ML21125A493
Text
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 System Design and Operating Parameters Unit 1 Unit 2 Original plant design life, years 1 40 40 Number of heat transfer loops 4 4 Design pressure, psig 2485 2485 Nominal operating pressure, psig 2235 2235 Approximate total RCS volume (including pressurizer and surge line) with 0% steam 12,540 ft.3 12,470 ft.3 generator tube plugging 2 Approximate system liquid volume (including pressurizer water) with 0% steam generator tube 11,640 ft.3 11,570 ft.3 plugging 2 Approximate system liquid volume (including pressurizer water) at maximum guaranteed power 11,990 ft.3 12,019 ft.3 with 0% steam generator tube plugging 3 12,283 x 106 12,283 x 106 Total Reactor heat output (100% power) Btu/hr (3600 MWt) (3600 MWt) 1 Licensed life is 60 years in accordance with Chapter 15 of the UFSAR.
2 This value is a best estimate based on ambient (70° F) conditions with 0% steam generator tube plugging. Refer to Westinghouse letter AEP-98-161 and IMP database SEC-SAI-4824-CO.
3 This includes a 3% volume increase (1.3% for thermal expansion and 1.7% for pipe connections to the reactor coolant loops, volume in the rod drive mechanisms and calculation inaccuracies).
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 System Design and Operating Parameters Unit 1 Unit 2 Bounding Bounding Conditions for Conditions for Rerating Rerating Lower/Upper 4 Lower/Upper 4 Reactor vessel coolant temperature at full power:
Inlet, nominal, ºF 511.7/549.3 511.7/549.3 Outlet, nominal, ºF 582.3/616.9 582.3/616.9 Coolant temperature rise in vessel at full power, 70.6/67.6 70.6/67.6 avg., ºF Total coolant flow rate, lb/hr x 106 139.5/133.2 139.5/133.2 Steam pressure at full power, psia 576/820 576/820 Steam Temp. @ full power, ºF 481.8/521.0 481.8/521.0 Approximate total RCS volume (including pressurizer and surge line) with 0% steam 12,540 ft.3 12,470 ft.3 generator tube plugging. 2 4
Limiting values based upon 3600 MWt rerating condition in WCAP-12135.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 Reac to r Co o l ant System Design Pressu re Settings Pressure (psig)
Unit 1 Unit 2 Design Pressure 2485 2485 Operating Pressure 2235 2235 Safety V alves 2485 2485 Power Relief V alves1 2335 2335 Pressurizer Spray V alves (Begin to Open) 2260 2260 Pressurizer Spray V alves (Full Open) 2310 2310 Pressurizer Pressure H igh - Reactor Trip 2385 2385 H igh Pressure Alarm 2310 2310 Pressurizer Pressure Low - Reactor Trip 1950 1950 Low Pressure Alarm 2210 2210 Pressurizer Pressure Low - Safety Inj ection 1815 1815 H ydrostatic Test Pressure 3106 31072 Back up H eaters On 2210 2210 Proportional H eaters (Begin to Operate) 2250 2250 Proportional H eaters (Full Operation) 2220 2220 1
During startup and shutdown, a manually energized safeguard circuit is in service while the reactor coolant system temperature is below 266°F for Unit 1 and 299°F for Unit 2. This allows automatic opening of that Unit' s two power relief valves at 435 psig for low temperature overpressure protection (LTOP) of the reactor vessel. This safeguard circuit ensures that the reactor pressure remains below the ASME Section III, Appendix G " Protection against Non-ductile Failure" limits in the case of an LTOP event.
2 Original design
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-3 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 Reac to r V essel Design Data Design Pressure, psig 2485 Operating Pressure, psig 2235 3107 Unit 2 and 31062 psig 1
H ydrostatic Test Pressure, psig Unit 1 Design Temperature, °F 650 Overall H eight of V essel and Closure H ead, ft-in. 43-911/16 (Unit 1)
(Bottom H ead O.D.
