LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Tables 5.2-1 to 5.2-42

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Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Tables 5.2-1 to 5.2-42
ML17046A352
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  • TABLE 5.2-1 REACTOR COOLANT SYSTEM DESIGN PRESSURE SETTINGS Design Pressure Operating Pressure Safety Valves Power Relief Valves Pressurizer spray Valves (Begin to Open) Pressurizer Spray Valves (Full Open) High Pressure Trip High Pressure Alarm Low Pressure Trip Low Pressure Alarm Hydrostatic Test Pressure Backup Beaters On Proportional Beaters (Begin to Operate) Proportional Heaters (Full Operation) 1 of 1 SGS-t1FSAR Pressure, 2485 2235 2485 2335 2260 2310 2385 2385 1865 1865 3107 2210 2250 2220 Psig Revision 15 June 12, 1996 TABLE 5.2-2 REACTOR COOLANT SYSTEM DESIGN PRESSURE DROP Unit 2 Across Pump Discharge Leg Pressure Drop, psi with DFBN (estimated) 3. 1 111 Across Vessel, Including Nozzles 4 9. 9 111 Across Hot Leg 1. 2 (1) Across Steam Generator 33. 8 111 Across Pump Suction Leg 2. 9 111 Total Pressure Drop 90. 9 111 Unit 1 Across Pump Discharge Leg 1.5 Across Vessel, Including Nozzles 52.0 Across Hot Leg 1.9 Across Steam Generator 35.67 Across Pump Suction Leg 1.8 Total Pressure Drop 92.87 I NOTES: Pressure Drop, psi with SDFBN (estimated) 121 3.1 49.3 1.2 34.4 2.9 90.9 3.1 49.3 1.1 35.8 2.9 92.2 1) Based on Best Estimate Flow (BEF) with Replacement Steam Generators RSGs at 0% tube plugging and Tave = 566° F. 2) Based on Best Estimate Flow (BEF) of 94,200 gpm/loop for Salem 1 and 94,800 gpm/loop for Salem 2 with Replacement Steam Generators RSGs at 0% tube plugging and Tave = 566° F. 1 of 1 SGS-UFSAR Revision 28 May 22, 2015
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  • TABLE 5.2-3 REACTOR VESSEL DESIGN DATA Design/Operating Pressure, psig Hydrostatic Test Pressure, psig Design Temperature, °F Overall Height of Vessel and Closure Heat, ft-in. (bottom head OD to top of control rod mechanism adapter) Thickness of Insulation, min., in. Number of Reactor Closure Head Studs Diameter of Reactor Closure Head Studs, in. ID of Flange, in. OD of Flange, in. ID at Shell, in. Inlet Nozzle ID, in Outlet Nozzle ID, in. Clad Thickness, min., in . 1 of 2 2485/2235 3107 650 3 54 7 172.5 205 173 29 5/32 Revision 7 July 22, 1987 TABLE 5.2-3 (Cont.) Lower Head Thickness, min., in. (base metal) Vessel Belt-Line Thickness, min., in. (base metal) Closure Heat Thickness, in. Reactor Coolant Inlet Temperature, °F Reactor Coolant Outlet Temperature, °F Reactor Coolant Flow, lb/hr 3 Total Water Volume Below Core, ft Water Volume in Active Core Region, ft3 3 Total Water Volume to Top of Core, ft Total Water Volume to Coolant Piping Nozzles Centerline, ft3 Total Reactor Vessel Water Volume, (with core and internals in place), Total Reactor Coolant System Volume, Total Reactor Coolant System Volume, SGS-UFSAR ft3 ft3 ft3 2 of 2 5-3/8 8.5 7 Low T High T avg avg 530.2 542.7{Unit 1) 530.3 542.8(Unit 2) 601.8 613.l(Unit 1) 601.7 613.1(Unit 2) 127.3x10 6 125.3x106(Unit 127.9x10 6 125.8xl06(Unit 1050 665 2164 2959 4945 12,076 (Unit 1) 13, 011 (Unit 2) Revision 24 May 11, 2009 1) 2)

TABLE 5.2-4 PRESSURIZER AND PRESSURIZER RELIEF TANK DESIGN DATA Pressurizer Design/Operating Pressure, psig Hydrostatic Test Pressure (cold), psig Design/Operating Temperature, °F Water Volume, Full Power, ft3* Steam Volume, Full Power, ft3 surge Line Nozzle Diameter, in. Shell ID, in. Electric Heaters Capacity, kW Heatup Rate of Pressurizer (using heaters only)°F/hr Maximum spray rate, gpm Pressurizer Relief Tank Design Pressure, psig Rupture Disc Release Pressure, psig Design Temperature, °F Normal Water Temperature, °F Total Volume, ft3 Total Rupture Disc Relief Capacity, lb/hr 2485/2235 3107 680/653 1080 720 14 84 1800 55 (approximately) 800 100 100 340 Containment Ambient (120°F max.) 1800 1.60 X 106 *60 percent of net internal volume (maximum calculated power) 1 of 1 SGS-tJFSAR Revision 16 January 31, 1998 TABLE 5.2-5 STEAM GENERATOR DESIGN DATA* # (AREVA NP Model 61/19T) -Unit 2 Only Number of Steam Generators 4 Design Pressure (Reactor coolant/steam), psig 2485/1185 Reactor Coolant Hydrostatic Test Pressure (tube side-cold), psig 3107 Design Temperature (reactor coolant/steam), OF 650/600 Low T avg Reactor Coolant Flow, lb/hr 31.97 X 106 Total Heat Transfer Surface Area, ft2 66,236 Heat Transferred, Btu/hr 2961 X 106 Steam Conditions at Full Load, Outlet Nozzle: Steam Flow, lblhr 3. 76 X 106 Steam Pressure, PSIA 805 Maximum Moisture Carryover, wt percent <0.1 Feedwater, OF 432.8 Shell OD (upperllower}Maximum, in. 175.75 I 135.0 Number of U-tubes 5048 U-tube OD, in. 0.750 Tube Wall Thickness (minimum), in. 0.043 Number of Manways/ID, in. 4/16 Number of handholes/ID, in. 4/6 Number of Inspection Ports/ID, in. 13 I 2.5 #Values are based on thermal design flow 1 of 2 SGS-UFSAR High T avg 31.46 X 106 2961 X 106 3.77 X 106 900 432.8 Revision 24 May 11, 2009 TABLE 5. 2-5 (Cant) STEAM GENERATOR DESIGN DATA*# (AREVA NP Model 61/19T) -Unit Rated Reactor Coolant Water Volume, ft3 1201 Primary Side Fluid Heat Content, Btu 42.34 Secondary Side Water Volume, ft3 2101 Secondary Side Steam Volume, ft3 3496 Secondary Side Steam Fluid Heat Content, Btu 6.113 #Values are based on thermal design flow 2 of 2 SGS-UFSAR 2 Only Load X 106 X 107 HISTORICAL INFORMATION No Load 1080 27.7 X 106 3524 2344 9.628 X 107 Revision 24 May 11, 2009

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  • TABLE 5.2-Sa STEAM GENERATOR DESIGN DATA* (Model F) -Unit 1 Only Number of Steam Generators Design Pressure (Reactor coolant/steam), psig Reactor Coolant Hydrostatic Test Pressure (tube side-cold), psig Design Temperature (reactor coolant/steam), °F Reactor Coolant Flow, lb/hr 2 Total Heat Transfer Surface Area, ft Heat Transferred, Btu/hr Steam Conditions at Full Load, Outlet Nozzle: Steam Flow, lb/hr Steam Temperature, °F Steam Pressure, PSIA Maximum Moisture Carryover, wt percent Feedwater, °F Overall Height, ft-in. Shell OD (upper/lower), in. Number of U-tubes U-tube OD, in. Tube Wall Thickness (minimum), in. Number of Inspection Openings/ID, in. Number of Manways/ID, in. Number of handholes/ID, in. Reactor coolant Volume, ft3 (Rated Load) Reactor coolant Volume, ft3 (No Load) *Quantities are for each steam generator 1 of 1 SGS-UFSAR 4 2485/1185 3107 650/600 Low T avg 6 31.83 X 10 55,050 2961 X 106 3.76 X 106 515.0 778 0.25 432.8 67-8 High T avg 6 31.33 X 10 2961 X 106 3.78 X 106 527.8 869 432.8 176.25 I 135.42 5626 0.688 0.041 4/2.7 4/16 6/6 966.1 966.1 Revision 19 November 19, 2001 I
