ML19294C023
| ML19294C023 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/31/1979 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML19294C021 | List: |
| References | |
| NUDOCS 8003060549 | |
| Download: ML19294C023 (100) | |
Text
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WISCONSIN ELECTRIC 2rf i,.f
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ANNUAL RESULTS AND I))
POWER COMPANY
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POINT BE ACH NUCLEAR PL ANT
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U.S. Nuclear Regulatory Commission Docket Nos. 50-266 and 50-301 Facility Operating License Nos.
80 0 3 0 00 5+ 7
PREFACE This Annual Results and Data Report for 1979 is submitted in accordance with Point Beach Nuclear Plant Unit Nos. 1 and 2, Technical Specification 15.6.9.1.B (Amendment Nos.
31 and 35 of 12-23-77, respectively) and filed under Docket Nos. 50-266 and 50-301 for Facility operation License Nos. DPR-24 and DPR-27, respectively.
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TABLE OF CONTENTS Page
1.0 INTRODUCTION
1 2.0 HIGHLIGHTS 2.1 Unit 1 1
2.2 Unit 2 1
3.0 FACILITY CHANGES, TESTS AND EXPERIMENTS 3.1 Amendments to Facility Operating Licenses 2
3.2 Facility or Procedure Changes Requiring NRC Approval 3
3.3 Tests or Experiments Requiring NRC Approval 3
3.4 Design Changes 4
3.4.1 Unit 1 4
3.4.2 Unit 2 11 3.4.3 Common 17 3.5 Procedure Changes 24 4.0 NUMBER OF PERSONNEL AND MAN-REM EXPOSURE BY WORK AND JOB FUNCTION 4.1 1978 (Corrected) 68 4.2 1979 69 5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION 5.1 Unit 1 70 5.2 Unit 2 80 a
a
1.0 INTRODUCTION
The Point Beach Nuclear Plant, Units 1 and 2, utilize identical pressurized water reactors rated at 1518 MWt each.
Each turbine generator is capable of producing 497 MWe net (524 MWe gross) of electrical power. The plant is located ten miles north of Two Rivers, Wisconsin, on the west shore of Lake Michigan.
2.0 HIGHLIGHTS 2.1 Unit 1 For the period 01-01-79 through 12-31-79, which included a 58-day refueling outage, four steam generator maintenance outages, a
moisture separator reheater repair outage, and an outage to modify a branch connection to the piping of each main feedwater line, Unit 1 operated at an average capacity factor of 70.5% and a net efficiency of 32.6%.
The unit and reactor availability were 76.2% and 77.3%,
respectively.
Unit 1 generated its 28 billionth kilowatt hour (gross) on 02-17-79, its 29 t' w*
Rilowatt hour on OE 17-79, and its 30 billionth kilowatt hour c.
79.
2.2 Unit 2 For the period 01-01-79 through 12-31-79, which included a 21 day 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 26 minute new Point Beach Nuclear Plant record refueling outage, a turbine overspeed trip test outage, an outage to modify safeguards logic and conduct a primary system natural circu.1 tion test, a feedwater nozzle volumetric examination outage, and a brief outage following a reactor trip due to loss of power to the red instrument bus, Unit 2 operated at an average capacity factor of 85.5% and a net efficiency of 32.6%.
The unit and reactor availability were 88.5%
and 89.1%, respectively.
Unit 2 geners ted its 23 billienth kilowatt hour (gross) on 02-12-79, its 24 billionth kilowatt hour on 05-29-79, its 25 billionth kilowatt hour on 09-08-79, and its 2J billirsin kilowatt hour on 12-01-79.
A 3.0 FACILITY CHANGES, TESTS AND EXPERIMENTS 3.1 Amendments to Facility Operating Licenses During the year of 1979, there were seven license amendments issued by the U. S. Nuclear Regulatory Commission to both Facility Operating License DPR-24 for Point Beach Unit 1 and Facility Operating License DPR-27 for Point Beach Unit 2.
These license amendments are listed by date of issue and summarized as follows.
3.1.1 01-30-79, Amendment 40 to DPR-27
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This amendment revised the Technical Specifications for the reactor coolant system pressure-temperature heatup and cooldown curves for Unit 2.
3.1.2 04-04-79, Amendment 35 to DPR-24, Amendment 41 to DPR-27 These amendments authc...:9 the installation and use of modified spent fuel storage racks which increase the allowable capacity for spent fuel storage from 351 assemblies to 1,502 assemblies.
3.1.3 04-06-79, Amendment 36 to DPR-24 This amendment extended the Unit 1 reactor coolant system pressure-temperature heatup and cooldown curves from seven to eleven EFPY.
3.1.4 04-13-79, Amendment 37 to DPR-24, Amendment 42 to DPR-27 These amendments incorporated the Point Beach Nuclear Plant Modified Amended Security Plan into the plant license.
3.1.5 05-11-79, Amendment 38 to DPR-24, Amendment 43 to DFR-27 These amendments issued a Technical Specification change which required the actuation of safety injection to be based on two-out-of-three channels of low pressurizer pressure and eliminated the coincident low pressurizer level signal from the logic. The amendments also required modification to the power supplies for the safety injection actuation channels.
3.1.6 08-02-79, Amendment 39 to DPR-24, Amendment 44 to DPR-27 These amendments added conditions to the licenses related to the completion of the fire protection modifications.
3.1.7 10-30-79, Amendment 40 to DPR-24, Amendment 45 to DPR-27 A new license condition was added by these amendments which requires a secondary water chemistry monitoring program to inhibit steam generator tube degradation.
3.1.8 11-13-79, Amendment 41 to DPR-24, Amendment 46 to DPR-27 These amendments increased the minimum fire brigade size from fcur to five members. The amendments were effective 90 days after issue.
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3.2 Facility or Procedure Changes Requiring Nuclear Regulatory Commission Approval There were no plant modifications or procedure changes during 1979, beyond those authorized with license amendments as noted previously, which required Nuclear Regulatory Commission approval.
3.3 Tests or Experiments Requiring Nuclear Regulatory Commission Approval There were no tests or experiments at Point Beach Nuclear Plant in 1979 which required Nuclear Regulatory Commission approval.
A 3.4 Design Changes 3.4.1 Unit 1 a.
Instrument Bus MG Sets (E-160)
The YO3 and Y04 instrument bus 7.5 KVA voltage regulator and 15 KVA isolation transformers were replaced with flywheel motor generators.
Summary of Safety Evaluation:
The instrument bus MG sets isolate the buses from line voltage spikes. Failure of the MG sets will move protection systems in the
~
conservative direction as trip circuits are activated.
b.
Security Related Modification (E-173)
No com.aent.
c.
Chemical and Volume Control System MOV-427 Control Switch at 1B42 (E-176)
A three-position spring return to center (off) control switch for letdown isolation was installed which parallels the open/close scheme of the control room switch.
Summary of Safety Evaluation:
The modification permits an operator to reestablish normal letdown in the event valve 427 closes on a pressurizer low level signal and primary system pressure and level are being controlled at the local control station.
d.
Turbine Lube Oil (E-181)
A redur..lant pressure switch to start the DC motor-driven lube oil pump was installed.
Summary of Safety Evaluation:
Not nuclear safety related.
e.
Reactor Coolant Pump Warning Lights (E-184)
Flashing lights were installed in each reactor coolant pump compartment to warn personnel that the pumps are operating without the flywheel covers installed.
Summary of Safety Evaluation:
Not nuclear safety related.
f.
Emergency Lighting (E-191)
Emergency lighting was installed at the charging pump local control station.
Summary of Safety Evaluation:
Not nuclear safety related.
g.
Auxiliary Feed Pumps (E-198)
The circuitry for the steam-driven auxiliary feedwater pump bearing service water valve was modified to permit operation of the pump and blowdown without having to bypass the solenoid valve or having to run the auxiliary feedwater pump without bearing service water.
Summary of Safety Evaluation:
The modification improves the integrity of service water cooling to the steam-driven auxiliary feedwater pump bearings.
h.
Addition of Safety Injection A0V (M-129)
The modification installed an air-operated valve in the refueling water storage tank suction to the safety injection pumps between valves 894 and 895.
The valve is automatically controlled such that when either 826B or C leave the shut condition, the new A0V uill shut.
Summary of Safety Evaluation: The modification prevents dilution of the boric acid storage tank while still guaranteeing the safety injection pumps stay wet at all times.
The modification does not affect operation of main safety injection pump suction valves 896A and B.
- i. Manipulator Gripper Air Cylinder (M.787)
The modification provides air actuation in both modes of gripper operation.
S_ummary of Safety Evaluation:
Not nuclear safety related.
j.
Reactor Coolant Vent and Drain (M-414)
A permanent three quarter inch reactor coolant vent header and drain piping were installed to improve venting of the system over the previous tygon tubing arrangement.
Summary of Safety Evaluation:
Not nuclear safety related.
k.
Gland Steam Drain Trap Relocation (M-439)
The drain trap was relocated from El. 20' to El. 5' to create a water deadleg of approximately 20' on the trap outlet.
Summary of Safety Evaluation:
Not nuclear safety related.
1.
Low Pressure Turbine (M-464)
The LPl and LP2 L-4 blade wheel was reshrouded per Westinghouse recommendation to harmonically regroup the blades.
Summary of Safety Evaluation:
Not nuclear safety related.
m.
Safety Injection System (M-475)
Temperature indicators were installed in the refueling water storage tank to record suction temperatures of the safety injection spray pumps during performance of ASME Section XI required testing.
Summary of Safety Evaluation:
Not nuclear safety related.
n.
Steam Generator Primary Manway and Pressurizer Manway Bolts (M-477)
The existing hex head bolts were changed out to a stud and nut arrangement to reduce the potential for galling the vessel drilled and tapped holes.
Summary of Safety Evaluation:
Design analyses showed that studs are an acceptable replacement for bolts. The equipment vendor also approved the change.
o.
Low Pressure Trap Header Pressure Protection (M-491)
The modification request upgraded the existing system by replacing the in-line valve with a higher pressure rated valve.
Summary of Safety Evaluation:
Not nuclear safety related.
p.
Containment cooler Throttle Valve (M-509)
SW-144, a carbon steel valve, was replaced with a correctly sized stainless steel valve.
Summary of Safety Evaluation:
Replacement of the throttled valve, which caused internal erosion, with more erosion resistant stainless steel valves improves system integrity.
q.
Crossover Steam Dump System (M-529)
Dump valve position indication was removed to eliminate through leakage and operational problems for A0V's 1 and 4.
Secondary indication is available to verify remote closure of the valves.
Summary of Safety Evaluation:
Not nuclear safety related.
Condensate Cooler Relief Protection (M-532J r.
The modification installed overpressure protection for the circulating water side of the coolers.
Summary of Safety Evaluation:
Not nuclear safety related.
s.
Extraction Piping (M-547)
The carbon steel extraction piping to No. 4 feedwater heater was replaced with stainless steel piping and high pressure drain traps were installed to reduce piping erosion caused by wet steam.
Summary of Safety Evaluation:
Not nuclear safety related.
t.
Charginc Pump Cubicles (M-551)
The cubicle doorways were widened to improve personnel and equipment access as well as improving ventilation.
Summary of Safety Evaluation:
Not nuclear safety related.
FilAandBDischargeStack(M-5521 u.
Multiport sample probes were installed with local power provided for a sample pump.
Summary of Safety Evaluation:
The r.ew samplers provide for more representative iodine and particulate sampling of discharges via the containment ventilation system.
v.
Snubber HS-14 (M-554)
The snubber support on the pressurizer relief valve discharge line was stiffened.
Summary of Safety Evaluation:
Additional stiffening minimizes deflection of the support under load will not compromise the snubber's ability to perform its intended function.
w.
Containment Sump "B" Sample Line (M-564)
A sample line off of ISI-D17 was installed to enable sampling which exempts the piping from a significant amount of Section XI ultrasonic testing.
Summary of Safety Evaluation: The modification provides residual heat removal sump B sampling and provides double boundary protection.
x.
Circulating Water Pumps (M-565)
Oil coolers and vent holes were installed in the circulating water pump motor upper oil reservoir.
Summary of Safety Evaluation:
Not nuclear safety related.
y.
Feedwater Heater Cooldown Connection (M-572)
Auxiliary connections to the steam generator blowdown system were added to circulate water through the feed-water heaters to more quickly cool down the units.
Summary of Safety Evaluation:
Not nuclear safety related.
z.
Security Related Modification (M-614)
No comment.
aa. Pressurizer Access (M-626)
The present access opening in the pressurizer compart-ment was enlarged and access platforms installed to enhance personnel safety in gaining access to the compartment.
Summary of Safety Evaluation:
Engineering evaluations concluded that the larger access hole will not deleteriously weaken the strength of the structure or increase the risk to the containment liner due to missiles generated in the pressurizer space.
bb. Extraction Steam Level Control (IC-129)
The existing level controller on the heating steam moisture separator was replaced by a
pneumatic controller of better quality.
Summary of Safety Evaluation:
Not nuclear safety related.
cc. Main Cont.rol Board Indication (IC-134)
Status lights for overpower and overtemperature AT runback were installed on panel C04 for each of the four channels.
Summary of Safety Evaluation:
Not nuclear safety related.
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dd. Power-Operated Relief Valve Redundant Pressure Channel (IC-147)
A redundant pressure signal channel was added to the controls for the power-operated relief valves when they are used to prevent overpressurization during low pressure and/or solid water conditions.
In addition, a backup gas / air supply was added for the power-operated relief valve air operators.
An interlock through the motor-operated isolation valves was provided for the key-switch indicators.
Summary of Safety Evaluation:
The modifications add redundancy to existing control channels.
ee. Feedwater Heater Drains (IC-164)
The air supply for the 5A and B heater drain and dump valve controllers was reconnected to the output of the turbine relay dump valve to cause the valves to open upon a turbine trip.
Summari of Safety Evaluation:
Not nuclear safety related.
ff. Pressurizer Spray Valves (IC-166)
The current-to-pressure transmitters, regulators, and gauges for 1PCV-431A and B were moved outside the shield wall to provide better access for maintenance and calibration as well as removing the transmitters from heat inside the cubicle.