43-10 (Unit 2) to top of Control Rod Mechanism Adapter)
Thickne ss of Insulation, min., in 3 Number of Reactor Closure H ead Studs 54 Diameter of Reactor Closure H ead Studs, in 7 ID of Flange, in 1721/2 OD of Flange, in 205 ID at Shell, in 173 Inlet Nozzle ID, in 271/2 Outlet Nozzle ID, in 29 5
Clad Thickne ss, min., in /32 Lower H ead Thickn ess, min., in (base metal) 5 V essel Belt-Line Thickn ess, min., in (base metal) 81/2 Closure H ead Thick ness, in 61/2 Total Water V olume Below Core, ft3 1050 Water V olume in Active Core Region, ft3 665 1
Original design 2
Steam G enerator Replacement Pressure
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-3 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 Reac to r V essel Design Data Total Water V olume to Top of Core, ft3 2352 Total Water V olume 2959 to Coolant Piping Nozzles Centerline, ft3 Total Reactor V essel Water V olume, ft3 46603 (estimated) (with core and internals in place)
Unit 1 Bounding Conditions Unit 2 for Rerating Lower / Upper Reactor Coolant Inlet Temperature, °F 514.9 / 545.2 541.27 Reactor Coolant Outlet Temperature, °F 579.1 / 607.5 606.35 Reactor Coolant Flow, lb/hr x 106 139.0 / 133.9 134.6 3
This volume is a general number that approximates either Unit' s V essel. The actual volume of either vessel is cycle dependent and can be obtained from the Westinghouse IMP database for the specific cycle.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-4 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 Pressu riz er and Pressu riz er Rel ief Tank Design Data Pressu riz er Design Pressure, psig 2485 Operating Pressure, psig 2235 H ydrostatic Test Pressure (cold), psig 3106 Unit 1, 31071 Unit 2 Design/Operating Temperature, °F 680/653 Pressurizer Water Level, Full Power2 49.9% (Unit 1) / 55% (Unit 2)
Total Internal V olume3, ft3 1800 Surge Line Nozzle Diameter, in. 14 Shell ID, in. 84 Electric H eater Capacity, k W 4 1800 kW H eatup rate of Pressurizer, °F/hr 55 (approx.)
Start-up Water Solid, °F/hr 40 H ot Standby Condition, °F/hr 70 Design Spray Rate for V alves Full Open, gpm 800 Continuous Spray Rate, gpm 1 1
Original design 2
Estimated values from operating data, actual values determined by Tave full-load.
3 This volume is a general number that approximates either Unit' s Pressurizer. The actual volume of either pressurizer can be obtained from the current Westinghouse IMP database for the Unit.
4 Some heaters may be removed from service to a practical limit, greater than the TS minimum capacity, which supports plant evolutions.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-4 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 Pressu riz er and Pressu riz er Rel ief Tank Design Data Pressu riz er Rel ief Tank Design Pressure, psig 100 Rupture Disc Release Pressure, psig 100 Design Temperature, °F 340 Containment Ambient Normal Water Temperature, °F (120°F Max.)
Normal Operating Pressure, psig 3 Normal Water V olume, ft3 1430 Normal G as V olume, ft3 370 Cooling time required Approx. 1 following design maximum discharge, hr.
Number of spray nozzles 5 Total Spray Flow, gpm 150 Total V olume, ft3 1800 Total Rupture Disc Relief Capacity, 1.6 x 106 saturated steam, lb/hr
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 18 D. C. COOK NUCLEAR PLANT Table: 4.1-5 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 STEAM GENERATOR DESIGN DATA Unit 1 Unit 2 Number of Steam Generators 4 4 Design Pressure, Reactor Coolant/Steam, psig 2485/1085 2485/1085 Reactor Coolant Hydrostatic Test Pressure(tube side-cold), psig 3106 3107 Design temperature, Reactor Coolant/Steam, °F 650/600 650/600 33.91 Reactor Coolant Flow, lb/hr x 106 33.7 38.22 Total Heat Transfer Surface Area, ft2 54,927 54,500 Rated Thermal Output/MWt) 816 852.75 Operating Parameters at 100% Load Primary Side:
27733 Heat Transfer Rate (per unit), Btu/hr x 106 2910 30704 Coolant Inlet Temperature, °F 582.3 - 616.9 606.4 Coolant Outlet Temperature, °F 511.7 - 549.3 541.3 33.9 (1)
Flow Rate, (per unit), lb/hr x 106 33.7 3.82 (2)
Pressure loss, psi. 32.05 26.1 Secondary Side:
Steam Temperature at full power, °F 481.8 - 521 521.1 3.53 (3)
Steam Flow, lb/hr x 106 3.685 3.9 (4)
Steam Pressure at full power, psia 575.8 - 819.7 820 Maximum moisture carryover, wt % 0.045 0.15 431.3 @ #6 Feedwater Temperature 440 @ nozzle Heater Outlet Fouling Factor, hr-ft2 °F/Btu 0.00005 0.00005 Overall Height, ft-in 67 - 7.25 67-8 Shell OD, upper/lower, in 175.75/135 175.9/135 Number of U-tubes 3496 3592
- Quantities are for each steam generator.
