M-76 Reactor recirculation piping1M-35 Reactor recirculation pump trip ATWS piping 1M-35Reactor recirculation piping2M-77 Reactor recirculation pump trip ATWS piping 2M-77SSMP diagram 1 & 2M-70 RCIC piping 2 M-89 RCIC piping 1 M-50 RHR piping 1M-37 RHR piping 1 M-39 RHR piping 2M-79 RHR piping 2M-81 RWCU piping 1M-47 RWCU piping 2M-88
QUAD CITIES - UFSAR Revision 7, January 2003 5.2-1 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY
This section addresses measures employed to provide and maintain the integrity of the reactor coolant pressure boundary (RCP B) for the plant design lifetime.
5.2.1 Compliance With Codes and Code Cases
The edition of applicable codes, addenda, and code cases for the pressure vessels, piping, valves and pumps of the RCPB components are listed in Section 3.2.
[5.2-1]
5.2.2 Overpressurization Protection
The Quad Cities Extended Power Uprate Proje ct included re-evaluating a broad set of most limiting transient events at the power upra ted conditions. The Limiting Transient Overpressure Events which are reanalyzed at 2957 MWt included the MSIV closure with direct scram, the single MSIV closure, the load rejection with bypass, the slow recirculation increase, and the fast recirculation increase. In addition, a Turbine Trip without bypass with a high flux scram was performed to reco nfirm that the MSIV closure with flux scram was the limiting event for the ASME overpressu re analysis. Specific diagrams showing the results of their transient are contained in Reference 5.
Overpressurization of the RCPB during rea ctor operations other than refueling is prevented by the design of the reactor control systems and the reactor safety system. These design features include:
[5.2-2]
A. High reactor pressure scram;
B. High neutron flux scram;
C. Turbine-generator load rejection scram;
D. Operation of the reactor core isolation cooling (RCIC) system; E. Operation of the turbine bypass system;
F. Operation of the dual-function safety/relief valve (SRV);
[5.2-3]
G. Operation of the relief valves;
H. Operation of the safety valves;
I. Operation of the high pressure coolant injection (HPCI) system ;and
J. Main steam isolation valve (MSIV) closure scram.
5.2.2.1 Design Bases
The purpose of the relief and safety valves is to prevent over-pressurizing of the RCPB including the reactor pressure vessel (RPV). Th e relief valves are also designed to rapidly depressurize the RPV in the event of a small break loss-of-coolant accident (LOCA) where HPCI malfunctions so that core spray QUAD CITIES UFSAR 5.2-2 Revision 12, October 2013 and the low pressure coolant injection (LPCI) mode of the residual heat removal (RHR) system will function to protect the fuel barrier.
Position indication of the safety/relief valve and the other four relief valves is required to be obtainable during reactor operation. To achieve these purposes, the relief, safety/re lief, and safety valves have the following capacities and setpoints:
[5.2-4]
Relief Valves (4)
[5.2-5] Capacity 558,000 lb m/hr each at 1120 psig Pressure Setting <=1115 psig (2)
(analytical limit) <=1135 psig (2)
Safety Valves (8)
Capacity 644,543 lb m/hr each at 1240 psig Pressure Setting 1240 psig (2) 1250 psig (2) 1260 psig (4)
Safety/Relief Valve (1) (Target Rock)
[5.2-6] Capacity 598,000 lb m/hr at 1080 psig Pressure Setting 1135 psig (safety function setpoint)
Pressure Setting <1135 psig (analytical limit for relief function)
The relief valves, which include the SRV, are size d to rapidly remove the generated steam flow upon closure of the turbine stop valves and coincident with failure of the turbine bypass
system.
The safety valves are sized to protect the RP V against overpressure during a MSIV closure without direct scram on valve position event, a turbine trip with a failure of the turbine bypass system and without direct scram on turbine stop valve position event, or a load reject with a failure of the turbine bypass system and without direct scram on turbine control valve
fast closure event (see Section 5.2.2.2.3 for fu rther details). The ASME Code requires that each vessel designed to meet Section III be pr otected from the consequence of pressure and temperature in excess of design conditions. Th e USAS B 31.1 Code for Pressure Piping also requires overpressure protection.
[5.2-7]
The relief and safety valve capacities are take n from the "Transient Protection Parameters Verification for Reload Licensing Analyses."
[5.2-8] The installation of the Acoustic Side Branch (ASB) in the inlet piping to the Electromatic Relief Valves and Safety Valves (see Section 3.9.
2.1) has increased the pressure loss coefficient upon valve actuation. The increased loss coeffi cient results in a reduction in flow through these valves for a given steam line pressure. The impact of flow reduction due to this
additional pressure loss is factored into the app licable events as part of each reload analysis.
The ASME capacity of the valves is not changed.
QUAD CITIES UFSAR 5.2-3 Revision 12, October 2013 5.2.2.2 Design Evaluation
5.2.2.2.1 Loadings and Analyses Steam generated following reactor isolation mu st be removed rapidly enough to prevent a large pressure rise. The relief valves are pr ovided to remove sufficient steam from the reactor, following a scram that includes MSIV closure, to prevent the safety valves from lifting.
[5.2-9]
In compliance with ASME Section III, the safety valves must be set to open no higher than 105% of design pressure, and at least one safety valve pressure setting shall not be greater than the design pressure of the vessel. The setpoints of the safety valves comply with the
ASME Code taking into account st atic heads and dynamic losses.
Studies have been made on plants that are g eometrically similar to Quad Cities on the loadings which the relief and safety valves place on the main steam line. The loadings
considered include:
[5.2-10]
A. The thermal expansion effects of the main steam and relief valve discharge piping.
B. The earthquake effects of the relief and safety valves and relief valve discharge piping.
C. The jet force exerted on the relief and safety valves during the first millisecond when the valve is open and steady state flow has not yet been established.
D. The dynamic effects of the kinetic energy of the piston disc assembly when it impacts on the base casting of the valve.
The piping system and supports were qualifie d for the following loading conditions which the relief valves placed on the main steam piping.
[5.2-11]
A. thermal and dead weight effects on the main steam and relief valve discharge piping B. Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) effects on the main steam and relief valve discharge piping
C. dynamic effects of the kinetic ener gy and the jet forces when the relief valves open (SRV) and safety valves open (SV)
These six load cases (dead weight, thermal, O BE, SSE, SRV and SV), reference Section 3.9, are then used in various combinations (i.e., one or more relief valves opening, one or more
safety valves opening) to provide maximum pi ping stress and support loading. This support loading includes the piping restraints inside the pressure suppression chamber (torus).
QUAD CITIES UFSAR 5.2-4 Revision 12, October 2013 Thermal expansion analyses were made for seve ral cases with the relief valve piping both cold and hot and with jet forces, and piston di sc impact forces, applied simultaneously to all valves. These studies show that the loads du e to relief valve operation have only a minor effect on the stress condition of the main steam piping. The greatest stress is found at the
branch connection below the valve. In no case has the stress at this point exceeded the maximum stresses allowed by the ASME Code.
[5.2-12]
An overview of the analysis performed for the Electromatic relief valves included the following:
[5.2-13]
A. Determine the valve nozzle loads and the valve center of gravity (C.G.)
accelerations from the piping analysis.
B. Using the loads of part (A), consider the following elements and find the stresses:
- 1. Calculate the valve body stress for internal steam pressure plus safe shutdown earthquake and relief valve operation (SSE+RV)
ABS valve nozzle loads from part (A).
- 2. Using (SSE + RV)
ABS "g" values, find the stresses in the turnbuckle and the pilot valve tube for these "g" values app lied to the extended structure C.G. in three directions simultaneously.
Assume a continuous solid circular cross-section of 1-1/32-inch outside diameter for the turnbuckle. This simplifies the analysis for the loads and
moments supported by the turnbuckle and pilot valve tube.
Include the internal steam pressure for the pilot valve tube.
- 3. The stresses of the solenoid asse mbly mounting bracket hold-down bolts are calculated for the (SSE + RV)
ABS acceleration of the solenoid switch assembly C.G. in three directions simultaneously.
- 4. The solenoid assembly mounting bolts (located at the top of the mounting bracket) are analyzed for the (SSE + RV)
ABS acceleration of the solenoid switch C.G. in three directions simultaneously.
- 5. Target Rock Safety Relief Valve Out-of-Service design basis information is addressed in the cycle-specific reload reports.
A summary of these stresses is shown in Table 5.2-1.
Pneumatic supply lines to the Target Rock SRV have been upgraded to seismic design as a result of IE Bulletin 80-01. This was done to ensure a pneumatic supply to the valve in the
event of a design basis earthquake.
[5.2-14]
QUAD CITIES UFSAR 5.2-5 Revision 12, October 2013 5.2.2.2.2 Relief Valve Sizing The relief valves are sized, based upon the original analysis, by assuming a turbine trip
with simultaneous reactor scram and with a failure of the turbine bypass system. This
transient is reanalyzed periodically as part of reload licensing analysis. Typical results for this transient from 2957 MWt operating condit ions are presented in Reference 5 and are shown in Figure 5.2-1. The sudden closure of the turbine stop valves with no initial bypass
flow effectively doubles the initial rate of in crease of primary system pressure. Scram is initiated from the stop valve closure.
[5.2-15]
The vessel pressure peaks at 1292 psig. Peak pr essure in the steam line at the safety valve location is approximately 1253 psig. Core coolin g, level, and pressure control are provided by the RCIC system. Analyses performe d at an uprated power level of 2957 MWt
[5] have shown that the safety valves may lift during th is transient if conservative parameters are assumed; however, this postulated ev ent would be a rare occurrence.
QUAD CITIES UFSAR 5.2-6 Revision 14, October 2017 5.2.2.2.3 Safety Valve Steam Flow Capacity For power uprate, the safety valves steam flow capacity is determined by assuming that the reactor is at 2957 Mwt when a MSIV closure o ccurs, the relief valves fail to open, direct reactor scram (based on MSIV position swi tches) fails, and the backup scram due to high neutron flux shuts down the reactor. This tr ansient is reanalyzed periodically as part of each reload license analysis. Pressure incre ases, following this reactor isolation, until limited by the opening of the safety valves. The peak allowable pressure is 1375 psig (according to ASME Section III equal to 110% of the vessel design pressure 1250 psig). The safety valves setpoints are spread in 10 psi increments between 1240 and 1260 psig. This satisfies the ASME Code specifications that the lowest safety valve be set at or below vessel design pressure, and the highest safety valve be set to open at or below 105% of vessel design pressure.
[5.2-17]
The total safety valve capacity is equal to approximately 43% of turbine design flow.
Typical resulting transients at 2957 MWt are shown in Reference 5 and in Figure 5.2-3.
The rapid pressurization caused by the isolat ion (about 100 psi/s) reduces the void content of the core and produces a sharp neutron flux spike before scram shuts down the reactor.