Summary of Safety Evaluation:
Not nuclear safety related.
gg. Volume Control Tank Vent Path (IC-179)
A vent path with two-valve isolation was installed in the nitrogen and hydrogen supply to the volume contro) tank.
_9_
Summary of Safety Evaluation:
The modification allows venting the regulated side of the nitrogen and hydrogen regulators supplying pressure to the volume control tank.
Venting is required to adjust the regulator setpoint under certain conditions.
Double-valve isolation with a
normally capped vent prevents inadvertent depressurization of either gas supply.
Small amounts of gas may be released in the auxiliary building;
- however, this gas is monitored by the auxiliary building vent stack radiation monitoring system.
hh. Pressurizer Low Pressure 2/3 Actuation (IC-192)
The c'odification replaced low pressurizer level actuation coincident with low pressurizer pressure, with a 2/3 actuation on low pressurizer pressure only.
Summary of Safety Evaluation:
Following restrictions by the NRC requiring removal of the level signai in the 1/3 level coincident with pressure safety injection logic, the pressure signal change to a logic of 2/3 is considered conservative in that it assures the required safety injection signal, but reduces the probability of inadvertent and unnecessary plant transients.
ii. Safeguards Instrumentation Power Supplies (IC-194)
The modification changed the power supplied to four safeguards pressure channels from plant AC to inverter power from the opposite unit.
Summary of Safety Evaluation:
Changing the power supplies from AC sources to DC sources reduces the possibility of safety injection occurring on both units at one time.
jj. Feedwater Control System Power Supply (IC-197)
The modification permits feedwater control to be operated in the red and blue channels so the potential problem of inverter loss will not affect both loops in the same unit. Prior to the modification, operators had to shift to manual feedwater control in both units to minimize the possibility of a trip upon loss of feedwater.
Summary of Safety Evaluation: The modification provides for more reliable plant operation.
kk. Safety Injection Initiation and Manual Reset (IC-199)
The modification installed a
separate reset and initiation switch for each train of safety injection.
Summary of Safety Evaluation:
The safeguards system single failure pushbutton problem is eliminated and direct manual initiation for all portions of the safeguards system is provided for.
- 11. Pressurizer Pressure 2/3 Safety Injection Annunciation (IC-202 First-out annunciation in the control room and on the computer alarm typewriters was provided for modification request IC-192.
Sumnary of Safety Evaluation:
The modification is necessary to provide proper annunciation and recording of a
safety injection caused by pressurizer low pressure.
mm. Pressurizer Safety Valve Acoustic Monitoring (C-207)
An acoustic monitor was added to each of the pressurizer safety valves to detect the passage of fluid through the valves when either a valve is opened or leakage exists.
A five-channel critical systems leak monitoring system providing separate alarm indication for each channel.
Summary of Safety Evaluation:
The modification provides direct valve position indication at all times as required by NUREG-0578.
nn. Reactor Coolant Pressure Control (IC-208)
An alarm was added to indicate then any power-operated relief valves or code safety relief valve on the pressurizer is nat shut per Three Mile Island Lessons Learned requirements.
Summary of Safety Evaluation:
Not nuclear safety related.
3.4.2 Unit 2 a.
High Pressure Turbine Lighting (E-130)
Light fixtures were installed in the doghouse to improve inspection and surveillance.
Summary of Safety Evaluation:
Not nuclear safety related.
b.
Equipment Hatch Power Disconnect (E-151)
Electrical receptacles and plugs were installed in the equipment hatch area to make hatch removal more efficient.
Summary of Safety Evaluation:
Not nuclear safety related.
c.
Instrument Bus MG Sets (E-161)
The Y03 and YO4 instrument bus 7.5 KVA voltage regulator and 15 KVA isolation transformers were replaced with flywheel motor generators.
Summary of Safety Evaluation:
The MG sets minimize instrument bus voltage spikes from line voltage perturbation.
Failure of the MG sets will move protection systems in the conservative direction as trip circuits are activated.
d.
Extension Building Power Supply (E-712)
A power supply from 2B02 breaker 43D was provided to the extension building.
Summary of Safety Evaluation:
Not nuclear s32^ty related.
e.
Chemical and Volume Control System MOV-427 Control Switch at 2B42 E-177)
A three position spring return to center (off) control switch for letdown isolation was installed which parallels the open/close scheme of the control room switch.
Summary of Safety Evaluation:
The modification permits an operator to reestablish normal letdown in the event valve 427 closes on a pressurizer low level signal and primary system pressure and level are being controlled at the local control station.
f.
Turbine Lube Oil (E-182)
A redundant pressure switch to start the DC motor-driven lube oil pump was installed.
Summary of Safety Evaluation:
Not nuclear safety related.
g.
Reactor Coolant Pump Warning Lights (E-184)
Flashing lights were installed in each reactor coolant pump compartment to warn personnel that the pumps are operating without the flywheel covers installed.
Summary of Safety Evaluation:
Not nuclear safety related.
h.
Emergency Lighting (E-192)
Emergency lighting was installed at the auxiliary feed pump local control station.
Summary of Safety Evaluation:
Not nuclear safety related.
- i. Heater Drain Tank Pump Seal Regulator Bypass (M-322)
A bypass throttle valve and downstream isolation valve were installed around PCV-2596A and B seal supply to the heater drain pumps. The change permits isolation of the pressure control valves while the unit is operating.
Summary of Safety Evaluation:
Not nuclear safety related.
j.
Circulating Water System (M-359)
A metal walkway was installed in each condenser waterbox to facilitate safe personnel access to perform condenser tube leak evaluations.
Summary of Safety Evaluation:
Not nuclear safety related.
k.
Charging System (M-370)
The two-inch valve used for isolation of bladder-type pulsation dampeners was removed because the dampeners are no longer installed.
Summary of Safety Evaluation:
Not nuclear safety related.
1.
Manipulator Gripper Air Cylinder (M-388)
This modification provides air actuation in both moder of gripper operation.
Summary of Safety Evaluation:
Not nuclear safety related.
m.
Charging Pumps (M-392)
Pressure pulsation dampeners were installed on the discharge of each charging pump.
Summary of Safety Evaluation:
The dampeners minimize vibration and pipe connection fatigue failure.
n.
Reactor Coolant Vent and Drain (M-415)
A permanent three-quarter inch reactor coolant vent header and drain piping was installed to improve venting of the system over the previous tygon tubing arrangement.
Summary of Safety Evaluation:
Not nuclear safety related.
o.
Gland Steam Drain Trap Relocation (M-432)
The drain trap was relocated from El. 20' to El. 5' to create a water deadleg of approximately 20' on the trap outlet.
Summary of Safety Evaluation:
Not nuclear safety
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related.
p.
Steam Generator Primary Manway and Pressurizer Manway Bolts M-478)
The existing hex head bolts were changed out to a stud and nut arrangement to reduce the potential for galling the vessel drilled and tapped holes.
Summary of Safety Evaluation:
Design analyses showed that studs are an acceptable replacement for bolts. The equipment vendor also approved the change.
q.
Low Pressure Trap Header Pressure Protection (M-492)
The modification request upgraded the existing system by replacing the in-line valve with a higher pressure rated valve.
Summary of Safety Evaluation:
Not nuclear safety related.
r.
Crossover Steam Dump System (M-529)
Dump valve position indication was removed to eliminate through leakage and operational problems for A0V's 1, 3, and 4.
Secondary indication is available to verify remote closure of the valves.
Summary of Safety Evaluation:
Not nuclear safety related.
s.
Condensate Cooler Relief Protection (M-533)
The modification installed overpressure protection for the circulating water side of the coolers.
Summary of Safety Evaluation:
Not nuclear safety related.
t.
Extraction Piping (M-548)
The carbon steel extraction piping to No. 4 feedwater heater was replaced with stainless steel piping and high pressure drain traps were installed to reduce piping erosion caused by wet steam.
Summary of Safety Evaluation:
Not nuclear safety related.
u.
F11A and B Discharge Stack (M-553)
Multiport sample probes were installed with local power provided for a sample pump.
Summary of Safety Evaluation:
The new samplers provide for more representative iodine and particulate sampling of discharges via the containment ventilation system.
~
v.
Circulating Water Pumps (M-5661 Oil coolers and vent holes were installed in the circulating water pump motor upper oil reservoir.
Summary of Safety Evaluation:
Not nuclear safety related.
w.
Feedvater Heater Cooldown Connection (M-573)
Auxiliary connections to the steam generator blowdown system were added to circulate water through the feed-water heaters to more quickly cool down the units.
Summary of Safety Evaluation:
Not nuclear safety related.
x.
Extraction Steam Level Control (IC-129)
The existing level controller on the heating steam moisture separator was replaced by a
pneumatic controller of better quality.
Summary of Safety Evaluation:
Not nuclear safety related.
y.
Power-Operated Relief Valve Redundant Pressure Channel (IC-148)
A redundant pressure signal channel was added to the controls for the power-operated relief valves when they are used to prevent overpressurization during low pressure and/or solid water conditions.
In addition, a backup gas / air supply was added for the power-operated relief valve air operators.
An interlock through the motor-operated isolation valves was provided for the key-switch indicators.
Summary of Safety Evaluation:
Not nuclear safety related.
-1S-
z.
Hydrogen Seal Oil System (IC-153)
A duplex strainer differential pressure alarm was installed in the air side seal oil system.
Summary of Safety Evaluation:
Not nuclear safety related.
aa. Feedwater Heater Drains (IC-165)
The air supply for the SA and B heater drain and dump valve controllers was reconnected to the output of the turbine relay dump valve to ecuse the valves to open upon a turbine trip.
Summary of Safety Evaluation:
Not nuclear safety related.
bb. Pressurizer Spray Valves (IC-167)
The current-to-pressure transmitters, regulators, and gauges for IPCV-431A and B were moved outside the shield wall to provide better access for maintenance and calibration as well as removing the transmitters from heat 1.iside the cubicle.
Summary of Safety Evaluation:
Not nuclear safety related.
cc. Pressurizer Low Pressure 2/3 Actuation (IC-193)
The modification replaced low pressurizer level actuation coincident with low pressurizer pressure, with a 2/3 actuation on low presaurizer pressure only.
Summary of Safety Evaluation:
Following restrictions by the NRC requiring removal of the level signal in the 1/3 level coincident with pressure safety injection logic, the pressure signal change to a logic of 2/3 is considered conservative in that it assures the required safety injection signal, but reduces the probability of inadvertent and unnecessary plant transients.
dd. Safeguards Instrumentation Power Supplies (IC-194)
The modification changed the power supplied to four safeguards pressure channels from plant AC to inverter power from the opposite unit.
Summary of Safety Evaluation:
Changing the power supplies from AC sources to DC sources reduces the possibility of safety injection occurring on both units at one time.
ee. Eeedwater Control System Power Supply (IC-198)
The modification permits feedwater control to be operated in the red and blue channels so the potential problem of inverter loss will not affect both loops in the same unit.
Prior to the modification, operators had to shift to manual feedwater control in both units to minimize the possibility of a trip upon loss of feedwater.
Summary of Safety Evaluation: The modification provides for more reliable plant operation.
ff. Safety Injection Initiation and Manual Reset (IC-200)
The modification installed a
separate reset and initiaton switch for each train of safety injection.
Sum. mary of Safety Evaluation:
The safeguards system single failure pushbutton problem is eliminated and direct manual initiation for all portions of the safeguards system is provided for.
gg. Pressurizer Safety Valve Acoustic Monitoring (IC-207)
An acoustic monitor was added to each of the pressurizer safety valves to detect the passage of fluid through the valves when either a valve is opened or leakage exists.
A five-channel critical systems leak monitoring system providing separate alarm indication for each channel.
Summary of Safety Evaluation:
The modification provides direct valve position indication at all times as required by NUREG-0578.
3.4.3 Common a.
Gai-tronics (E-180)
A speaker was installed in the service water pump enclosure located in the pumphouse.
Summary of Safety Evaluation:
Not nuclear safety related.
b.
Battery Room Ventilatlan (E-189)
A flow switch was instalied with alarming in the control room for each of the battery room exhaust fans.
Summary of Safety Evaluatior:
Not nuclear safety related.
c.
Gas Turbine (E-194)
A cutoff switch was installed to silence the continuous alarm received during routine fire protection Cardox system testing.
Summary of Safety Evaluation:
Not nuclear safety related.
d.
Maintenance Office Addition (M-275)
The modification removed several internal se rvice building walls to permit more efficient conduct of Maintenance Group supervisory activities by creating one large office in lieu of several small offices.
Summary of Safety Evaluation:
Not nuclear safety related.
e.
Water Treatment Crossconnection (M-418)
The modification installed an isolation valve and associated piping to utilize HX-95B as well as HX-95A as a hot water supply tank for anion regeneration.
Summary of Safety Evaluation:
Not nuclear safety related.
f.
Water Treatment System Acid Flow (M-486)
A throttling valve was installed for manually controlling acid flow to improve system reliability.
Summary of Safety Evaluation:
Not nuclear safety related.
g.
Security Related Modification (M-501)
No comment.
h.
Waste Disposal (M-505)
The modification installed a waste evaporator feed filter drain to improve changeout of the filter.
Summary of Sarety Evaluation:
Not nuclear safety related.
i.
Chemical and Volume Control System Gas Analyzer (M-513)
A water trap in the sample lines to the gas analyzer was installed to prevent entry of water into the analyzer.
Summary of Safety Evaluation:
Not nuclear safety related.
j.
Emergency Diesel Generator (M-515)
The exhaust manifolds on 3D and 4D were modified to provide an exhaust screen inspection port.
Summary of Safety Evaluation:
Not nuclear safety related.
k.
Water Treatment Acid Piping (M-520)
The existing sulfuric acid system piping was replaced with an all-welded sulfuric acid resistant stainless steel alloy to improve reliability and personnel safety.
Summary of Safety Evaluation:
Not nuclear safety related.
1.