1 RCS flow rate based on Thermal Design Flow of 88500 gpm at 536°F.
2 RCS flow rate based upon Mechanical Design Flow of 99700 gpm at 535°F.
3 Heat transfer rate and steam flow for 816 MWt per steam generator.
4 Heat transfer rate and steam flow for 900 MWt per steam generator.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 18 D. C. COOK NUCLEAR PLANT Table: 4.1-5 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 STEAM GENERATOR DESIGN DATA U-tube outer Diameter, in 0.875 0.875 Tube Wall Thickness, (minimum), in 0.044 0.050 2/18 Number of manways/ID, in 4/16 2/16 Number of handholes/ID, in 6/6 6/6 Number of inspection ports/ID, in 2/6 2/4 Unit 1 Rated Load No Load Reactor Coolant Water Volume, ft3 1141.15 1141.1 (5)
Primary Side Fluid Heat Content, Btu x 106 29.86 29.27 Secondary Side Water Volume, ft3 20358 3235 Secondary Side Steam Volume, ft3 3583 (8) 2315 Secondary Side Fluid Heat Content, Btu x 107 5.69 (8) 8.789 Unit 2 Rated Load No Load Reactor Coolant Water Volume,** ft3 1112 1112 Primary Side Fluid Heat Content, Btu 29.0 x 106 28.46 x 106 Secondary Side Water Volume, ft3 2077 3351 Secondary Side Steam Volume, ft3 3589 2315 Secondary Side Fluid Heat Content, Btu 5.18 x 107 8.44 x 107
- Quantities are for each steam generator.
Values may change subject to steam generator tube plugging.
5 Hot condition @ power. Volume at ambient temperature is approximately 1130 ft. 3.
6 Based upon hot volume and primary fluid @ 567.8°F and pressure of 2250 psia.
7 Based upon hot volume and primary fluid @547°F and pressure of 2250 psia.
8 Based upon secondary side fluid @ 515.2°F (saturated conditions).
9 Based upon secondary side fluid @ 547°F (saturated conditions).
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 4.1-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 REACTOR COOLANT PUMPS DESIGN DATA1 Number of Pumps 4 Design Pressure/Operating Pressure, psig 2485/2235 Hydrostatic Test Pressure (cold), psig 3106 Unit 1, 31072 Unit 2 Design Temperature (casing), oF 650 RPM at Nameplate Rating 1189 Suction Temperature, oF 536.3 (Unit 1)/541.003 (Unit 2)
Required net positive suction head, ft 170 Developed Head, ft 277 Capacity, gpm 88,500 Seal Water Injection, gpm 8 Seal Water Return, gpm 3 Pump Discharge Nozzle ID, in 27.5 Pump Suction Nozzle ID, in 31 Overall Unit Height, ft-in. 27-0 Water Volume, ft3 81 Pump-Motor Moment of Inertia, lb-ft2 82,000 MOTOR DATA:
AC Squirrel Cage Induction, Single Type Speed, Air Cooled Voltage 4000 Insulation Class F Phase 3 1
Quantities are for each pump.
2 Original design.
3 Original design power capability parameter.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 4.1-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 MOTOR DATA:
AC Squirrel Cage Induction, Single Type Speed, Air Cooled Frequency, Hz 60 Starting Current, amp 4800 Input (hot reactor coolant), kw 4337 Input (cold reactor coolant), kw 5663 Power, HP (nameplates ) 6000 Pump Weight, lb. (dry) 175,200
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-7 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 Reac to r Co o l ant Piping Design Parameters Reactor inlet piping, ID, in 27.5 Reactor inlet piping, minimum1 thick ness, in 2.56 Reactor outlet piping, ID, in 29 Reactor outlet piping, minimum1 thickne ss, in 2.69 Coolant pump suction piping, ID, in 31 Coolant pump suction piping, minimum1 thickne ss, in 2.88 Pressurizer surge line piping, ID, in 11.188 Pressurizer surge line piping, nominal thickne ss, in 1.406 Design pressure, psig 2485 Operating Pressure, psig 2235 H ydrostatic test pressure (cold), psig 3106 (Unit 1) / 31072 (Unit 2)
Design temperature, ºF 650 Design temperature (pressurizer surge line)ºF 680 3
Design pressure, pressurizer relief line, psig 1
Design temperature, pressurizer relief lines, ºF Water volume (all 4 loops without surge line), ft3 1185 Surge line volume, ft3 43 1
Original procurement minimums 2
Original design 3
From pressurizer to safety valve: 2485 psig, 650°F; From safety valve to pressurizer relief tank : 500 psig, 470°F
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 27.0 D. C. COOK NUCLEAR PLANT Table: 4.1-8 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 Pressu riz er V al v es Design Parameters Pressu riz er Spray Con trol V al ve s Number 2 Design pressure, psig 2485 Design temperature, ºF 650 Design flow for valves full open, each, gpm 400 Fluid temperature, ºF 530-550 Position after failure of actuating force Closed Pressu riz er Saf ety V al ve s Number 3 Relieving capacity, lb/hr 420,000 Set pressure, psig 2485 Fluid Saturated steam Constant backpr essure:
