L-2017-074, St. Lucie, Unit 1, Updated Final Safety Analysis Report, Amendment No. 28, Chapter 1, Introduction and General Description of Plant Site Characteristics
Text
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT CHAPTER 1 LIST OF TABLES Table Title Page 1.3-1 Plant Parameter Comparison 1.3-7 1.6-1 Materials Incorporated By Reference 1.6-2 1.7-1 Design Compliance 1.7-2
UNIT 1 1-iii Amendment No. 27 (04/15)
b) The reactor fuel is slightly enriched uranium dioxide contained in Zirconium alloy tubes.
c) Minimum departure from nucleate boiling ratio (DNBR) during normal operation and anticipated transients is not below that value which could lead to fuel rod failure. The maximum center line temperature of the fuel, evaluated at the design overpower condition, is below that value which could lead to fuel rod failure.
The melting point of the UO 2 is not reached during routine operation and anticipated transients.
d) Fuel rod clad is designed to maintain cladding integrity throughout fuel life. Fission gas release withi n the rods and other factors affecting design life is considered for the maximum expected exposures.
e) The reactor and control systems are designed so that any xenon transients will be adequately damped.
f) The reactor is designed to accommodate safely and without fuel
UNIT 1 1.2-2 Amendment No. 27 (04/15)
1.2.3 OPERATING CHARACTERISTICS AND SAFETY CONSIDERATIONS 1.2.3.1 Nuclear Steam Supply System The reactor core is fueled with uranium dioxide pellets enclosed in Zirconium alloy tubes with welded end plugs. The tubes are fabricated into assemblies in which end fittings prevent axial motion and grids prevent lateral motion of the tubes. The control element assemblies (CEAs) consist of Inconel clad boron carbide absorber rods which are guided by Zirconium alloy tubes located within the fuel assembly. The core consists of 217 fuel assemblies loaded with multiple U-235 enrichments.
The reactor vessel and its closure head are fabricated from manganese moly steel internally clad with stainless steel. The vessel and its internals are designed so that the integrated neutron flux (E > 1 Mev) at the vessel wall will be less than 4.7 x 10 19 n/cm 2 over a 60-year period.
The internal structures include the core support barrel, the core support plate, the core shroud, and the upper guide structure assembly. The core support barrel is a right circular cylinder supported from a ring flange from a ledge on the reactor vessel. The flange carries the entire weight of the core. The core support plate transmits the weight of the core to the core support barrel by means of vertical columns and a beam structure. The core shroud surrounds the core and minimizes the amount of coolant bypass flow. The upper guide structure provides a flow shroud for the CEAs and prevents upward motion of the fuel assemblies during pressure transients. Lateral motion limiters or snubbers are provided at the lower end of the core support barrel assembly.
The reactor coolant system is arranged as two closed loops connected in parallel to the reactor vessel. Each loop
consists of one 42-inch ID outlet (hot) pipe, one steam generator, two 30-inch ID inlet (cold) pipes and two pumps. An electrically heated pressurizer is connected to the hot leg of one of the loops and a safety injection line is con-nected to each of the four cold legs.
The reactor coolant system operates at a nominal pressure of approximately 2235 psig. The reactor coolant enters near the top of the reactor vessel, and flows downward between the reactor vessel shell and the core support barrel into the lower plenum. It then flows upward through the core, leaves the reactor vessel, and flows through the tube side of the two vertical U-tube steam generators where heat is transferred to the secondary system. Reactor coolant pumps return the reactor coolant to the reactor vessel.
The two steam generators are vertical shell and U-tube units. The steam generated in the shell side of the steam generator flows upward through moisture separators and scrubber plate dryers which reduce the moisture content to less than 0.2 percent. All surfaces in contact with the reactor coolant are either stainless steel or NiCrFe alloy in order to minimize corrosion.
The reactor coolant is circulated by four electric motor driven single-suction vertical centrifugal pumps. The pump shaft leakage is minimized by mechanical seals. Each pump motor is equipped with an anti-reverse mechanism to prevent reverse rotation of any pump that is not in operation.
1.2.3.2 Engineered Safety Features and Emergency Systems
Engineered safety features systems protect the public and plant personnel in the highly unlikely event of an accidental release of radioactive fission products from the reactor system, particularly as the result of a LOCA. The safety features function to localize, control, mitigate, and terminate such accidents to hold exposure levels below applicable limits.
UNIT 1 1.2-6 Amendment No. 27 (04/15)
The fuel handling building contains the spent fuel pool and new fuel storage facilities, as well as the cooling and purification equipment for the fuel pool. The fuel is transferred from the reactor containment building to the fuel
handling building through the fuel transfer tube.
The turbine building houses the turbine generator, condensers, feedwater heaters, condensate and feedwater pumps, turbine auxiliaries and certain of the switchgear assemblies.
The service building provides offices, shop and warehouse space, and is located next to the turbine building unloading bay.