Peak fuel surface heat flux is significantly slower, reaching a pe ak of 129% at about 3 seconds. Vessel dome pressure reaches about 1336 psig with the peak at the bottom of the vessel near 1358 psig. Therefore, the 43% capa city safety valves provide adequate margin below the peak allowable vessel pressure of 1375 psig.
Overpressurization protection analysis is performed using the NRC approved transient code(s) each cycle. A description of the over pressurization protection methodology used for Westinghouse reloads can be found in Secti on 9.3.2 of Reference 6 for Westinghouse analyses and can be found in cycle-specific reload analyses for AREVA analyses. The MSIV closure without direct scram on valve position event, a turbine trip with a failure of the turbine bypass system and without direct scram on turbine stop valve position event, and a
load reject with a failure of the turbine bypa ss system and without direct scram on turbine control valve fast closure event are evaluated ea ch reload to ensure the ASME overpressure limit is not exceeded. Also, for the turbine bypass valves equipment out-of-service option, the Feedwater Controller Failure (FWCF) ev ent and the Inadvertent High Pressure Coolant Injection (IHPCI) event are analyzed.
To satisfy the ASME criterion for the maximum vessel peak pressure and the steam dome pressure safety limit for the FWCF and IHPCI events with turbine bypass valves equipment out-of-service option, a power restriction may apply depending on the availab ility of the safety valves. Results for the limiting event are presented in the reload licensing report. [5.2-18]
5.2.2.2.4 Relief Valve Discharge Line Restraint Analysis
Earlier experience at some BWR plants with Ma rk I containments revealed inadequacies in the relief valve piping restraints inside the pr essure suppression chamber. In some cases, original restraints were replaced. The NRC had requested that the relief valve piping and restraints inside the pressure suppression cham ber of Quad Cities Station, Units 1 and 2, be inspected for signs of damage and analyzed to confirm the adequacy of the original design. [5.2-19]
In the original design, each of the units had five main steam Electromatic relief valves and associated discharge lines which were constructe d of 8-inch schedule 80 pipe material per ASTM A-106, Grade B. These five lines entered separate bays of the suppression chamber
through the drywell-to-suppression chamber ve nt tubes, and terminated in a ramshead configuration at the suppression chamber ce nterline approximately two-thirds of the distance below the normal suppr ession chamber water level.
[5.2-20]
QUAD CITIES UFSAR 5.2-7 Revision 8, October 2005 To mitigate the pressure spike following certain postulated transients, which related to the projected rate of reactivity in sertion following a scram, a modi fication was planned to replace the Electromatic relief valves with faster-acting Target Rock SRVs. The complete modification was never installed. Instead a partial versio n, known as the Scram Reactivity-Interim Fix modification, was installed on both units. This modification replaced only one of the original
main steam Electromatic relief valves with a Ta rget Rock SRV on each unit. An extensive analysis was performed to justify installation of the faster-acting Target Rock SRV without upgrading the discharge line restraints.
[5.2-21]
The analysis, which was quite conservative, showed that the stresses in the restraint members resulting from all loads combined were below yiel d in all cases. Analysis of the many welded and bolted connections demonstrated that all of th em had acceptable stresses. A special visual inspection of the highest stressed connection wa s made on Unit 2 on October 14, 1975. The inspection revealed no damage to the connection.
[5.2-22]
The first part of the analysis consisted of find ing the hydraulic forces on the piping that resulted when a relief valve was opened.
The computer code SRVA used advanced calculational techniques to model vent clearing phenomena, and the results were inherently very conservative. The calculation revealed t hat the hydraulic transient force was the largest single force acting on the pipe; all other forces were much smaller.
The original restraint design did not consider the forces related to clearing water from the vent line. Fortunately, the restraints were de signed with enough additional strength to prevent their failure under the revised load combination.
The maximum forces predicted to occur in vari ous members of restraint structures are given in Table 5.2-2. The maximum calculated fiber stresses and shear stresses in various members
are also shown. The stresses in the Target Rock valve discharge line restraints were predicted to be higher than those in the Electromatic valve discharge line restraints.
[5.2-23]
Table 5.2-2 shows that the stresses in some memb ers of the supporting structures were higher than the AISC allowable stresses. However, the fact that about 215 total blowdowns had occurred prior to the inspection without any appa rent damage, and also that the major portion of the hydraulic transient load is of a signif icantly short duration (lasting only about 0.25 seconds, with peak load occurring only for a fraction of this duration), suggest that the hydraulic transient load used in the analysis was conservatively estimated, and a higher allowable stress in the members (due to a high strain rate) was being realized.
[5.2-24]
Presently, each unit has four relief valve lines with Electromatic relief valves and one line with a Target Rock dual-function SRV which open s faster and allows marginally greater flow than the Electromatic valves. An extensive hi story of Electromatic relief valve discharges exists for the plant. Subsequent inspections have revealed only isolated cases of minor damage to pipe supports, and these were attributed to installation deficiencies. The discharge lines equipped with the faster-acting Target Rock SRVs have QUAD CITIES - UFSAR 5.2-8 Revision 13, October 2015 shorter discharge histories; however, inspection of one line after two discharges revealed no damage to pipe supports.
[5.2-25]
The adequacy of the Electromatic relief valves (including the Target Rock SRV) and their associated discharge lines and restraints fo r liquid and two-phase flow was demonstrated as part of a post TMI test activity conducte d for the BWR Owners Group. The test results were documented in NEDE-24988-P
[1] which was provided to the NRC in September, 1981.
[5.2-26]
Subsequent to the analysis described above, addi tional loadings on the discharge lines were defined in the Mark I Containment Program.
T-quencher devices were installed on the exits of the five relief valve discharge lines to reduce hydrodynamic loads in the suppression pool and promote stable steam condensation. Di scharge line supports were redesigned. In addition, a new larger vacuum breaker valve was installed on each discharge line in the
drywell approximately 20 feet upstream of the jet deflector plate which protects the vent tube. Finally, relief valve setpoints were adjust ed and a lock-out timer was installed on the low-set valve on each unit to prevent repeated actuation concurrent with an elevated water level in the discharge line. The larger vacuum breaker, set-point adjustments, and lock-out timer were designed to reduce transient loads on the discharge line and its supports, and hydrodynamic loads on the suppression chamber and attached structures.
[5.2-27]
The discharge lines and the associated supports were reanalyzed for the new hydrodynamic loads identified in the Mark I Program (see Se ction 6.2.1.3.4), and were shown to comply with the applicable acceptance criteria. The analysis applicable to the current discharge line configuration is described in the Plant Unique Analysis Report (PUAR).
5.2.2.3 Piping and Instrumentation Diagrams
The piping and instrumentation diagrams (P&I Ds) for the nuclear boiler system including the overpressure protection devices and thei r routing are shown on P& ID M-13, M-35 for Unit 1; and M-60, M-77 for Unit 2.
5.2.2.4 Equipment and Component Description
The Electromatic relief valves are actuated autom atically by a high reactor vessel pressure signal, or they can be operated manually from the control room. To add protection in case of a small line break or for certain degraded tr ansients, actuation of the relief valves to depressurize the reactor vessel will occur aut omatically when certain permissives are satisfied. Logic sequences resulting in aut omatic actuation are discussed in Sections 6.3.2.4, 7.3.1.4, and 15.6.
[5.2-28]
QUAD CITIES UFSAR 5.2-9 Revision 8, October 2005 The reactor relief valves are located on the main steam lines upstream of the first isolation valve and discharge directly to the pressure suppression pool. There are two independent sensor systems supplying the signals to all valv es to operate, and all valves are powered by the same safety-grade power source which is separate from the HPCI power source. An additional power source is also available and is automatically switched over upon loss of the primary power source.
[5.2-29]
The reactor safety valves are located on the main steam lines inside the primary
containment. They are balanced, spring-loaded safety valves that discharge directly to the drywell atmosphere. The safety valves are the final protection against overpressurizing the vessel and are sized to prevent reactor pressu re from exceeding the pressure limitations specified in the ASME Code.
[5.2-30]
The dual-function Target Rock SRV discharges within the primary containment system, to the suppression chamber. This valve operat es automatically on high reactor pressure approximately 100 - 135 psi above the operating pressure, but below the setting of the
safety valves. It also serves in the automatic relief system and can be actuated on either automatic or manual signal. This valve will pa ss approximately 7% of turbine design steam flow. [5.2-31]
The relief valves and the dual-function SRV will also function to automatically depressurize the reactor, under certain conditions, followi ng a loss-of-coolant accident as part of the automatic depressurization system (ADS).
The ADS is discussed in Section 6.3. These valves normally open automatically on rea ctor overpressures of between 115 and 135 psi, and then close at lower, preset pressure levels.
An additional function of these valves is to open and remain open below their preset cl osing pressures when signalled to do so following a LOCA that does not pressurize the dr ywell. This "remain open" signal is based on sustained signals indicating low-low water level simultaneously with pump operation for either the LPCI or the core spray function.
[5.2-32]
By remaining open, these valves reduce the reactor pressure to the point where the LPCI system and/or the core spray system can acco mplish reflooding of the reactor core. This permits the activation of the reactor core reflooding systems required for the various break sizes. To protect against a faulty "remain op en" signal, a short time delay (120 seconds) is provided during which the operat or can override the signal.
[5.2-33]
To prevent inadvertent ADS actuation from a fire, power to the ADS logic can be removed by turning the ADS inhibit switch to the INHI BIT position. For severe fires which prevent safe shutdown by normal means, the ADS inhibi t switch is turned to the INHIBIT position; and, as an additional precautionary measure, th e power supply to the ADS logic circuit is subsequently opened in the 125 Vdc distribution panels.
[5.2-34]
Another means to prevent inadvertent actuation of relief valves is to pull their fuses. Refer to the Safe Shutdown Report fo r a detailed description of relief valve actuation and it's affect on time lines for reactor water ma keup during an Appendix R scenario.
[5.2-35]
The relief valves and their discharge lines are also capable of being used in an alternate
shutdown cooling mode. In this mode of core cooling, water would be supplied to the RPV by a core spray or LPCI pump which would fill the vessel above the steam line discharge nozzles. The water would then flow through the steam lines and the open relief valves back
to the suppression pool and the suction side of the core spray or LPCI pump. Based on
extensive research performed in response to NUREG 0737 item II.D.1, ERV's are capable of being used in the alternate shutdown cooling mode and all pipe QUAD CITIES UFSAR 5.2-10 Revision 13, October 2015 stresses and support loads for this mode of co re cooling are within design allowables.
[5.2-36]
Each of the four relief valves and the d ual-function SRV discharge to the pressure suppression chamber via dedica ted (one per valve) discharge lines. Analyses have shown that upon valve closure, steam remaining in the discharge line can condense, thereby creating a vacuum which will draw suppression p ool water up into the discharge line. This "elevated water leg" condition is quickly alleviated by operation of the vacuum breaker on the discharge line; however, the condition is of concern since a subsequent actuation in the presence of an elevated water leg can result in unacceptably high thrust loads on the
discharge piping.