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  • TABLE 5.2-6 REACTOR COOLANT PUMPS DESIGN DATA (Model 93A) Number of Pumps Design Pressure/Operating Pressure, psig Hydrostatic Test Pressure (cold), psig Design Temperature (casing), °F RPM at Nameplate Rating Suction Temperature, °F Developed Head, ft Capacity, gpm Seal Water Injection, gpm Seal Water Return, gpm Pump Discharge Nozzle ID, in. Pump Suction Nozzle ID, in. Overall Unit Height, ft-in Water Volume, ft3 Pump-Motor Moment of Inertia, 1b-ft2 Motor Data: Type Voltage Insulation Class Phase Frequency, cps Starting Current, amp Input (hot reactor coolant), kW Input (cold reactor coolant), kW Power, Hp (nameplate) Pump Weight, lb (dry) 1 of 1 SGS-UFSAR 4 2485/2235 3107 650 1180 559 277 8 3 27 1/2 31 25-5 1/4 56 82,000 AC Induction Single Speed, Air Cooled 4160 B Thermalastic Epoxy 3 60 4800 4260 5690 6000 169,200 Revision 6 February 15, 1987
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  • TABLE 5.2-7 REACTOR COOLANT PIPING DESIGN PARAMETERS Reactor Inlet Piping ID, in. Reactor Inlet Piping Nominal Thickness, in. Reactor Outlet Piping ID, in. Reactor Outlet Nominal Thickness, in. Coolant Pump Suction Piping ID, in. Coolant Pump Suction Piping Nominal Thickness, in . Pressurizer Surge Line Piping ID, in. Pressurizer Surge Line Piping Nominal Thickness, in. Design/Operating Pressure, psig Hydrostatic Test Pressure (Cold), psig Design Temperature, °F Design Temperature (pressurizer surge line), °F 1 of 2 SGS-UFSAR 27.5 2.38 29 2.50 31 2.66 (1) (2) 2485/2235 3107 650 680 Revision 7 July 22, 1987 TABLE 5.2-7 Water Volume, (all 4 loops including surge line)/ ft3 Design Pressure (pressurizer relief lines) 1 psig Design 'I'emperature (pressurizer relief lines) , 0E' (1) Onit 111.500", Unit 2 11.188" (2) Unit 1 1.25°, Unit 2 1.406" (3)
  • From pressurizer to safety valve 2485 psig 650°F. 1455 (3) (3) From safety valve to pressurizer relief tank 600 psig 600°F. 2 of 2 SGS-UFSAR Revision 22 May 5, 2006 I TABLE 5.2-8 PRESSURIZER VALVES DESIGN PARAMETERS PRESSURIZER SPRAY CONTROL VALVES Number of Valves Design Pressure Design Temperature Design Flow (valves full open, each) Fluid Temperature Position (after failure of actuating force) SAFETY VALVES 1. VALVE PARAMETERS Number of Valves Manufacturer Type Point Size Rated Capacity (Saturated Steam) Design Pressure and Temp. Constant Back Pressure Normal Developed Inlet Flange Rating Discharge Flange Rating 2. INLET PIPING PARAMETERS Diameter Length Loop 3 Loop 4 Loop 5 POWER OPERATED RELIEF VALVES Number of Valves Manufacturer Type Set Point Size Rated Capacity (Saturated Steam) Design Pressure and Temp. Valve SGS-UFSAR 1 of 2 2/Unit 2485 psig 650°F 400 gpm 545°F Closed 3/Unit Crosby Valve and Gage Co. Crosby HB-BP-86 6M6 Safety Valve (Steam Internals) 2485 psig 6" Inlet x 6" Outlet Orifice Size = 2.154 (3.644 sq. in.2) 420,000 lb/hr each 2485 psig and 650°F 3-5 psig 350 psig 1500 #ASA 600 #ASA 6" Sch 160 Unit 1 14.553' 12.873' 12.309' 2/Unit Unit 2 12.054' 12.241' 11.719' Copes-Vulcan Division Diaphragm Operated Relief Valve *2335 psig 2" Valve with 3" inlet and outlet BW connection Orifice 2" 210,000 lb/hr at 2335 psig 2485 psig and 680° F 1500 #ASA Revision 20 May 6, 2003 PORV BLOCK VALVES Number of Valves Valve Manufacturer Operator Manufacturer Type Valve Rating TABLE 5.2-8 (Cont.) 2/Unit Velan Engineering Co. Limitorque 3" Motor Operated Gate Valve 3GM56FN with BW ends and SMB-00-15 motor operator 1500 #ASA
  • Pressurizer Relief Valves lift at 2335 psig and reset at 2315 psig. 2 of 2 SGS-UFSAR Revision 20 May 6, 2003 I Component Reactor Vessel Replacement Reactor Vessel Closure Head Steam Generator* F/L CRDMs RC Pump Pressurizer Przr Relief Tank Przr Safety Valves RC SGS-UFSAR ASME III ASME III ASME III ASME III No Code 1\..SME III ASME III ASME III ASA B31.1 TABLE 5.2-9A UNIT 1 REACTOR SYSTEM -CODES Date & Addenda 1965 & all thru Winter 1965 1998 & all thru Summer 2000 1971 & all thru Summer 1973 1965 & all thru Summer 1966 (Design per ASME III, Article 4) 1965 & all thru Winter 1966 1968 & all thru Summer 1968 1968 & all thru Summer 1968 1955 1 of 2 Code Cases All applicable in effect prior to 4/26/66 All applicable in effect prior to 1971, 1484-3, 1528-3 & N474-l All applicable in effect at the time All applicable in effect at the time Applicable portions of ASA N-7 and N-10 Revision 22 Hay 5, 2006 I Sys tern Valves TIS!\ r< i 1 . 1 ASl\ R I . or 1 964 MSS-SP-66, or 1964 ASME III 1968 & Applicable portions of ASA N-7 and N-10 portions of N-10
  • The steam generators were anrl ir1stalled in accordance with NRC GL 89-09 to meet ASME III Section III Class 1 requirements. Lower narrow range level taps conform to 1989 ASME Section III Class 1 reconciled to the 1nal construction code. The tube side and the shell side conform to the rements of ASME Section III for Class l vessels. The steam generators were NPT stamped by the manufacturer prior to hydrostatic test . The tube side and the shell side were subsequently hydrostatic pressure tested prior to installation at Unit 1. The primary piping to steam generator primary inlet and outlet welds conform to the requirements of the 1989 Edition of the ASME Code Section III for Class l piping. Applicable Code Cases are N-416-1 and N-389. 