Sewage Treatment Effluent Sump (M-524)
A flushing line was installed to cleaning the sump via the fire water system.
Summary of Safety Evaluation:
Not nuclear safety related.
m.
Cryogenic Noble Gas Sample Connection (M-525)
A connection was added to the nitrogen purge addition tubing to purge the compressors after maintenance and to sample the compressor outlet or charcoal decay tank during operation.
Summary of Safety Evaluation:
Not nuclear safety related.
n.
Fire Water System (M-526)
The modification installed six new one-inch hose reels to meet upgraded fire protection requirements.
Summary of Safety Evaluation:
Not nuclear safety related.
o.
Controlled Side Hot Shower Room (M-530)
The shower room was modified to provide additional space and more efficient utilization of existing space.
Summary of Safety Evaluation:
Not nuclear safety related.
p.
Fire System Main Loop Heater (M-535)
Post-indicating valves were installed to provide individual fire hydrant isolation.
Summary of Safety Evaluation:
Not nuclear safety related.
q.
Waste Solidification (M-539)
A viewing window with lighting near the control board was installed to enhance operator observation.
Summary of Safety Evaluation:
Not nuclear safety related.
r.
Fire Protection System (M-544)
A pumper truck connection was added to enable the Two Creeks Volunteer Fire Department to obtain a suction from the Unit 2 seal well.
Summary of Safety Evaluation:
Not nuclear safety related.
s.
Truck Access Condensate Return (M-555)
An outside vent and drain condenser was installed to enable steam to condense on the cold sidewalls and return to the receiving tank.
Summary of Safety Evaluation:
Not nuclear safety related.
t.
Spent Fuel Pit Leak Detection (M-571)
Sample connections were added to the one-inch leakoffs to collect and measure spent fuel pit leakage.
Summary of Safety Evaluation:
Not nuclear safety related.
u.
Fire Protection Oxygen Manifold (M-574)
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An oxygen manifold was installed for charging Bio-Pak self-contained breathing apparatus.
Summary of Safety Evaluation:
Not nuclear safety related.
v.
Removal of TCV-LW-9 Service Water Valve (M-580)
The distillate cooler temperature control valve was removed and replaced with a spoolpiece because it was not required.
Summary of Safety Evaluation:
Not nuclear safety related.
w.
Gatehouse Hot Water Heater (M-588)
The heater drain line was modified to drain at floor level and a curb was installed between the heating ventilating and air conditioning equipment base and the north side of the wall between the north and south rooms.
Summary of Safety Evaluation:
Not nuclear safety related.
x.
Fire Protection (M-589)
A seal was installed on the control room to the cable spreading room door to prevent smoke and fumes in the control in the e. vent of a fire in the cable spreading room.
Summary of Safety Evaluation:
Not nuclear safety related.
y.
Heating Ventilating and Air Conditioning Fire Dampers (M-613)
The new dampers were installed at existing penetrations in (1) the lube oil storage reservoir room; (2) the vital switchgear room: (3) the auxiliary feed pump room; (4) each of the battery rooms to upgrade plant fire protection.
Summary of Safety Evaluation:
Not nuclear safety related.
z.
Sewage Treatment (M-617)
A roughing screen and cleanout flange was installed at the inlet to the aeration basin to prevent clogging of the comminutor.
Summary of Safety Evaluation:
Not nuclear safety related.
aa. Sewage Treatment (M-618)
A channel was installed on the bottom of the weir compartment to reduce flow under the weir plate.
Summary of Safety Evaluation:
Not nuclear safety related.
bb. Sewage Treatment (M-619)
The return sludge line was rerouted to the north end of the aeration basin.
Summary of Safety Evaluation:
Not nuclear safety related.
ec. Radwaste Steam (M-638)
A high point vent was installed on the steam supply line near the component cooling heat exchangers in order to hydrostatically test the lines after maintenance has been performed on the system.
Summary of Safety Evaluation:
Not nuclear safety related.
dd. Circulating Water Temperature (IC-149)
Four additional temperature detectors were installed near the intake crib and level transmitters were installed on the two circulating water surge chambers.
Readout for these. instruments as well as for the ambient temperature detectors and circulating water inlet and discharge temperatures was provided to the control roon.
Summary of Safety Evaluation:
Not nuclear safetv related.
ee. Water Treatment (IC-178)
Ultrasonic level detectors were installed on the acid and caustic tanks to provide more reliable level indication.
Summary of Safety Evaluation:
Not nuclear safety related.
ff. Circulating Water Level (IC-185)
The forebay level indication system was modified to install a three-pen recorder in place of the existing two-pen recorder.
Summary of Safety Evaluation:
Not nuclear safety related.
gg. Auxiliary Feedwater (IC-201)
Flow meters for 1 and 2P29 and P38A and B were installed on the main control boards.
Summary of Safety Evaluation: The modification provides operators with auxiliary feedwater flow information which in conjunction with steam generator level, will indicate proper operation of the auxiliary feedwater system during accident conditions.
O b
4 3.5 Procedure Chang?s All of the following procedures were reviewed per the requirements of 10 CFR 50.59 and evaluated for their safety implications. They were all approved by the Manager's Supervisory Staff.
3.5.1 OP-1A, Cold Shutdown to Low Power Operation, Revision 18, 1_2-26-78 (performed 03-11-79, Unit 1 Steps 4.19 and 4.20, securing RHR and drawing a bubble in the pressurizer, respectively, were performed prior to Step 4.14, in anticipation for the RCS leak test.
The primary-to-secondary pressure differential was held to a minimum.
To accomplish this, the RCS was heated up to approximately 520 F prior to performing a leak test minimizing RCS pressure.
Step 4.21, returning the safety injection system to normal lineup, was done after Step 4.1.A as primary system pressure was increased.
Step 4.18 was performed after Step 4.24 such that the leak test was performed in a hot condition.
Step 4.26, forming a hydrogen blanket in the volume control tank, was deleted as the blanket was already established. Step 4.18.1 was deleted as the steam header pressure was above the setpoint for safety injection.
Step 4.18.3 was deleted because pressure was above 1000 psig during heatup and the manways were not leaking.
Step 4.18.4 was deleted because it was not necessary to secure either a reactor coolant pump or a pressurizer heater group in the hot condition. Step 4.18.5 was changed raise pressure with the pressurizer heaters rather than the charging pumps since a bubble in the pressurizer had already been established.
Step 4.18.6 was deleted as heatup had been completed and depressurization was not necessary. A new Step 4.18.6 was inserted to reduce pressure to normal operating pressure.
Step 4.18.7 was deleted as the safety injection low-low steam pressure circuit had not been defeated since Step 4.18.1 had been deleted.
In Step 4.18.9 a note was added that if it was necessary to depressurize a cooldown should be performed.
(Temporary) 3.5.2 OP-1A, Cold Shutdown to Low Power Operation, Revision 18, 12-26-78 (performed 07-19-79, Unit 2)
Steps 3.1.1, 3.1.2, 3.1.3, 3.1.4, 3.1.5, 3.1.6, 3.1.7, 4.16, 4.17, and all steps in 4.18 were deleted to reflect operation from a short cold shutdown, non-drained condition to a low power condition.
A note was added to Step 4.12 to ensure that the Furmaniting of valve 700 is complete before securing RHR.
In Step 4.13, completion of inservice test IT-125 was added.
In Step 4.14 a note was added to isolate the nitrogen bottle supply to A0V-439 and 431C for low pressure overpressurization protection. New Step 4.21.8 was added to conduct a secondary system hydrostatic test at 1356 psig to qualify feed line work if required; however performance of the step was not necessary.
(Temporary) 3.5.3 OP-1A, Cold Shutdown to Low Power Operation, Revision 19, 07-26-79 Typographical errors were corrected in Steps 2.13, 2.13.1, 4.18.1, 4.18.2, and 4.24.1.
Step 2.24 was deleted as it was consolidated into Step 2.20..
Step 2.23.3 was clarified to note that containment integrity must be established when the reactor vessel head is removed the reactor is in a refueling shutdown condition.
In Step 3.1.9, the checklist number for main steam, CL-13A, was added. In Step 4.3.1,
" remainder" was substituted for " conclusion" for clarification.
In Step 4.14 isolation of the nitrogen bottle supply to A0V-430 and 431C was added to provide low pressure overpressurization protection.
In Step 4.18.8 the specific M0V's, 515 and 516, were indicated as required to be open versus in the previous issue a general statement not identifying specific valve numbers had ueen used.
A qualifying note was added to Step 4.22 to set the NIS recorder to recora the count rate of the highest source range channel and an intermediate range channel unless directed by Reactor Engineering to record both source range channels.
(Permanent) 3.5.4 OP-1A, Cold shutdown to Low Power Operation, Revision 19, 07-26-79 (performed 08-18-79, Unit 1)
In Step 3.1 certain checkoff lists were not required to be completed because the systems had not been manipulated.
Step 4.20 (drawing a pressurizer bubble) was advanced to before Step 4.12 to facilitate a timely recovery from the cold shutdown.
In Step 4.15 the reference to the nitrogen bottle supply was deleted because the system was not yet installed.
Step 4.18 was revised to perform the primary leak test with a pressurizer bubble rather than in a solid conditic.i.
(Temporary) 3.5.5 OP-1A, Cold Shutdown to Low Power Operation, Revision 19, 07-26-79 (performed 08-18-79, Unit 1)
In Step 3.1 certain checkoff lists were not required to be completed because the systems had not been manipulated.
Step 4.20 (drawing a pressurizer bubble) was advanced to before Step 4.12 to facilitate a timely recovery from the cold shutdown.
(Temporary) 3.5.6 OP-1A, Cold Shutdown to Low Power Operation, Revision 19, 07-26-79 (performed 09-01-79, Unit 1 Step 3.1.10, performance of required portions of CL-1B, Containment Integrity Check List, was added.
Step 4.17 was deferred until af ter Step 4.26 because the leak test was performed hot; thus Step 4.18 was deleted.
(Temporary) 3.5.7 OP-1A, Cold Shutdown to Low Power Operation, Revision 20, 08-22-79 Step 3.1.9 was clarified to note the specific pages of CL-13A (main steam checkoff list) which must be performed.
A new Step 3.1.10 was incorporated to reference completion of required portions of CL-1B, Containment Integrity Checkoff List.
In Step 4.24.4 a typographical error was corrected.
(Permanent) 3.5.8 OP-1A, Cold Shutdown to Low Power Operation, Revision 20, 08-22-79 (performed 11-19-79, Unit 1)
Steps 2.1 and 4.24.1 were modified to permit startup and physics testing with a positive moderator temperature coefficient because this condition was anticipated.
Step 2.15 was changed to reflect reduced primary system operating temperature.
In Step 4.27.2 a sentence was added to bypass the cryogenic delay tanks until RCS nitrogen levels were acceptable per Chemistry.
In Step 4.30 the new value for no-load Tavg was inserted.
Step 4.32 was added so an inservice leak inspection of the steam generator first-off vent valves (MS-211 and MS-212) could be conducted.
(Temporary) 3.5.9 OP-1B, Reactor Startup, Revision 8, 07-22-77 (performed 03-12-79, Unit 1)
In Step 4.11 a sustained startup rate of 0.5 dpm was changed to 0.1 dpm as the 0.1 dpm rate was only achievable with the rods available.
One dilution was necessary because the ERP was unreliable.
Startup was performed using an ICRR as required.
(Temporary) 3.5.10 OP-1B, Reactor Startup, Revision 8, 07-22-77 (performed 03-23-79, Unit 2 In Step 4.13, a.5 dpm startup instruction was given by the Training Supervisor during operator training startups; this number being agreed to by the Duty shift Supervisor.
(Temporary) 3.5.11 OP-1C, Low Power Operation to Normal Power Operation, Revision 15, 04-19-78 (performed 08-05-79, Unit 1)
The procedure was terminated at Step 4.50 because the "A" steam generator tube leak required shutdown before normal power levels were achieved.
(Temporary) 3.5.12 OP-1C, Low Power Operation to Normal Power Operation, Revision 15, 04-19-79 (performed 12-01-79, Unit 1)
Step 3.3 was corrected to indicate the current value for no-load Tavg (522 F).
(Temporary) 3.5.13 OP-3A, Normal Power Operation to Low Power Operation, Revision 5, 10-21-74 (performed 08-05-79, Unit 1)
Steps 4.1, 4.2 and 4.3 were deleted because the unit was at low power; Unit 1 being in the process in increasing load when the "A" steam generator tube leak forced the shutdown.
(Temporary) 3. 5. lt.
OP-3A, Normal Power Operation to Low Power Operation, Revision 5, 10-21-74 (performed 10-05-79, Unit 1 Step 4.2 which performs a heat balance on the primary system was deleted because it was not required per WMTP1 6.2.
(Temporary) 3.5.15 OP-3C, Hot Shutdown to Cold shutdown, Revision 20, 09-29-78 performed 03-10-79, Unit 1 Steps 2.4 and 3.2 calling for reactor coolant pump and/or the RHR system in operation for boron changes were not performed.
Due to operational conditions with the spray valves
- open, RCP operation was suspended to prevent depressurization during the attempted repair of the pressurizer spray valves. At the same time, boration became necessary to remain within shutdown margin limits.
Steps 4.4.1 and 4.6 calling for cold shutdown required inservice tests were deleted.
The shutdown was considered to be of a short expected duration (i.e.,
less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />), which permitted the waiving of all cold shutdown inservice tests.
In Step 4.11 the handwheels on valves 841A&B were not red tagged because of containment entry requirements due to the prevailing atmosphere (no pert:nnel were working in the area at the time).
In Step 4.16, the temperature at which the RCP s were secured, was changed to 195 F from 160 F as the work required on the pressurizer spray valves only required a reduction below 200 F.
(Temporary) 3.5.16 OP-3C, Hot Shutdown to Cold Shutdown, Revision 20, 09-29-78 (performed 03-14-79, Unit 1 Step 4.4.1 was changed in sequence to Step 4.15.1 (performance of main steam nonreturn valve inservice test) as system conditions permitted.