[5.2-37]
To prevent these unacceptable loads, the setpoint s and control logic for the relief valves and the SRV have been designed to ensure that ea ch valve which closes will remain closed until the normal water level in the discharge line is restored. This is accomplished by first establishing opening and closing setpoints su ch that all pressure induced subsequent actuations (after the "first pop") are limited to the two lowest set valves. These two valves are equipped with additional logic which function s in conjunction with the setpoints to inhibit valve reopening (via reactor repressurization or the ADS) for at least 10.5 seconds following each closure. (This compares with a calculated worst case elevated water leg duration time of 6.3 seconds). This combination of setpoint se lection and control logic design satisfies the single failure criterion, and is sufficient to en sure that no credible scenario can result in actuation of a relief or SRV in the presence of an elevated water leg.
The relief and safety valves are provided with acoustic monitors. Vibration from steam discharging through the valve triggers an alarm in the control room if the valve is open.
Indicating lamps and a test function are al so provided for the acoustic monitors.
The relief valves and Target Rock SRV have indi cating lamps in the control room which light if an opening signal is present. These inform th e operator if the valve is receiving a signal to open.
Both the relief and safety valves are equippe d with temperature elements and acoustic monitors that signal alarms in the control room if one of these valves opens. In addition, the
safety valves are equipped with a 10-psig rupture disc in the discharge line. In the event that the temperature elements and acoustic monitors failed to detect a leak in the safety valves discharge lines, an inspection during a refueling outage would reveal it.
5.2.2.5 Mounting of Pressure-Relief Devices
The relief valves, safety valves, and the Target Rock SRV are mounted on the main steam lines. FSAR Figure 10.3-1 and P&IDs M-13, M-60, M-34, and M-76 show the distribution of these pressure relieving devices on the four main steam lines. Information on thrust, bending, and other loads plus the resulting stre sses is contained in Tables 5.2-1 for Unit 1 and 5.2-2 for Unit 1 and 2.
QUAD CITIES UFSAR 5.2-11 Revision 13, October 2015 5.2.2.6 Applicable Codes and Classification
The structural integrity of the reactor coolant pressure boundary is maintained at the level required by ASME Section XI, 2007 Edition th rough 2008 Addenda for the fifth inservice inspection (ISI) interval.
[5.2-38]
5.2.2.7 Material Specification
Reactor coolant pressure boundary materials, including overpressurization protection materials, are discussed in Section 5.2.3.1.
5.2.2.8 Process Instrumentation
Process instrumentation is shown on P&ID M-13 (Unit 1) and P&ID M-60 (Unit 2).
5.2.2.9 System Reliability
Safety valve sizing (see 5.2.2.2.3) uses ve ry conservative assumptions on relief valve availability and method of scram. Further discussions on failures and their effects are
contained in Section 15.2.
5.2.2.10 Testing and Inspection
The relief valves and safety valves are inspecte d for cyclic strain and thermal stress. The safety valves are bench checked periodically for the proper setpoint. See Section 3.9 for inservice inspection and inservice testing (IST) of valves.
[5.2-39]
5.2.3 Reactor Coolant Pressure Boundary Materials
5.2.3.1 Material Specifications
The principal pressure retaining materials and the appropriate material specifications for the reactor coolant pressure boundary componen ts are defined in GE design and purchase specifications, or the spec ifications of other suppliers of RCPB components.
[5.2-40]
QUAD CITIES UFSAR 5.2-12 Revision 8, October 2005 5.2.3.2 Compatibility with Reactor Coolant The importance of establishing and maintainin g appropriate water chemistry conditions in the reactor coolant of boiling water reactor (B WR) nuclear power plants is well established.
During the past decade, most operating BWRs (including Dresden and Quad Cities Stations) have experienced unanticipated pipe cracking problems that have resulted in a significant loss of availability, and have incre ased the total personnel radiation exposure associated with inspection and repair. The c ause of these problems has been intergranular stress corrosion cracking (IGSCC) which resu lts from the simultaneous occurrence of an aggressive environment, particular materials, and stress conditions. A contributing cause of these problems has been the formation of locally corrosive environments as the result of
the ingress of impurities during operation.
Exelon Generation Company (EGC) recognizes that if IGSCC is to be controlled, the approp riate water chemistry must be maintained in the primary system of the company's BWR plants.
[5.2-41]
5.2.3.2.1 Boiling Water Reactor Water Chemistry
The BWR water chemistry control program establishes achievable ranges for water chemistry parameters where IGSCC is suppr essed. Compliance with the program's impurity concentrations has been shown to reduce the rate of IGSCC and reduce the probability of initiating new cracks. Data fr om Dresden Unit 2 indicates that IGSCC in reactor recirculation piping can be suppre ssed by controlling impurity concentrations within the achievable ranges combined with injection of approximately one ppm hydrogen into the feedwater to reduce free oxygen. This approach to the prevention of cracking is called hydrogen water chemistry (HWC).
In addition to reducing IGSCC, research shows that the appropriate control of water chemistry will also assist in controlling radiation buildup, minimizing fuel failure, and
minimizing damage to the turbine caused by chemistry.
Specific corporate water chemistry control requirements have been implemented at Quad Cities. These requirements reflect the curre nt understanding of the role of chemical transport, impurity concentration, materials of construction, corrosion behavior, chemical analytical methods, and industry practice re garding the operation and integrity of the primary system. The specific requirements were primarily taken from the BWR Owners Group (BWROG) and Electric Power Resea rch Institute (EPRI) Water Chemistry Guidelines, existing GE chemistry guide lines, and known or suspected contaminant concerns at EGC's BWRs. These specific re quirements are provided in system chemistry procedures.
Noble Metal Chemical Addition (NMCA) has been developed by General Electric as a method to enhance the effectiveness of Hydrog en Water Chemistry (HWC) in mitigating Intergranular Stress Corrosion Cracking (I GSCC) in vessel internal components.
Additionally, use of NMCA allows lower injection rates of HWC which in turn reduces plant radiation exposure over the life of the plant. NMCA process deposits a very thin
discontinuous layer of noble metals onto all we tted surfaces during the injection process.
The treated surfaces will behave cataly tically and promote oxidant-hydrogen recombination. This results in low corrosion potential of components at low hydrogen injection rates. Higher reacto r water conductivity is anticipated during the application due to the effect of non-corrosive noble metals on the measured conductivity (reference Table 5.2-4).
QUAD CITIES UFSAR Revision 5, June 1999 5.2-13 5.2.3.2.1.1 Training
A training program for personnel involved in water chemistry control is required to implement the corporate policy. The goal of this program is to improve the overall awareness among station personnel of the need for chemistry control. Personnel required to be trained include all chemistry staff, chemistry technicians, licensed operators, and selected systems engineering and maintenance personnel.
5.2.3.2.2 Compatibility of Construction Materials with Reactor Coolant The materials of construction exposed to th e reactor coolant consist of the following:
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A. Solution-annealed austenitic stainless steels (both wrought and cast) Types 304, 304L, 316, 316L, and XM-19, B. Nickel base alloys - Inconel 600 and Inconel 750X,
C. Carbon steel and low alloy steel,
D. Some 400 series martensitic stainless steel (all tempered at a minimum of 1100°F), and
E. Colmonoy and Stellite hardfacing material.
All of these materials of construction, with the possible exception of Inconel 600, are
resistant to stress corrosion cracking in the reactor coolant. General corrosion of these materials, except carbon and low alloy steel , is negligible. Conservative corrosion allowances are provided for all exposed su rfaces of carbon and low alloy steels.
Contaminants in the reactor coolant are contro lled to very low limits set by the reactor water quality specifications. No detrimental effects will occur on any of these materials from allowable contaminant levels in high puri ty reactor coolant. Radiolytic products have no adverse effects on these materials.
5.2.3.2.3 Compatibility of Construction Materi als with External Insulation and Reactor Coolant The materials of construction exposed to external insulation are:
[5.2-43]
- Solution annealed austenitic stainless steels (Types 304, 304L, and 316), and
- Carbon and low alloy steel.
QUAD CITIES UFSAR Revision 6, October 2001 5.2-14 Two types of external insulation are employed at Quad Cities. The first, reflective metal insulation, does not contribute to any surface contamination and has no effect on
construction materials. The second, nonmetallic insulation, is used on stainless steel piping and components and complies with the require ments of the following industry standards:
- ASTM C692-71, Standard Methods for Ev aluating Stress Corrosion Effects of Wicking Type Thermal Insulation on Stainless Steel (Dana Test); and
- RDT-M12-1T, Test Requirements for Ther mal Insulating Materials for Use on Austenitic Stainless Steel, Section 5 (KAPL Test).
Chemical analyses are required to verify that the leachable sodium, silicate, and chloride in this insulation are within acceptable levels. The insulation is packaged in waterproof
containers to avoid damage or contam ination during shipment and storage.
Since there are no additives in the reactor c oolant, leakage would expose materials to high purity, demineralized water. Exposure to de mineralized water would cause no detrimental effects.
5.2.3.3 Fabrication and Processing of Ferritic Materials
This subsection describes how Appendix G requires the determination of pressure -
temperature limits for ferritic materials to ac hieve acceptable stresses, and the adjustment of these limits to account for the effects of accumulated neutron irradiation.
5.2.3.3.1 Fracture Toughness - Reactor Pressure Vessel
Title 10 CFR Part 50, Appendix G, "Fractu re Toughness Requirements," requires that pressure-temperature limits be established for reactor coolant system heatup and cooldown operations, inservice leak and hy drostatic tests, and normal reactor operation. These limits are required to ensure that the stresses in th e RPV remain within acceptable limits. They are intended to provide adequate margins of safety during any condition of normal operation, including anticipat ed operational transients.
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Specific pressure-temperature limits are pres ented and discussed in Section 5.3.2 which indicate EGC's commitments regarding 10 CFR 50, Appendix G and Regulatory Guide 1.99, Revision 2.
5.2.3.3.2 Fracture Toughness - Reactor C oolant Pressure Boundary Minus Reactor Pressure Vessel The relief valves and the safety valves were exempted from fracture toughness requirements because Section III of the 1965 ASME Code did not require impact testing on valves with inlet connections of 6 inches or less nominal pipe size.
[5.2-45]
QUAD CITIES UFSAR Revision 5, June 1999 5.2-15 Main steam isolation valves were also ex empted from fracture toughness requirements because Section III of the 1965 ASME Code with Summer 1965 Addenda did not require brittle fracture testing on ferritic pressu re boundary components when the system temperature was in excess of 250°F at 20% of the design pressure.
The recirculation pumps were exempted from the ASME Code and the USAS Code for pressure piping because of their classification as machinery. This is more completely discussed in Section 5.4.1.1
5.2.3.3.3 Control of Welding
5.2.3.3.3.1 Control of Electroslag Weld Properties
Electroslag welding of longitudinal seams of the RPV was performed in accordance with ASME Section III, Code Case 1355. This code case and other code cases applying to materials and fabrication are identified in the vessel manufacturer's fabrication report and on the manufacturer's data report, Form N-1A. A detailed description of the electroslag welding process used on Quad Cities is contained in Amendment 13/14, Appendix F, Dresden FSAR
[2]. The electroslag welding process was not used in the fabrication of the recirculation pump casings.