Normal, psig 3 Expected during discharge, psig 350 Pressu riz er Po w er Rel ief V al ve s Number 31 Design pressure, psig 2485 Design temperature ºF 680 Design capacity at nominal set pressure 2350 psia, (each) lbm/hr 210,000 Saturated steam Fluid or water 1
Only two required. Third valve is considered an installed spare.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 29.0 D. C. COOK NUCLEAR PLANT Table: 4.1-9 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 Reactor Coolant System Design Pressure Drop Pressure Drop, psi (estimated)
Unit 1(1) Unit 2(2)
Across Pump Discharge Leg 1.3/1.1 1.4 /1.2 Across Reactor Vessel, Including Nozzles 52.0/44.6 50.4 / 48.5 Across Hot Leg 1.2/1.0 1.2 / 1.1 Across Steam Generator, Including Nozzles 33.4/50.9 33.1 / 36.1 Across Pump Suction Leg 3.1/2.6 3.2 / 2.9 Total Pressure Drop 91.0/100.2 89.3 / 89.9 Note that the first value provided coincides with the maximum Best Estimate Flow (minimum steam generator tube plugging, minimum reactor vessel average temperature, TPR) and that the second value provided coincides with the minimum Best Estimate Flow (maximum steam generator tube plugging, maximum reactor vessel average temperature, TPI).
(1)
Data updated as a result of new Best Estimate Flows calculated in 2018.
(2)
Data updated as a result of new Best Estimate Flows calculated in 2017.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 4.1-10 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 3 Design Thermal and Loading Cycles Item Transient Cycles 1 Level A Limits (Normal) 1 Heatup at 100 °F/hr. 200 2 Cooldown at 100°F/hr. (Pressurizer @ 200 °F/hr.) 200 3 Unit Loading at 5% of full power/min. 18,300/11,680 2 3 4 Unit Unloading at 5% of full power/min. 18,300/11,680 2 3 5 Step Load Increase of 10% of full power 2,000 4 6 Step Load Decrease of 10% of full power 2,000 4 7 Large Step Decrease in load (with steam dump) 200 1
For Unit 1 Model 51R replacement steam generator manway and handhole stud preloads, the design considers 100 cycles each of tensioning and detensioning or torquing and detorquing, as appropriate.
2 Unit 1 rerating to 3600 MWt.
3 The Unit 1 Model 51R replacement steam generators have been structurally designed for the lower cycle limit for both 3264 MWt and the 3600 MWt power uprate condition. The RCS average temperature and steam temperature will deviate +/- 3 °F in one minute. The corresponding RCS pressure variation will be +/- 100 psi.
4 WCAP-17588-P, D. C. Cook Unit 1 Lower Radial Support Clevis Insert Acceptable Minimum Bolting Pattern Analysis, used 200 Step Load Increase of 10% of full power and 200 Step Load Decrease of 10% of full power transients to qualify the minimum bolting pattern. A new procedural limit was set to account for the lower number of transients allowed for the Unit 1 Clevis Insert Bolts. WCAP-17588-P does not impact any other analyses performed using the transients described in Table 4.1-10.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 4.1-10 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 3 Design Thermal and Loading Cycles Item Transient Cycles 1 8 Hot Standby Operation 18,300 5 9 Turbine Roll Test 10 10 Steady State Fluctuations Infinite 6 Level B Limits (Upset) 11 Loss of Load (without immediate turbine or Reactor trip) 80 Loss of Power (blackout with natural circulation in Reactor 12 40 Coolant System) 13 Loss of Flow (partial loss of flow one pump only) 80 14 Reactor Trip From Full Power 400 a) Operational Basis Earthquake (20 events of 20 cycles each 400 event), except Reactor Vessel 15 b) Operational Basis Earthquake, Reactor Vessel only (10 events 200 of 20 cycles each event)
Level C Limits (Emergency) 5 Applies to steam generator only. Reflects cyclic limit for the feed ring of a rapid injection of cold feedwater.
6 Reactor coolant system average temperature is assumed to increase and decrease a max. of 6°F in one minute. The corresponding reactor coolant pressure variation is less than 100 psi.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 4.1-10 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 3 Design Thermal and Loading Cycles Item Transient Cycles 1 None Level D Limits (Faulted) 16 Reactor Coolant Pipe Break (LOCA) 1 17 SSE 1 18 Steam Pipe Break 1 Test Conditions 19 Primary Side Hydrostatic Tests Before Initial Startup @ 3107 psig 57 20 Primary Side ASME Section XI/Field Tests 10 8 Secondary Side Hydrostatic Test Before Initial Startup at 1356 21 5 / 20 8 9 psig 22 Primary to Secondary Leak Test 50 / 90 8 23 Secondary to Primary Leak Test 120 8 7
Unit 1 Model 51R replacement steam generator shop hydro was 3106 psig.