1.2-15 Am. 4-7/86 (DELETED)1.2-16 Am. 4-7/86 (DELETED)1.2-17 Am. 4-7/86 (DELETED)1.2-18 Am. 4-7/86 (DELETED)1.2-19 Am. 4-7/86 (DELETED)1.2-20 Am. 4-7/86 (DELETED)1.2-21 Am. 4-7/86
Refer to drawing 2998-G-058 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 SITE PLAN FIGURE 1.2-1 Amendment No. 15 (1/97)
Refer to drawing 2998-G-059 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 ENLARGED PLOT PLAN FIGURE 1.2-2 Amendment No. 15 (1/97)
Refer to drawing 8770-G-060 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
GROUND FLOOR PLAN FIGURE 1.2-3 Amendment No. 15 (1/97)
Refer to drawing 8770-G-061 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
MEZZANINE FLOOR PLAN FIGURE 1.2-4 Amendment No. 15 (1/97)
Refer to drawing 8770-G-062 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
OPERATING FLOOR PLAN FIGURE 1.2-5 Amendment No. 15 (1/97)
Refer to drawing 8770-G-063 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
SECTIONS SHEET 1 FIGURE 1.2-6 Amendment No. 15 (1/97)
Refer to drawing 87708-G-064 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT TURBINE BUILDING -
PLAN & SECTIONS SH2 FIGURE 1.2-7 Amendment No. 15 (1/97)
Refer to drawing 8770-G-065 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR BUILDING -
FLOOR PLANS SH1 FIGURE 1.2-8 Amendment No. 15 (1/97)
Refer to drawing 8770-G-066 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR BUILDING -
FLOOR PLANS SH 2 FIGURE 1.2-9 Amendment No. 15 (1/97)
Refer to drawing 8770-G-067 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR BUILDING -
SECTIONS SHEET 1 FIGURE 1.2-10 Amendment No. 15 (1/97)
Refer to drawing 8770-G-068 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGMEENT REACTOR BUILDING --
SECTIONS SHEET 2 FIGURE 1.2-11 Amendment No. 15 (1/97)
Refer to drawing 8770-G-069 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 1 FIGURE 1.2-12 Amendment No. 15 (1/97)
Refer to drawing 8770-G-070 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 2 FIGURE 1.2-13 Amendment No. 15 (1/97)
Refer to drawing 8770-G-071 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN SHEET 3 FIGURE 1.2-14 Amendment No. 15 (1/97)
Refer to drawing 8770-G-072 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUIDLING SECTIONS SHEET 1 FIGURE 1.2-15 Amendment No. 15 (1/97)
Refer to drawing 8770-G-075 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGMENT REACTOR AUXILIARY BUILDING SECTIONS SHEET 2 FIGURE 1.2-16 Amendment No. 15 (1/97)
Refer to drawing 8770-G-076 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING MISCELLANEOUS PLANS FIGURE 1.2-17 Amendment No. 15 (1/97)
Refer to drawing 8770-G-073 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGEMENT FUEL HANDLING BUILDING PLANS FIGURE 1.2-18 Amendment No. 15 (1/97)
Refer to drawing 8770-G-074 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GENERAL ARRANGMENT FUEL HANDLING BUILDING SECTIONS FIGURE 1.2-19 Amendment No. 15 (1/97)
- *
- LOGIC DIAGRAM LEGEND SYMBOL DESIGNATION AND GATE ACTION GIVES OUTPUT ONLY IF ALL INPUTS ARE PRESENT =EB= -m-OR GATE NOT GATE TIME DELAY UNIT MEMORY UNIT TIME DELAY (WIPED OUT) UNIT GIVES OUTPUT IF ANY INPUT IS PRESENT GIVES OUTPUT ONLY IF IN PUT IS NOT PRESENT PRODUCES OUTPUT SAME AS INPUT AFTER DELAY (SEC.) A DEVICE WHICH RETAINS ITS STA TE UNTIL OUESTED TO ALTER ITS STATE OUT PUT CHANGES ON APPLICATION OF INPUT AND IS WIPED OUT AFTER SECS. 0 INDICATING LAMP-RED © INDICATING LAMP-GREEN , ' @ PUSH BUTTON MATCHING SYMBOLS I OUTPUT MAIN CONTROL ROOM ANNUNCIATOR H-HIGH DESIGNATION PANEL NO. LOCAL PANEL ANNUNCIATOR H-HIGH c:;:) XX-)(XX-PANEL NO. L-LOW --------__ ,.. __ .,.. __ ...., ....... SWITCH BLOCK TAG NO.--+-I TRIP SIGNAL LINE START OR PERMISSIVE SIGf'>AL l..INE TIEBACK PNEUMATIC OR HYDRAULIC SIGNAL LINE SERVICE LOCATION FUNCTIONAL DESCRIPTION I POSITION OR SWITCH CONDITION EQUIPMENT BLOCK EQUIPMENT DESCRIPTION
& NO. EQUIPMENT (RUN STOP) DWG. NO. !1770-6-276 FLORIDA POWER !!. LIGHT COMPANY St. Lucie Plant CONTROL AND BLOCK DIAGRAi.I FIGURE 1.2-21
- *
- l9U I PMEllT i I I I I HEAT MIXER EXCHANGER POSIT I YE 0 MANWAY OR DISPLACEMENT SS HATCH PUMP --EY CENTRIFUGAL -p EDUC TOR PUMP -<§r FAN. _[_ BLOWER OR VEH COMPRESSOR -ET CAMMED y *:ENTR I FUGAL EJECTOR PUMP ctJ TANK
- I \'.I Y-STRAINER STRAINER -El--IN-LINE STRAINER -HJ-TEMPORARY STRI I MER --0-f ILTER --§-IMPULSE TRAP -B-DWG. NO. 8770-8-270 lUClf I I f lOAI I RAP FLORIDA POWER & LIGHT COMPANY St. Lucie Plant EQUIPMENT 5YllBOL5 -SHEET 2 FIGURE 1. 2-)(1
1.3 COMPARISONS
Comparisons contained herein were considered valid at the time the operating license for St Lucie 1 was issued, and are being retained in the Updated FSAR for document completeness and historical record. No present or future
update of this section is required.
1.3.1 COMPARISON WITH SIMILAR FACILITY DESIGNS Table 1.3-1 presents a summary of the characteristics of the St. Plant. The table includes similar data for Calvert Cliffs Units 1 and 2, Turkey Point Units 3 and 4, and Palisades Unit 1, as published in the FSAR for those units. The parameters as listed in the Hutchinson Island PSAR are shown in parentheses for ease of comparison.
The Calvert Cliffs Units 1 and 2 design was selected for comparison because of the basic similarity of the reactor core and the nuclear steam supply system and to enable the reviewer to apply to the St. Lucie Plant the information gained from the review of Calvert Cliffs since the FSAR submittal in early 1971. Palisades was selected for comparison because of the similarity of the reactor coolant system and because it was the first large pressurized
water reactor plant designed by Combustion Engineering to be licensed for operation. The Turkey Point station is
included as representing a contemporary plant designed by another supplier.
The principal differences to be observed between the parameters listed for Calvert Cliffs and those for St. Lucie include the range of boron concentration, the number of control element assemblies (CEA's), and the reactor vessel minimum clad thickness. These differences have no effect on plant safety.
The difference in boron concentration and the number of CEA's reflects the evolution of the core design, with improved calculations providing a more accurate representation of the reactivity control capability and requirements
of the CEA's and the worth of dissolved boron.
The containment concept used for this plant is similar to that which has received a construction permit at Prairie Island (Docket 50-282 and 50-306) and Kewaunee (Docket 50-305) except that the diameter is larger and the thickness is greater for this plant. Since the shell thickness is greater than 1 1/2 inches, field stress relieving of the
complete vessel was necessary. Extensive experience exists in field stress relieving of large vessels, including
1.3-1 1.3.2 COMPARISON OF PRELIMINARY AND FINAL DESIGN Comparisons contained herein were considered valid at the time the operating license for St. Lucie Unit 1 was issued, and are being retained in the updated FSAR for document completeness and historical record. No present or future update of this section is required.