(Temporary) 3.5.17 OP-3C, Hot Shutdown to Cold Shutdown, Revision 21, 03-19-79 Step 4.4.1 was changed in sequence to Step 4.15.1 (performance of main steam nonreturn valve inservice test) as system conditions permitted.
(Permanent) 3.5.18 OP-3C, Hot Shutdown to Cold Shutdown, Revision 21, 03-19-79 (performed 06-30-79, Unit 2)
Steps 2.9, 4.15, 4.17, 4.18, and 4.19 were not performed due to special conditions involving the feed line inspection.
Step 4.21 was delayed until startup.
(Temporary) 3.5.19 OP-3C, Hot Shutdown tc Cold shutdown, Revision 22, 07-26-79 Precaution 2.1.1 was deleted because it is a redundant statement of Precaution 2.1.2.
Step 2.4 was clarified to permit boration or degassing without an RCP or RHR pump running but requires at least one RCP or RHR pump in oper-ation during dilution. Step 2.11 was changed to observe the effects of boric acid addition in terms of excore detector response, not only source range counts.
Step 2.22 was added
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to log all releases on CHP-70, Contaminated Atmospheric Steam Dump" when a shutdown is performed with known primary-to-secor <iary leakage.
In Step 4.11 clarification wording was addei; instruction numbers were corrected; and a note was added o open the isolation valves on the backup nitrogen supply to the low pressure overpressurization protection system A note was added to Step 4.14 to assure MOV-515 and 516 are open and A0V-430 and 431C are in auto to assure the low pressure overpressurization protection system is operable.
Step 4.15.1 was modified to indicate the appropriate inservice test to be conducted for each unit.
(Permanent) 3.5.20 OP-3C, Hot Shutdown to Cold Shutdown, Revision 22, 07-26 _7_9 (performed 08-04-79, Unit 1)
The inservice tests requirements of the procedure were deleted because they were not required in view of the shutdown being less than a 48-hour duration and Tavg being 2200 F.
(Temporary) 3.5.21 OP-3C, Hot Shutdown to cold Shutdown, Revision 22, 07-26-79 (performed 08-29-79, Unit 1)
Step 4.6 was deleted because IT-140 had been performed during an earlier cold shutdown within the three-month period.
Step 4.10 was deleted for IT-130 and IT-200; Step 4.15.1 for IT-220; Step 4.20 for IT-110; and Step 4.31 for IT-120, all for the same reason.
In Step 4.11 the reference to the nitrogen bottles was deleted because the modification is not yet complete.
(Temporary) 3.5.22 OP-3C, Hot Shutdown to Cold Shutdown, Revision 22, 07-26-79
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[ performed 12-12-79, Unit 1 Steps 4.6, 4.10, 4.15.1, 4.20, and 4.21 were deleted because the subject inservice tests had already been performed within the required time intervals.
In Step 4.1.1, the letdown gas strippers were secured at greater than 10 cc/kg hydrogen because it was not possible to reduce concentrations to less than 10 cc/kg.
In Step 4.18 the valve number for SI-894 was reassigned as SI-825C.
(Temporary) 3.5.23 OP-3C, Hot Shutdown to Cold shutdown, Revision _22, 07-26-7C (performed 10-05-79, Unit 1)
Step 3.4 was added to accomplish in-loop inspections and checking of the containment cleanup fans prior to cooldown of the RCS.
Step 4.1.2 was added to perform IT-210, LCV-427 test with pressurizer bubble and normal letdown lineup as required per modification request.
Step 4.13 was modified to point out that OP-7A as temporarily changed on 10/06/79
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should be utilized.
(Temporary)
OP-4A, Filling and Venting Reactor Coolant System, 3.5.24 Revision 12, 09-21-78 (performed 05-06-79, Unit 2)
In Step 3.1 the page number was specified to reference the required equipment and valve checkoff sheets which must be completed as an initial condition.
Steps 4.3 and 4.15 were deleted and Steps 4.12 and 4.14 were modified to reflect the upgrading to the hard pipe reactor coolant system vent system.
In Step 4.27,
" disconnect" was substituted for
" remove" for clarification.
(Temporary) 3.5.25 OP-4A, Filling and Venting the Reactor Coolant System, Revision 12, 09-21-78 (performed 08-17-79, Unit 1)
Although Revision 13 the current revision of the procedure, Revision 12 was performed because the new revision incorporates installation of the RCS hard pipe vent system.
This system had not yet been installed on Unit 1.
(Temporary) 3.5.26 OP-4A, Filling and Venting Reactor Coolant System, Revision 13, 07-26-79 In Step 3.1 the page number was specified to reference the required equipment and valve checkoff sheets which must be completed ar an initial condition.
Steps 4.3 and 4.15 were deleted and steps 4.12 and 4.14 were modified to reflect the upgrading to the hard pipe RCS vent system.
In Step 4.27,
" disconnect" va. substituted for " remove" for clarificatio..
(Permanent) 3.5.27 OP-4A, Filling and Venting Reactor Coolant System, Revision 14, 08-27-79 Figure 1 was revised to reflect installation of a hard pipe vent system.
Various typographical < rrors were corrected throughout the procedure.
Step 4.28.as added which allows venting of the reactor coolant drain tmnk to the pressurizer relief tank prior to establishing prcter pressurizer relief tank atmosphere.
(Permanent) 3.5.28 OP-4B, Reactor Coolant Pump Operation, Revision 14, 07/17/79 Precaution 2.17 was added to reflect installation of the RCP flywheel cover warning lights (industrial safety improvement modification).
(Permanent) 3.5.29 OP-4C, Pressurizer Relief Tank Operation, Revision 3, 08-27-79 Steps 41, 4.3, and 4.4 were expanded and clarified for operator information.
Step 4.5 was added to set the nitrogen pressure regulator appropriately to accomplish completion of the operation expeditiously.
(Permanent) 3.5.30 CP-4D, Draining the Reactor Coolant System, Revision 12, 09-20-78 (performed 03-15-79, Unit 1)
The sequence for burping and purging the steam generators was changed such that "A"
steam generator was done first rather than "B"
as is called out in the procedure.
A nitrogen purge was placed on the "A"
steam generator for approximately eight hours as as a precautionary measure.
(Temporary) 3.5.31 OP-4D, Draining the Reactor Coolant System, Revision 13, 08-27-79 The procedure was revised to accommodate system venting in accordance with the new hard pipe vent system.
Steps 3.12, 6.1, 6.2, and 6.5 were clarified to coordinate purging of the steam generators with the Maintenance Group.
In Steps 5.27 and 5.28 actual instrument numbers were inserted for clarification.
(Permanent) 3.5.32 OP-4D, Draining the Reactor Coolant System, Revision 14, 07-17-79 (performed 08-30-79, Unit 1)
Revision 12, dated 09/20/78 was utilized per permission granted by the Staff in MSSM 79-37 because the reactor coolant hard pipe vent system is not yet operational.
In addition Step 3.11 on containment purging was deferred until later during procedure performance.
Section 5.0 was not performed because it was not applicable; the unit not in refueling shutdown.
(Temporary) 3.5.33 OP-5D, Chemical Addition and Control, Revision 6,05-12-3 (performed 03-23-79, Unit 2 In Section C,
a new Step 4.5 was added to degas the pressurizer before taking the RCS solid. Subsequent steps were renumbered.
(Temporary) 3.5.34 OP-5D, Chemical Addition and Control, Revision 7, 04-05-79 In Section C,
a new Step 4.5 was added to degas the pressurizer before taking the RCS solid.
Subsequent steps were renumbered.
(Permanent) 3.5.35 OP-7A, Placing Residual Heat Removal System in Operation, Revision 13, 05-12-79
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New Steps 2.1 and 4.1 were added to prevent opening of SI-851A&B, sump "B" valves, during RHR operation to prevent depressurization of the reactor coolant system and flooding of the containment 8' level. (Permanent) 3.5.36 OP-7A, Placing the Residual Heat Removal System in Operation, Revision 13, 05-12-79 (performed 10-05-79)
The procedure will be conducted during the upcoming U1R7 shutdown.
The procedure was changed to meet the requirements of the ASME Code to qualify the operation as an inservice operational pressure test:
Step 4.9 was modified to meet the four-hour requirement at pressure and full inspection of all pressurized piping. Step 4.11 was expanded to open M0V-852A & B (core deluge valves) to expand the pressurized area. Step 4.15 was expanded to run both pumps to maximize test pressure and the note following this step was expanded to meet the four-hour pressure requirement and then shut MOV-852A&B to return them to their normal lineup.
(Temporary) 3.5.37 OP-7B, Residual Heat Removal System Removing from Operation, R_evision 8, 05-12-79 New Step 4.1 was added to prevent opening of SI-851A&B, sump "B" valves, during RHR operation to prevent depressurization of the reactor coolant system and flooding of the containment 8' level.
(Permanent) 3.5.38 OP-7B, Residual Heat Removal System Removing from Operation, Revision 8, 05-12-79 performed 11-19-79, Unit 1)
In Step 4.8.1 a typographical error was corrected to indicate the correct valve, RH-713A.
(Temporary) 3.5.39 OP-7B, Residual Heat Removal System Removing from Operation, Revision 9, 11-26-79 In Step 4.8.1 a typographical error was corrected to indicate the correct valve, RH-713A.
(Permanent) 3.5.40 OP-9A, Liquid Waste Process System Operation, Revision 7, 08-27-79 The procedure was revised to reflect experience gained and actual practice by waiving a 15 minute, 1000 gallon recirc of "B"
waste distillate tank for sampling if a sample is obtained within two hours of transferring tan <
"A" to "B" via the polishing demineralizers.
Steps 4.4.2 and 4.3 were clarified for operator information.
(Pe rmanen'.)
3.5.41 OP-9C, Containment Venting and Purging, Revision 8, 02-20-79 In Section 1.0 an incorrect reference was corrected.
In Step 4.1.2, references to certain computer procedures were corrected.
In attached Figure #2, the note at the bottom of the page was clarified.
(Permanent) 3.5.42 OP-9C, Containment Venting and Purging, Revision 9, 04-25-79 New Precaution and Limitation Step 2.4 was added to prevent opening of the purge supply and exhaust valves except during a cold shutdown condition.
This precaution was added as a result of the manufacturer's notice concerning the possibility that the valves will not shut under accident conditions.
(Permanent) 3.5.43 OP-13A, Secondary Systems Startup and Shutdown, Revision 15, 01-04-79 (performed 03-23-79 [ shutdown section], Unit 2 In Step 5.6 the main steam stops were shut with no steam flow because of a rapid loss of vacuum.
(Temporary) 3.5.44 OP-13A, Secondary Systems Startup and Shutdown, Revision 16, 05-21-79 The valve checklist was revised to prevent backflow from the operating unit's R15, air ejector discharge, to the unit being lined up for startup.
Step 4.10.2 was revised allow starting of the main air ejectors with the air ejector delay duct charcoal filter bypass valve, AR-26 open.
In Step 2.5 the word "stop" was corrected to " trip" (Permanent) 3.5.45 OP-13A, Secondary Systems Startup and shutdown, Revision 16, 05-21-79 (startup performed 08-04-79, Unit 1)
In Step 4.5.4 a valve number typographical error was corrected.
(Temporary) 3.5.46 OP-13A, Secondary Systems Startup and Shutdown, Revision 17, 08-08-79 In Step 4.5.4 a valve number typographical error was corrected.
(Permanent) 3.5.47 OP-13A, Secondary Systems shutdown and Startup, Revision 17, 08-08-79 performed 09-29-29, Unit 2 Step 5.6 was modified because problems closing MS-2017 required packing adjustments.
Steps 5.12.1 through 5.12.4 were not performed because the steam generator feed pumps were not required to be shut down.
Steps 5 13 and 5.18 were deleted because the equipment was not secured during this shutdown.
(Temporary) 3.5.48 E0P-3A, Steam Generator Tube Rupture, Revision 9, 03-23-78
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Step 4.4 currently specifies a no-load Tavg of 547"F.
Because Tavg will be lower because of Unit 1 operating at reduced primary system te;nperature, the specific temperature was deleted and was teplaced with "no-load Tavg" so the procedure can be used by either unit if required.
(Temporary) 3.5.49 E0?-9D, Emergency Weather Conditions, Revision 2, 11-08-79 The procedure was retitled (formerly "High Winds") and reuritten to encompass all severe weather conditions.
An arrangement has been made to receive storm warnings via the Appleton dispatcher and/or police radio in addition to receipt of field forcing alarms and evaluation of wind speed recorder data.
Steps have been incorporated to notify personnel of impending severe weather conditions via the plant Gai-tronics such that ongoing non-essential work activities may be terminated; unnecessary equipment can be secured; and surveillance of essential equipment and oper-ating parameters can be increased.
(Permanent) 3.5.50 E0?-10A, Control Room Inaccessibility, Revision 7, 07-26-79 A note was added following Step 4.12.2 that if letdown isolation on low level occurs, reopen MOV-427 with the local control switch mounted inside the MOV breaker cubicle when pressure level returns to normal. The note was necessitated by completion of modification request E-177.
(Permanent) 3.5.51 E0P-11A, Post-Accident Containment Ventilation System, Revision 2, 11-09-79 In Step 1.2 the containment exhaust gauge vent and stop valves were added.
Steps 4.2 and 4.6 were revised to reference Figure 1, a new chart which provides post-accident containment vent rate data.
(Permanent) 3.5.52 RP-1A, Preparation for Refueling, Revision 0, 01-23-79 (performed 10-11-79, Unit 1)
In Step 3.3 the actual limit for tritium (5.08 x 10 1) was inserted.
Step 4.30 regarding having all personnel not involved in the RV head lift exit containment was deleted because it was not required for this lif t.
Step 4.41 was revised to include performance of both ORT #1 and ORT #2.
The original note by the Superintendent -Operations only required ORT #2 to be performed.
(Temporary) 3.5.53 RP-1B, Recovery from Refueling, Revision 6, 06-23-77 (performed 04-04-79, Unit 2)
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In Steps 4.2 and 4.5, the reactor vessel internals were installed without notification to secure and then restore RHR flow after installation.