8 Unit 1 Model 51R replacement steam generator.
9 Unit 1 Model 51R replacement steam generator not subjected to secondary side shop hydro. Leakage test performed after installation in accordance with Code Case N-416-1.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.1 D. C. COOK NUCLEAR PLANT Table: 4.1-11 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 3
SUMMARY
OF PLANT OUTAGE FOR YANKEE-ROWE (1964 - 1969)
Starting Date Duration Days/Hours Outage Type Case Equipment/System 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 Switch yard Electric 3/5/65 - Scheduled Switch yard 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
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.1 D. C. COOK NUCLEAR PLANT Table: 4.1-11 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 3
SUMMARY
OF PLANT OUTAGE FOR YANKEE-ROWE (1964 - 1969)
Starting Date Duration Days/Hours Outage Type Case Equipment/System 7/10/66 - 3.68 Forced Reactor Scram 8/25/66 - 2.40 Forced Reactor Scram 10/4/66 34 10.23 Scheduled Refueling 2/24/66 - 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 M.C. Pump Stator Cap Repair No. 4 Main Coolant Pump Thermal Barrier Leak and other 11/8/68 6 16.42 Scheduled 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
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.1 D. C. COOK NUCLEAR PLANT Table: 4.1-11 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 3
SUMMARY
OF PLANT OUTAGE FOR YANKEE-ROWE (1964 - 1969)
Starting Date Duration Days/Hours Outage Type Case Equipment/System 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
UFSAR Revision 30.0 INDIANA MICH IG AN POWER Revision: 24.0 D. C. COOK NUCLEAR PLANT Table: 4.1-12 UPDATED FINAL SAFETY ANALY SIS REPORT Page: 1 of 2 UNIT 1 AND UNIT 2 - REACTOR COOLANT SYSTEM CODES1 UNIT 1 Co mpo nent Co de Addenda and Co de Cases 1965 Edition through 1966 Winter Addenda, Reactor V essel ASME III 2 Class A Code Cases 1332-2, 1358, 1339-2, 1335, 1359-1, 1338-3, 1336 Reactor V essel Closure H ead (RV CH ) Class 1 (RV CH ) 1995 Edition through 1996 Addenda (RV CH )
1965 Edition through 1966 Winter Addenda Full Length Control Rod Drive Mechanisms ASME III 2 Class 1 1995 Edition through 1996 Addenda (RV CH )
Steam G enerators 1965 Edition through 1966 Winter Addenda ASME III 2 Class A (OSG Model 51 Steam Dome Shell) Code Cases 1401 and 1498 1989 Edition (No Addenda), Code Cases N-20-3, Steam G enerators (RSG Model 51R) ASME III 2 Class 1 N-71-15, N-411-1, N-474-1, 2142-1, 2143-1, N-401-1, and N-416-1 Reactor Coolant Pump Casings No Code (Designed with ASME III 2 Article 4 as a G uide) 1968 Edition 1965 Edition through Winter 1966 Addenda, Pressurizer ASME III 2 Class A Code Cases 1401, 1459 Pressurizer Safety V alves ASME III 2 1968 Edition Power Operated Relief V alves B-16.5 Main Reactor Coolant System Piping B31.1 1 1967 Edition Reactor Coolant System V alves B-16.5 or MSS-SP-66, and ASME III 1968 Edition 2 1
Repairs and replacement for pressure retaining components within the code boundary, and their supports, are conducted in accordance with ASME Section X I.