2 of 2 SGS-UFSAR Revision 18 April 26, 2000 Component Reactor Vessel Reactor Vessel Closure Head Steam Generator F/L CRDMs Reactor Coolant Pump Casing Pressurizer Pressurizer Relief Tank Pressurizer Safety Valves Reactor Coolant Piping System Piping and Fittings System Valves TABLE 5.2-9B REACTOR COOLANT SYSTEM -CODES UNIT 2 ASME III ASME III ASME III ASME III ASME III ASME III ASME III ASME III USAS B31.1.0 USAS B31.1.0 B16.5, or MSS-SP-66, or ASME III Date and Addenda Code Cases 1965 and all addenda through All applicable in effect Winter 1966 prior to 4/3/67 1998 with addenda through 2000 1995 and all addenda through N-20-4 1996 1998 with addenda through 2000 1965 and all addenda through All applicable in effect Winter 1966 at the time 1968 and all addenda through All applicable in effect Summer 1968 at the time 1968 and all addenda through Summer 1968 1967it Applicable portions of ASA N-7 and N-10 1967# Applicable portions of ASA N-7 and N-10 1964 Applicable portions of N-10 1964 1968 #RCS fabrication, installation, welding, and examination involved in installing the Unit 2 Replacement Steam Generators utilized ASME Section XI (1998 Edition with 2000 Addenda) and ASME Section III, Subsection NB (1995 Edition with 1996 Addenda) . Both of these later codes are NRC-endorsed per 10CFR 50. 55a and were reconciled to the original construction codes. 1 of 1 SGS-UFSAR Revision 24 May 11, 2009 I I I TABLE 5.2-10 DESIGN THERMAL AND LOADING CYCLES* (ll AREVA NP Model 61/19T SG -Unit 2 1. Heatup at 100°F/hr Cooldown at 100°F/hr (Pressurizer 200°F/hr) 2. Unit Loading at 5 Percent of Full Power/Min Unit Unloading at 5 Percent of Full Power/Min 2a. Unit Loading at 5 Percent of Full Power/Min Unit Unloading at 5 Percent of Full Power/Min 3. Step Load Increase of 10 Percent of Full Power Step Load Decrease of 10 Percent of Full Power 4. 50 Percent Step Decrease in Load (with steam dump) 5. Loss of Load (without immediate turbine or reactor trip) 6. Loss of Power (blackout with natural circulation in the RCS) 7. Loss of Flow (partial loss of flow one pump only) 8. Reactor Trip From Full Power 9. Turbine Roll Test 10. Hydrostatic Test Conditions a. Primary Side Hydrostatic Test Shop and Field b. Secondary Side Hydrostatic Test Before Initial Startup 11. Primary Side Leak Test 12. Accident Conditions a. Reactor Coolant Pipe Break b. Steam Pipe Break c. Steam Generator Tube Rupture 1 of 2 SGS-UFSAR 200 200 18/300 18,300 14,500 14,500 21000 2,000 200 80 40 80 400 10 10 10 50 1 1 1 (Unit 2 RRVCH) (Unit 2 RRVCH) Revision 24 May 11, 2009 TABLE 5.2-10 (Cont} DESIGN THERMAL AND LOADING CYCLES* (ll AREVA NP Model 61/19T SG -Unit 2 Design Cycles** 13. Steady State Fluctuations -the reactor coolant average temperature for purposes of design is assumed to increase and decrease a maximum of 6°F in one minute. The corresponding reactor coolant pressure variation is less than 100 psi. It is assumed that an infinite number of such fluctuations will occur. 14. Design Earthquake Cycles * ** (1) a. Operating Basis Earthquake b. Design Basis Earthquake The ASME Section III Nuclear Power Plant Components Code is inapplicable to the Salem Station; hence, the normal, upset, emergency, and faulted conditions terminology does not apply to the transients identified in this table. However, since the RCS vessels (reactor and steam generators) are basically standard components, analysis on these vessels with the more recent ASME Code conditions (normal, upset, emergency, and faulted) have been performed as discussed in Sections 5.1.2.8.1 and 5.1.2.8.2. Estimated for equipment design purposes (40-year life) and not intended to be an accurate representation of actual transients or to reflect actual operating experience. 50 10 Note that the actual transient definition for design purposes may be more stringent than listed in this table. 2 of 2 SGS-UFSAR Revision 24 May 11, 2009
1. 2. TABLE 5.2-lOa DESIGN THERMAL AND LOADING CYCLES* Model F SG Unit 1 Heatup at 100°F'/hr Cooldown at 100°F/hr (Pressurizer 200°F/hr) Unit Loading at 5 Percent of Full Power/Min Unit Unloading at 5 Percent of Full Power/Min 2a. Unit Loading at 5 Percent of Full Power/Min Unit Unloading at 5 Percent of Full Power/Min 3. 4. Step Load Increase of 10 Percent of Full Power Load Decrease of 10 Percent of Full Power 50 Percent Step Decrease in Load (with steam dump) Design Cycles** 200 200 13,200*** 13,200*** 14,500 (Unit 1 RRVCH) 14,500 (Unit 1 RRVCH) 2,000 2,000 200 5. Loss of Load (without immediate turbine or reactor trip) 80 6. 7. 8. 9. Loss of Power (blackout with natural circulation in the RCS) Loss of Flow (partial loss of flow one pump only) Reactor '!'rip From Full Power Turbine Roll Test 10. Hydrostatic Test Conditions a. b. Primary Side Hydrostatic Test Before Initial Startup Secondary Side Hydrostatic Test Before Initial Startup 11. Primary Side Leak Test 12. Accident Conditions a. Reactor Coolant Pipe Break b. Steam Pipe Break c. Steam Generator 'I'ube Rupture 1 of 2 SGS-UFSAR 40 80 400 10 5 5 50 1 1 1 Revision 22 May 5, 2006 I TABLE 5.2-10a (Contl DESIGN THERMAL AND LOADING CYCLES* Model F SG -.Unit 1 Design Cycles** 13. Steady State Fluctuations -the reactor coolant average temperature for purposes of design is assumed to increase and decrease a maximum of 6°F in one minute. The corresponding reactor coolant pressure variation is less than 100 psi. It is assumed that an infinite number of such fluctuations will occur. 14. Design Earthquake Cycles a. Operating Basis Earthquake b. Design Basis Earthquake The ASME Section III Nuclear Power Plant Components Code is inapplicable to the Salem Station; hence, the normal, upset, emergency, and faulted conditions terminology does not apply to the transients identified in this table. However, since the RCS vessels (reactor vessel, pressurizer, and steam generators} are basically standard components, analysis on these vessels with the more ASME Code conditions (normal, upset, emergency, and faulted} have been performed as discussed in Sections 5.1.2.8.1 and 5.1.2.8.2. Estimated for equipment design purposes (40-year life) and not intended to be an of actual transients or to reflect actual operating experience. Model F steam generators on Unit 1 are designed to 13,200 cycles. 2 of 2 50 10 SGS-UFSAR Revision 18 April 26, 2000 TABLE 5. 2-11