[5.2-46]
5.2.3.4 Fabrication and Processing of Austenitic Stainless Steels
This section provides information relative to fabrication and processing of austenitic stainless steel for components in the RCPB.
5.2.3.4.1 Avoidance of Stress Corrosion Cracking
The methods and actions regarding stress corro sion cracking are discussed in subsequent sections.
5.2.3.4.1.1 Avoidance of Significant Sensitization
Sensitization of austenitic stainless steels duri ng fabrication can induce residual stresses that promote IGSCC.
The Quad Cities pressure vessels were manuf actured with some components of furnace-sensitized stainless steel material. In parti cular, vessel nozzle safe-ends were sensitized because they had been attached to the vessel prior to furnace annealing.
[5.2-47]
QUAD CITIES UFSAR Revision 12, June 2013 5.2-16 All the nozzles having furnace-sensitized safe-ends were replaced with non-sensitized components and are listed below as follows:
[5.2-48]
A. Recirculation inlet,
B. Recirculation outlet,
C. Core spray,
D. Six-inch instrumentation (top head),
E. Head vent,
F. Jet pump instrumentation,
G. Control rod drive hydraulic system re turn (safe-end removed and nozzle capped),
H. Core differential pressu re/standby liquid control, and
I. Two-inch instrumentation.
In each case, except for the two-inch instru ment nozzles, the replacement process left a portion of the sensitized stainless steel butteri ng on the nozzle when the sensitized safe-end was cut off. This buttering was overlay clad on the inside of the nozzle. The two-inch
instrument nozzles were inconel and the sa fe-end and entire weld were removed and replaced.
During Q2R21 refueling outage, a leak was discovered from the Unit 2 N-11B two-inch instrument nozzle during the system pressure test. A portion of the old nozzle was removed and a new nozzle was welded to the outside of the vessel. The repair is reconciled to the original code of construction.
The pipe socket on the inboard end of the co re differential pressure/standby liquid control nozzle was cut off and a new one formed by weld build-up and machining.
The upper steam dryer guide rod brackets had b een furnace-sensitized and were replaced.
The weld pads and surrounding cladding were overlay clad. Although the following equipment had not itself been furnace-sensit ized, the weld pads and surrounding cladding were overlay clad:
A. Steam dryer lower guide rod brackets,
B. Surveillance specimen holder brackets,
C. Core spray sparger brackets,
D. Steam dryer support brackets,
E. Feedwater sparger brackets,
F. Shroud head guide rod brackets, and
G. Jet pump riser brace.
QUAD CITIES UFSAR Revision 7, January 2003 5.2-17 In order to avoid partial and/or local sensitizat ion of austenitic stainless steel during heat treatments and welding operations for reactor core structural members and reactor coolant system pipe components, the control of heat input was carefully monitored and procedurally controlled with maximum interpass temperature limited to 350°F. Heat treating of reactor core structural members during manufacturing was limited to a maximum of 800°F. No
heat treating was permitted during field erection. General Electric quality control inspectors ensured conformance to approved pr ocedures both at the vendor's shop and at the site. The types of weld metal used for safe-ends of components within the RCPB are inconel and stainless steel.
5.2.3.4.1.2 Electroslag Welds (Regulatory Guide 1.34)
Refer to 5.2.3.3.3.1 for information on control of electroslag weld properties.
5.2.3.5 Intergranular Stress Corrosion Cracking
Generic Letter 88-01, "NRC position on IGSCC in BWR Austenitic Steel Piping", dated January 25, 1988, provides the NRC staff's positi ons and guidelines concerning the piping materials used for the reactor coolant pressure boundary. Subsequently, the final safety evaluation of EPRI Report TR-113932 ("BWR Vesse l and Internals Project, Technical Basis for Revisions to Generic Lette r 88-01 Inspection Schedules (BWRVIP-75)"), dated May 14, 2002, revised the Generic Letter (GL) 88-01 insp ection schedules. The BWRVIP-75 revised inspection schedules were based on consid eration of inspection results and service experience gained by the industry since issuance of GL 88-01, and includes additional knowledge regarding the benefits of improved BWR water chemistry. The information that follows is a point-by-point comparison of the requirements of the generic letter and the measures in place at Quad Cities. The numbe ring corresponds to the generic letter, and exceptions are indicated.
5.2.3.5.1 Programs to Mitigate IGSCC
Exelon Generation Company has an integrated program to mitigate IGSCC which includes the following:
[5.2-49]
A. Hydrogen water chemistry (HWC) (s ee Section 5.4.3 for information on HWC);
B. Stress improvement through induction heat stress improvement (IHSI) and mechanical stress improvement program (MSIP);
C. Weld overlays, including overlays with pipelocks, for flaw indications in excess of ASME Section XI, Subsection IWB-3500 limits (overlays meet
NUREG-0313, Rev. 2 requirements); and
D. System modifications, which include removal of the head spray line, removal and capping of the control rod drive re turn line, and replacement of reactor water cleanup piping with conforming material.
QUAD CITIES UFSAR 5.2-18 Revision 12, October 2013 5.2.3.5.2 Augmented Inspection Program
Exelon Generation Company's augmented in spection program conforms basically to positions on inspection schedules, methods and personnel, and sample expansion delineated in GL 88-01 and BWRVIP-75 as approved by the NRC. Exceptions are for welds that are not accessible for non-destru ctive examination (NDE).
5.2.3.5.3 NRC Notifications
Exelon Generation Company will notify the N RC for flaw indications that exceed IWB-3500 limits or changes in welds with flaw indicati ons following in-house determinations and/or recommendations.
5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary Inservice examination and tests of ISI Class 1, 2, 3, and MC. (See Section 3.9.6 for explanation of ISI Class vs. ASME Class. Only the RPV and skirt are ASME Class 1 in the RCPB). Components will be performed in acco rdance with Section XI of the ASME Code and applicable Addenda as required by 10 CF R 50.55a(g), except where specific written relief has been granted by the NRC. Certain requirements of later editions or addenda of Section XI are impracticable to perform on Q uad Cities because of the design, component geometry, and materials of construction. Fo r this reason, 10 CFR 50.55a(g)(6)(i) authorizes the NRC to grant relief from certain requireme nts after making the necessary findings.
[5.2-50]
During the 1990 refueling outage, cracks were discovered in the Unit 2 vessel head. A supplemental reactor head and upper shell inspection plan has been implemented for Unit 2 in lieu of the successive exami nation requirements of ASME Section XI.
The inservice testing of pumps and valves is discussed in Section 3.9.6. The ISI/IST of
Class 2, 3, and MC components is discussed in Section 6.6. This ISI program for Class 1, 2, and 3 is based on the requirements of Section XI of the ASME Code 2007 Edition through 2008 Addenda. The program for Class MC is based on the requirements of Section XI of the ASME Code 2001 Edition through the 2003 Addenda. Table 5.2-3 lists the systems included in the ISI program.
[5.2-51]
QUAD CITIES UFSAR Revision 6, October 2001 5.2-19 5.2.4.1 System Boundary Subject to Inspection
In addition to the RPVs and their support sk irts, components and supports within ASME Section III, Class 1 boundaries are subject to the requirements of ISI per ASME Section XI.
The P&IDs and IWE (MC) program drawings defi ne the applicability of the ISI Class 1, 2, 3, MC, and NC IST program for systems subje ct to the requirements of ASME Section XI.
These ISI boundaries on the P&IDs are limited to safety-related systems which contain water, steam or radioactive materials and, in accordance with Regulatory Guide 1.26, this includes some non-RCPB components.
[5.2-52]
Some pumps and valves not included in the IS I Class 1, 2 or 3 boundaries marked on the P&IDs have been included in the IST program in recognition of their importance to safe plant operation. These pumps and valves are noted as ISI Class "NC" meaning not classified ISI Class 1, 2, or 3 but having augmented quality requirements. Section 3.9 contains a discussion of ISI/IST for pumps and valves.
5.2.4.2 Arrangement and Accessibility
The "as built" Quad Cities design does not permit ready access for volumetric examination of RPV shell welds in accordance with the re quirements of ASME Section XI - Rules for Inservice Inspection of Nuclear Power Plant Components. Exelon Generation Company recognizes the importance of inspecting welds that are presently inaccessible and will study and implement, if practicable, new means to include these welds within the ISI program as
such means become available.
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5.2.4.3.1 Examination Techniques and Procedures The methods, techniques, and procedures used for ISI comply with subarticle IWA-2200 of ASME Section XI.
[5.2-54]
Liquid penetrant or magnetic particle methods will be used for surface examinations, and radiographic and/or ultrasonic methods w ill be used for volumetric examinations.
Periodic visual inspections are also made of no zzle-to-vessel weld joints to assure that no deterioration is occurring.
[5.2-55]
5.2.4.4 Inspection Intervals
As defined in subarticle IWA-2430 of ASME Section XI, the inspection interval for ISI Class I components will be 10 years. The interval may be extended by as much as one year to
permit inspections to be conc urrent with plant outages.
[5.2-56]
The inspection schedules are in accordance with IWB-2400, IWC-2400, IWD-2400, IWE-2400 for Class 1, 2, 3, and MC respectively.
It is intended that inservice examinations be performed during normal plant outages such as refueling shutdowns or maintenance shutdowns occurring during the inspection in terval. No examinations will be performed which require draining of the RPV or remova l of the core solely for the purpose of accomplishing the examinations.
QUAD CITIES UFSAR Revision 12, October 2013 5.2-20 A supplemental reactor head and upper she ll inspection plan for Unit 2 has been implemented in lieu of the successive examination requirements of ASME
Section XI IWB-2420(b).
5.2.4.5 Examination Categories and Requirements
The extent of the examinations performed and the methods utilized (e.g., volumetric, surface, visual) are in accordance with ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF 2500-1 for Class 1, 2, 3, and MC components and
component supports.
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5.2.4.6 Evaluation of Examination Results
The standards for examination evaluation meet the requirements of Section XI, IWB-3000, "Acceptance Standards." The program for flaw evaluation follows Table IWB-3410-1, "Acceptance Standards."
[5.2-58]
The program regarding repairs of unacceptable indications or replacement of components containing unacceptable indications meets the requirements of Section XI, IWA-4000, "Repair Procedures."
[5.2-59]
5.2.4.7 System Leakage and Hydrostatic Pressure Tests
System leakage and hydrostatic tests are co nducted in accordance with IWB-5000, "System Pressure Tests." Section 5.3.2.2 pr ovides additional requirements.
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5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary The reactor vessel head is flanged to the vesse l and sealed with two concentric O-rings. The area between the two O-rings is monitored to provide an indication of leakage from the inner O-ring seal.