1.3.2.1 General This section contains a discussion of all significant changes that have been made in the plant design since submittal of the PSAR. Changes considered as significant include changes in design bases or criteria for safety related structures, systems or components, plant arrangement, mode of system operation, type of equipment or gross
changes in component or system capacity.
1.3.2.2 Site Characteristics No significant site characteristic changes have been brought to light that would reflect a design change since submittal of the PSAR.
1.3.2.3 Design Criteria a)AEC General Design Criteria The PSAR contained a comparison of plant design to the Proposed (70) AEC General Design Criteria. It was on the basis of these GDC that the pre liminary plant design was formulated and presented in the PSAR. Since the receipt
of the construction permit (July 1, 1970) the (64) AEC General Design Criteria have been published as Appendix A to
10CFR50 (July 7, 1971). As far as was practicable, depending on engineering, construction and equipment
procurement schedules, the criteria given in Appendix A were reflected in the plant design. A comparison of the final
design with the present GDC is given in Section 3.1.
b)Design Codes It was stated in the PSAR that the design code for plant piping systems would be USAS B31.7 Nuclear Power Piping Code for reactor coolant system piping and USAS B31.1 Power Piping Code for other safety related piping.
Procurement schedules allowed the use of B31.7 for all safety related plant piping and this code was applied in the
design.Also, although the PSAR contains no reference to its application, the Draft ASME Nuclear Pump and Valve Code was used in the design of safety related pumps and valves. The application of design codes used in the final design
is discussed in Section 3.2.
c)Design Tornado Criteria It was stated in the PSAR that the design tornado windspeed for this site would be 337 mph. This value was changed at the request of the AEC before the construction permit was issued, and the design tornado used in the design of structures, systems and components consisted of 300 mph translational and 60 mph rotational wind speeds with a 3 psi pressure differential. The tornado design criteria are discussed in Section 3.3.2.
1.3-2 1.3.2.4 Reactor The following changes have been made to the reactor to improve the operating characteristics.
a)Control Element Drive Mechanisms Magnetic jack drive mechanisms are provided for positioning the control element assemblies instead of rack and pinion drive mechanisms. The magnetic jack control element drive mechanism is completely sealed by a pressure boundary, eliminating the need for seals. Motion of the control element drive shaft is accomplished by sequencing
five solenoid coils located around the pressure boundary. The control element drive mechanisms are described in
Section 4.2.3.1.
Combustion Engineering, Inc., has supplied identical control element drive mechanisms on previous plants, including Maine Yankee (AEC Docket No. 50-309) and Calvert Cliffs Units 1 and 2 (AEC Docket No. 50-317 and 50-318).
b)Number of Control Element Assemblies and Drive Mechanisms There are 81 CEA in the reactor compared to 85 CEA shown in the PSAR design while the number of drive
mechanisms has changed from 65 to 69. This results in an increase in the number of single CEA (45 to 57. including
8 part-length CEA) and a reduction in the number of dual CEA (40 to 24), thereby providing greater flexibility for
optimization of CEA programming and fuel management.
c)Burnable Poison Shims Burnable poison shims have been added to the fuel assemblies, replacing some fuel. These shims permit lowering of the initial boric acid concen tration in the coolant. This provides additional assurance that the moderator temperature
coefficient, at power at beginning of life, will not be positive.
1.3.2.5 Reactor Coolant System No significant changes have been made to the reactor coolant system design since issuance of the PSAR.
1.3.2.6 Engineered Safety Features No significant changes have been made to the safety injection system design since issuance of the PSAR.
Other than design code changes as discussed in Section 1.3.2.3, there have been no significant changes in the design of the containment, containment spray system, containment cooling system or shield building ventilation
systems from that described in the PSAR.
Changes have been made in the final design of the containment hydrogen purge system from that described in the PSAR, primarily in the location 1.3-3 of the containment isolation valves and purge line piping. The PSAR (Amendment 4) hydrogen purge report showed the location of the isolation valves and penetrations inside the shield building amulus at the top of the containment vessel with the purge lines inside the annulus. The present arrangement has the isolation valves located outside the
shield building in the HVAC equipment room in the reactor auxiliary building. The purge lines are inside the
containment vessel and penetrate the containment vessel at elevation 43'. The present design offers improved
system reliability due to greater accessibility of the isolation valving and the reduced length of exposed piping outside the containment. In addition two (2) hydrogen recombiners have been added inside containment. The final design of the containment hydrogen control system is discussed in Section 6.2.5.
1.3.2.7 Instrumentation and Control The PSAR stated that the actuating signals for the Shield Building Ventilation System (SBVS) and Containment Cooling System would be Containment Spray Actuation Signal (CSAS). The actuating signal for the SBVS was
subsequently changed to Containment Isolation Signal (CIS) and the actuating signal for the containment cooling
system was changed to Safety Injection Actuation Signal (SIAS). These signals offered greater diversity and were
more directly related to the functions of the systems being actuated. The final design of the engineered safety
feature actuation system is discussed in Section 7.3.
1.3.2.8 Electrical Systems The diesel generator sizing criteria stated in answer to Question 5.6, Amendment 6 of the PSAR stated that that the maximum post-accident load on the diesel generator could exceed the continuous rating of the unit. These criteria
were subsequently changed to ensure that the maximum load was within the continuous rating. The final sizing
criteria are discussed in Section 8.3.1.1.7.
The PSAR stated that the on-site diesel oil fuel storage capacity would be sufficient for thirty (30) days full load operation of one diesel generator. This capacity was subsequently reduced to a 8.0 day supply based on post-LOCA load profile. This time is sufficient to allow resupply in the event of an emergency condition requiring prolonged operation of the emergency diesel generators. The oil storage system design and capacity are discussed in Section
9.5.4.1.3.2.9 Auxiliary Systems a)Circulating Water System The PSAR stated that cooling water for normal operation and emergency plant conditions would be taken from the Atlantic Ocean through a diked area of Big Mud Creek. This area was to have served as the ultimate heat sink in providing a source of emergency cooling water for a long term residual heat removal following a LOCA. This design was subsequently changed due to the restriction in access to and use of Big Mud Creek which would be caused by erection of a dike at the Indian River inlet to Big Mud Creek. The present circulating water system design provides for 1.3-4
TABLE 1.3- CORE MECHANICAL DESIGN PARAMETERS ST. LUCIE Unit 1 Reference Section CALVERT CLIFFS Units 1 and 2 PALISADES Unit 1 TURKEY POINT Units 3 and 4 Fuel Assemblies Design CEA 4.2 CEA Cruciform RCC Rod Pitch, in.