2 ASME Boiler and Pressure V essel Code, Section III-Nuclear V essels
UFSAR Revision 30.0 INDIANA MICH IG AN POWER Revision: 24.0 D. C. COOK NUCLEAR PLANT Table: 4.1-12 UPDATED FINAL SAFETY ANALY SIS REPORT Page: 2 of 2 UNIT 1 AND UNIT 2 - REACTOR COOLANT SYSTEM CODES UNIT 2 Co mpo nent Co de Addenda and Co de Cases 1968 Edition (1968 Summer Addenda),
ASME III 2 Class A Reactor V essel Code Cases 1335-4 and N-60-5 ASME III, Class 1 (RV CH )
1995 Edition through 1996 Addenda Full Length Control Rod Drive Mechanisms ASME 2 Class 1 1995 Edition through 1996 Addenda 1968 Edition through Winter 1968 Addenda, Code Cases 1401, 1498 for upper assemblies and Steam G enerators ASME III 2 Class A 1983 Edition through Summer 1984 for replacement lower assemblies Reactor Coolant Pump Casings No Code (Designed with ASME III 2 Article 4 as a G uide) 1968 Edition through Summer 1969 Addenda Pressurizer ASME III 2 Class A 1965 Edition through Winter 1966 Addenda Pressurizer Safety V alves ASME III 2 1968 Edition Power Operated Relief V alves B16.5 Main Reactor Coolant System Piping B31.1 1 1967 Edition Reactor Coolant System V alve B-16.5 or MSS-SP-66, and ASME III, 1968 Edition 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 20.1 D. C. COOK NUCLEAR PLANT Table: 4.1-13 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 COMPONENT TRANSIENT LIMITS1 Component Cyclic Or Transient Limit Design Cycle Or Transient Reactor Coolant 200 heatup cycles @ 100ºF/hr and Heatup cycle - Tavg from 200ºF to 547ºF.
System 200 cooldown cycles @ 100ºF/hr. Cooldown cycle - Tavg from 547ºF to 200ºF.
(pressurizer cooldown @ 200ºF/hr) 80 loss of load cycles Without immediate turbine or reactor trip 40 cycles of loss of offsite AC electrical power Loss of offsite AC electrical power source supplying the onsite Class 1E distribution system 80 cycles of loss of flow in 1 reactor coolant loop Loss of only 1 reactor coolant pump 400 reactor trip cycles 100% to 0% of RATED THERMAL POWER 200 large step decreases in load 100% to 5% of RATED THERMAL POWER with steam dump Operating basis earthquake 400 cycles - 20 earthquakes of 20 cycles each (except Reactor Vessel) 200 cycles - 10 earthquakes of 20 cycles each (Reactor Vessel only) 50 leak tests Pressurized to 2500 psia 5 hydrostatic pressure tests Pressurized to 3107 psig (3106 psig for Unit 1 Model 51R)
Secondary System 1 steam line break Break in a steam line 5.5" equivalent diameter 5 hydrostatic pressure tests Pressurized to 1356 psig 1
A log of the actual number of transients is maintained.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 4.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 MATERIALS OF CONSTRUCTION OF THE REACTOR COOLANT SYSTEM COMPONENTS Component Section Material Unit 1 Unit 2 Reactor Vessel Pressure Plate ASTM A-533 Grade B Class 1 ASTM A-533 Grade B Class 1 Pressure Forgings (excl. RVCH) ASTM A508, Class 2 ASTM A508, Class 2 RVCH SA-508 Grade 3, Class 1 SA-508 Grade 3, Class 1 Type 316 forging overlaid on I.D. and O.D. with Type 316 forging overlaid on I.D. and O.D.
Primary Nozzle Safe Ends Type 308L and Inconel weld metal after final with Type 316 weld metal prior to final post-post-weld heat treatment weld heat treatment Combination of Type 308, Type 309 and Type Cladding, Stainless Type 308L, Type, 309L 312 Stainless Weld Rod Type 308, Type 309 Type 308L, Type 309, Type 309L, Type 316 O-Ring Head Seals Inconel - 718 Inconel - 718 CRDM's Inconel and Stainless Type 304 Inconel and Stainless Type 304 Studs SA - 540 Grade B - 24 SA - 540 Grade B - 24 Instrumentation Nozzles Inconel and Stainless End Type 304 Inconel and Stainless End Type 304 Insulation Stainless Steel Stainless Steel ASTM A 533 Grade A Class 1 for upper Steam Generator Pressure Plate ASTM A 533 Grade A Class 1 assembly (steam dome), ASTM Pressure Forgings Tubesheets SA-508 Class 3a ASTM A - 508 Class 2 A Transition Cone & Stub Barrels SA-508 3a ASTM A - 508 Class 3 Stainless steel weld metal - carbon steel to Primary Nozzle Safe Ends SA-336 Class F316N/F316LN stainless steel juncture on O.D. overlaid with Type 309 and 308L weld metal Cladding, Stainless ER 308L, ER309L Type ER 309L Stainless Weld Rod Type 308L, Type 309 Type 308L, Type 309L Cladding for Tube Sheets UNS NO6082 Inconel Tubes SB-163 Alloy 690 TT Inconel - 690 (TT)