[5.2-61]
The primary reactor coolant leak detection system consists of four different and independent means by which leakage can be detected. These are monitoring the:
[5.2-62]
A. Sumps in the containment;
B. Drywell temperature and pressure changes;
C. Drywell air coolers cooling wate r differential temperature changes; and
D. Drywell atmosphere activity level changes.
There were no startup (preop) tests to verify system operability or sensitivity, nor was the drywell subjected to leaks to determine response.
QUAD CITIES UFSAR Revision 12, October 2013 5.2-21 Once a leak has been detected within the dryw ell by any of the methods covered in this section, it becomes necessary to locate, and if possible, determine its magnitude and rate of change with time. The smaller the leak the more difficult it becomes in locating its source.
For example, through the use of a continuous air sample system, it is possible to detect changes in radioactive nuclides from one 24-hour period to the next. Very small leaks are thus possible to detect.
[5.2-63]
The systems described in this section would be used by the operator to find the source of leakage. The systems are remote in nature and provide the operator a method of cross checking to locate the source of leakage or the area in the drywell in which the leak has developed.
5.2.5.1 Containment Sumps
One method of detecting leakage from the prim ary coolant pressure boundary is monitoring the flow out of the drywell floor drain su mp and the equipment drain sump. All free unidentified leakage from the primary coolant pressure boundary will drain to the floor drain sump.
[5.2-64]
All controlled identified leakage (seals, etc.) is piped to the equipment drain sump and monitored separately. Therefore, the flow rate from the floor drain sump is the total
unidentified leakage of reactor coolant and is the principal leakage of concern from a safety standpoint.
The normal background flow out of the floor drain sump is quite low, approximately 1 gal/min. Therefore, leaks from the primary sy stem on the order of 1 to 3 gal/min, over the period of about 1 day, can be detected and me asured even when large allowances are made for variations in background leakage.
5.2.5.2 Drywell Temperature and Pressure
Temperature sensors are located at 34 different poin ts in the drywell. Six points feed into a 12-point recorder in the Control Room, and othe r 28 points feed into a 30-point recorder on the Drywell Environs Rack. The steady-sta te temperature in the drywell would be increased about 1°F for a 2 gal/min leak from the primary system.
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The drywell pressure is monitored and indica ted in the control room. The steady-state pressure in the drywell would be increased abo ut 0.03 psi for a 1 gal/min leak of reactor coolant.
5.2.5.3 Air Coolers Temperature Differential
Temperature sensors are installed in the rea ctor building closed cooling water discharge from each of the seven drywell air coolers.
The temperatures are recorded. The inlet water temperature of the reactor building closed c ooling water header to the coolers is also recorded. An increase in the differential temperature across the coolers of 5°F would
typically be equivalent to approximately a 3 ga l/min steam leak or a 7.5 gal/min liquid leak.
QUAD CITIES UFSAR Revision 12, October 2013 5.2-22 5.2.5.4 Other Drywell Leakage Monitors
The following systems are used by the operator to determine that leakage exists within the drywell. These various systems, operating to gether or singly, provide the information to the operator that a possible problem has developed within the drywell.
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5.2.5.4.1 Air Sample System
Each drywell is equipped with 22 air samplin g points, and 3 sample lines to the oxygen analyzer. Each suppression chamber has one sampling point. The air sample can be drawn
through 1/2 inch tubing from the various samp le points. Primary containment isolation is provided by either redundant manual or automat ic isolation valves. At the local rack, the samples may be filtered using a filter cartridge holder and the air returned to the drywell, or a grab sample may be taken at the rack for laboratory analysis. A continuous air monitor is provided from one of the oxygen analyzer points. This air monitor will count gross beta activity which will be recorded, and will alarm on an increase. This provides an indication that a leak has occurred.
5.2.5.4.2 Pressure Switches
A pressure switch will alarm if failure of the i nner O-ring takes place on the reactor vessel.
5.2.5.4.3 Acoustic Monitors
An acoustic monitor is mounted at each safety and relief valve (including the SRV) in the drywell, 13 in all. Leaks from these valves wo uld cause vibrations to be picked up by the monitor, sounding an alarm in the control room.
Each monitor has its own set of indicating lights in the control room.
5.2.5.5 Leakage Rate Limits
The limiting leakage conditions included in th e Technical Specifications are that with the plant in MODES 1, 2 and 3: there shall be no pressure boundary leakage; unidentified coolant leakage into the primary containment s hall be less than or equal to 5 gal/min; total leakage in the primary containment shall be le ss than or equal to 25 gal/min averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and unidentified leakage increases shall be less than or equal to 2 gal/min within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (in MODE 1).
[5.2-67]
In the event that the unidentified leakage lim it, or total leakage lim it is exceeded, the Technical Specifications allow 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to reduce the leakage to within limits. In addition, if the unidentified leakage increase limit is exc eeded, the Technical Specifications allow 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to reduce the leakage increase to within limits or to identify that the source of the unidentified leakage increase is not IGSCC susc eptible material. In the event that the required actions cannot be performed within th e 4-hours period, or if pressure boundary leakage exists, the affected unit must be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
QUAD CITIES UFSAR Revision 12, October 2013 5.2-23 In addition, a limit is procedurally administ ered for the volume pumped from the drywell equipment drain sump (identified leakage). This limit is 9600 gallons in an 8-hour period
(20 gal/min).
5.2.5.6 High/Low Pressure Interfaces
The core spray, (RHR)/LPCI, and the RHR shut down cooling suction are all monitored for reactor coolant system leakage into the system by pressure switches located in the pump discharge lines. These switches activate a high pressure alarm in the main control room
when the line pressure exceeds the alarm setp oint. These lines are protected by relief valves which are bench tested at least every third refueling outage.
[5.2-68]
5.2.5.7 Compliance With Regulatory Guide 1.45
The various leak detection systems and cap abilities, as described herein, detect RCPB leakage, both identified and unidentified.
These sensitive and diverse systems meet the acceptance criteria of Regulatory Guide 1.45.
[5.2-69]
5.2.6 Detection of Leakage Beyond th e Reactor Coolant Pressure Boundary Provisions have been made in the design of Quad Cities to detect leakage from vital
fluid-carrying systems beyond the limits of the RCPB. Included in these provisions, which are discussed below, are floor drain sumps, area radiation monitoring, area temperature monitoring, and visual inspection.
[5.2-70]
5.2.6.1 Floor Drain Sumps
Floor drain sumps and pumps are provided within the secondary containment (reactor building). Leakage from fluid-carrying sy stems would be detected by an increased frequency of sump pump operation, increased input to the radwaste floor drain collector system, or high water level in the sump with ultimate annunciation in the control room.
5.2.6.2 Area Radiation Monitoring
Area radiation monitors are provided througho ut the plant equipment and operating areas.
These monitors can detect leakage from radioactive sources. Activity levels are indicated in the control room. Leakage would be detected by an increased level of activity beyond normal operating background with ultimate hi gh activity annunciation in the control room.
Further information on area radiation mo nitoring is contained in Section 12.3.
QUAD CITIES UFSAR Revision 6, October 2001 5.2-24 5.2.6.3 Area Temperature Monitoring
Area temperature monitors are provided in ap propriate areas and equipment spaces of the plant. These monitors will detect leakage from high temperature fluid-carrying systems.
Temperature indication is provided in the co ntrol room. Leakage would be detected by an increase in the normal operating temperature of the area with ultimate high temperature annunciation in the control room and, in some cases, automatic isolation of the system.
Systems provided with automatic isolation on detection of high area temperatures are HPCI, RCIC, RWCU and the MSIVs.
[5.2-71]
5.2.6.4 Visual Inspection of Equipment and Operating Areas
Access to equipment spaces to permit normal ro utine visual inspection is provided to those areas of controlled occupancy as well as those of continuous occupancy, radiation levels permitting.
[5.2-72]
QUAD CITIES UFSAR Revision 10, October 2009 5.2-25 5.2.7 References
- 1. NEDE-24988-P, "Analysis of Generic BW R Safety/Relief Valve Operability Test Results," General Electric.
- 2. Dresden 2 and 3 FSAR, Amendment 13/14, Appendix F.
- 3. Deleted. 4. Deleted.
- 5. GE-NE-A22-00103-10-01, "Dresden and Quad Cities Extended Power Uprate, Task T0900: Transient Analysis, Revision 0."
- 6. "Reference Safety Report for Boiling Water Reactor Reload Fuel," Westinghouse Topical Report CENPD-300-P-A (Proprietary), CENPD
-300-NP-A (Non-Proprietary), July 1996.
(Sheet 1 of 1)
Revision 6, October 2001 QUAD CITIES - UFSAR
Table 5.2-1
SUMMARY
OF STRESSES ON RELIEF VALVE PARTS FOR UNIT ONE
Relief Valve Part Allowable Stress at Design Temperature Maximum Stress 1. Valve Body
1.0S = 18,760 psi 1.5S = 28,140 psi m = 1,533 psi* m + b = 22,954 psi* 2. Turnbuckle 1.0S yield = 28,100 psi = 15,332 psi 3. Pilot Valve Tube 1.0S yield = 23,580 psi = 22,115 psi 4. Solenoid Assembly Mounting Bracket Hold-down Bolts 1.0S yield = 105,000 psi = 3,036 psi 5. Solenoid Assembly Mounting Bolts 1.0S yield = 105,000 psi = 3,584 psi
- Note that the ERV valve body stresses were originally qualified by enveloping the existing Dresden and Quad Cities ERV pipe loadings. The numbers in this table
reflect this original condition. The pr esent ERV valve body stress is qualified based on the qualification of the ERV pipe stress.
See UFSAR Section 3.9 for more details on the current pipe stress summary.
(Sheet 1 of 1)
Revision 8, October 2005 QUAD CITIES - UFSAR
Table 5.2-2
FORCES AND STRESSES IN SUPPORTING STRUCTURE AT QUAD CITIES 1 & 2 (HISTORICAL)
Member Axial (kip) Shear (kip) Moment (kip-ft)
Major Minor Axis Axis Maximum Fiber Stress (ksi)
Shear Stress (ksi)
A B Target Rock SRV Line: ** Beam 906-903
8WF58 0.5 20.0 123.020.541.954.48 1.2260.971 Beam 601-610 12FW45 1.5 8.0 57.418.229.561.97 0.8640.684 Beam 400-406 12WF36 2.1 1.0 17.03.710.340.27 0.3020.239 Beam Column A900-802 6WF25 17.0 13.0 20.50.518.116.38 0.5290.419 Electromatic Valve Lines
- Beam 906-903
8WF58 0.5 16.0 98.717.334.23.59 1.0000.792 Beam 601-610 12WF45 2.2 6.5 48.515.725.41.61 0.7420.588 Beam 400-406 12WF36 2.0 1.5 13.73.79.50.40 0.2770.220 Beam Column A900-802 6WF25 13.5 10.5 16.70.214.35.15 0.4180.331
____________________________
A = Shear Stress/(0.95)(F y) B = Shear Stress/F y*
- These values are being retained for historical purposes only. Refer to UFSAR Section 3.9 for current SRV/ERV pipe support stress summaries.