SUMMARY

OF PLANT OUTAGE FOR YANKEE ROW {1964 to 1969)

  • Starting Duration Date Outage Tne Case 1/17/64 3.1 Forced Turbine Trip 2/12/64 21.8 Scheduled Control Rod Drop Testing 3/11/64 4.5 Forced Moisture Separator Level Switch Tripped due to Vibration 3/26/64 4 Forced Control Valves Sticking 5/18/64 5.4 Forced Low Condensate Pump Discharge pressure 8/2/64 35 Scheduled Refueling and General Maintenance 9/9/64 2.4 Scheduled Check of Overspeed Trip 9/11/64 14.7 Forced Spurious Reactor Trip 10/18/64 12.2 Forced Condenser Noise
  • 10/22/64 22.4 Forced Neutron Counter Gain Control 2/12/65 15.2 Forced Switchyard Electric 3/5/65 Scheduled Switchyard Electric 8/9/65 93 6 Scheduled Refueling 11/26/65 2 20 Scheduled Turbine Repair-Physics Testing 2/4/66 3.12 Forced Reactor Scram 4/4/66 89.5 Scheduled Leaking Pressurizer Safety Valves 7/10/66 3.68 Forced Reactor Scram 8/25/66 2.40 Forced Reactor Scram
  • 1 of 2 SGS-UFSAR Revision 6 February 15, 1987 TABLE 5.2-11 (Cant)
  • Starting Duration Date Days/Hours Outage Type Case Equipment/System 10/4/66 34 10.23 Scheduled Refueling 12/24/66 2.88 Forced Reactor Scram 12/28/66 2.12 Forced Reactor Scram 3/8/67 11 21 Scheduled Steam Generator Leak Repair 5/12/67 16.87 Scheduled Condenser Cleaning 7/9/67 17 1.5 Scheduled Steam Generator Leak Repairs 10/28/67 9 Scheduled AEC Operator Examinations 10/13/67 2.6 Forced Reactor Scram 3/23/68 38 Scheduled Core VI-VII Refueling and
  • Maintenance 7/20/68 1 10 Scheduled Repair Leak from No. 1 Main Coolant Pump Stator Cap 11/8/68 6 16.42 Scheduled Repair No. 4 Main Coolant Pump Thermal Barrier Leak and other Maintenance 1/18/69 1 2.1 Scheduled Operator Training 2/15/69 1 1.8 Scheduled Operator Training 3/1/69 11 Scheduled AEC Operator Examination 4/11/69 4 18 Forced Repair Reactor Instrument Leak 7/17/69 4.8 Forced Reactor Scram 8/2/69 53 18.5 Scheduled Refueling Maintenance 10/16/69 6.1 Forced Reactor Scram
  • 10/29/69 12 Scheduled Turbine Valve Flange Steam Leak Repair 2 of 2 SGS-UFSAR Revision 6 February 15, 1987 TABLE 5.2-12 LOAD COMBINATIONS AND STRESS LIMITS Load Combination 1. Normal (deadweight, thermal and 2. Normal and Operation Basis Earthquake 3. Normal and Design Basis Earthquake 4. Normal and 5. Normal and Design Basis Earthquake and Pipe Rupture Stress Limit* Normal Conditions 'Jpset Condition Faulted Condition Faulted Conditicn Faulted Conditicn 1. Any condition in the course of system power range and system shutdown, in the absence of Upset, Emergency, or Faulted Conditions. 2. deviations from Normal Conditions anticipated to occur enough design should include a capability to withstand the conditions without operational impairment. The Upset Condition includes those transients caused by a fault in a component its isolation from the transients due to a loss of load or power and any system upset not resulting in a forced outage. The estimated duration of an Upset Condition shall be included in the Design The Conditions include the effect of the for which the system must remain or must regain its operational status. 3. Emergency Condition -Any deviations from Normal Conditions which shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any in the The total number of postulated occurrences for such events shall not exceed twenty-five. 4. Those combinations of conditions associated with postulated events whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent where considerations of public health and safety are involved. Such considerations with criteria as may be a specified by jurisdictional authorities. Among the Faulted Conditions may be a specified earthquake for which safe shutdown is *Definition of Operating Condition categories from Summer 1968 Addenda to the ASME Boiler and Pressure Vessel Code,Section III. 1 of 1 SGS-UFSAR Revision 25 October 26, 2010 TABLE 5.2-13 LOADING CONDITIONS AND STRESS LIMITS: PRESSURE VESSELS Stress Intensity Limits (a) p ::;s m m 1. Normal Condition (b) p (or PL) + PB ::; 1. 58 m m 1 (c) p (or PL) PB + Q m 2 ::; 3.08 m 2. Upset Condition (a) p ::; s m m (b) p (or PL) + PB ::; 1.58 m m 1 (c) p (or PL) PB + Q m 2 ::; 3.08 m 3. Emergency Condition (a) P ::; 1.258 or s , m m y whichever is larger or 1.5S , whichever is larger y 4. Faulted Condition Design Limit Curves as discussed 4,5 in the text and attached. For the where: p m s u SGS-UFSAR Unit 2 RSG, ASME criteria were Applied. primary general membrane stress intensity primary local membrane stress intensity primary bending stress intensity secondary stress intensity stress intensity value for ASME B and PV Code,Section III, Nuclear Vessels minimum specified material yield (ASME B and PV Code,Section III, Table N-421 or equivalent) 1 of 4 Revision 24 May 11, 2009 I TABLE 5.2-13 (Cont.) LOADING COMBINATIONS AND STRESS LIMITS: PRESSURE PIPING I.oading Conditions Stress Intensity Limits 1. Normal Conditions (a} p c:;s m (b) p m (or P1) + P8 c:;s (a) p ::::;1.25 m 2. Upset Conditions (b) p m (or P1) + P8 ::::;1. 28 (a) p ::::1.2s m 3. Emergency Conditions (b) p (or P1) + PB s(1.5)(1.2) s m 4. Faulted Conditions Design Limit Curves as discussed in the text and attached. OR Maximum stress s 2.4 S where: P primary general membrane stress intensity m PL primary local membrane stress intensity P8 primary bending s.tress intensity S allowable stress from USASI 831.1 Code for Pressure Piping 2 of 4 SGS-UFSAR Revision 22 May 5, 2006