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 4.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 MATERIALS OF CONSTRUCTION OF THE REACTOR COOLANT SYSTEM COMPONENTS Component Section Material Unit 1 Unit 2 Channel Head Castings SA-508 Class 3A ASTM A - 216 Grade WCC (Unit 1 Forging)
Pressurizer Shell SA - 533 Grade A (Class 1) SA - 533 Grade A (Class 2)
Heads SA - 216 Grade WCC SA - 533 Grade A (Class 2)
Support Skirt SA - 516 Grade 70 SA - 516 Grade 70 Nozzle Weld Ends SA - 182 F316 SA - 182 F316 Inst. Tube Coupling SA - 182 F316 SA - 182 F316 Cladding, Stainless Type 308, Type 309 (modified) Type 308 Type 309 (modified)
Nozzle Forgings Integrally cast with head SA - 508 Class 2 Mn - Mo Heater Support Baffle Plate SA - 240 Type 304 SA - 240 Type 304 Inst. Tubing SA - 213 Type 316 SA - 213 Type 316 Heater Well Tubing SA - 213 Type 316 Seamless SA - 213 Type 316 Seamless Heater Well Adaptor SA - 182 F316 Pressurizer Relief Shell ASTM A- 285 Grade C ASTM A- 285 Grade C Tank Heads ASTM A-285 Grade C ASTM A -285 Grade C Internal Coating Amercoat 55 Amercoat 55 ASTM A-351 Grade CF8M ASTM A-351 Grade CF8M Pipe Pipes ASTM A - 376 Grade TP 304 or TP 316 ASTM A- 376 Grade TP 304 or TP 316 Fittings ASTM A-351 Grade CF8M ASTM A-351 Grade CF8M Nozzles ASTM A- 182 Grade F316 ASTM A- 182 Grade F316 Pump Shaft ASTM A-182 Grade F347 ASTM A- 182 Grade F347 Impeller ASTM A-351 Grade CF8M ASTM A- 351 Grade CF8M Casing ASTM A-351 Grade CF8M ASTM A- 351 Grade CF8M ASTM A-351 Grade CF8M and ASTM A-182 ASTM A- 351 Grade CF8M and ASTM Valves Pressure Containing Parts Grade F316 A- 182 Grade F316
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 4.2-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 REACTOR COOLANT WATER CHEMISTRY SPECIFICATION Electrical Conductivity Determined by the concentration of boric acid and alkali present. Expected range is 1 to 40 µMhos/cm at 25ºC.
Solution pH Determined by the concentration of boric acid and alkali present. Expected values range between 4.2 (high boric acid concentration) to 10.5 (low boric acid concentration) at 25ºC.
Oxygen, ppm, max. 0.10 Chloride, ppm, max 0.15 Fluoride, ppm, max. 0.15 Hydrogen, cc (STP)/kg H20 25 - 50 Total Suspended Solids, ppm, max. 1.0 pH Control Agent (Li7OH) Reactor coolant pH is controlled during power operation by adjusting lithium as a function of the coolant boron concentration.
Boric Acid as ppm B Variable from 0 to 4000
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 21 D. C. COOK NUCLEAR PLANT Table: 4.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1
SUMMARY
OF ESTIMATED PRIMARY PLUS SECONDARY STRESS INTENSITY FOR COMPONENTS OF THE REACTOR VESSEL (UNIT 1)1 Stress Intensity (psi)
Allowable Stress (psi)
Item Original Rerated (at Operating Temperature)
Control Rod Housing 55,300 66,050 69,900 Head Flange 50,400 54,380 80,100 Vessel Flange 45,350 65,850 80,100 Primary Nozzles Inlet 48,400 49,860 80,000 Primary Nozzles Outlet 54,060 59,580 80,000 Stud Bolts 95,870 83,320 104,400 Core Support Pad 40,800 69,700 69,900 Bottom Head to Shell 34,100 34,530 80,000 Bottom Instrumentation 53,400 51,490 69,900 Vessel Wall Transition 37,900 33,570 80,000 1
The vessel stress intensities for Unit 2 are available in the Unit 2 Stress Report.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 21 D. C. COOK NUCLEAR PLANT Table: 4.3-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1
SUMMARY
OF ESTIMATED CUMULATIVE FATIGUE USAGE FACTORS FOR COMPONENTS OF THE REACTOR VESSEL (UNIT 1)1 Usage Factor 2 and 3 Item Original Rerated Control Rod Housing .06 0.81 Head Flange .015 0.185 Vessel Flange .005 0.092 Stud Bolts .310 0.449 Primary Nozzles Inlet 0.020 .098 Primary Nozzles Outlet 0.028 .063 Core Support Pad (lateral) 0.015 .693 Bottom Head to Shell 0.003 .018 Bottom Instrumentation 0.142 .122 Vessel Wall Transition 0.002 .007 1
The usage factors for Unit 2 are available in the Unit 2 Stress Report.