(Sheet 1 of 1)
Revision 5, June 1999 QUAD CITIES - UFSAR
Table 5.2-3
LIST OF SYSTEMS INCLUDED IN THE ISI PROGRAM
System Class Control Rod Drive 1 & 2 Residual Heat Removal (RHR) 1 & 2 RHR Service Water 3 Standby Liquid Control (SBLC) 1 & 2 Reactor Water Cleanup 1 Core Spray 1 & 2 High Pressure Coolant Injection (HPCI) 1 & 2 Main Steam 1 Feedwater 1 & 2 Diesel Generator Cooling Water 3 Reactor Recirculation 1 Reactor Core Isolation Cooling (RCIC) 1 Control Room HVAC 3
Drywell MC Suppression Chamber MC Vent System MC
(Sheet 1 of 1) Revision 7, January 2003 QUAD CITIES -UFSAR
Table 5.2-4
REACTOR COOLANT SYSTEM CHEMISTRY LIMITS OPERATIONAL MODE(s)
Chlorides Conductivity (µmhos/cm @ 25 o C) 1 < 0.2 ppm <
1.0 2 and 3 < 0.1 ppm <
2.0**
- During Noble Metal Chemical Addition (NMCA), <=10.0
µmhos/cm @ 25 o C is the limit
QUAD CITIES - UFSAR Revision 6, October 2001 5.3-1 5.3 REACTOR VESSELS
This section presents pertinent data on the Q uad Cities reactor pressure vessels (RPVs).
Unless otherwise noted, the information presente d applies to both Unit 1 and Unit 2 RPVs.
5.3.1 Reactor Vessel Materials
The RPV materials and fabrication methods co nform to the ASME Boiler and Pressure Vessel Code (ASME Code) 1965 Edition and th e Summer 1965 Addendum as referenced in Section3.2.8.4. Inservice inspection (ISI) techniques conform to ASME Section XI with approved exceptions as noted in Section 5.2.4.
[5.3-1]
5.3.1.1 Material Specifications
Reactor vessel material specifications are di scussed in Section 5.2.3.1. Additional information on RPV materials is contained in Section 5.3.3.2.
5.3.1.2 Special Processes Used for Manufacturing and Fabrication
The Quad Cities Unit 1 RPV was fabricated entirely in the United States by Babcock &
Wilcox (B&W). The Unit 2 RPV was fabricated by several different vendors, including one in Holland, as noted in the following paragraphs.
[5.3-2]
Fabrication work on the Unit 2 bottom he ad assembly and lower shell course was performed by the Rotterdam Dockyard Company (RDM) in Rotterdam, Holland. These two pieces were seam-welded together and returned to the United States as a fully completed subassembly including control rod drive (CRD) stub tubes, shroud support skirt, and vessel support skirt.
[5.3-2a]
The CRD stub tube material is Inconel SB167, Code Case 1336, Paragraph 1. The stub
tubes were joined to the vessel bottom by a we ld on the Inconel-clad surface which makes a full penetration of the stub tube wall as specified in Figure N-462.4(e) of the ASME Code, 1965,Section III. The toe of this weld was removed by the finished counterbore.
All work on Unit 2 was performed and docume nted in accordance with ASME Section III.
The procedures required by the attachment to the National Board of Boiler and Pressure Vessel Inspectors' letter of July 24, 1968, were implemented by providing the Illinois State Board of Boiler Rules with the required docu mentation. This documentation included copies of all welder qualification test reports and performance test reports for each welder.
All other components of the Unit 2 core inte rnals and primary system were of domestic manufacture. For example, B&W completed the circumferential seam weld which attached the upper shell course to the RPV flange.
QUAD CITIES - UFSAR 5.3-2 Chicago Bridge and Iron (CB&I), which complete d fabrication of the Unit 2 RPV prior to its shipment to the plant site, provided a certifi cation comparable to the ASME Code N-1A form. The following footnote was included in that certification:
[5.3-3]
"This unstamped vessel was built as a `State Special' based on agreements between the State of Illinois and Commonwealth
Edison Company. A portion of the vessel was fabricated by
Rotterdam Dockyard Company.
This vessel was not stamped because Rotterdam Dockyard Company does not hold an ASME
certificate of authorization.
Procedures equivalent to the requirements of the ASME Code were used."
Electroslag welding of longitudinal seams of the RPV was performed by B&W in accordance with ASME Section III, Code Case 1355 (See Section 5.2.3.3.3.1 for further details).
[5.3-4]
5.3.1.3 Special Methods for Nondestructive Examination
Standard methods, in use at that time, were used for nondestructive examinations except for inspection of the CRD stub tubes as explained in the following paragraphs.
Inspection of the CRD stub tube shop welds was accomplished by progressive and final dye penetrant inspection and by ultrasonic (UT) insp ection from the finished counterbore, all as required by ASME Section III, Paragraph N-462.4(e). The UT inspection exceeded the
ASME Code requirements in that it covered th e weld metal in addition to the base metal, heat-affected zone, and weld cla dding. Also, the sensitivity used for UT testing was "high gain" and exceeded the ASME Code requirements.
[5.3-5]
A similar "high gain" UT test was applied to the CRD stub tube field welds in addition to progressive dye penetrant testing.
5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels
Regulatory Guides, as such, did not exist at th e time the Quad Cities RPVs were fabricated.
Information related to specific Regulatory Gu ides (as requested in Regulatory Guide 1.70, Rev. 3) is provided below to correlate actual past practice with current requirements.
Unless otherwise stated, there has been no commitment to the Regulatory Guides.
Electroslag welding of longitudinal seams wa s performed in accordance with ASME Section III, Code Case 1355 as discussed in Section 5.2.3.3.3.1.
[5.3-6]
Section 5.2.3.4.1.1 discusses the use of sensitized stainless steel and the actions to
remove/control sensitized components.
QUAD CITIES - UFSAR 5.3-3 Revision 8, October 2005 Regulatory Guide 1.50
Preheat temperatures used when welding lo w alloy steel components (shells, flanges, plates) met applicable requirements or had con tract variations approved by GE, the vendor responsible for supplying the RPV.
[5.3-7]
Section 5.3.2.1 contains information on comp liance with the methodology in Regulatory Guide 1.99, Rev. 2.
Regulatory Guide 1.190 Regulatory Guide (RG) 1.190 provides stat e of the art calculation and measurement procedures that are acceptable to the NRC fo r determining Reactor Pr essure Vessel neutron fluence. RPV fluence has been evaluated using a method in accordance with the recommendations of RG 1.190. Future evaluat ions of RPV fluence will be completed using a method in accordance with the recommendatio ns of RG 1.190 (as noted in Reference 3).
5.3.1.5 Fracture Toughness
Sections 5.2.3.3.1 and 5.3.2.1 describe fractu re toughness provisions for the Quad Cities Units 1 and 2 RPVs.
5.3.1.6 Material Surveillance
Vessel material surveillance samples are lo cated within the reactor vessel to enable periodic monitoring of changes in material pr operties with exposure. The samples include specimens of the base metal, weld zone meta l, heat affected zone metal, and standard specimens. These specimens receive neutro n exposures more rapidly than the vessel wall material of interest (i.e., the innermost 25% of vessel wall thickness) and therefore lead it in integrated neutron flux. The neutron exposu re rate of the average specimen at the core midplane is approximately 1.2 times the exposure rate of the adjacent inside surface of the vessel wall. There were 401 samples initially inserted in the vessel. Table 5.3-1 provides the location and status of the material specimens.
[5.3-8]
In 2003, the NRC approved Quad Cities participation in the BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Pr ogram (ISP) as described in BWRVIP-78 and BWRVIP-86 in Reference 2. The NRC approved the ISP for the industry in Reference 2 and approved Quad Cities participation in Refere nce 3. The ISP meets the requirements of 10 CFR 50 Appendix H and provides several adv antages over the original program. The surveillance materials in many plant-specific programs do not represent the best match with the limiting vessel beltline materials sinc e some were established prior to 10 CFR 50 Appendix H requirements. Also, the ISP a llows for better comparison to unirradiated material data to determine actual shifts in toughness. Finally, for many plants, ISP data will be available sooner to factor into plant oper ations since there are more sources of data.
The current withdrawal schedule for both units is based on the NRC-approved revision of BWRVIP-86 (Reference 2). Based on this sc hedule, Quad Cities is not scheduled to withdraw an additional material specimen.
[5.3-9]
QUAD CITIES - UFSAR 5.3-4 Revision 8, October 2005 5.3.1.7 Reactor Pressure Vessel Fasteners
The top head of the RPV is secured to the vesse l with studs, nuts, and spherical washers.
Nut torquing and detorquing is accomplis hed using a stud tensioner. Technical Specifications require that the RPV head bo lting studs (or closure studs) not be under tension unless the metal temperature of th e vessel shell immediately below the vessel flange is at or above 83°F. This value (83° F) comes from the reference temperature (RT NDT) and ASME Code considerations as discussed in Section 5.3.2.
[5.3-10]
A fatigue usage analysis dated May 1999, demons trated that the cumulative fatigue usage factor (CFUF) for the RPV closure studs woul d remain below 1.0 for current forty-year design life. A previous fatigue usage analys is dated March 1990, using then current duty cycle values, originally predicted that the RP V closure studs would reach the CFUF limit of 1.0 in 1998. This prediction was recalculated using actual cycle data through November 1997, to demonstrate that the CFUF limit would be reached in 2002. The purpose of the
May 1999 analysis was to reduce conservatism used in the original vessel closure stud analysis and to qualify the studs for a forty-year design life using an updated fatigue evaluation. The primary means of reducing the fatigue usage was to use the actual number
of operating cycles and perform new cycle pa iring based on stress ranges and number of occurrences. Further reduction in fatigue us age was accomplished by using the appropriate ASME Code fatigue curve of 2.7Sm versus 3Sm.
The results of this analysis show that the CFUF for the vessel closure studs is less than 1.0 for both Units 1 and 2 at the end of the forty-year design life. Further analysis was performed in 2003 in support of flood-up of the RPV using Feedwater/Condensate resulting in additional limitation on the number of several vessel stress cycles as described in Tabl e 3.9-1A. The results of these analyses show that the usage factor meets the allowable limit of 1.0 established in the ASME Section III Code and as a result justifies forty years of operation.
[5.3-11]
5.3.2 Pressure - Temperature Limits
Fast (>1 MeV) neutron irradiation above 10 17 nvt begins to affect the mechanical properties of ferritic steel. The most important considerat ion is that of the change in the temperature at which ferritic steel breaks in a brittle rath er than a ductile mode (referred to as the Nil Ductility Transition Temperature or NDTT ). The NDTT increases with increasing
irradiation. ASME Section III, N-446 specifies the design conditions for determination of the NDTT. Extensive tests have established the magnitude of changes in the NDTT as a function of the integrated neutron dosage.
[5.3-12]
The SA 302B steel, with fabrication procedures specified by the ASME Code and by GE, is relatively insensitive to neutron irradiation.