0.58 4.2 0.58 0.550 0.563 Cross-Section Dimensions, in.
7.98 x 7.98 4.2 7.98 x 7.98 8.1135 x 8.1135 8.426 x 8.426 Fuel Weight (as U O 2), pounds 207,200 (217,600) 4.2 207,269 210,524 176,200 Total Weight, pounds 271,280 (284,000) 4.2 282,570 295,800 226,200 Number of Grids per Assembly 8 8 8 7 Fuel Rods Number 36,896 (38,192) 4.2 36,896 43,168 32,028 Outside Diameter, in. 0.44 4.2 0.4 4 0.4135 0.422 Diametral Gap, in.
0.0085 (0.0065) 4.2 0.0085 0.0065 0.0065 Clad Thickness, in.
0.026 4.2 0.026 0.022 0.0243 Clad Material Zircaloy or M5 4.2 Zircaloy or M5 Zircaloy or M5 Zircaloy Fuel Pellets Material UO 2 Sintered 4.2 UO 2 Sintered UO 2 Sintered UO 2 Sintered Diameter, in.
0.3795 (0.3815) 4.2 0.3795 0.359 0.367 Length, in.
0.650 (0.600) 4.2 0.650 0.600 0.600 Control Assemblies Neutron Absorber B 4C/SS (B 4 C) 4.2 B 4 C Cd-In-Ag (5-15-80%) Cruciform Cd-I n-Ag (5-15-80%) Cladding Material NiCrFe Alloy 4.2 NiCrFe Alloy 304 SS Tubes, E.B. welded to 13.5 in. span 304 SS Cold Worked Clad Thickness 0.040 4.2 0.040 0.016 0.019 Number of Assembly, full/part length 78/8 (85) 4.2 7 7/8 41/4 53 Number of Rods per Assembly 5 4.2 5 117 Tubes per Rod 20 Core Structure Core Barrel I.D./O.D., in.
148/151.5 (148/151) 148/149.75 149.75/152.5 133.875/137.875 Thermal Shield I.D./O.D., in.
156.75/162.75 (156/162) N one None 142.625/148.0 NUCLEAR DESIGN DATA Structural Characteristics Core Diameter, inches (Equivalent) 136 4.2 136 136.71 119.5 Core Height, inches (Active Fuel) 136.7 (137) 4.2 136.7 132 144 Reflector Thickness and Composition Top Water plus steel 10 10 10 10 Bottom Water plus steel 10 10 10 10 Side Water plus steel 15 15 15 15 H 2O/U, Unit Cell (Cold) 1.63 (3.35) 4.3 3.44 3.50 4.18 Number of Fuel Assemblie s 217 4.2 217 204 157 UO 2 Rods per Assembly, unshimmed/shimmed 176/164 (176) 4.2 176/164 212/208 204 Batch A 176 4.2 -- -- 176 Batch B 164 4.2 -- -- 164 Batch C (176/164/164) 4.2 -- -- (176/164/164)
Performance Characteristics Loading Technique 3 Batch Mixed Central Zone (3 Batch) 4.3 3 Batch Mixed Central Zone 3 Batch Mixed Central Zone 3 Regions Central Zone Fuel Discharge Burnup, MWD/MTU Average First Cycle 12,800 (11,900) 4.3 13,775 10,180 13,000 First Core Average 22,000 22,550 17,600 24,500 UNIT 1 1.3-8 Amendment No. 27 (04/15)
TABLE 1.3-1 (Cont'd)REACTOR COOLANT SYSTEM - CODEREQUIREMENTSST. LUCIEUnit 1ReferenceSectionCALVERT CLIFFSUnits 1 and 2PALISADES Unit 1TURKEY POINT*
Units 3 and 4ComponentReactor VesselASME III class A5.2ASME III class AASME III class AASME III class ASteam Generator Tube SideASME III class A5.2ASME III class AASME III class AASME III class AShell SideASME III class A5.2ASME III class AASME III class AASME III class APressurizerASME III class A5.2ASME III class AASME III class AASME III class APressurizer Relief (or Quench) TankASME III class C5.2ASME III class CASME III class CASME III class CPressurizer Safety ValvesASME III5.2ASME IIIASME IIIASME IIIRead or Coolant PipingUSAS B 31.7(USAS B 31.1)5.2USAS B 31.7USAS B 31.1USAS B 31.1PRINCIPAL DESIGN PARAMETERS FOR THECOOLANT SYSTEMOperating Pressure, psig 22355.1 2235 2085 2235Reactor Inlet Temperature, o F539.7(549)5.1544.5 545546.2Reactor Outlet Temperature, o F595.1(601)5.1599.4591.1602.1Number of Loops 25.1 2 2 3Design Pressure, psig 24855.1 2484 2485 2485Design Temperature, o F 6505.1 650 650 650Hydrostatic Test Pressure (cold), psig 3110 3110 3110 3110Total Coolant Volume, cu. Ft11,10111,10110,809 9088PRINCIPAL DESIGN PARAMETERS of THE REACTOR VESSELMaterial**SA-533, Grade B, Class I, low alloy steel, internally clad with type 304 austenitic SS.5.1SA-533, Grade B, Class I, low alloy steel, internally clad with Type 304 austenitic SS.SA-302, Grade B, Class I, low alloy steel, internally clad with Type 304 austenitic SS.SA-302, Grade B, Class I, low alloy steel, internally clad with Type 304 austenitic SS.Design Pressure, psig 24855.4 2485 2485 2485Design Temperature, o F 6505.4 650 650 650Operating Pressure, psig 2235 2235 2085 2235Inside Diameter of Shell, in.
1725.4 172 172 172Outside Diameter across Nozzles, in.
253 253 254 236Overall Height of Vessel and Enclosure Head, ft. - in. to top of CRD Nozzle41-11-3/4(42' 1")5.441-11-3/440-1-13/16 41-6Minimum clad thickness, in.5/165.41/83/165/32PRINCIPAL DESIGN PARAMETERS OF THE STEAMGENERATORSNumber of Units 25.5 2 2 3TypeVertical U-Tube with integral moisture separator5.5Vertical U-Tube with integral moisture separatorVertical U-Tube with integral moisture separatorVertical U-Tube with integral moisture separator *Operating Level
- Replacement RVCH is a one piece forging SA-508 Class 3 low alloy steel,internally clad with Type 308L/309L austentic stainless steel.1.3-10 Amendment No. 21 (12/05)
1.4 IDENTIFICATION OF AGENTS AND CONTRACTORSThe Florida Power & Light Company is the applicant for the operating license for Hutchinson Island Nuclear Power Unit 1. Florida Power & Light Company is responsible for the design, engineering review, construction and operation
of the plant.