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  • TABLE 5.2-13 (Cont) LOADING CONDITIONS AND STRESS LIMITS: EQUIPMENT SUPPORTS Loading Conditions Stress Intensity Limits 1. Normal Condition Working Stresses or Applicable Factored Load Design Values 2. Upset Condition Working Stress or Applicable Fac-tored Load Design Values 3. Emergency Condition Within yield after load redistribution 4. Faulted Condition Permanent Deflection of Supports Limited To Maintain Supported Equipment Within Design Limit Curves as Discussed in the Text. Note 1: Note 2: SGS-UFSAR The limits on local membrane stress intensity (PL <1.5Sm) and primary membrane plus primary bending stress intensity (PL (or PL) + PB .s_l.5Sm) need not be satisfied at a specific location if it can be shown by means of limit analysis or by tests that the specified loadings do not exceed of the lower bound collapse load as per paragraph N-417.6 (b) of the ASME Band PV Code,Section III, Nuclear Vessels. In lieu of satisfying the local membrane (PL .s_1.5S) stress intensity (PL + specific requirements for the or the primary plus secondary PB + O.s_3Sm) at a specific location, the structural action may be calculated on a plastic basis and the design will be considered to be acceptable if shakedown occurs, as opposed to continuing deformation, and if the deformations which occur prior to shakedown do not exceed specified limits, as per 3 of 4 Revision 6 February 15, 1987 Table 5.2-13 (Cont.) LOADING CONDITIONS AND STRESS LIMITS: EQUIPMENT SUPPORTS Note 3: Note 4: Note 5: SGS-UFSAR paragraph N-417. 6 (a) (2) of the ASME B and PV Code,Section III, Nuclear Vessels. The limits on local membrane stress intensity (PL and primary membrane plus primary bending stress intensity :s;1. 58 ) need not be satisfied at a specific location if it can be m shown by means of limit analysis or by tests that the loadings do not exceed 120 percent of 2/3 of the lower bound collapse load as per paragraph N-417.10 (c) of the ASME Band PV Code,Section III, Nuclear Vessels. As an alternate to the design limit curves which represent a pseudo plastic instability analysis, a plastic instability analysis may be performed in some cases considering the actual strain-hardening characteristics of the but with yield strength adjusted to correspond to the tabulated value at the temperature in Table N-424 or N-425, as per paragraph N-417.11 (c) of the ASME B and PV Code,Section III, Nuclear Vessels. These cases will be justified on an individual basis. For the Unit 2 RSG, the faulted condition criteria is as provided in ASME Code Section III, Division 1, Sub-section NB, Appendix F, Sub-paragraph 1331.1. 4 of 4 Revision 24 May 11, 2009 I
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  • TABLE 5.2-14

SUMMARY

OF PRIMARY PLUS SECONDARY STRESS INTENSITY RANGES FOR COMPONENTS OF THE REACTOR VESSEL Control Rod Housing Head Flange Vessel Flange Primary Nozzles Stud Bolts Vessel Support Core Support Pads Bottom Head to Shell Bottom Instrumentation Penetrations Vessel Wall Transition Range of Stress Intensity (psi) 77,760

  • 65,260 60,040 57,090 109,400 ** 47,240 34, 690 67,110 33,730 Allowable Stress (psi) (at Operating Temperatures) 69,900 80,100 80,100 80,100 110, 400 80,100 69,900 80,100 69,900 80,100
  • Justified by simplified elastic-plastic analysis ** Lower than primary nozzle stress SGS-UFSAR 1 of 1 Revision 19 November 19, 2001 TABLE 5.2-15

SUMMARY

OF CUMULATIVE FATIGUE USAGE FACTORS FOR COMPONENTS OF THE REACTOR VESSEL Item Control Rod Housing Replacement Head E'lange RVLIS/RVHV Penetrations Vessel Flange Stud Bolts Primary Nozzles Vessel Support Pads Core Support Pads (iateral) Bottom Head to Shell Bottom Instrumentation Penetrations *Covers all transients 0. 71. 0.263 0.75 0.183 0.89 0.1510 0.05 0.012 0.0118 0.1002 **As defined in Section III of the ASME Boiler and Pressure Vessel Code, Nuclear Vessels. 1 of 1 SGS-UFSAR Revision 22 May 5, 2002 Note -TABLE 5.2-16 STRESS DUE TO MAXIMUM STEAM GENERATOR TUBE SHEET PRESSURE DIFFERENTIAL (2485 PSIG) Model 61/19T Unit 2 Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG Design Report," (Reference 31) for the required stresses and stress factors for the Salem Unit 2 Model 61/19T steam generators. 1 of 1 SGS-UFSAR Revision 24 May 11, 2009 Note -SGS-UFSAR TABLE 5.2-17 RATIO OF ALLOWABLE STRESS TO COMPUTED STRESSES FOR A STEAM GENERATOR TUBE SHEET PRESSURE DIFFERENTIAL OF 2485 PSIG Model 61/19T -Unit 2 Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG -Design Report,u (Reference 31) for the stress analyses associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 Revision 24 May 11, 2009 Note -TABLE 5.2-18 STEAM GENERATOR PRIMARY-SECONDARY BOUNDARY COMPONENTS Model 61/19T Unit 2 Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG -Design Report, 11 (Reference 31) for tabulations associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 SGS-UFSAR Revision 24 May 11, 2009 Note -SGS-UFSAR TABLE 5.2-19 STEAM GENERATOR PRIMARY-SECONDARY COMPONENTS Model 61/19T -Unit 2 Refer to VTD 900013 Sheet 11 "Salem Unit 2 RSG Design Report," (Reference 31) for tabulations associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 Revision 24 May 11, 2009 TABLE 5.2-20 STEAM GENERATOR PRIMARY-SECONDARY BOUNDARY COMPONENTS Model 61/19T -Unit 2 Note Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG -Design Report," (Reference 31) for tabulations associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 SGS-UFSAR Revision 24 May 11, 2009 Note -TABLE 5.2-21 STEAM GENERATOR PRIMARY -SECONDARY BOUNDARY COMPONENTS Model 61/19T -Unit 2 Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG -Design Report," (Reference 31) for tabulations associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 SGS-UFSAR Revision 24 May 11, 2009 Note -SGS-UFSAR TABLE 5.2-22 MODEL 61/19T STEAM GENERATOR USAGE FACTORS (INDIVIDUAL TRANSIENTS) PRIMARY AND SECONDARY BOUNDARY COMPONENTS UNIT 2 Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG -Design Report," (Reference 31) for tabulations associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 Revision 24 May 11, 2009 Note -SGS-UFSAR TABLE 5.2-23 MODEL 61/19T STEAM GENERATOR USAGE FACTORS (INDIVIDUAL TRANSIENT) UNIT 2 Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG-Design Report," (Reference 31) for tabulations associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 Revision 24 May 11, 2009 Note -SGS-UFSAR TABLE 5.2-24 TUBE SHEET STRESS ANALYSIS RESULTS FOR MODEL 61/19T STEAM GENERATORS UNIT 2 Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG -Design Report," (Reference 31) for tabulations associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 Revision 24 May 11, 2009 Note -TABLE 5.2-25 LIMIT ANALYSIS CALCULATION RESULTS TABLE OF STRAINS, LIMIT PRESSURES, AND FATIGUE EVALUATIONS FOR MODEL 61/19T STEAM GENERATORS UNIT 2 Refer to VTD 900013 Sheet 1, "Salem Unit 2 RSG Design Report," (Reference 31) for tabulations associated with the Salem Unit 2 Model 61/19T steam generators. 