In fact, no change in the Adjusted Reference Temperature (ART) is expected to occur at neutron exposures less than 4.0 x 10 17 nvt.
Originally, the flux levels were calculated using a modified Albert-Welton point kernel[1] which was originally developed as an approx imate method of calculating the attenuation from a point fission source in water. The method represents fast neutron attenuation in water by a function which was experimenta lly fitted to data obtained for neutron attenuation by hydrogen in water. The no nhydrogenous portion of the attenuation was approximated by energy independent removal cro ss sections. The attenuation coefficients were obtained by fitting the kernel to the Oak Ridge Bulk Shielding Reactor water centerline data. This QUAD CITIES - UFSAR 5.3-5 Revision 9, October 2007 method was incorporated into a computer progra m that integrated the contribution of discrete source points in the reactor volume to ea ch point where flux was to be calculated.
The form of the Albert-Welton point kernel was developed to calculate dose rate. To obtain flux densities using this kernel, the normalization constant, was converted to The conversion was made by normalizing the number of fission neutrons above 1 MeV per watt in the fission spectrum to rep per watt of thermal power using the Hurst dosimeter response curve to obtain dose rates from neutron flux.
The intensity of the discrete source points used to describe the reactor volume was determined by a computer program using power functions fitted to the gross radial and axial fission distributions. The absolute power yielde d by integrating these points was normalized to the peak reactor thermal power of 2511 MWt.
The nonhydrogenous removal cross sections used in the calculations were taken from "Effective Removal Cross Sections for Shielding," G. T. Chapman and C. L. Storrs, Oak Ridge National Laboratory AECD 3978. The values used were:
UO 2 = 0.100 cm
-1 Zr = 0.100 cm
-1 Fe = 0.168 cm
-1 The value for UO 2 was reduced from 0.110 to 0.100 whic h results in some conservatism.
Attenuation in the water region was included in the point kernel. This attenuation is controlled by the nonhydrogenous oxygen removal cross section and the relative density of the water regions. The oxygen remo val cross section was taken as r = 0.033 cm
-1. The water densities used for the core and shield regions were consistent with core and coolant flow analysis.
The projected end-of-life fluences include data from the 10 CFR 50 Appendix H metal
surveillance capsules removed from the RPVs with neutron fluences representative of approximately 1/4 of RPV life. These projected peak fluences for the Quad Cities RPVs range from 3.5 x 10 17 to 4.9 x 10 17 nvt. [5.3-13] More recently, the NRC issued Regulatory Guid e (RG) 1.190, which provides state of the art calculation and measurement procedures that are acceptable to the NRC for determining Reactor Pressure Vessel neutron fluence. Quad Cities RPV fluence has been evaluated using a method in accordance with the recommendations of RG 1.190. Future evaluations of RPV fluence will be completed using a method in a ccordance with the recommendations of RG 1.190 (as noted in Reference 3).
watt hr-rep m 10 x 7.29 = 9 (5.3-1) watt sec--cm neutrons 10 x 1.67 = 2 11 (5.3-2)
QUAD CITIES - UFSAR Revision 9, October 2007 5.3-6 5.3.2.1 Limit Curves
The reactor vessel is a primary barrier agains t the release of fission products to the environs. In order to provide assurance that th is barrier is maintained at a high degree of integrity, pressure-temperature (P-T) limit s have been established for the operating conditions to which the reactor vessel can be subjected. Figures 5.3-4a through 4c present the P-T curves for those operating conditions
- Pressure Testing (Curve A), Non-Nuclear Heatup/Cooldown (Curve B), and Core Critical Operation (Curve C). These curves have been established to be in conformance with Appendix G to 10 CFR 50 and Regulatory Guide 1.99, Revision 2, and take into account the change in NDTT as a result of neutron
embrittlement. The adjusted reference temper ature (ART) of the limiting vessel material is used to account for irradiation effects. In addition, the NRC has approved an exemption to 10CFR10.60(a), "Acceptance criteria for fractu re prevention measures for lightwater nuclear power reactors for normal operatio n." The approved exemption allows the application of ASME Code Case N-588 and AS ME Code Case N-640 in the development of the P-T curves described below.
[5.3-14]
5.3.2.1.1 Beltline, Nonbeltline, and Closure Flange Regions
Four vessel regions are considered for the de velopment of the P-T curves: 1) the core beltline region, 2) the nonbeltline region (other than the closure flange region and the
bottom head region), 3) the closure flange re gion, and 4) the bottom head region. The beltline region is defined as that region of the reactor vessel that directly surrounds the effective height of the reacto r core, and is subject to an RT NDT adjustment to account for radiation embrittlement. The nonbeltline, cl osure flange, and bottom head regions receive insufficient fluence to necessitate an RT NDT} adjustment. These regions contain components which include the reactor vessel nozzles, cl osure flanges, top and bottom head plates, control rod drive penetrations, and shell plat es that do not directly surround the reactor core. Although the closure flange and bottom he ad regions are nonbeltline regions, they are treated separately for the development of th e P-T curves to address 10 CFR 50 Appendix G requirements.
Boltup Temperature
The limiting initial RT NDT of the main closure flanges, the shell and head materials connecting to these flanges, connecting welds, and the vertical electroslag welds which terminate immediately below the vessel flange, is 23°F. Therefore, the minimum allowable boltup temperature is established as 83°F (RT NDT + 60°F), which includes a 60°F conservatism required by the orig inal ASME Code of Construction.
Curve A- Pressure Testing
As indicated in Curve A of (Figure 5.3-4a) for pressure testing, the minimum metal
temperature of the reactor vessel shell is 83°F for reactor pressures less than 312 psig. This 83°F minimum boltup temperature is based on an RT NDT of 23°F for the electroslag weld immediately below the vessel flange and a 60°F conservatism required by the original ASME Code of Construction. The bottom head region limit is established as 68°F, based on moderator temperature assumptions for shutdown margin analyses.
At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 113°F. The 113°F minimum temperature is based on a closure flange region
RT NDT of 23°F and a 90°F conservatism required by 10 CFR 50 Appendix G. The P-T limits for pressure testing are valid to 54 effective full power years (EFPY).
QUAD CITIES - UFSAR Revision 6, October 2001 5.3-7 Figure 5.3-4a is governing for applic able pressure testing with a maximum heatup/cooldown rate of 20°F/hour.
Curve B - Non-nuclear Heatup/Cooldown
Curve B of Figure 5.3-4b applies during heat ups with non-nuclear heat (e.g., recirculation pump heat), and during cooldowns when the re actor is not critical (e.g., following a scram).
The curve provides the minimum reactor vesse l metal temperatures based on the most limiting vessel stress. The maximum heatup/cooldown rate of 100°F/hour is applicable.
Curve C - Core Critical Operation
CurveC, the core critical operation curve shown in Figure 5.3-4c, is generated in accordance with 10 CFR 50 Appendix G which requires core critical P-T limits to be 40°F above any pressure testing or non-nuclear heatup/cooldown limits.
The actual shift in RT{NDT} of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-82 and 10 CFR Part 50, Appendix H, irradiated reacto r vessel material specimens installed near the inside wall of the reactor vessel in the core area. The irradiated specimens are used in
predicting reactor vessel material embrittl ement. The operating limit curves of Figures 5.3-4a through 4c, Curves A throughC, s hall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.
5.3.2.2 Operating Procedures
Pressure-temperature limit curves in the Tec hnical Specifications are established to the requirements of 10 CFR 50, Appendix G, to a ssure that brittle fracture of the RPV is prevented. Further description of these limit curves is in Section 5.3.2.1.
[5.3-15]
10 CFR 50 Appendix G stipulates, "Pressure test s and leak tests of the reactor vessel that are required by Section XI of the ASME Code mu st be completed before the core is critical." However, pressure testing of Class 1 piping components following non-welded repair/replacement of non-RPV related Class 1 piping components may be conducted with the reactor core critical when th e following conditions are met.
- A valid Class 1 periodic pressure test wh ich meets the requirements of ASME Section XI Table IWB-2500-1, Examination Category B-P exists for the current operating cycle.
- The replacement activities are performed in accordance with a controlled work process.
A system leakage test at operating pressure is performed on the primary system following each removal and replacement of the RPV head. The system is checked for leaks and
abnormal conditions which are then corrected before reactor startup. The minimum RPV temperature during the system leakage test is in accordance with Figure 5.3-4a.
The reactor coolant system was given a system hydrostatic test in accordance with ASME Code requirements prior to init ial reactor startup. Before pressurization the system was heated to NDTT +60°F. Piping and support hangers were checked while thermal expansion was in progress. Recirculation pump operation was also checked.
QUAD CITIES - UFSAR 5.3-8 5.3.3 Reactor Vessel Integrity
Section 5.3.3.1 summarizes the RPV's purpose and the factors that contribute to RPV
integrity.
The following vendors participated in the desi gn and/or fabrication of the Quad Cities Unit 2 RPV:
[5.3-16]
A. Babcock and Wilcox was a supplier to GE for:
- 1. All fabrication of the bottom head, including the first cylindrical shell course with nozzles and CRD housings, vessel sk irt, and internal shroud support.
- 2. All remaining shell courses with no zzles attached. No circumferential seams were welded by B&W except the upper shell course to the vessel flange.
- 3. The complete closure head.
B. Rotterdam Dockyard Company, Rotterd am, Holland as a subcontractor to B&W, completed the work under A.1. above fo r B&W. Rotterdam Dockyard Company installed all stub tubes and welded the bo ttom head to the lower cylindrical shell course.
C. Chicago Bridge and Iron, Memphis, Tennessee completed all the remaining work as a contractor to GE. This work includ ed such items as welding and post weld stress relief of circumferential seams, hydr otesting and post hydro examination.
Babcock and Wilcox, RDM and CB&I's quality a ssurance organizations were engaged in their respective work. In addition, B&W aud ited RDM's performance. General Electric Company audited all three vendors' performance, engaging their Quality Control Procured Equipment organization for all work performe d within the U.S.A., and using the European based GE Technical Services Company (GETSC O) quality control organization as an agent for the RDM fabrication. Hartford had the re sponsibility for third party inspection at B&W, RDM and CB&I and signed both the partial data reports and the N-1A form.
Documentation was provided by B&W to dire ct RDM as to the remaining fabrication and testing operations to be performed.
General Electric Company's quality control organization audited this documentation.
Documentation regarding material and status was also provided by B&W to GE for all components shipped to CB&I. After review, GE forwarded this information to CB&I.
Records for the B&W and RDM fabrication are located at B&W. Records for the CB&I
fabrication are located at CB&I.