Although omission of these steps constituted a
procedural violation, the Staff determined that the incident was not reportable.
Mr. Rhodes stated he intends to discuss the matter in more detail during the pending refueling critique meeting.
(Temporary) 3.5.54 IT-05 (Unit 1), Inservice Testing of Spray Pumps and Eductor Supply Check Valves 847A & B, Revision 3, 09-19-78 Ihe procedure was expanded to utilize the test as a Section XI pressure test.
(Temporary) 3.5.55 IT-07, Inservice Testing and Rotation of Service ' dater Pumps P32A, B, C, D, E and F, evision 4, 10-03-78 (conducted 08-15-79)
The pump lineups were changed (more pumps added) to provide additional cooling for Unit 2; that containment temperature being high.
Unit 1 was in a cold shutdown condition at the time.
(Temporary) 3.5.56 IT-08 (Unit,1), Inservice Testing of Turbine-Driven Auxiliary Feed Pump, Revision 3, 05-01-79 A note was added to Step 3.2 to record both the control board and local indication of the position for valve 4000.
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Steps 3.8, 3.9 and 3.24 which used gate valve MS-126 to warm up the pump were deleted and old Step 3.11 was modified to use the trip valve to warm up the pump since that valve is better suited for the service.
(Permanent) 3.5.57 IT-08 (Unit 1), Inservice Testing of urbine-Driven Auxiliary Feed Pump, Revision 4, 06-01-79 In Steps 3.10 and 3.21 pressure gauge numbers were corrected for recording pump suction pressure.
Step 3.21 was also resequenced to isolate the suction pressure gauge just after securing the pump.
(Permanent) 3.5.58 IT-09 (Unit 2), Inervice Testing of Turbine-Driven Auxiliary Feed Pump, Revision 3, 05-01-79 In Steps 3.10 and 3,21 pressure gauge numbers were corrected for recordEg pump suction pressure.
Step 3.21 was also resequenced to isolate the suction pressure gauge just after securing the pump.
/ Permanent) 3.5.59 IT-09 (Unit 2), Inservice Testing of Turbine-Driven Auxiliary Feed Pump, Revision 4, 06-01-79 In Steps 3.10 and 3.21 pressure gauge numbers were corrected
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for recording pump suction pressure.
Step 3.21 was also resequenced to isolate the suction pressure gauge just efter securing the pump.
(Permanent) 3.5.60 IT-10, Inservice Testing of Electrically-Driven Auxiliary Feed Pumps, Revision 3, 06-01-79 Steps 3.3, 3.4, 3.10, and 3.21 were revised to indicate the appropriate pump suction pressure gauge.
Section 4.0 was revised to incorporate the new title of Mr. Rhodes to Superintendent -Operations.
(Permanent) 3.5.61 IT-10, Inservice Testing of Electrically-Driven Auxiliary Feed Pumps, Revision 4, 09-11-79 A
typographical error was corrected in Step 3.4.
(Permanent) 3.5.62 IT-10, Inservice Testing of Electrically-Driven Auxiliary Feed Pumps, Revision 4, 09-11-79 (performed on 11-09-79)
Steps 3.8.1 to 3.8.6 and 3.19.1 to 3.19.6 were added to qualify the test as a Section XI pressure test.
(Temporary) 3.5.63 IT-40 (Unit 1), Inservice Testing of Safety Injection Valves
! Quarterly), Revision 1, 09-07-79 New Step 4.9.2 was added to open valve SI-D17A (Whitey Vlave) to draw a 500 m1 sample, return the valve to its condition and deliver the sample to chemistry.
(Permanent) 3.5.64 IT-40 (Unit 1), Inservice Testing of Safety Injection valves (Quarterly), Revision 4, 08-03-79 Step 4.2.2 was added to replace the handwheel on SI-D17A (Whitey valve); remove the cap and draw a 500 m1 sample and shut the valve.
The sample is to be labeled and delivered to Chemistry.
The step was added to reflect completion of modification request M-564.
(Permanent) 3.5.65 1T-65 (Unit 2), Inservice Testing of Containment Isolation Valves (Quarterly'[, Revision 2, 07-17-79 In Step 4.29.2 a dim status light on isolation panel "B"
should be checked rather than a bright light; the test being appropriately corrected.
(Permanent) 3.5.66 IT-70 (Unit 1), Inservice Testing of Service Water Valves Quarterly), Revision 1, 08-02-79 Steps 4.12 through 4.15 were added for testing of valves H0V-2816 and MOV-2930A&B per inservice test requirements; it being recognized during a review of the test that these valves had not been originally included because of an oversight.
(Permanent) 3.5.67 IT-75 (Unit 2), Inservice Testing of Service Water Valves (Quarterly), Revision 2, 08-02-79 Steps 4.12 through 4.14 were added to include valves H0V-2817, A0V-LW-61 and A0V-LW-62 for the reasons described above for IT-70.
(Permanent) 3.5.68 IT-80 (Unit 1), Inservice Testing of Main and Radwaste Steam Valves, Quarterly, Revision 2, 08-23-79 Valves 2015 and 2016, atmospheric steam dumps, were deleted from the Section XI inservice test program because no credit is taken for the valves in the FFDSAR.
(Pe rmanent) 3.5.69 IT-85 (Unit 2), Inservice Testing of Main Steam Valves, Quarterly, Revision 2, 08-23-79 Valves 2015 and 2016, atmospheric steam dumps, were deleted from the Section XI inservice test program because no credit is taken for the valves in the FFDSAR.
(Permanent) 3.5.70 IT-90 (Unit 1), Inservice Testing of Auxiliary Feedwater Valves (Quarterly), Revision 1, 05-01-79 New Steps 4.2.1 and 4.3.1 were added and existing Steps 4.2.4 and 4.3.4 were rephrased to record the local position indication on the valve prior to stroking; then returning the valve to the recorded position.
(Permanent) 3.5.71 IT-95 (Unit 2), Inservice Testing of Auxiliary Fcedwater Valves (Quarterly), Revision 0, 08-01-77 (performed 04-22-79)
In Steps 4.2.3 and 4.3.3 the opening percentage value of the valves was changed to provide rated flow.
(Temporary) 3.5.72 IT-95 (Unit 2), Inservice Testing of Auxiliary Feedwater Valves (Quarterly), Revision 1, 05-01-79 New Steps 4.2.1 and 4.3.1 were added and existing Steps 4.2.4 and 4.3.4 were rephrased to record the local position indication on the valve prior to stroking; then returning the valve to the recorded position.
(Permanent) 3.5.73 IT-110 (Unit 1), Inservice Testing of Reactor Coolant,
~
Chemical and Volume Control System and Component Cooling Valves (Cold Shutdown
, Revision 2, 09-27-79 (performed 03-14-79)
Steps 4.13 and 4.14 were added to test valve LCV-427 as required by Section XI.
(Temporary) 3.5.74 IT-110 (Unit 1), Inservice Testing of Reactor Coolant, Chemical and Volume Control System and Component Cooling Valves (Cold Shutdown), Revision 3, 03-19-79 Steps 4.13 and 4.14 were added to test valve LCV-427 as required by Section XI.
(Temporary) 3.5.75 IT-115 (Unit 2), Inservice Testing of Reactor Coolant, Chemical and Volume Cotrol System and Component Cooling Valves (Cold Shutdown
, Revision 3, 03-19-79 Steps 4.13 and 4.14 were added to test valve LCV-427 as required by Section XI.
(Temporary) 3.5.76 IT-140 (Unit 1), Inservice Testing of Auxiliary Feedwater System Check Valves (Cold Shutdown), Revision 4, 05-01-79 The rest was rewritten to change it from a full-flow test of the auxiliary feedwater pumps to a full-flow test of the auxiliary feed check valves ac required by Section XI.
Emphasis was added on valve positioning to attain the required flow.
The use of local valve position indication on the steam-driven auxiliary feed pumps was added to obtain consistent data. New steps were also added to assure return of the system to its proper valve lineup.
(Permanent) 3.5.77 IT-145 (Unit 2), Inservice Test of Auxiliary Feedwater System Check Valves (Cold Shutdown), Revision 4, 05-01 'i9 The rest was rewritten to change it from a full-flow test of the auxiliary feedwater pumps to a full-flow t(st of the auxiliary feed check valves as required by Section XI.
Emphasis was added on valve positioning to attain the required flow.
The use of local valve position indication on the steam-driven auxiliary feed pumps was added to obtain consistent data. New steps were also added to assure return of the system to its proper valve lineup.
(Permanent) 3.5.78 IT-160 (Unit 1), Inservice Testing of Purge Supply and Exhaust Valves (Cold Shutdown), Revision 0, 12-16-77 The inservice test was canceled because the valves are tagged out-of-service per NRC requirements.
(Canceled) 3.5.79 IT-165 (Unit 2), Inservice Testing of Purge Supply and Exhaust Valves (Cold Shutdown), Revision 0, 12-16-77 The inservice test was canceled because the valves are tagged out-of-service per NRC requirements.
(Canceled) 3.5.80 IT-200 (Unit 1), Inservice Testing of Safety Injection Accumulator D charge Check Valves 842A and B (Cold Shutdown), Ru..sion 3, 03-09-79 The test frequency was changed from a refueling shutdown interval to a cold shutdown interval in accordance with Section VI, Subsection IWV, and a commitment made to the NRC in the submittal of the inservice test plan.
(Permanent) 3.5.81 IT-205 (Unit 2), Inservice Testing of Safety Injection Accumulator Discharge Check Valves 842A and B (Cold Shutdown), Revision 3, 03-09-79 The test frequency was changed from a refueling shutdown interval to a cold shutdown interval in accordance with Section XI, Subsection IWV, and a commitment made to the NRC in the submittal of the inservice test plan.
(Permanent) 3.5.82 IT-220 (Unit 1) and IT-225 (Unit 2), Incervice Testing of Main Steam Line Non-Return Valves (Cold Shutdown Requirement), Revision 1, 02-19-79 The tests were revised to measure the initial breakaway torque required for each valve.
(Permanent) 3.5.83 IT-230 (Unit 1), Leak Test of Class 1 Components Following a Refueling Shutdown, Revision 1, 04-20-79 Typographical errors were corrected in Steps 4.4, 4.9, 4.19, and 4.31.
(Permanent) 3.5.84 IT-235 (Unit 2), Leak Test of Class 1 Components Following a Refueling Shutdown, Revision 0, 10-09-78 (performed 04-10-79)
Typographical errors were corrected in Steps 4.4, 4.9, 4.19, and 4.31.
(Temporary) 3.5.85 IT-235 (Unit 2), Leak Test of Class 1 Components Following a Refueling Shutdown, Revision 1, 04-20-79 Typographical errors were corrected in Steps 4.4, 4.9, 4.19, and 4.31 (Permanent) 3.5.86 IT-240, Inservice Operational Precsure Test of High Head Safety Injection System, Revision 0, 10-05-79 (performed 11-19-79)
In Step 2.2, " reactor coolant system pressure of >1800 psig with the reactor in hot shutdown" was changed to be
>1600 psig with the hot shutdown requirement deleted to permit the test to be conducted during performance of IT-230 (Class 1 leak test).
In Step 3.6 a sentence was added to use an installed pressure test gauge, if available. One was not available so test gauge #21 was installed.
In Step 3.20 a
valve number typographical error was corrected.
(Temporary) 3.5.87 ORT #1 (Unit 1), Flow Test of High Head Safety Injection Pumps, Revision 9, 10-09-78 (performed 10-11-79)
Step 4.2.8 was modified to procedurally ensure that the portable vibration monitor being used has been calibrated within the last year.
(Temporary) 3.5.88 ORT #1 (Unit 1), Flow Test of High Head Safety Injection Pumps, Revision 10, 11-14-79 Step 4.3 was added to procedurally ensure that the portable vibration monitor being used has been calibrated within the last year.
(Permanent) 3.5.89 ORT #1 (Unit 2), Flow Test of High Head Safety Injection Pumps, Revision 7, 08-21-78 (performed 03-31-79)
Valves 825A&B were not opened in the initial part of the test because 825A was already open and an electrical relay failure prevented closure of 825B when the test was being conducted.
(Temporary) 3.5.90 ORT #1 (Unit 2), Flow Test of High Head Safety Injection Pumps, Revision 8, 04-05-79 Steps 5.8 and 5.9 were added to alternate suction valves from the refueling water storage tank to the suction of the safety injection pumps to verify full flow through the individual suction valves.
(Permanent) 3.5.91 ORT #1 (Unit 2), Flow Test of High Head Safety Injection Pumps, Revision 9, 05-21-79 A topographical error was corrected in Step 6.3. (Permanent) 3.5.92 ORT #1 (Unit 2), Flow Test of High Head Safety Injection Pumps, Revision 10, 11-14-79 Step 4.3 was added to procedurally ensure that the portable vibration monitor being used has been calibrated within the last year.
(Permanent) 3.5.93 ORT #2 (Unit 1), Flow Test of Low Head Safety Injection Pumps, Revision 3, 09-29-78 (performed 10-14-79 Step 3.3.7 was modified to ensure the portable vibration monitor being used has been calibrated within the last year.
(Temporary) 3.5.94 ORT #2 (Unit 1), Flow Test of Low Head Safety Injection Pumps, Revision 4, 11-14-79 Step 3.4 was added to procedurally ensure that the portable vibration monitor being used has been calibrated within the last year.
(Permanent) 3.5.95 ORT #2 (Unit 2), Flow Test of Low Head Safety Injection Pumps, Revision 4, 11-14-79 Step 3.4 was added to procedurally ensure that the portable vibration monitors being used have been calibrated within the last year.
(Permanent) 3.5.96 ORT #3 (Unit 1), Safety Injection Actuation with Loss of Engineered Safeguards AC, Revision 8, 08-23-79 The procedure was revised to reflect completion of modification request IC-199, reset circuitry changes; deletion of the reference to pressurizer level / pressure coincidence; and inclusion of valves A0V-825C, the MOV-825A&B bypass.