1 of 1 SGS-UFSAR Revision 24 May 11, 2009 TABLE 5.2-26 REACTOR COOLANT SYSTEM QUALITY ASSURANCE PROGRAM ComEonent RT* UT* PT* MT* ET* l. Steam Generator 1.1 Tube Sheet 1.1.1 Forging yes yes 1.1. 2 Cladding yes(1) yes(2) 1.2 Channel Head 1. 2 .1a Casting (Unit 1) yes yes 1. 2 .lb Forging (Unit 2) yes yes 1. 2. 2a Cladding (Unit 1) yes 1. 2. 2b Cladding (Unit 2) yes yes 1.3 Secondary Shell and Head 1.3.1a Plates (Unit 1) yes 1. 3. lb Forging (Unit 2) yes yes 1.4 Tubes yes yes 1.5 Nozzles {forgings) yes yes 1.6 Weldments 1. 6.1 Shell, longitudinal (Unit 1) yes yes 1. 6. 2a Shell, circumferential yes yes (Unit 1) 1. 6. 2b Shell, circumferential yes yes yes (Unit 2) 1. 6. 3a Cladding (channel head-tube sheet joint clad-ding restoration) (Unit 1) yes 1. 6. 3b Cladding (channel head-tube sheet joint clad-ding restoration) (Unit 2) yes yes 1. 6. 4a Steam and Feedwater Nozzle to Shell (Unit 1) yes yes 1. 6. 4b Steam and Feedwater Nozzle to Shell (Unit 2} yes yes yes 1. 6. 5a Support brackets (Unit 1) yes 1. 6. 5b Support brackets (Unit 2} yes 1. 6. 6a Tube to Tube Sheet (Unit 1) yes 1. 6. 6b Tube to Tube Sheet (Unit 2) yes yes yes 1.6.7a Instrument Connections (Unit 1) (secondary} for lower NR level taps. No primary connections. yes yes yes 1. 6. 7b Instrument Connections (Unit 2) (primary and secondary). yes 1 of 4 SGS-UFSAR Revision 24 May 11, 2009 TABLE 5.2-26 Com12onent RT* 1.6.8a Temporary Attachments After Removal (Unit 1) 1. 6. Bb Temporary Attachments After Removal (Unit 2) 1.6.9a After Hydrostatic Test (all welds and complete channel head -where accessible) (Unit 1) 1.6.9b After Hydrostatic Test (all welds and complete channel head -where accessible) (Unit 2) 1. 6.10 Nozzle Safe Ends (weld deposit) yes 2. Pressurizer 2.1 Heads 2 .1.1 Casting yes 2.1.2 Cladding 2.2 Shell 2.2.1 Plates 2.2.2 Cladding 2.3 Heaters 2.3.1 Tubing(4) 2.3.2 Centering of element yes 2.4 Nozzle 2.5 Weldments 2.5.1 Shell, longitudinal yes 2.5.2 Shell, circumferential yes 2.5.3 Cladding 2.5.4 Nozzle Safe End yes (if forging) 2.5.5 Nozzle Safe End (if weld deposit) 2.5.6 Instrument Connections 2.5.7 Support Skirt 2.5.8 Temporary Attachments After Removal 2.5.9 All Welds and Cast Heads After Hydrostatic Test 2.6 Final Assembly 2.6.1 All Accessible Surfaces After Hydrostatic Test 2 of 4 SGS-UFSAR (Cont) UT* yes yes yes yes yes yes yes yes yes yes yes yes yes yes MT* ET* yes yes yes yes yes yes yes yes yes yes Revision 24 May 11, 2009 LT*

TABLE 5.2-26 (Cant) Corneonent RT* UT* 3. Piping 3.1 Fittings and Pipe (Castings) yes 3.2 Fittings and Pipe (Forgings) yes 3.3 Weldments 3.3.1 Circumferential yes 3.3.2 Nozzle to Runpipe yes (No RT for nozzles less than 4 inches) 3.3.3 Instrument Connections yes 4. Pumps 4.1 Casting yes 4.2 Forgings 4.2.1 Main Shaft yes 4.2.2 Main Studs yes 4.2.3 Flywheel (Rolled Plate) yes 4.3 Weldments 4.3.1 Circumferential yes 4.3.2 Instrument Connections 5. Reactor Vessel 5.1 Forgings 5 .1.1 Flanges yes 5.1. 2 Studs yes 5.1.3 Head Adapters yes 5.1. 4 Head Adapter Tube yes 5.1. 5 Instrumentation Tube yes 5.1. 6 Main Nozzles yes 5 .1. 7 Nozzle Safe Ends yes (if forging is employed) 5.2 Plates yes 5.3 Weldments 5.3.1 Main Stearn yes 5.3.2 CRD Head Adapter Connection 5.3.3 Instrumentation Tube Connection 5.3.4 Main nozzles yes 5.3.5 Cladding yes(3) 5.3.6 Nozzle-Safe Ends yes (if forging) 3 of 4 SGS-UFSAR PT* yes yes yes yes yes yes yes yes yes yes yes yes yes yes yes yes yes yes MT* ET* yes yes yes yes yes yes Revision 24 May 11, 2009 LT*

Component 5.3.7 Nozzle Safe Ends (If weld deposit) 5.3.8 Head Adaptor Forging to Head Adapter Tube 5.3.9 All Welds After 6. 6.1 6.2 * (1) (2) (3) (4) Hydrotest Valves Castings Forgings (No UT for valves and smaller) RT -Radiographic UT -Ultrasonic PT -Dye Penetrant two inch MT -Magnetic Particle ET -Eddy Current LT -Leak Flat Surfaces Only Weld Deposit Areas Only OT of Clad Bond-to-Base Metal Or a UT and ET SGS-UFSAR TABLE 5.2-26 (Cont) RT* UT* yes yes yes yes 4 of 4 yes yes yes yes MT* yes Revision 24 May 11, 2009 ET* I Component Reactor Vessel Replacement Reactor Vessel Closure Head (Units 1 & 2) Steam Generator Unit 2 AREVA NP Model 61/19T Steam Generator Unit 1 -Model F Pressurizer Pressurizer Relief Tank Pipe Pump Valves SGS-UFSAR TABLE 5.2-27 MATERIALS CONSTRUCTION OF THE REACTOR COOLANT SYSTEM COMPONENTS Plate -Unit 2 Pressure Plate -Unit 1 Pressure Forgings Cladding, Stainless Stainless Weld Rods Head Seals Studs Instrumentation Nozzles Insulation Monoblock forging CRDM Lower Tube Pressure Plate Pressure Forgings Cladding for Heads Grade B Class 1 SA-302-B Grade B ASTM A-508 Class 2 Type 304 or equivalent Type 308, 309, or 312 Inconel 718 SA-540 Grade B-24 Inconel SB 167 Stainless Steel SA-508 Gr. 3 Cl. 1 SA-182 Type 304 SB-167, A690 (UNS N06690) SA-533, Type B, CL 2 SA-508, GR 3, CL 2 Stainless Steel 308L/309L Stainless Weld Rod Stainless Steel 308L/309L Cladding for Tube Sheets Inconel 600 Tubes Inconel 690 Thermally Treated Channel Head Divider Plate SB-168 UNSN 06690 Shell Material Forgings Cladding for Heads, Stainless Weld Rod Cladding for Sheets Tubes Channel Head Castings Shell Heads Support Skirt Nozzle Weld Ends Inst. Tube Coupling Cladding, Stainless Internal Plate Inst. Tubing Heater Well Tubing Heater Well Adaptor Shell Heads Internal Coating Pipes Fittings Nozzles Shaft Impeller Casing Pressure Parts 1 of 1 SA-533 Class 2 SA-508 Class 2a Type 308 or 309 ss Type 308L or 3091 Inconel Inconel SB-163, Code Case 1484-3 SA-216 Grade WCC SA-533 Class 1 SA-216 Grade WCC SA-516 Grade 70 SA-182 F316 SA-182 F316 Type 304 or equivalent SA-240 Type 304 SA-213 Type 304 SA-213 Type 316 Seamless SA-182 F316 ASTM A-285 Grade C ASTM A-285 Grade C Amercoat 55 ASTM A-376 Type 316 ASTM A-351 Grade CF8M ASTM A-182 Grade F316 ASTM A-182 Grade F347 ASTM A-351 Grade CF8 ASTM A-351 Grade CF8 ASTM A-351 Grade CF8M and ASTM A-182 Grade F316 Revision 25 October 26, 2010 TABLE 5.2-28 REACTOR COOLANT WATER CHEMISTRY SPECIFICATION Conductivity