Florida Power & Light Company has engaged Combustion Engineering, Inc. to design, manufacture and deliver the Nuclear Steam Supply System and nuclear fuel for the first core and three reload batches to the site. The Nuclear
Steam Supply System includes the reactor coolant system, reactor auxiliary system components, nuclear and
certain process instrumentation, and the reactor control and protective system. In addition, C-E will furnish technical
assistance for erection, initial fuel loading, testing and initial startup of the Nuclear Steam Supply Systems.
Ebasco Services Incorporated has been engaged by the Owners as the Engineer-Constructor for this project and as such has performed engineering and design work for the balance-of-plant equipment systems, and structures not
included under the C-E scope of supply.
Ebasco has been engaged to perform onsite construction of the entire
plant with technical advice for installation of the reactor components to be provided by C-E.
Following issuance of the facility operating license, Ebasco Services Incorporated was retained as Engineer-Constructor for St. Lucie Unit 1 plant backfit, retrofit and maintenance activities, under the direction of Florida Power
& Light Company.
Presently, Bechtel Power Corporation has been engaged to perform site backfit construction activities under the direction of Florida Power & Light Company.
1.4-1 Amendment 15, 1/97
Item 7: The final loads and ratings of the emergency diesel generators so that we can be assured that the ratings comply with our current criteria.
Status: The information requested was submitted with PSAR Amendment 10 (4/26/71) and is also contained in Section 8.3.1.
Item 8: The analysis of the relative humidity in the shield building annulus after accidents to assure that it would be reduced below 70 percent in a short time and thus assure an acceptable efficiency of the shield building annulus filter system.
Status: The analysis of annulus relative humidity is presented in Section 6.2.1.
Item 9: The design of equipment which would be the primary means to control hydrogen buildup in the containment following a loss-of-coolant accident in addition to the purging system which is presently
proposed.Status: The design of the containment hydrogen control system is discussed in Section 6.2.5.
Item 10: The results of studies of means to prevent common failure modes from negating scram action.
Status: In conformance with the requirements of WASH-1270 "Technical Report on Anticipated Transients Without Scram", CE has analyzed the reactor protective system to identify areas that may be
particularly vulnerable to common mode failures. Topical reports CENPD-149 "Review of Reactor
Shutdown System (RPS Design) For Common Mode Failure Susceptibility" describes the common
mode failure review of the reactor shutdown system.
Item 11: The results of studies of consequences of failure to scram during anticipated transients and design features which would make such failure tolerable.
1.5-2 Status: Combustion Engineering has analyzed the response of pressurized water reactors which are of the St Lucie type to demonstrate the response of the plant to anticipated transients without scram (ATWS)
The Combustion Engineering report, entitled "Topical Report on Anticipated Transients Without Scram (Proprietary)" was submitted to the AEC on January 10, 1972 Evaluations are performed in this report
based upon the incredible assumption that no CEA's are inserted into the core during the course of the
following transients: CEA withdrawal CEA drop, idle loop start-up, loss of flow, boron dilution, excess
load, loss of load, loss of feedwater, sample line break, and pressurizer safety valve failure.
The transient resulting from loss of normal on-site and off-site power is also analyzed, but with a
conservative one percent negative reactivity insertion assumed following reactor trip signal generation, since for this case the failures which initiate the transient would also remove power from the CEDM, allowing the CEA to insert. Applicant's letter of 3/31/75 to the NRC refer enced CENPD-158, 'Anticipated
Transients Without Reactor Trip," December, 1974, as the analysis applicable to St. Lucie Unit 1
except where qualified in the 3/31/75 letter. The generic sol utions to NRC concerns for ATWS wil l be implemented on St. Lucie Unit 1. A schedule for implementation will be provided to the NRC Staff
following their review and acceptance of the CE report.
Item 12 The results of studies of the consequences of secondary system accidents particularly relative to fuel clad damage during a steamline break accident and to multiple steam generator tube failures Status: The results of the studies of the consequences of secondary system accidents is presented in Sections 15.4.3 and 15.4.5.
Item 13 The anal ysis of the consequences to turbine failure including (1) the ability of the structures and systems which are important to safety to withstand the effects of turbine missiles inview of the results of
the recent turbine missile calculations and (2) radiological consequences of a turbine missile entering
the fuel storage pool and damaging fuel Status: The analysis of the potential turbine missiles is presented in Section 3.5 Item 14: The status of the Research and Development items identified on Pages 45-5 0 of our Hutchinson Island Safety Evaluation dated April 13, 1970.
Status: The items referred to appear in two groups in the AEC Safety Evaluation: Section 15.1, Paragraphs (a) through (e), and Section 15.2, Paragraphs (a) through (i). A statement of each item follows:
1.5-3 15.1.a Fuel Assembly Flow Tests The Staff Safety Evaluation states: "Tests are being conducted on fuel assemblies (1) to verify flow mixing factors and (2) to establish the characteristics of fuel rods in axial flow".
1.5-4 Status: The Flow Mixing Tests have been completed and are described in Section 4.4.4.4 and Division 1, Section 1, of Reference 1.
The results of the hot flow tests at reactor flow conditions using full length partial cross section fuel assemblies are discussed in Section 4.4.3.6 and Division 1, Section 1, of Ref. 1.
15.1.b Mechanical Testing of Control Element Assemblies (CEA's)
The Staff Safety Evaluation states: "A Series of tests have been completed on single CEA's demonstrating functional feasibility of the CEA concept under all possible combinations of misalignments, dynamic loading, bowing and thermal effects." Status: A series of tests have been completed on both single and dual CEA's in a cold water, low pre ssure facility to satisfy the following objectives:
a)Determine the mechanical and functional feasibility of the CEA-type control rod concept.
b)Experimentally determine the relationship between CEA drop time and CEA drop weight, annular clearance between CEA fingers and guide tubes and coolant flow rate within the guide
tube.c)Experimentally determine the relationship between flow rate and pressure drop within the guide
tube as a function of CEA axial position and of finger-to-guide tube clearance.
d)Determine the effects on drop time of adding a flow restriction or of plugging the flow holes in
the lower portion of a guide tube (as might occur under incident conditions).
e)Determine the effects of misalignment within the CEA guide tube syst em on drop time.