QUAD CITIES - UFSAR Revision 6, October 2001 5.3-9 5.3.3.1 Design
5.3.3.1.1 General Parameters
The purpose of the RPV is to support and cont ain the reactor core, th e reactor internals, and the reactor core coolant-moderator and to serve as a high integrity barrier against leakage of radioactive materials to the drywell. To achieve these purposes, the RPV was
designed using the following general parameters:
[5.3-17]
A. Design pressure 1250 psig
B. Nominal operating pressure 1005 psig
C. Base metal SA-302 Grade B in accordance with Code Case 1339 (RPV SHELL)
D. Cladding Weld deposited Type ER308L electrode
E. Design codes ASME B & PV Code Sec. III, Class A 1965 Edition and Summer 1954 Addendum
The nominal operating pressure of 1005 psig was based upon economic analyses for boiling
water reactors. The design pressure of 1250 ps ig was determined by an analysis of margins required to provide a reasonable range for mane uvering during operation, with additional allowances to accommodate transients above the operating pressure without causing operation of the safety valves.
The strength required to withstand external and internal loads, while maintaining a high degree of corrosion resistance, dictated the us e of a high-strength low alloy steel SA-302, Grade B, with an internal cladding of Type ER 308 stainless steel applied by weld overlay.
The reactor vessel was designed for a 40-year op erational life. During this period, it will not be exposed to more than 10 19 nvt of neutrons with energies exceeding 1 MeV.
ASME Section III, Class A, pressure vessel desi gn criteria provide assurance that a vessel designed, built, and operated within its desi gn limits has an extremely low probability of failure due to any known failure mechanism.
5.3.3.1.2 Specific Criteria
The specific stress limit criteria of the re actor coolant pressure boundary for loading combinations of operating loads plus maxi mum earthquake load, and operating loads plus maximum earthquake load plus loads resulting from a design basis accident (DBA), are
discussed below.
[5.3-18]
For the RPVs:
A. Stress intensities do not exceed ASME Section III (1965 Edition and Summer 1965 Addendum) allowable stress intens ity limits for Design Loads.
B. Primary membrane stresses do not ex ceed 90% of the yield strength of the material.
QUAD CITIES - UFSAR Revision 5, June 1999 5.3-10 For the CRD housings and the incore monitor housings, the additional stresses caused by the DBA are very small. The stress limits fo r the combination of operating loads plus maximum earthquake loads are therefore controllin
- g. For this condition, the stress limit does not exceed 1.5 times the hot allowable stress (1.5 S m). [5.3-19]
For the jet pump instrumentation penetration seal, the stresses caused by the maximum earthquake and the DBA are very small. The stress limits for operating conditions, as specified in ASME Section III, "Nuclear Vessels
," are therefore used as the limits for these accident conditions.
[5.3-20]
5.3.3.1.3 Temperature and Pressure Cycles
See Section 3.9 for RPV temperature and pressure cycles information.
5.3.3.1.4 Dynamic Loads
For the loading case consisting of operating loads plus maximum earthquake loads, the
stresses in the reactor vessel, the support ski rt, and the internal components which support and position the core are within limits which assure essentially elastic behavior. Use of these stress limits results in no gross deformat ion of parts which could affect control blade insertability.
[5.3-21]
5.3.3.2 Materials of Construction
The reactor vessel is a vertical cylindrical pre ssure vessel as shown in Figure 5.3-5. The RPV shell base plate material is high-streng th low alloy steel SA-302, Grade B, in accordance with Code Case 1339. The vessel in terior is clad with weld deposited ER308L stainless steel electrode. The main steam outl et lines are from the vessel body, below the reactor vessel flange.
[5.3-22]
The reactor vessel was designed and built in accordance with ASME Section III, Class A.
General Electric Company specified additio nal requirements. Records of material properties were developed and are retained for future evaluation of the RPV over its operating lifetime.
[5.3-23]
The CRD housings and the incore instrumentat ion thimbles are welded to the bottom head of the reactor vessel. The incore flux monito r housings are made of Type 304 stainless steel and designed to ASME Section III.
[5.3-24]
The RPV is supported by a steel skirt welded to the bottom of the vessel.
A preservice inspection of the components was conducted after site ere ction to assure that the RPVs were free of gross defects and to pr ovide a reference base for later inspections.
The ISI programs provide for continuing inspe ctions during refueling outages. See Section 5.2.4 for discussion of the ISI of the RPVs.
[5.3-25]
QUAD CITIES - UFSAR Revision 6, October 2001 5.3-11 5.3.3.3 Fabrication Methods
Sections 5.3.1.2 and 5.3.3 provide informatio n on the fabrication methods for the Quad Cities RPVs.
5.3.3.4 Inspection Requirements
As required by the ASME Code, the reactor cool ant system was hydrostatically tested prior to initial criticality. Hydrostatic tests are also performed after any modification to the system. The hydrostatic test pressure and te sting conditions are detailed in ASME Section XI, and the USAS B 31.1 Code for Pressure Piping.
[5.3-26]
The reactor vessel for Unit 1 is stamped with a Code N symbol which signifies that the hydrostatic test and all other required inspe ction and testing has been satisfactorily completed, final certification has been issu ed, and all applicable ASME Code requirements have been met.
5.3.3.5 Shipment and Installation
Several shipments of vessel components, as well as shipment of the completed RPV, occurred during the fabrication process. As noted in Section 5.3.3, several vendors in various locations participated in RPV fabricat ion. General Electric QA assured that all shipments and installation met appropriate regulations and requirements.
[5.3-27]
5.3.3.6 Operating Conditions
Section 5.3.2.1 specifies the operating conditio ns used to show conformance to Regulatory Guide 1.99, Rev. 2.
5.3.3.7 Inservice Surveillance
Section 5.2.4 summarizes the inservice surveillanc e or ISI for the Quad Cities Units 1 and 2 RPVs.
QUAD CITIES - UFSAR 5.3-12 Revision 8, October 2005
5.3.4 References
- 1. "Reactor Handbook," 2nd Edition, Vol. III Part B, Shielding, pages 72 and 80.
- 2. BWRVIP-86-A: "BWR Vessel and Inter nals Project, Updated BWR Integrated Surveillance Program (ISP)," Final Report, October 2002.
- 3. C. F. Lyon letter to J. L. Skolds, "Quad Cites Nuclear Power Station, Units 1 and 2 -
Issuance of Amendments Re: Reactor Ve ssel Specimen Removal Schedule," dated August 28, 2003.
(Sheet 1 of 1) Revision 8, October 2005 QUAD CITIES - UFSAR Table 5.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE WITHDRAWAL SCHEDULE
UNIT 1 HOLD NUMBER LOCATION AZIMUTH REMOVAL YEAR STATUS NEUTRON DOSIMETER MOUNTED SIDE OF PART #7 95° 1974 REMOVED 2 TOP GUIDE 0
° 1974 REMOVED 3 WALL 35° 1974 REMOVED 4 TOP GUIDE 90
° 1979 REMOVED 5 WALL 65° STANDBY 6 TOP GUIDE 180
° 1982 REMOVED 7 WALL 95° STANDBY 8 WALL 215
° 1982 REMOVED 9 WALL 245
° STANDBY 10 WALL 275
° STANDBY UNIT 2 HOLD NUMBER LOCATION AZIMUTH REMOVAL YEAR STATUS NEUTRON DOSIMETER MOUNTED SIDE OF PART #7 95° 1975 ------- 12 TOP GUIDE 0
° 1975 REMOVED 13 WALL 35
° 1975 REMOVED 14 TOP GUIDE 90
° 1979 REMOVED 15 WALL 65
° STANDBY 16 TOP GUIDE 180
° 1981 REMOVED 17 WALL 95
° STANDBY 18 WALL 215
° 1981 REMOVED 19 WALL 245
° STANDBY 20 WALL 275
° STANDBY NOTE: 0° IS DUE WEST.
(Sheet 1 of 1)
Revision 7, January 2003 QUAD CITIES - UFSAR Table 5.4-1 JET PUMP CHARACTERISTICS
General Number of jet pumps 20 Throat ID 8.1 in. Diffuser ID 20 in. Nozzle internal diameter 3.31 in.
Hydraulic Parameters Diffuser exit velocity 13.2 ft/s Driving flow (per pump) 4720 gal/min, 1.79 x 10 6 lb/hr Suction flow (per pump) 8160 gal/min, 3.106 x 10 6 lb/hr M ratio 1.729 MN efficiency (includes 180
° bend) 30.8% Jet pump head (discharge to suction) 67.6 ft
(Sheet 1 of 1)
Revision 9, October 2007 QUAD CITIES - UFSAR Table 5.4-2 HYDROGEN WATER CHEMISTRY SYSTEM TRIPS
Area hydrogen concentration high Reactor scram
- Low power level (main steam flow)
Operator request (manual)
- - Bypassed during plant star t and early power ascension.
(Sheet 1 of 1)
QUAD CITIES - UFSAR Table 5.4-3 REACTOR CORE ISOLATION COOLING SYSTEM EQUIPMENT SPECIFICATIONS
Pump Number 1 Discharge pressure 525 - 2800 ft Flow rate 400 gal/min NPSH 20 ft
Turbine Steam pressure inlet 150 - 1120 psia Steam pressure exhaust 25 psia Power 80 - 500 HP Steam flow rate 6,000 - 16,500 lb/hr
(Sheet 1 of 1) QUAD CITIES - UFSAR Table 5.4-4 SAFE SHUTDOWN MAKEUP PUMP SYSTEM EQUIPMENT SPECIFICATIONS
Pump Number 1 Discharge pressure 2885 ft Flow rate 400 gal/min NPSH 15 ft
Motor Voltage 4000 V Phase 3 Cycles 60 RPM 3550 Power 600 HP
(Sheet 1 of 1)
Revision 3, December 1995 QUAD CITIES - UFSAR Table 5.4-5 RESIDUAL HEAT REMOVAL EQUIPMENT DESIGN PARAMETERS
Pumps, Main System Number 4 (Note 1)
Type Single stage-vertical-centrifugal Seals Mechanical Drive Electric Motor Power source Normal aux iliary or standby diesel Speed 3600 rpm Pump casing Cast steel Impeller Stainless steel Shaft Stainless steel Code ASME Section III, Class C
Note:
1 The parameters used in the integrated ECCS performance LOCA analysis required by 10 CFR 50 Appendix K are discussed in Section 6.3.3 (SAFER/GESTR).
(Sheet 1 of 1)
Revision 4, April 1997 QUAD CITIES - UFSAR Table 5.4-6 RESIDUAL HEAT REMOVAL HEAT EXCHANGER DESIGN PARAMETERS
Heat Exchangers Quantity 2 Heat Load 105 x 10 6 Btu/hr each Primary side flow (containment water) 10,700 gal/min Secondary side flow (river water) 7,000 gal/min*
Design temperatures River water 95°F Containment water 165°F Design pressure Primary (shell) 450 psi Secondary (tube) 350 psi
Design codes Code (shell) ASME Section III, Class C Code (tube) ASME Section VIII
- Note: Heat exchanger design parameter is 7,000 gal/min; however, only 3,500 gal/min (one RHR service water pump) is needed during accident conditions. Shutdown cooling mode was sized for 7,000 gal/min; however, typical operation in this mode uses only one pump at reduced flow.