  • Dissolved Oxygen , ppm, max. Chloride, ppm, max Fluoride, ppm, max. Sulfate, ppm, max Hydrogen, cc {STP)/kg H20 Suspended Solids, ppm, max. pH Control Agent (Li70H) Boric Acid as ppm B
  • 1.0 *** Up to 3.5 +/- 0.15 ppm steady state, in accordance with Station Lithium Program Variable from 0 to Limit not applicable with temperature < 250°F Control range during operation. Hydrogen may be reduced to 15 cc/kg 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a planned outage. * *
  • Limit < 350 ppb. SGS-UFSAR ONLY in Modes 1 & 2. Goal for continuous RCP 1 of l Revision 26 May 21, 2012 is
  • *
  • SGS-UFSAR TABLE 5.2-29 THIS PAGE INTENTIONALLY BLANK 1 of 1 Revision 12 July 22, 1992
  • *
  • Weld Salem 1 2-042 A/8/C 9-042 3-042 A/B/C Salem 2 2-442 A/B/C 9-442 3-442 A/B/C SGS-OFSAR TABLE 5.2-30 RPV 8ELTLINE REGION WELD CHEMISTRY FOR SALEM UNITS 1 AND 2 Weld Wire Heat/Flux Lot 348009/3692 & 348196/3692 Ni-200 Wire 13253/3791 348009/3708 + Ni-200 wire 13253/3833 & 20291/3833 90099/3977 21935/3889 & 12008/3889 Cu (w/o) 0.18 0.22 0.19 0.219 0.197 0.213 Ni (w/o) 1.04 0.73 1.04 0.735 0.060 o. 867 1 of 1 Basjs CE NPSD-1039, Rev 2 CE NPSD-1119, Rev 1 CE NPSD-1039, Rev 2 CE NPSD-1119, Rev 1 CE NPSD-1039, Rev 2 CE NPSD-1119, Rev 1 Revision 19 November 19, 2001
  • TABLE 5.2-31 RPV BELTLINE REGION WELD MECHANICAL PROPERTIES FOR SALEM UNITS Unit Weld Weld Wire Initial RT NOT Ill Weld Flux Seam Heat/Flux Lot Type Salem 1 2-042 348009/3692 -56 1092 A/8/C 398196/3692 -56 1092 Ni-200 Wire 9-042 13253/3791 -56 1092 3-042 34B009/3708 -56 1092 A/B/C + Ni-200 Wire Salem 2 2-442 13253/3833 -56 1092 A/B/C 20291/3833 -56 1092 9-442 90099/3977 -56 0091 3-442 21935/3889 -56 1092 A/B/C 12008/3889 -56 1092 (1) Generic RTNDT forC-E's SAW Weld; u: 17 1 of 1 SGS-UFSAR
  • 1 AND 2 Unirradiated USE (ft-lb} 96.2 112 112 96.2 99.7 114 Revision 19 November 19, 2001
  • Salem 1 82402-1 82402-2 82402-3 82403-1 82403-2 82403-3 Salem 2 84712-1 84712-2 84712-3 84713-1 84713-2 84713-3
  • TABLE 5.2-32 RPV BELTLINE REGION PLATE MATERIAL CHEMICAL AND MECHANICAL PROPERTIES FOR SALEM UNITS 1 AND 2 h . . . {1) C Compos1t1on Cu (w/o) Ni (w/o) 0.24 0.53 0.24 0.53 0.22 0.51 0.19 0.48 0.19 0.49 0.19 0.48 0.13 0.56 0.12 0.62 0.11 0.57 0.12 0.60 0.12 0.57 0.12 0.58 Mechanical Properties(!) Initial RTNDT (F) 45 3 4 18 6 0 12 10 8 8 10 Unirradiated USE (ft-lb) 91 98 104 93 83 85 106 97 107 98 103 121 (1) Measured data. All tests performed,by Westinghouse. 1 of 1 SGS-UFSAR
  • I Revision 19 November 19, 2001 TABLE 5.2-33 STRESS RESULTS OF UNIT 1 TUBESHEET AND SHELL JUNCTIONS ANALYSIS MODEL F SG Thin Cast Head Model LOCATION (l) 1 3 4 CONDITION: 0.11121 0.41 13) 0. 43 ( 3) 1.0(4){7) Normal and Upset (6) 0.97(5) ( 6) Fatigue Usage <0.71 <0.38 <0.41 Emergency o.os12) 0.2813) 0.51 {)) 0. 72141 Fault:ed 0.03'21 0.27()) 0. 34131 0. 7714) :'es-0. 07 {2) 0. 37 :31 0.65'"" 0.95(4) N:.::es: ( 1) See Figure 5.2-22 12) P (3) /Allowable ( 4) (PL .;. Pb) /Allowable ( 5) !PL + pb + Ql /Allowable : of 2 6 7 0.40(3) 0.28 '2) 0.61141 0. 71 {5) 0.65(S) <0.25 <0.25 0.23(3) 0.10(Z) 0.2514) 0.33'31 0. 61 ( 41 0.29 (3) 0.08(2) 0.23(4) Revision 18 April 26, 2000 TABLE 5.2-33 (Contl Notes [Cant): SGS-UFSAR (6) The 3SM limit on PL + Pb + Q stress intensity range was exceeded. However, the provisions of Paragraph NB-3228.3 (Simplified elastic-plasLic analysis) of Reference 1 were satisfied. (7) Satisfied 2/3 the lower bound collapse load of NB-3228.2 of Reference 1. 2 of 2 . " Revision 18 April 26, 2000 TABLE 5.2-34 UNIT 1 SECONDARY SHELL AND TRANSITION CONE STRESS RESULTS MODEL F SG SECTIONS {Figure 5.2-23) A-A 8-B c-c D-D E-E CONDITION: Design 0. 94 12) 0.33(3) 0. 94121 o. 78 (3} 0.9412' 0. 72 (4) 0.76(S) 0. 64 (S} Normal & ( 6) Inside 0. 41 o. 47 0. 68 0. 78' 0.71 Upset* Outside 0.44 0.57 0.85 (7) 0.87 0.87 Fatigue Inside <0.01 <0.03 <0.01 <0.03 <0.01 <0.03 Outside <0.01 <0.01 <0.01 <0.02 <0.01 Emergency 0. 48 12) 0.17 13) 0.47(2) 0.40()) 0.48(Z) Faulted 0.47(2) 0.16{)) 0.47{2) 0.39(3) 0.47'2) 0.88(4) 0.77(S) 0.76(S) Test 0.63{2) 0.22(3) 0.63(2) 0.53131 0.63'21 , _. A.-::. upper Lateral Load Pad location. Not shown in Figure 5. 2-23. ( 4) I::: ; -'. P )/Allowable c !6) (PL + Pb + Ql /Allowable (7) The maximum primary + secondary stress intensity range exceeds the allow-able stress limit. Therefore, a simplified elastic-plastic analysis was per:ormed. This analysis is reflected in the cumulative usage factor calculations. 1 of 1 SGS-UFSA.R Revision 18 April 26, 2000 TABLE 5.2-35 UNIT 1 STRESS RESULTS OF TUBE ANALYSIS MODEL F SG LOCATION(!) A-A Q::.Q CONDITION: Design 0.60{2) 0.6212) o.sa13) 0.99713) Normal/Upset 0.96 0.9214} 0.85(4) 0.6714' Fatigue Usage 0.88 0.53 0.46 0.22 Emergency 0.67(2) 0.69'21 0.74(2) o.so12) Faulted (LOCA + SSE) 0.17121 0.17 (2) 0.99131 0.96131 Faulted (FLB + SSE) 0.47(2) 0.4a12' 0.51()) 0.55(3) .... Test 0 :.,..91. (2) 0. 94 12) 0.99121 0.68121 Nor.es: (1) See Figure 5.2-24 {2) P ./Allowable f'l ( 3) + "M P b l /Allowable ( 4) fl=> + p + Ql /Allowaple "L b 1 of 1 SGS-UFSAR E-E 0.60(2) 0.84 0.22 0.6712' 0.17 12' 0.4712) 0.91 12) Revision 18 April 26, 2000 Table 5.2-36 UNIT 1 TUBE ANALYSIS FOR EXTERNAL MODEL F SG Allowable 0 Actual Criteria y (ps{) CONDITION (ksi} D.P A-A Design 670 p 780 a Emergency 537(2) 1.2P 936 a Faulted 35.3 985(3) 0.9P 2602 c ' Test 38.9 1491 (4) O.BP 2549 \, c ... -.. '* Notes: ( 1 I
  • I See Figure 5.2-24 ; 2) Small LOCA \ ...}, :..a:::-ge LOCA .;.:; : Secondary Hydrotest 1 of 1 SGS-:JFSAE PRESSURE Pressure Differential, psi at these Sections{!) a.:..e c-c 780 780 936 936 2523 2424 2471 2374 D-D E-E 780 780 936 936 1531 2602 1500 2549 Revision 18 April 26, 2000
  • *