The results of these tests were used as the basis for selecting the final CEA and guide tube geometries.
The tests have demonstrated that the five finger CEA concept is mechanically and functionally feasible
and that the CEA design has met the criteria established for drop time under the most adverse
conditions. The testing has also verified that the analytical model used for predicting the drop times
gives uniformly conservative results.
The effects on drop time of all possible combinations of frictional restraining forces in the CEDM, angular and radial misalignment of the CEDM bowing of the guide tubes and misalignments of the CEA have
been experimentally investigated and defined. The conditions tested simulated all the effects of
tolerance buildup, dynamic loadings and thermal effects. The tests demonstrated that misalignments
and distortions in excess of those expected from tolerance buildup or any other anticipated cause would
still result in acceptable drop times.
1.5-5
b)A thimb le assembly consisting of the instrument plate, three in-core instrumentation thimbles and the lifting sling.
c)An upper guide tube with the guide tube attached to the thimble extension and the detector cable
partially inserted in the guide tube.
Insertion and withdrawal tests were performed to determine the frictional forces of a multitude instrument
thimble assembly during insertion and withdrawal from a set of fuel bundles. This test simulated the
operation that will be performed during the refueling of the reactor. To determine whether jamming of the
thimbles would occur during this operation, bending loads were applied to the thimble assembly by
tilting the instrument plate in 0
.5 degree increments up to a total of five degrees horizontal. Guide tubes were filled with water. The assembly was raised and lowered approximately five times for each tilt
setting. Results showed no discernible difference in the friction forces for the various tilt settings. The
tests demonstrated that the repeated insertions and withdrawal of in-core instrumentation thimble
assemblies into the fuel bundle guides can be accomplished with reasonable insertion forces.
Life cycle tests were performed to determine if the frictional forces increase as a result of 40 insertions and withdrawals. An automatic time was installed in the crane electrical circuitry to automatically cycle
the thimble assembly between the fully inserted and withdrawn position. The instrument plate was set
for five degrees tilt and the assembly was cycled 60 times. The insertion and withdrawal forces were
measured during the first and last five cycles. No discernible difference was noticed.
An off-center lift test was performed to determine if the thimble assembly could be withdrawn from the core region while lifting the assembly from an extreme off-center position. For a lifting point 11 inches
off-center, insertion was accom plished without incident. The flexibility of the thimble is such that
jamming of the assembly due to off-center lifting does not occur.
Cable insertion tests were performed to determine the forces required to completely insert and withdraw a detector cable from the in-core instrumentation thimble assembly. The guide tube routing included typical bends equal to, or worse than those found in the reactor. The detector cable was passed through
the guide tubing and into a thimble. In all cases, the insertion and withdrawal forces were reasonable for
hand insertion.
15.2.a Effect of Fuel Rod Failure on ECCS Performance 1.5-7 Status: C-E has conducted experimental and analytical investigations of fuel rod failures under simulated LOCA conditions. The analytical work provides indications of the actual conditions to be expected in the core during a transient, in terms of potential clad heating rates, internal pressures and transient duration. The
experimental work, described in Section 16 of Reference 2, applies these parameters in various
combinations to establish the nature of fuel rod deformation which might occur under accident conditions. This subject has been covered comprehensively in the Statement of Affirmative Testimony
and Evidence of Combustion Engineering in the Matter of Rulemaking Hearing for the Acceptance
Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors, Docket
No. RM-50-1. Further details are given in Section 15.4.1.2.
15.2.b Effects of Fuel Bundle Flow Blockage Status: That portion of the program responding to the ACRS concern related to flow blockage during operation has now been completed and the results are summarized in Section 4.4.3.6 and in ASME paper 68-
WA/HT-34 presented at the December 1968 Winter Annual Meeting.
15.2.c Verification of Fuel Damage Limit Criterion Status: The basis for C-E design is described in Sectio n 4.2.1. The staff notes (Page 46) that-C-E does not have an experimental program directed towards establishing these limits. Since the Staff's Safety Evaluation
Report was issued, CE has conducted a series of fuel irradiation tests in the Battelle Research, MZFR, GETR and Halden Reactors to determine the densification characteristics of CE fabricated fuel. The results of the tests conducted through 1973 are summarized in Section 12 of Reference 2.
15.2.d Effects of Blowdown Forces on Primary System Compo nents Status: The dynamic response of reactor internals resulting from hydro-dynamic blowdown forces under a postulated LOCA condition is discussed in a proprietary Combustion Engineering Topical Report, CENPD-42 which was submitted on August 1, 1972. This report contains a complete description of the
theoretical basis for methods of analysis for the various reactor components, as well as documentation
of computer program and the respective analyti cal structural models.
Reactor vessel internal structures will be analyzed to ensure the required structural integrity during abnormal operating conditions, including the effects of blowdown, pressure drop and buckling forces. For
the LOCA, the CEFLASH-4 computer program is used to define the flow transient and the
WATERHAMMER program determines the corresponding dynamic pressure load distribution. The
dynamic response of the reactor vessel internals to the space and the time-dependent pressure loads
will be obtained through the use of a number of structural 1.5-8 dynamic analysis codes. Lateral and vertical dynamic response of the internals will be considered, as well as the transient response and dynamic buckling of a core support barrel in shell modes. Both the
CEFLASH-4 and WATERHAMMER models are being evaluated against the LOFT program results.
1.5-9 The loads resulting from the LOCA will be added to the loads resulting from normal operation and the design basis earthquake for each critical component and the component deflections and stresses analyzed to ensure compliance with the criteria specified in Section 3.9.1.
15.2.e ECCS Thermal Effect on Rods Status: ACRS has asked that information be developed to show that the "--melting and subsequent disintegration of a portion of fuel assembly-- will not lead to unacceptable conditions." They refer
specifically to the "--effects in terms of fission product release, local high pressure production and the
possible initiation of failure in adjacent fuel elements--".
Inquiry has been made as to whether that accident conditions might occur which cause clad temperatures to reach such high temperatures that embrittlement occurs and whether subsequent
quenching operations will cause the embrittled portions to disintegrate and thereby prevent a sufficient flow of emergency core coolant to the remainder of the core.
Fuel damage of the magnitude indicated is prevented by the inherent nuclear and thermal characteristics of the UO 2 core and by the provision of engineered safety features.
With regard to the nonexcursion mechanisms leading to the conditions described by ACRS, the following might be conjectured:
a)Fuel bundle inlet flow blockage during fuel power operation and subsequent overheating of the coolant starved fuel, or b) Loss of reactor coolant.