(Permanent) 3.5.97 ORT #3 (Unit 1), Safety Injection with Loss of Engineered Safeguards AC, Revision 8, 08-23-79 (conducted on 10-26-79)
In Steps 4.5 and 4.20.4 typographical errors were corrected.
In Step 4.10 1W1C1 fan breaker was operated in the test position because the fan was out-of-service for maintenance at the time of the test.
(Temporary) 3.5.98 ORT #3 (Unit 1), Safety Injection with Loss of Engineered Safeguards AC, Revision 9, 11-05-79 Typographical errors in Steps 4.5 and 4.20.4 were corrected.
(Permanent) 3.5.99 ORT #3 (Unit 2), Safety Injection Actuation with Loss of Engineered Safeguards AC, Unit 2, Revision 8, 01-30-79 The test sequence was reorganized to provide a more logical order.
In addition, clarifying notes and equipment numbers were provided.
(Permanent) 3.5.100 ORT #3 (Unit 2), Safety Injection Actuation with Loss of Engineered Safeguards AC, Revision 8, 01-30-79 (performed 04-04-79)
In Steps 3.8.5 and 4.10, the fan W1A1 breaker was in the test position because the fan was out-of-service. In Step 4.14 service air compressor K3B was inoperative at the time of the test.
In Step 4.19.1 the "A"
train manual safety injection master relay failed to remain latched (see
~
MSSM 78-08 for an earlier discussion of this test).
(Temporary) 3.5.101 ORT #3 (Unit 2), Safety Injection Actuation with Loss of Engineered Safeguards AC, Revision 9, 09-12-79 In Steps 3.2 and 3.3 the proper Unit 2 designation was added.
In Step 3.9 the reference to pressurizer level and pressure coincidence was deleted.
In Step 4.5 main feed pump interlock bypass information was corrected per instructions by the Maintenance Supervisor (Electrical).
In Step 4.9 the order of steps was changed to be similar to Unit 1.
Step 4.18 was changed to note count-down should proceed via the Gai-tronics.
In Step 4.21 valve A0V-825C was added to prevent boric acid tank dilution.
The Step 4.25.1 note was expanded to note that A0V's will also close.
Steps 4.26 and 4.29 were modified per completion of modification requests IC-199 and IC-200 (Train "A"
and "B" reset buttons located on the rear of panel C01.
(Permanent) 3.5.102 ORT #4 (Unit 1), Main Turbine Mechanical overspeed Trip Device, Revision 1, 04 20-79 In Step 3.5, references to the fourth through sixth trip tests were removed because only three trip tests are performed.
Step 3.6.1 was restated to relatch the unit by resetting " setter" and " reference" to turbine RPM rather than zero.
(Permanent) 3.5.103 ORT #4 (Unit 2), Main Turbine Mechanical Overspeed Trip Device, Revision 0, 01-30-79 (performed 04-13-79)
In Step 3.5, references to the fourth through sixth trip tests were removed because only three trip tests are performed.
Step 3.6.1 was restated to relatch the unit by resetting " setter" and " reference" to turbine RPM rather than zero.
(Temporary) 3.5.104 ORT #4 (Unit 2), Main Turbine Machanical Overspeed Trio Device, Revision 1, 04-20-79 In Step 3.5, references to the fourth through sixth trip tests were removed because only three trip tests are performed.
Step 3.6.1 was restated to relatch the unit by resetting " setter" and " reference" to turbine RPM rather than zero.
(Permanent) 3.5.105 ORT #5 (Unit 1), Sump "B" to Residual Heat Removal Pump Suction Valves' Hydro, Revision 3, 08-30-79 In Section 1.0 tyographical corrections were made.
In Step 3.5 the test pressure was changed from 60 psig to 75 psig to qualify the test as a Class 2 Section XI hydro.
(Temporary) 3.5.106 ORT #5 (Unit 1), Sump "B" to Residual Heat Removal Pump Suction Valves' Hydro, Revision 4, 10-10-79 Typographical errors were corrected in Section 1.0.
(Permanent) 3.5.107 ORT #6 (Unit 1), Containment Spray Sequence Test, Revision 3, 08-24-79 The test was revised to reflect completion of modification request IC-199, containment spray reset switches for each train.
(Permanent) 3.5.108 ORT #6 (Unit 2), Operations Refueling Test, Containment Spray Sequence, Revision 3, 09-12-79 Step 3.4 was added per completion of modification request IC-200 to reset both trains of containment spray on the rear of panel C01.
(Permanent) 3.5.109 ORT #7 (Unit 2), Operation of Backdraf t Dampers, Revision 1, 02-02-76 (performed 04-07-79)
The test was rewritten to better document damper
" operability" as desired by the NRC. (Temporary) 3.5.110 ORT #7 (Unit 2), Operation of Backdraft Dampers, Revision 2, 04-09-79 The test was rewritten to better document damper
" operability" as desired by the NRC.
(Permanent) 3.5.111 ORT #8, Visual Check of Residual Heat Removal System, Revision 2, 08-30-77 This procedure was performed as modified during the U1R7 shutdown.
The test was changed to qualify the visual inspection as an inservice operational pressure test as required by Section XI: A sentence was added to Step 2.1 to meet the four-hour minimum pressurization of the system.
In the comment section a data section was added.
(Temporary) 3.5.112 ORT #15 (Unit 2), Fuel Manipulator and Fuel Transfer System Checkout, Revision 0, 05-15-78 (performed 04-01-791 Typographical errors in Steps 2.1.1 and 2.1.2 were corrected. (Temporary) 3.5.113 ORT #15 (Unit 2), Fuel Manipulator and Fuel Transfer System Checkout, Revision 1, 04-05-79 Typographical errors in Steps 2.1.1 and 2.1.2 were corrected.
(Permanent) 3.5.114 ORT #15 (Unit 1), Fuel Manipulator and Fuel Transfer System Checkout, Revision 2, 04-05-79
~
Typographical errors in Steps 2.1.1 and 2.1.2 were corrected. (Permanent) 3.5.115 ORT #25 (Unit 1), IVLT, Penetration #9, Revision 4, 01-30-79 (performed 10-13-79)
Steps 3.1.3 and 3.2.3 were modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test.
This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.116 ORT #26 (Unit 1), IVLT, Penetration #10, Revision 4, 08-21-78 (performed 10-21-79)
Step 3.2 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.117 ORT #27 (Unit 1), IVLT, Penetration #11, Revision 2, 05-15-76 (performed 10-17-79)
Step 3.3 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.118 ORT #28 (Unit 1), IVLT, Penetration #12A, Revision 1, 11-14-75 (performed 10-14-79)
Step 3.3 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage af ter the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.119 ORT #29 (Unit 1), IVLT, Penetration #12B, Revision 1, 11-14-75 (performed 10-11-79)
Step 3.2 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of tLe test of any piping leakage or the inability to perform the pneumatic pressure test.
This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.120 ORT #30 (Unit 2), Isolation Valve Leak Test, Penetration #12C, Revision 1, 02-02-76 (performed 04-07-79)
The secondary local leak test section was omitted because it had been performed previously on 03/29/79; the~efore only the primary test was required to be performed in the retest.
(Temporary) 3.5.121 ORT #32 (Unit 1), IVLT, Penetration #14C, Revision 2, 05-15-76 (performed 10-11-79)
Step 3.2 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.122 ORT #33 (Unit 1), IVLT, Penetration #25C, Revision 4, 10-03-79 (performed 10-16-79)
Step 3.2 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.123 ORT #34 (Unit 1), IVLT, Penetration #26, Revision 1, 11-14-75 (performed 10-18-79)
Step 3.4 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary)
~
3.5.124 ORT #39 (Unit 1), IVLT, Penetration #29A, Revision 2, 05-15-76 (performed 10-17-79 Step 3.3 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.125 ORT #40 (Unit 1), IVLT, Penetration #29B, Revision 4, 10-09-78 (performed 10-17-79)
Step 3.3 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.126 ORT #42 (Unit 1), IVLT, Penetration #30C, Revision 3, 08-30-77 (performed 10-23-79)
Steps 3.2 and 3.4 were modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test.
This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.127 ORT #43 (Unit 1), IVLT, Penetration #31B, Revirion 3, 08-30-77 (performed 10-15-79)
St:p 3.2 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test.
This change permitted the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.128 ORT #44 (Unit 1), IVLT, Penetration #31C, Revision 4, 01-30-79 (performed 10-15-79)
Step 3.3 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test.
This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.129 ORT #47 (Unit 1), IVLT, Penetration #33A, Revision 3, 10-09-78 (performed 10-16-79)
Steps 3.1 and 3.2 were modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test.
This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.130 ORT #48 (Unit 1), IVLT, Penetration #33B, Revision 3, 10-09-78 (performed 10-18-79)
Step 3.1 was modified to test at a.ressure of 40 psia (75 psig) 15 psi and inspect the line r~ce leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits
~
the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.131 ORT #49 (Unit 1), IVLT, Penetration #33C, Revision 2, 05-15-76 (performed 10-28-79)
Step 3.4 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressare test.
This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.132 ORT #49 (Unit 1), IVLT, Penetration #33C, Revision 2, 05-15-76 (performed 11-15-79)
The data sheet comments section and the drawing were modified to qualify this test as a section XI pressure test.
(Temporary) 3.5.133 ORT #50 (Unit 1), IVLT, Penetration #34A, Revision 2, 05-15-76 (performed 10-11-79)
Step 3.4 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the
~
piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.134 ORT #53 (Unit 1), IVLT, Penetration #34D, Revision 2, 05-15-76 (performed 10-15-79)
Steps 3.2 and 3.4 were modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test.
This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.135 ORT #58 (Unit 1), IVLT, Penetration #56, Revision 3, 1_0-09-78 (performed 10-25-79)
Step 3.2 was modified to test at a pressure of 90 psia (75 psig) 15 psi and inspect the line for leakage after the piping has been pressurized for at least 10 minutes.
An instruction was added to place a note in the comments section of the test of any piping leakage or the inability to perform the pneumatic pressure test. This change permits the ORT to qualify as a Section XI pressure test.
(Temporary) 3.5.136 ORT #60 (Unit 1), IVLT, Penetration #55, Revision 2, 05-15-76 (performed 10-18-79 Step 3.2 was modified to correct a typographical error in that valve 869B should have been indicated rather than 860B.
(Temporary) 3.5.137 ORT #60 (Unit 1), IVLT, Penetration #55, Revision 3, 11-14-79 In Step 3.2 a typographical error was corrected such that valve 869B was indicated rather than valve 860B.
(Permanent) 3.5.138 ORT #65 (Unit 1), IVLT, Penetration #X2, Revision 5, 01-30-79 The drawings were corrected to reflect as-built conditions (removal of valve #3 internals and addition of check valve 3A).
(Temporary) 3.5.139 ORT #65 (Unit 1), IVLT, Penetration #X2, Revision 6, 11-14-79 The drawings were corrected to reflect as-built conditions (removal of valve #3 internals and addition of check valve 3A). (Permanent) 3.5.140 ORT #69 (Unit 2), IVLT Penetrations #16 and #18, Revision 5, 04-25-79 Valve D10 was deleted from the Page 7 instructions because it was no longer applicable.
(Permanent) 3.5.141 ICP 2.1 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog Channels I through IV, Revision 12, 09-18-78 (performed 05-16-79)
A new step was added to Section 6.0 to place the appropriate level trips to normal for the duration of the test to prevent safety injection for occurring during testing.
In Section 8.0 a step was added to return the aforementioned level trips to the trip condition per the requirements of SMP 11.3.
(Temporary) 3.5.142 ICP 2.1 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog Channels I through IV, Revision 12, 09-18-78 (performed 07-11-79)
Steps 6.140 and 6.222 were changed such that turbine power
>X% to 513.38 inc. and load limit reduction low to 537.08 dec. because first stage pressure was changed from 530 psi to 550 psi, and load limit reduction lo should be 5 since it is desired to run the turbine back to less than 80% power when called for.
(Temporary) 3.5.143 ICP 2.1 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 13, 07-13-79 Steps 6.140 and 6.222 were changed such that turbine power
>X% to -<13.38 inc. and load limit reduction low to <37.08 dec. because first stage pressure was changed from 530 psi to 550 psi, and load limit reduction lo should be < since it is desired to run the turbine back to less than 80% power when called for.
(Temporary) 3.5.144 ICP 2.1 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 13, 07-13-79 (performed 11-20-79)
Steps 6.22, 6.23, 6.24, 6.26, 6.38, 6.107, 6.108, 6.109, 6.111, 6.123, 6.188, 6.189, 6.190, 5.192, 6.204, 6.280, 6.281, 6.282, 6.284, 6.296, and 6.320 were changed to reflect unit operation at reduced primary system temperatures.
Steps 6.69, 6.159, 6.245, and 6.315 were changed to reflect a setpoint change approved for low-low steam generator level.
Steps 6.53, 6.74, 6.163, 6.249, 6.254, and 6.325 were changed to provide protection against spurious safeguards actuation resulting from channel noise incurred during reduced power operation.
(Temporary) 3.5.145 ICP 2.1 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 13, 07-13-79 (performed 11-28-79)
Steps 6.22, 6.23, 6.24, 6.26, 6.38, 6.107, 6.108, 6.109, 6.111, 6.123, 6.188, 6.189, 6.190, 5.192, 6.204, 6.280, 6.281, 6.282, 6.284, 6.296, and 6.320 were changed to reflect unit operation at reduced primary system temperatures.
Steps 6.69, 6.159, 6.245, and 6.315 were changed to reflect a setpoint change approved for low-low steam generator level.
Steps 6.43, 6.74, 6.163, 6.249, 6.254, and 6.325 were changed to provide protection against spurious safeguards actuation resulting from channel noise incurred during reduced power operation.
(Temporary) 3.5.146 ICP 2.1 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 13, 07-13-79 (performed 12-06-79)
This cbsolete revision of the procedure (see 3.5.148 below) was utilized in order install safeguards trip setpoint changes (Steps 6.53, 6.74, 6.163, 6.249, 6.254 and 6.325) for low Tavg operation.