2 Covers all transients.
3 As defined in Section III of the ASME Boiler and Pressure Vessel Code, Nuclear Vessels.
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 4.5-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 4 REACTOR COOLANT SYSTEM QUALITY CONTROL PROGRAM Component RT* UT** PT*** MT**** ET*****
- 1. Steam Generator 1.1 Tube Sheet 1.1.1 Forging yes yes 1.1.2 Cladding yes(+) yes(++)
1.2 Channel Head 1.2.1 Casting yes yes 1.2.2 Forging (Unit 1 Model 51R) yes yes 1.2.3 Cladding yes 1.3 Secondary Shell & Head 1.3.1 Plates yes 1.3.2 Forgings (Unit 1 Model 51R) yes yes 1.4 Tubes yes yes 1.5 Nozzles (forgings) yes yes 1.6 Weldments 1.6.1 Shell, longitudinal yes yes 1.6.2 Shell, circumferential yes yes 1.6.3 Cladding (Channel Head-Tube Sheet joint yes cladding restoration) 1.6.4 Steam and Feedwater Nozzle to shell yes yes 1.6.5 Support brackets yes 1.6.6 Tube to tube sheet yes 1.6.7 Instrument connections (primary and yes secondary)
Radiographic Ultrasonic Dye Penetrant Magnetic Particle Eddy Current
(+)
Flat Surfaces Only
(++)
Weld Deposit Areas Only
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 4.5-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 4 REACTOR COOLANT SYSTEM QUALITY CONTROL PROGRAM Component RT* UT** PT*** MT**** ET*****
1.6.8 Temporary attachments after yes removal 1.6.9 After hydrostatic test (after welds and complete channel head - yes where accessible) 1.6.10 Nozzle safe ends (if forgings) yes yes 1.6.11 Nozzle safe ends (if weld deposit) yes
- 2. Pressurizer 2.1 Heads 2.1.1 Casting yes yes 2.1.2 Cladding yes 2.2 Shell 2.2.1 Plates yes yes 2.2.2 Cladding yes 2.3 Heaters 2.3.1 Tubing(+++) yes yes 2.3.2 Centering of element yes 2.4 Nozzle yes yes 2.5 Weldments 2.5.1 Shell, longitudinal yes yes 2.5.2 Shell, circumferential yes yes 2.5.3 Cladding yes 2.5.4 Nozzle Safe End (if forging) yes yes 2.5.5 Nozzle Safe End (if weld deposit) yes 2.5.6 Instrument Connections yes 2.5.7 Support Skirt yes 2.5.8 Temporary Attachments after removal yes
(+++)
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 4.5-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 4 REACTOR COOLANT SYSTEM QUALITY CONTROL PROGRAM Component RT* UT** PT*** MT**** ET*****
2.5.9 All welds and cast heads after hydrostatic test yes 2.6 Final Assembly 2.6.1 All accessible surfaces after hydrostatic test yes
- 3. Piping repairs and replacements are conducted in accordance with ASME Section XI 3.1 Fittings and Pipe (Castings) yes yes 3.2 Fittings and Pipe (Forgings) yes yes 3.3 Weldments 3.3.1 Circumferential yes yes 3.3.2 Nozzle to runpipe (except no RT for nozzles yes yes less than 4 inches) 3.3.3 Instrument connections yes
- 4. Pumps 4.1 Castings yes yes 4.2 Forgings 4.2.1 Main Shaft yes yes 4.2.2 Main Studs yes yes 4.2.3 Flywheel (Rolled Plate) yes 4.3 Weldments 4.3.1 Circumferential yes yes 4.3.2 Instrument connections yes
- 5. Reactor Vessel 5.1 Forgeries 5.1.1 Flanges yes yes 5.1.2 Studs yes yes 5.1.3 Head Adapters yes yes 5.1.4 Head Adapter Tube yes yes
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 4.5-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 4 of 4 REACTOR COOLANT SYSTEM QUALITY CONTROL PROGRAM Component RT* UT** PT*** MT**** ET*****
5.1.5 Instrumentation Tube yes yes 5.1.6 Main Nozzles yes yes 5.1.7 Nozzle safe ends (if forging is employed) yes yes 5.2 Plates yes yes 5.3 Weldments 5.3.1 Main Steam yes yes 5.3.2 CRD Head Adapter Connection yes 5.3.3 Instrumentation tube connection yes 5.3.4 Main nozzles yes yes Yes 5.3.5 Cladding (++++) yes 5.3.6 Nozzle-safe ends (if forging) yes yes 5.3.7 Nozzle safe ends (if weld deposit) yes yes 5.3.8 Head adaptor forging to head adaptor tube yes yes 5.3.9 All welds after hydrotest yes
- 6. Valves 6.1 Castings yes yes 6.2 Forgings (No NDE for valves two inches and yes yes smaller)
(++++)
UT of Clad Bond-to-Base Metal