Condition (a), inlet flow blockage during fuel power operation and subsequent overheating and melting of
the fuel, is not considered possible because open (non shrouded) fuel bundles are used, thereby
providing cross flow to the flow starved channel even if some of the inlet holes were blocked. Details and
conclusions of the tests performed at C-E on the influence of inlet geometry on flow in the entrance
region are presented in ASME paper 68-WA/HT-34 delivered at the December 1968 Winter Annual
Meeting. Further analysis of these tests showed that if a group of four flow holes in the core support
plate at the base of the fuel bundle were blocked, the subchannels above the blocked region would have an inlet velocity about 21 percent of the core average bulk inlet velocity. Because of cross flow from the
surrounding nonblocked regions, the net effect of this flow shortage, using conservative calculations, is to increase the enthalpy rise of the blocked region by a maximum of 35 percent. At nominal conditions, the hot channel DNB ratio would drop from 2.0 to 1.4, assuming that the blockage occurred directly
below the design hot channel.
1.5-10
1.6 MATERIALS INCORPORATED BY REFERENCE Topical reports listed in Table 1.6-1 are used and/or referenced as part of this application. Table 1.6-1 also includes documents submitted to the AEC in other applications that are incorporated in this application by reference.
Topical reports incorporated by reference were valid at the time of application to the NRC, and are being retained in the updated FSAR for document completeness and historical record. No present or future update of this section is required.1.6-1 Am. 1-7/83 TABLE 1.6-1 MATERIALS INCORPORATED BY REFERENCE Title CE Report No.
Topical Report Issue Date FSAR Reference Section CENPD-61 December 8, 1972 3.10 Suppl 1 1/23/73 Seismic Qualification of Category I Electric Equipment for Nuclear
Steam Supply Systems Suppl 2 3/14/73 Suppl 3 5/10/73 Suppl 4 7/9/73 Suppl 5 8/24/73 CENPD-42 August 1972 3.9.1.3 3.9.1.4.3 Dynamic Analysis of Reactor Vessel
Internals Under Loss of Coolant
Conditions with Application of
Analysis to CE 800 MWE Class Reactors Coast Code Description CENPD-98A July 5, 1973 15.1 15.2 CENPD-118P September 3, 1974 4.2.1.4.15 Densification of Combustion
Engineering Fuel CENPD-118 CENPD-132-P August 1974 6.3.3.6 Calculational Methods for the CE Large Break LOCA Evaluation Model CENPD-132 Rev 01 CENPD-133-P August 1974 6.3.3.6 CENPD-133 Rev 01 CEFLASH 4A, A Fortram IV Digital Computer Program for Reactor
Blowdown Analysis CENPD-133 Pm Suppl 1 8/74 CENPD-133 Suppl 1 CENPD-133 P Suppl 2 CENPD-133 Suppl 2 1.6-2 TABLE 1.6-1 (Continued)
Title CE Report No.
Topical Report Issue Date FSAR Reference Section CENPD-134 P September 1974 6.3.3.6 COMPERC-II, A program for Emergency Refill -Reflood of the core CENPD-134 Rev 01 CENPD-134 P Suppl 1 2/75 CENPD-134 Suppl 1 CENPD-135 P September 1974 6.3.3.6 STRIKEN II, A Cylindrical Geometry Fuel Rod Heat Transfer Program CENPD-135 Rev 01 CENPD-135 P Suppl 2 2/75 CENPD-135 Rev 01, Suppl 2 CENPD-136 P August 1974 6.3.3.6 CENPD-136 Rev 01 High Temperature Properties of Zircaloy and UO 2 for Use in the CE LOCA Evaluation Model CENPD-137-P September 1974 6.3.3.6 Calculational Methods for the CE
Small Break LOCA Evaluation Model CENPD-137 Rev 01 CENPD-138-P September 1974 6.3.3.6 CENPD-138 Rev 01 PARCH, A Fortran IV Digital Computer Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup.CENPD-138 P Suppl 1 2/75 CENPD-138 Rev 01 Suppl 1 CE Fuel Evaluation Model Topical Report CENPD-139 P-A September 1974 4.2.14 CENPD-139-A CENPD-155 P-A September 1974 5.4.4 CENPD-155-A CE Procedures for Design, Fabrication, Installation and Inspection of Surveil-lance Specimen Brackets 1.6-3 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 CONTROL WIRING DIAGRAM INDEX Figure 1.6-2 Amendment No. 16, (1/98) 1.6-4 1.7 DESIGN COMPLIANCE WITH AEC SAFETY GUIDES, INFORMATION GUIDES AND CODE OF FEDERAL REGULATIONS CROSS REFERENCES Information contained herein was valid at the time the operating license for St. Lucie 1 was issued, and is being retained in the updated FSAR for document completeness and historical record. No present or future update of
this section is required.
Conformance with the AEC General Design Criteria is discussed in Section 3.1. The criteria requirements and cross references indicated in the AEC Safety Guides and Information Guides are presented in Table 1.7-1.
1.7-1 TABLE 1.7-1 DESIGN COMPLIANCEDocument/Title/GDC ReferencesCriteriaRequirementsCompliance -
FSAR SectionSafety Guide 1 -
Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps - GDC 411.Emergency core cooling and containmentheat removal system pumps should have adequate NPSH at maximum expected pumped fluid temperatures assuming no increase in containment pressure over that present prior to a postulated LOCA.
6.2.2.2; 6.3.2Safety Guide 2 -
Thermal Shock to Reactor Pressure Vessels - GDC 35 5.41.Data collection and research on theproperties of reactor pressure vessel material should be continued to verify that non-brittle behavior can be assured throughout the vessel lifetime under postulated accident conditions.2.No significant changes in the approvedcore or reactor pressure vessel designs will occur. This negates the need to review in this individual case the potential thermal shock problem.3.The-vessel design does not precludeannealing should the safety margin to brittle failure appear to become unacceptable under the assumption of cooling operation.Safety Guide 3 -1.Not applicable.Safety Guide 4 -
Assumptions Used for
Evaluating the Potential Radiological Consequences of a LOCA for Pressurized Water Reactors1.The radioactive material releaseatmosphere diffusion and dose conversion assumptions are followed in evaluating the design basis LOCA.
15.41.7-2
TABLE 1.7-1 (Continued)Document/Title/GDC ReferencesCriteriaRequirementsCompliance -
FSAR Sectionrestored within 90% and frequency within 98% in < 40% of load sequence time interval.4. Prototype qualification data and preop testsshould be confirmed on each diesel.Safety Guide 10 -
Mechanical (Cadweld)
Splices in Reinforcing Bars of Concrete Containments - GDC 11. Each member of the splicing crew shouldprepare two qualification splices using the same materials as those used in the structure. The qualification and production splices should meet the requirements specified by the designer and approved by the licensee.