(Temporary) 3.5.147 ICP 2.1 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 14, 12-03-79 Steps 6.22, 6.23, 6.24, 6.26, 6.38, 6.107, 6.108, 6.109, 6.111, 6.123, 6.188, 6.189, 6.190, 5.192, 6.204, 6.280, 6.281, 6.282, 6.284, 6.296, and 6.320 were changed to reflect unit operation at reduced primary system temper-atures. Steps 6.69, 6.159, 6.245, and 6.315 were changed to reflect a setpoint change approved for low-low steam generator level.
Steps 6.53, 6.74, 6.163, 6.249, 6.254, and 6.325 were changed to provide protection against spurious safeguards actuation resulting froia channel noise incurred during reduced power operation.
(Permanent) 3.5.148 ICP 2.1 (Unit 2), Reactor Protection and Safeguaids Analog Channels I through IV, Revision 9, 09-18-78 (performed 04-10-79 Steps 6.32 and 6.33 were changed to reflect different voltage values due to the installation of a new Channel 41 power range detector.
(Temporary) 3.5.149 ICP 2.1 (Unit 2), Reactor Protection and Safeguards Analog Channels I through IV, Revision 10, 04-17-79 Steps 6.32 and 6.33 were changed to reflect different voltage values due to the installation of a new Channel 41 power range detector.
(Permanent) 3.5.150 ICP 2.1 (Unit 2), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 11, 07-11-79 In Step 6.222 a typographical error was corrected such that load limit reduction lo is less than or equal to 37.08 dec.
In Step 2.290 and 6.291 voltage was changed to 2.78 and 3.64, respectively, because of Channel 44 detector changeout.
(Permanent) 3.5.151 ICP 2.1 (Unit 2), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 11, 07-11-79 (performed 09-25-79)
The procedure was modified in order to accomplish ICP 2.2.
In Steps 6.69, 6.85, 6.159, 6.245, 6.315, and 6.320 the ideal setting for steam generator lo-lo level bistables was changed from 10% to 15% (>16.00 ma dec.) for reference leg heatup compensation in accordance with IE Bulletin No. 79-21.
(Temporary) 3.5.152 ICP 2.1 (Unit 2), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 12, 09-26-79 Steps 6.69, 6.85, 6.159, 6.245, 6.315, and 6.320 were modified such that steam generator lo-lo level bistables were changed from 10% to 15% (>16.00 ma dec.) in accordance with IE Bulletin No. 79-21.
(Permanent) 3.5.153 ICP 2.2 (Unit 2), Periodic Test, Reactor Protection and Safeguards Analog, Channels I Through IV, Revision 8, 09-19-78 (performed 03-17-79)
In Steps 6.19, 6.20, 6.21, 2.59, 6.60, 6.61, 6.100, 6.101, 6.102, 6.141, 6.142, and 6.143 null temperature resistances were changed because of stretch operation.
Likewise, Tavg bistable settings in Steps 6.23, 6.63, 6.104, and 6.105 were changed for the same reason.
(Temporary) 3.5.154,IrF 2.2 (Unit 2), Periodic Test, Reactor Protection and aafeguards Analog Channels I through IV, Revision 8, 09-18-78 (performed 05-11-79)
A new step was added to Section 6.0 to place the appropriate level trips to normal for the duration of the test to prevent safety injection for occurring during testing.
In Section 8.0 a step was added to return the aforementioned level trips to the trip condition per the requirements of SMP 11.3.
(Temporary) 3.5.155 ICP 2.2 (Unit 2), Periodic Test, Reactor Protection and Safeguards Logic, Channels I through IV, Revision 8, (performed 06-01-79)
In Step 6.29 Qu volts was changed to 2.78 volts because a new detector was installed in NIS Channel 41.
In Step 6.30 Q1 volts was changed to 3.64 volts for the same reason.
(Temporary) 3.5.156 ICP 2.2 (Unit 2), Periodic Test, Reactor Protection and Safeguards Logic Channels I through IV, Revision 9, 06-04-79 In Step 6.29 Qu volts was changed to 2.78 volts because a new detector was installed in NIS Channel 41.
In Step 6.30 Q1 volts was changed to 3.64 volts for the same reason.
(Permanent) 3.5.157 ICP 2.2 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog Channels I through IV, Revision 11, 09-18-78 (performed 05-01-79 A new section was added entitled, " Preliminary", between Sections 5.0 and 6.0 to place all pressurizer level trips to normal for the duration of the test to prevent safety injection from occurring during the test. A new Section 8.0 was added to return the appropriate pressurizer level trips to the trip condition per the requirements of SMP 11.3.
(Temporary) 3.5.158 ICP 2.2 (Unit 1)), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 11, 09-18-79 (performed 12-11-79)
Steps 6.19, 6.20, 6.21, 6.23, 6.25, 6.59, 6.60, 6.61, 6.63, 6.75, 6.100, 6.101, 6.102, 6.104, 6.116, 6.141, 6.142, 6.143, 6.145, and 6.157 were revised to reflect unit operation at reduced primary system temperatures.
(Temporary) 3.5.159 ICP 2.2 (Unit 1), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 12, 12-10-79 Steps 6.19, 6.20, 6.21, 6.23, 6.25, 6.59, 6.60, 6.61, 6.63, 6.75, 6.100, 6.101, 6.102, 6.104, 6.116, 6.141, 6.142, 6.143, 6.145, and 6.157 were revised to reflect unit operation at reduced primary system temperatures.
(Permanent) 3.5.160 ICP 2.2 (Unit 2), Periodic Test, Reactor Protection and Safeguards Analog Channels I through IV, Revision 9, 06-04-79 (performed 07-03-79)
In Steps 6.19, 6.20 and 6.75, typographical errors were corrected.
In Steps 6.151 and 6.152 voltage was changed to 2.78 and 3.64, respectively, because of Channel 44 detector changeout.
In Steps 6.24 and 6.64 the phrase "trans analog" was changed to " analog simulator", the proper terminology.
(Temporary) 3.5.161 ICP 2.2 (Unit 2), Periodic Test, Reactor Protection and Safeguards Analog, Channels I through IV, Revision 10, 07-11-79 In Steps 6.19, 6..^ 0, and 6.75, typographical errors were corrected.
In Steps 6.151 and 6.152 voltage was changed to 2,78 and 3.64, respectively, because of Channel 44 detector changeout.
In Steps 6.24 and 6.64 the phrase "trans analog" was changed to " analog simulator", the proper terminology.
(Temporary) 3.5.162 ICP 2.3, Periodic Test, Reactor Protection System Logic, Revision 4, 09-20-79 New Precaution 5.4 was added to check that all RT relays are energized by observing the relay " button" is pulled in after testing is completed in either train and before the trip breakers are return to normal configuration.
Precaution 5.5 was added to verify by individual operation that operation of a single relay will not cause the trip to occur after testing of all trip combinations of relays listed under eacl.
protective function.
Step 6.1.17.e was moved to be Step 6.1.e because it was not appropriate in its original position.
A caution note was added to Steps 6.1.20 and 6.2.20 to perform a power range low setpoint trip (Train A) only it reactor power is below P10 because if the check would be performed above P10 a turbine runback could occur.
In Notes 1 and 2 at the end of the procedure the word " white light" was substituted for " recorder" (Permanent) 3.5.163 ICP 2.5 (Unit 1), Periodic Test, Safeguards System Logic, Revision 7, 10-10-77 performed 05-17-79)
Steps 6.1.1.d and 6.3.1.d were changed to a 2/3 logic check between pressure channels PC-429C, PC-430E, and PC-431G because modification request IC-192 modified the logic to delete the pressurizer pressure / level coincident trip and created a 2/3 pressure trip.
Steps 6.1.1.e, 6.1.1.f, 6.3.1.e, and 6.3.1.f were deleted for the aforementioned reason.
In Steps 6.2.1.a, 6.2.1.c, 6.4.1.a, and 6.4.1.c PC-430E (B3) and PC-431G (C3) were deleted for the same reason.
(Temporary) 3.5.164 ICP 2.5 (Unit 1), Periodic Test, Safeguards System Logic, Revision 8, 05-22-79 Steps 6.1.1.d and 6.3.1.d were changed to a 2/3 logic check between pressure channels PC-429C, PC-430E, and PC-431G because modification request 1C-192 modified the logic to delete the pressurizer pressure / level coincident trip and created a 2/3 pressure trip.
Steps 6.1.1.e, 6.1.1.f, 6.3.1.e, and 6.3.1.f were deleted for the aforementioned reason.
In Steps 6.2.1.a, 6.2.1.c, 6.4.1.a, and 6.4.1.c PC-430E (B3) and PC-431G (C3) were deleted for the same reason.
(Permanent) 3.5.165 ICP 2.5 (Unit 1), Periodic Test, Safeguards System Logic, Revision 9, 06-13-79 In Step 6.4.8.a, a typographical error was corrected such that the test point value is "approximately 900 ohms" versus the 625 ohms indicated.
(Temporary) 3.5.166 ICP 2.5 (Unit 1), Periodic Test, Safeguards System Logic, Revision 9, 05-22-79 (performed 06-12-79)
In Step 6.4.8.a, a typographical error was corrected such that the test point value is "approximately 900 ohms" versus the 625 ohms indicated.
(Temporary) 3.5.167 ICP 2.5 (Unit 1), Periodic Test, Safeguards System Logic, Revision 10, 08-29-79 Step 5.4 was added to not perform the safety injection or containment spray test on the effective train of safeguards equipment when a Diesel is out-of-service for inspection (Train "A" when the "B"
Diesel is out, and Train "B" when the "A" Diesel is out).
The precaution was added because a train of safeguards equipment of out-of-service while the Diesel inspection is in progress.
If the effective train of safeguards would be tested under these conditions, all safeguards equipment would be momentarily secured during the test.
(Permanent) 3.5.168 ICP 2.5 (Unit 1), Periodic Test, Safeguards Systen Logic, Revision 11, 09-20-79 Precaution 5.5 was added such that after testing all trip combinations of relays listed under each protection function, cperation of a single relay will not cause a trip to occur.
The verification is to be performed by operation of each individual relay.
(Permanent) 3.5.169 ICP 2.5 (Unit 2) Periodic Test, Safeguards System Logic, Revision 6, 03-02-78 (performed 05-12-79)
Step 6.1.1.d was be changed to a 2/3 logic check between pressure channels PC-429C, PC-430E and PC-431G.
IC-193 modified the logic to delete pressure / level coincident trip and create a 2/3 pressure trip.
Steps 6.1.1.e,
6.1.1.f, 6.2.1.c, 6.3.1.d, 6.3.1.e, and 6.3.1.f are to be deleted for this same reason.
Steps 6.2.1.a, 6.2.1.c, 6.4.1.a and 6.4.1.c are changed to delete PC-430E (B3) and PC-431G (B3) for the same reasons noted for 6.1.1.d. (Temporary) 3.5.170 ICP 2.5 (Unit 2), Periodic Test, Safeguards System Logic, Revision 7, 05-11-79 Step 6.1.1.d was changed to a 2/3 logic check between pressure channels PC-429C, PC-430E and PC-431G.
IC-193 modified the logic to delete pressure / level coincident trip and create a 2/3 pressure trip.
Steps 6.1.1.e, 6.1.1.f, 6.2.1.c, 6.3.1.d, 6.3.1.e. and 6.3.1.f are to be deleted for this same reason.
Steps 6.2.1.a, 6.2.1.c, 6.4.1.a and 6.4.1.c are changed to delete PC-430E (B3) and PC-431G (B3) for the <:ame reasons noted for 6.1.1.d.
(Permanent) 3.5.171 ICP 2.5 (Unit 2), Periodic Test, Safeguards System Logic, Revision 8, 08-29-79 Step 5.4 was added to not perform the safety injection or containment spray test on the effective train of safeguards equipment when a Diesel is out-of-service for inspection (Train "A" when the "B"
Diesel is out, and Train "B" when the "A" Diesel is out).
The precaution was added because a train of safeguards equipment of out-of-service while the Diesel inspection is in progress.
If the effective train of safeguards would be tested under these conditions, all safeguards equipment would be momentarily secured during the test.
(Permanent) 3.5.172 ICP 2.5 (Unit 2), Periodic Test, Safeguards System Logic, Revision 9, 09-20-79 Precaution 5.5 was added such that after testing all trip combinations of relays listed under each protection function, operation of a single relay will not cause a trip to occur.
The verification is to be performed by operation of each individual relay.
In Step 6.1.1.d, the coordinates A3 were added for PC-429C because they had been inadvertently omitted during the last revision.
(Permanent) 3.5.173 ICP 2.7, Periodic Test, Nuclear Instrumentation Power Range Channels N41, N42, N43, and N44, Revision 4, 10-24-78 (performed 03-17-79 Currents used in Steps 6.32.1 and 6.32.2 were changed to those currents existing at the time of the test.
In Step 6.32.3 ideal voltages were changes to reflect the currents used for the test.
These changes were required because power range summing amplifier gain changes made reflecting thermal power will not allow the appropriate currents to be added to each detector existing current without exceeding the range of the summing amplifier output.
(Temporary) 3.5.174 ICP 2.7, Periodic Test, Nuclear Instrumentation Power Range Channels N41, N42, N43, and N44, Revision 5, 03-27-79 Step 6.10 was revised to clarify that the appropriate P10 status lights are illuminated on the trip status panel on C04 because the permissive status lights require a 2/4 logic.
(Permanent) 3.5.175 ICP 2.7 (Unit 1), Periodic Test, Nuclear Instrumentation Power Range Channels N41, N42, N43, and N44, Revision 5, 03-27-79 (performed 11-16-79)
New Condition 4.2 was added to note that this test includes the checks performed in ICP 2.14 and may be accomplished in total in place of 2.14.
(Temporary) 3.5.176 ICP 2.7 (Unit 1), Periodic Test, Nuclear Instrumentation Power Range Channels N41, N42, N43, nd N44, Revision 6, 11-20-79 New Condition 4.2 was added to note that this test includes the checks performed in ICP 2.14 and may be accomplished in total in place of 2.14.