  • TABLE 5.2-37 SALEM 1 PREDICTED EOL RT NDT FOR REACTOR VESSEL BELTLINE MATERIALS Initial RTNDT (I) EOL RTNDT 1.41'2) EOL RTNDT %1'2) Material (OF) (OF) (OF) Intermediate Shell B2402-l 45 215 171(3} Intermediate Shell B2402-2 -5 151 113 Intermediate Shell B2402-3 -3 119 89 Lower Shell B2403-1 4 166 129 Lower Shell B2403-2 18 181 144 Lower Shell B2403-3 6 168 131 Intermediate to Lower Shell Circumferential Weld Seam -56 197 143 9-0:42 Intermediate Shell Longitudinal Weld Seams -56 205 144 2-042 A & B Intermediate Shell Longitudinal Weld Seam -56 172 118 2-042 c Lower Shell Longitudinal Weld Seams -56 206 144 3-042 A & B Lower Shell Longitudinal Weld Seam -56 232(J) 16S 3-042 c (1) Values from Tables 5.2-31 and 5.2-32 (2) EOL RT NDT is also tenned Adjusted Reference Temperature (ART). Values from reference 31. lf4T and 3.4T represent 25% and 75% vessel wall thickness. EOL is 32 EFPY. (3) Limiting material SGS-UFSAR 1 of 1 Revision 19 November 19,2001
  • *
  • TABLE 5.2 .. 38 SALEM 2 PREDICTED EOL RTNI>t FOR REACTOR VESSEL BELTLINE MATERIALS Material Initial RTNnT(l) EOL RTNI>1114t<2> EOL RTNl>'l 3/4t<2> (OF) (OF) (OF) Intennediate Shell 84712-1 0 125 100 Intennediate Shell 84712-2 12 145 117 Intennediate Shell 84712-3 10 119 98 Lower Shell 84713-1 8 126 103 Lower Shell 84713-2 8 126 102 Lower Shell 84713-3 10 128 104 Intennediate to Lower Shell Circumferential Weld Seam -56 102 76 9-442 Intennediate Shell Longitudinal Weld Seam -56 153 104 2-442A Intennediate Shell Longitudinal Weld Seams -56 181 128 2-442 B& C Lower Shell Longitudinal Weld Seams -56 199°) 14Q<3l 3*442 A & C Lower Shell Longitudinal Weld Seam -56 168 114 3-442 B (I) Values from Tables 5.2-31 and 5.2-32 (2) EOL RTNDT is also tenned Adjusted Reference Temperature (ART). Values from reference 32. Y.JT and %T represent 25% and 75% vessel wall thickness. EOL is 32 EFPY. (3) Limiting material SGS-UFSAR I of 1 Revision 19 November 19,2001
  • *
  • TABLE 5.2-39 SALEM 1 PREDICTED RTm FOR REACTOR VESSEL 8ELTLINE MATERIALS<1> Fluence(2> (1019nlcm2, ARTPTS Margin RT <3> EOL RTPTS<*> RT ns Screening Material NI>T(u) E>l.OMeV} (OF) (OF} eF} eF> Criteria (DF) Intennediate Shell I 1.64 175.1 17 45 237 270 lntennediate Shell B2402-2 1.64 160.1 17 -5 172 270 lntennediate Shell B2402-3 1.64 121.0 17 -3 135 270 Lower Shell 82403-l L64 146.8 34 4 185 270 Lower Shell 82403-2 1.64 148.1 34 18 200 270 Lower Shell 82403-3 1.64 146.8 34 6 187 270 lntennediate to Lower Shell 1.64 214.9 65.5 -56 224 300 Circumferential Weld Seam 9--042 lntennediate Shell Longitudinal Weld Seams 1.18 228.1 65.5 -56 238 270 2-042A, 8&C Lower Shell Longitudinal Weld 1.64 254.9 65.5 -56 264(S) 270 Seams 3-042 A, B & C from reference 29 unless otherwise noted. (2) Values from Table :5.4-7 for clad-metal interface and are predicted EOL values. All welds assumed to have maximum fluence for the weld seam group. (3) Values from Tables 5.2-31 and 5.2-32. (4) EOL is 32 EFPY. (5) Limiting materiaL SGS-UFSAR 1 of 1 Revision 19 November 19,2001
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  • TABLE 5.2-40 SALEM 2 PREDICTED RT PTS FOR REACTOR VESSEL BELTLINE MA TERIALS(1) Fluence<2J (I 019n/cm2, ARTprs Margin RT <3> EOL RTPTS<4> RT PTS Screening Material NDT(u) E>l.OMeV) eF> eF) (OF) (OF) Criteria COF) Intennediate Shell B4712-I L77 104.2 34 0 138 270 Intermediate Shell B4712-2 1.77 112.8 34 12 159 270 Intennediate Shell B4712-3 1.77 85.5 34 10 130 270 Lower Shell 84713-1 1.77 96.3 34 8 138 270 Lower Shell B4713-2 1.77 95.6 34 8 138 270 Lower Shell B4173-3 1.77 95.8 34 10 140 270 Intennediate to Lower Shell 1.77 106.0 Circumferential Weld Seam 9-442 65.5 -56 116 300 Intennediate Shell Longitudinal Weld Seams 1.20 198.5 65.5 -56 208 270 2-442A,B&C Lower Shell Longitudinal Weld 1.20 219.0 65.5 -56 229(S) 270 Seams 3-442 A, B & C (1) Values from reference 30 unless otherwise noted. (2) Values from Table 5.4-8 for clad-metal interface and are predicted EOL values. All welds assumed to have maximum tluence for the weld seam group. (3) Values from Tables 5.2-31 and 5.2-32. (4) EOL is 32 EFPY. (5) Limiting material. SGS-UFSAR 1 of 1 Revision 19 November 19,2001
  • *
  • TABLE 5.2*41 SALEM 1 USE PROJECTIONS FOR REACTOR VESSEL BELTLINE MATERIALS<'> Material Fluence at V..T <2> Unirradated USE O> Projected USE Projected EOL USE (nlcm2, E>l MeV) (ft-lb) Decrease(%) (ft-lb) Intennediate Shell 9.8 X 1018 91 19 74 B2402-l Intermediate Shell 9.8 X 1018 98 15 83 B2402-2 Intennediate Shell 9.8 X 1018 104 16 87 B2402-3 Lower Shell 9.8 X 1018 93 29 66 Lower Shell 9.8 X 1018 83 29 59 B2403-2 Lower Shell 9.8 x 1018 B2403-3 85 29 60 Intermediate to Lower Shell 9.8 X 101g 112 36 72 Circumferential Weld Seam 9-042 Intermediate Shell Longitudinal Weld 9.8 X 1013 96.2 32 65 Seams 2-042 A, B & C Lower Shell Longitudinal Weld 9.8 X 1018 112 32 76 Seams 3-042 B& C (l) Values from reference 31 unless otherwise noted. (2) Fluence calculated for EOL (32EFPY) from Table 5.4-7. All welds assumed to have peak fluence. (3) Values from Tables 5.2.;31 and *5.2-32. 1 of 1 SGS-UFSAR Revision 19 November 19,2001
  • *
  • TABLE 5.2-42 SALEM 2 EOL USE PROJECTIONS FOR REACTOR VESSEL BELTLINE MA TERIALS<1l Material Fluence at Y4T <,> Unimldated USE (ll Projected USE Projected EOL USE (nlcm1) (ft-lb) Decrease(%) (ft-lb) Intermediate Shell l.06x 1019 106 22 83 84712-1 Intermediate Shell 1.06 X 1019 97 14.5 83 B4712-2 Intermediate Shell 1.06 X 1019 B4712-3 107 20 86 Lower Shell 1.06 X 1019 84713-1 98 21 77 Lower Shell 1.06 X 1019 B4713-2 103 21 81 Lower Shell 1.06 X 1019 121 21 96 B4713-3 Intermediate to Lower Shell 1.06 X 101!1 Circumferential 99.7 35 65 Weld Seam 9-442 lntennediate Shell Longitudinal Weld 1.06 X 1019 96.2 37 61 Seams 2-442A, B &C Lower Shell Longitudinal Weld 1.06 X 1019 114 37 72 Seams 3-442 A, B& C (1) Values from reference 32 unless otherwise noted. (2) Fluence calculated for EOL (32EFPY) from Table 5.4-8. All welds assumed to have peak fluence. (3) Values from Tables 5.2-31 and 5.2-32. 1 of 1 SGS-UFSAR Revision 19 November 19,2001
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