12.1.22. The qualification and production splicesshould be inspected at both ends of the splice sleeve and at the center tap hole in accordance with the requirements specified by the designer and approved by the licensee.3.Qualification and production splice samplesfor tensile testing should test equal to or in excess of 125 percent of the minimum yield strength according to ASTM Standards, and the average tensile strength of each group of 15 consecutive samples should equal or exceed the guaranteed ultimate strength specified for the reinforcing bar.4. Tensile Test Frequency should followingthe Guide schedule for production splices and sister splices, if provided.5. The Procedure for Sub-Standard TensileTest Results should follow the Guide schedule for retesting, cessation of mechanical splicing, independent laboratory analysis, balance of production rejection and reduced strength acceptability.1.7-5
TABLE 1.7-1 (Continued)Document/Title/GDC ReferencesCriteriaRequirementsCompliance -
FSAR Sectiond.Axes orientations orthogonal.
e.Locations in vertical line.
f.Locations accessible.
g.Rigidly attached to the containmentstructure.2. Peak deflection accelerographs should beinstalled on other selected category I structures, the need to be evaluated on a case basis.3. If different soil conditions underlieindependent foundations containing Category I components, additional instrumentation should be provided, the need to be evaluated on a case by case basis. If needed, a free-field accelerograph should be installed.4. The peak acceleration level in the reactorcontainment basement should be available to the control room operator a few minutes after a postulated earthquake.5. The accelerographs should be designed tofunction in the expected range of environment conditions.6. A plan for utilization of any data recordedshould be available.Safety Guide 13 -
Fuel Storage Facility Design Basis - GDC 611. The spent fuel facility should meetCategory I seismic requirements.
9.1; 15.42. The facility should be designed to preventtornado winds or missiles from causing significant pool water loss or fuel damage.3. Interlocks should be provided to preventcranes passing over or within striking distance of fuel when fuel handling is not in progress.4. The fuel pool enclosure should be providedfor radioactive particulate cleanup and controlled leakage. The ventilation and filtration capability is assessed on the basis that one fuel bundle clad-gap activity might be released.1.7-7 TABLE 1.7-1 (Continued)Document/Title/GDC References Criteria Requirements Compliance -
FSAR Section 5. The fuel pool structure should be designedto withstand a cask drop without substantial water loss or fuel damage.- or -The cranes capable of carrying heavy loads (including the spent fuel cask) should be designed to provide single-failure-proof handling of heavy loads.6. Water cleanup, makeup and drainconnections should be designed to preclude substantial water loss in case of maloperation or line rupture.7. Low water and high radiation conditionsshould be alarmed in control room.8. The water makeup system should beCategory I and redundant backup water makeup arrangements should be available.
The backup system need not be permanently installed.9.1-2.2.2Safety Guide 14 -
Reactor Coolant Pump Flywheel Integrity -
GDC 4 5.51.Flywheel material should meet the followingtest criteria:
- a. NDT 10 Fb. Three specimens exhibit C v (WR) 50ft-lbc. Minimum fracture toughness atoperating temperature equivalent to K Icdynamic 100 ksi in.d. 100% U.T. using equivalentacceptance criteria of ASME B & PV Code Section III.e. Flame cut flywheels have 1/2 in. of stockleft for machining to size.f. Finish bores, keyways and drill holesMP or LP tested.2.The flywheel should be designed to meetthe following criteria:
- a. Normal operative speed stresses 1/3minimum specified yield strength.b. Design overspeed 110% anticipatedmaximum overspeed. The basis for the assumed design overspeed should be stated.c. Design overspeed stresses 2/3minimum specified yield strength.1.7-8Amendment No. 20 (4/04)
TABLE 1.7-1 (Continued)
Document/Title/
GDC References Criteria Requirements Compliance -
FSAR Section instruments calibrated to same standards
used for procedure qualification.
5.If item 4.
is not met, additional side bend
examinations should be done.
6.Production welds heat input requirements in
item 1 should be met.
Safety Guide 32 -
Use of IEEE STD-308-
1971, "Criteria for Class
1E Electric Systems..."
- GDC 17;8.2 IEEE STD 308-1971 may be used in
implementing GDC 17 except where conflict in
resolution occurs; 1.Availability of offsit e Power - a preferred design includes two immediate access
circuits from the transmission network; an
acceptable design would substitute a
delayed access circuit if it conforms to
Criterion 17 and the other circuit is
immediately accessible.
2.The capacit y of the battery charger supply
should be based on the largest combined
demands of steady-state loads and the
charging capacity to restore the battery to
the fully charged state.
Safety Guide 33 -
Quality Assurance
Program Requirements (operation) - 10 CFR50, Appendix B The requirements for administrative controls to
safely operate nuclear power plants given in
proposed standard ANS-3.2 (Nov. 2, 1972) and
NSI N45.2 - 1971 are generally acceptable for
complying with 10 CRF50, Appendix B.
Appendix A to Safety Guide 33 should be used
to assure minimum plant operating and
maintenance procedures coverage.
17.2 1.7-19 Amendment No. 17, (10/99)
TABLE 1.7-1 (Continued)Document/Title/GDC ReferencesCriteriaRequirementsCompliance -
FSAR Sectionb.List and discussion of systemsdissimilar along with safety significance evaluation of differences3. Seismic design criteria 3.10a.Protective action initiation during DBE.
b.ESF operation after accident plus DBE
- c. Documentation of equipmentqualification tests4. Quality assurance description as applied toreactor protection, ESF and emergency power equipment 7.1; 17.1 7.1; 8.3.1.25. Criteria and bases to establish minimumindependence and redundance of subjectsystems and their provision quality assurance including electrical cable.
a.Derating b.Routing in hostile areas c.Vital and non-vital sharing in trays d.Fire detection and protection e.Marking and tray marking, and f.Spacing of wires and components atterminations6. Design criteria to protection from potentialeffects of normal and accident radiation levels. Documentation of equipment radiation qualification tests.
3.117. Description of and documentation ofqualification tests on safety related equipment within the containment to withstand the DBA environment.
3.118. Identify limiting temperature oninstrumentation and control equipment that requires reactor shutdown, expected worst case temp.-humidity environment and criteria to loss of HVAC will 3.111.7-21