(Permanent) 3.5.177 ICP 2.7, Periodic Test, Instrumentation Power Range Channels N41, N42, N43, and N44, Revision 5, 03-27-79 (per-formed 11-29-79, Unit 1)
Step 6.25 was modified reduce the ideal setting for overpower trip high range channels at 80% until after unit startup.
(Temporary) 3.5.178 ICP 2.8, Power Range Channels Axial offset Calibration, Revision 2, 10-24-78 (performed 04-16-79)
Step 6.1 used currents for 77% power instead of 105% power to avoid having one channel in the trip condition for a long period of time.
In Steps 7.11 and 7.15 voltage was changed from 2.187 (105%) to 1.604 (77%) for the same reasons noted for Step 6.1.
Likewise, in Step 7.18 voltage was changed from 8.75 to 6.417, and in Steps 7.19 and 7.20 indication was changed accordingly. (Temporary) 3.5.179 ICP 2.8, Periodic Test, Nuclear Instrumentation System Power Range Channels - Axial Offset Calibration, Revision 2, 10-25-78 (performed 12-02-79, Unit 1)
In Steps 7.11, 7.15, 7.18, and 7.20 voltages and power levels were adjusted to reflect power conditions under which the detectors were calibrated.
(Temporary) 3.5.180 ICP 2.10 (Unit 2), Periodic Test, Nuclear Instrumentation 3
System Source Range Channels N31 and N32, Revision 2, 06-05-79 In Step 6.20 an asterisk which had been inadvertently omitted from the previous revision was reinserted.
(Permanent) 3.5.181 ICP 2.14, Periodic Test, Nuclear Instrumentation Powc r Range Channels N41, N42, N43, and N44, Revision 1, 03-27-79 Step 6.6 was revised to clarify that the appropriate P10 status lights are illuminated on the trip status panel on C04 because the permissive status lights require a 2/4 logic.
(Permanent) 3.5.182 ICP 4.1 (Unit 1), Calibration Procedure, Reactor Protection and Safeguards Analog Racks, Revision 1, 10-03-77 (performed 11-06-79)
Changes were made to the data sheets to reflect existing zero resistance values for temperature channels and to include testpoint locations for steam generator flow channels.
Calibration data for additional indication instrumentation was also included.
(Temporary) 3.5.183 ICP 4.1 (Unit 1), Calibration Procedure, Reactor Protection and Safeguards Analog Racks, Revision 2, 11-20-79 Changes were made to the data sheets to reflect existing zero resistance values for temperature channels and to include testpoint locations for steam generator flow channels.
Calibration data for additional indication instrumentation was also included.
(Permanent) 3.5.184 ICP 4.2, Calibration Procedure, Reactor Coolant Flow Trans-mitters, Revision 2, 04-04-79 A review of procedural documentation disclosed that the approvals for Revision 1 could not be found.
Therefore, this procedure was resubmitted for approval.
In addition, in Step 5.2.2, the calibration points checked in this procedure were standardized by the inclusion of formatted data sheet.
(Permanent)
~56-
3.5.185 ICP 4.2, Calibration Procedure, Reactor Coolant Flow Transmitters, Revision 2, 04-04-79 performed 10-16-79, Unit 1 The zero reading for all flow transmitters was readjusted sc indicated reactor coolant flow is 100% at lower temperature operation. This was required because coolant density change causes the indication to read high.
(Temporary) 3.5.186 ICP 4.3, Calibration Procedure, Pressurizer Level Transmitters, Revision 4, 10-20-78 (performed 10-08-79, Unit 1)
In Step 6.3 the terminology used was corrected to indicate opening of the sealed leg bellows vent cap.
(Temporary) 3.5.187 ICP 4.3, Calibration Procedure, Pressurizer Level Transmitters, Revision 5, 11-20-79 In Step 6.3 the terminology used was corrected indicate opening of the sealed leg bellows vent cap.
(Permanent) 3.5.188 ICP 4.4, Calibration Procedure, Pressurizer Pressure Transmitters, Revision 0, 08-17-72 performed 10-09-79 Unit 1)
Step 2.1.e was added to perform calibration of PT-420, a new instrument, and a corresponding data sheet was added to the procedure.
(Temporary) 3.5.189 ICP 4.4, Calibration Procedure, Pressurizer Pressure Transmitters, Revision 1, 11-20-79 Step 2.1.e was added to perform calibration of PT-420, a new instrument, and a corresponding data sheet was added to the procedure.
(Permanent) 3.5.190 ICP 4.13, Calibration Procedure, Accumulator Level and Pressure Rack Calibration, Revision 2, 06-12-79 The data sheet for channel L934 was corrected to reflect the proper terminals for LC-934A and LC-934B.
(Permanent) 3.5.191 ICP 4.14, Calibration Procedure, Boric Acid Control System, Revision 2, 04-13-78 (performed 10-22-79, Unit 1)
Incorrect tolerances listed on the data sheets were corrected.
Incorrect units of measure for ourput were also corrected to be consistent with the the specified tolerances.
(Temporary) 3.5.192 ICP 4.14, Calibration Procedure, Boric Acid Control System, Revision 3, 11-20-79 Incorrect tolerances listed on the data sheets were corrected.
Incorrect units of measure for ourput were also corrected to be consistent with the the specified tolerances.
(Permanent) 3.5.193 ICP 4.16, Calibration Procedure, Boric Acid Tank Level Calibration, Revision 2, 06-12-79 In Steps 2.1.4, 2.1.5 and 2.1.6, and in Step 6.1, asterisks were added for L-171, L-102 and L-189, respectively to check the footnote.
A footnote was added at the bottom of Page 1 to emphasize that the channels are calibrated only during Unit I refueling calibrations.
The data sheets for these channels were similarly revised because the three channels are only for Unit 1.
(Permanent) 3.5.194 ICP 4.18, Calibration Procedure, Volume Control Tank Rack Calibration, Revision 4, 06-14-80 New Steps 6.1, 6.2, 6.3, and 6.14, 6.15, 6.16 were added to prevent unwanted control action while calibrating channels 112 and 141.
(Permanent) 3.5.195 ICP 4.21, Calibration Procedure, Charging Flow, Revision 0, 09-18-72 (performed 10-22-79, Unit 1)
In Step 5.9 calibration of the control room meter as well as the charging pump control station was added.
The data sheet was likewise updated.
(Temporary) 3.5.196 ICP 4.21, Calibration Procedure, Charging Flow, Revision 1, 11-10-79 In Step 5.9 calibration of the control room meter as well as the charging pump control station was added.
The data sheet was likewise updated.
(Temporary) 3.5.197 ICP 4.23, Calibration Procedure - Radiation Monitoring System (Electronic), Revision 2, 02-19-79 Step 5.2 was expanded to indicate which channels will cause control functions upon actuation of the high alarm. Step 4.1 was revised to delineate calibration just prior to a refueling shutdown versus cold shutdown.
Step 6.3 was added to complete Appendix C for channel R20 only (Unit 1 only).
A new Appendix C was added to calibrate channel R20 for Unit 1.
Step 10.0 was added to Appendix A to notify Health Physics so source calibration of an instrument can be conducted subsequent to the electrical calibration. New Step 2.4 of Appendix B is the same as Step 10.0 for Appendix A.
(Permanent) 3.5.198 ICP 4.23, Calibration Procedure, Radiation Monitoring System (Electronic
, Revision 2, 02-19-79 performed 10-02-79, Unit 1)
Steps 5.2, 6.1, 6.2, and 6.3 concerning channel control functions were modified to reference more detailed information which is contained in the attachments to the procedure. (Temporary) 3.5.199 ICP 4.23, Calibration Procedure, Radiation Monitoring System (Electronic
, Revision 3, 11-20-79 Steps 5.2, 6.1, 6.2, and 6.3 concerning channel control functions were modified to reference more detailed information which is contained in the attachments to the procedure. (Permanent) 3.5.200 ICP 4.25, Calibration Procedure, Nuclear Instrumentation Intermediate Range Channels, Revision 4, 01-23-79 (performed 10-10-79, Unit 1)
In Step 6.7.3 the high level rod stop and high level trip voltages were currected to reflect previously approved setpoint changes.
(Temporary) 3.5.201 ICP 4.25, Calibration Procedure, Nuclear Instrumentation Intermediate Range Channels, Revision 5, 11-20-79 In Step 6.7.3 the high level rod stop and high level trip voltages were corrected to reflect previously approved setpoint changes.
(Permanent) 3.5.202 ICP 4.27, Calibration Procedure, Nuclear Instrumentation Auxiliary Channels, Revision 0, 09-20-72 (performed 10-14-79, Unit 1)
A new Step 6.2, with accompanying data sheets, was added to calibrate the following NIS recorders:
NR45, overpower recorders N41/43 and N42/44, and delta flux recorders N41 through 44.
(Temporary) 3.5.203 ICP 4.27, Calibration Procedure, Nuclear Instrumentation Auxiliary Channels, Revision 1, 11-20-79 A new Step 6.2, with accompanying data sheets, was added to calibrate the following NIS recorders:
NR45, overpower recorders N41/43 and N42/44, and c'. elta flux recorders N41 through 44.
(Permanent) 3.5.204 ICP 4.28, Calibration Procedure, Independent Overspeed Protection System, Revision 0, 09-19-72 (performed 10-27-79, Unit 1)
New Precaution 5.2 was added to open the operating unit's sliders before testing to prevent an operating unit trip caused by noise from the DC power supply.
(Temporary) 3.5.205 ICP 4.28, Calibration Procedure, Independent Overspeed Protection System, Revision 1, 11-20-79 New Precaution 5.2 was added to open the operating unit's sliders before testing to prevent an operating unit trip caused by noise from the DC power supply.
(Permanent) 3.5.206 ICP 4.29, Calibration Procedure, Analog Rod Position, Revision 0, 09-19-72 performed 10-26-79, Unit 1)
RPI data sheets were added which contain standardized calibration points.
(Temporary) 3.5.207 ICP 4.29, Calibration Procedure, Analog Rod Position, Revision 1, 11-20-79 RPI data sheets were added which contain standardized calibration points. (Permanent) 3.5.208 ICP 6.24, Emergency Diesel Calibration Procedure, Revision 2, 04-05-79 The procedure was reclassified to a " major" procedure.
In Section 5.1 preliminary steps were added for clarification.
Throughout the test steps were added to provide for the calibration of switches and alarms, which although are not Technical Specification-related, can be conveniently calibrated at that time.
(Permanent) 3.5.209 ICP 6.24, Emergency Diesel Calibration Procedure, Revision 2, 04-05-79 (performed 06-15-79)
A new Step 5.10.1 was added to jumper terminals 6 and 2 of the LWD Agastat relay in order to calibrate the raw water switch.
Step
" 10.4 was added to remove the jumper installed in 5.10.1.
Other existing steps in the section were renumbered accordingly.
(Temporary) 3.5.210 ICP 6.24, Emergency Diesel Calibration Procedure, Revision 3, 06-22-79 A new Step 5.10.1 was added to jumper terminals 6 and 2 of the LWD Agastat relay in order to calibrate the raw water switch.
Step 5.10.4 was added to remove the jumper installed in 5.10.1 Other existing steps in the section were renumbered accordingly.
(Permanent) 3.5.211 TS-1 (Unit 1), Emergency Diesel Generator 3D, Revision 7, 09-30-77 (performed 08-27-79)
A special test was performed after 2A52-73 4160 V breaker repairs.
The one hour load test was waived to facilitate red tagging 4D out-of-service for its annual inspection and maintenance.
3D had undergone a full load test earlier in the same day.
(Temporary) 3.5.212 TS-3 (Unit 1), Main Turbine Stop and Governor Valves with Turbine Trip Test, Revision 7, 12-10-79 In Step 7.2 a data space was added to indicate DC oil pump start pressure for the redundant automatic start pressure switch per completed modification request E-181.
(Permanent) 3.5.213 TS-10 (Unit 2), Local Leak Test of Personnel Hatches, Revision 1, 09-30-77 (performed 04-07-79 Testing of the upper hatch (66' level) was not required at this time and therefore was omitted.
The lower hatch was tested after it was opened by Maintenance personnel.
(Temporary) 3.5.214 TS-10 (Unit 2), Local Leak Test of Personnel Hatches, Revision 1, 09-30-77 (performed 06-06-79)
On Page 3 data sheet for the hatch located on the 26' level, the acceptance criteria was changed within conservative limits for total containment leakage.
(Temporary) 3.5.215 TS-13 (Unit 2), Monthly Fire Equipment Surveillance, Unit 2 Turbine Room, Revision 4, 06-01-79 On Page 6, Step 46 FP-237 (shut) was added per completion of a modification request.
(Permanent) 3.5.216 TS-14, Monthly Fire Equipment Surveillance, Outside Areas, Maintenance Portable Welding Carts, Revision 3, 06-01-79 In Step 5, items d, e and h; Step 5.1, and Step 6.a and b were deleted because the extinguishers at these locations were removed.
(Permanent) 3.5.217 TS-14, Monthly Fire Equipment Surveillance Outside Areas, Maintenance and Portable Welding Carts, Revision 4, 08-28-79 The procedure was revised to incorporate addition of new fire fighting equipment installed and to correct existing tag numbers of equipment renumbered.
(Permanent) 3.5.218 TS-15 (Unit 1), Monthly Fire Equipment Surveillance, Unit 1 Turbine Room and Service Building, Revision 4, 06-01-79 A typographical error was corrected in Step 35 and in Step 41 FP-236 (shut) was added because modification request M-526 was completed.
(Permanent) 3.5.219 TS-15, Monthly Fire Equipment Surveillance, Unit 1 Turbine Room and Service Building, Revision 5, 08-28-79 The test as revised to reflect the addition of new equipment in the subject areas.
(Permanent) 3.5.220 TS-18, Annual Pressure Testing of Fire Hose, Revision 3, 07-19-79 Step 3.1 was deleted to require an annual test of spare hoses being stored in the warehouse.
The hoses shall be tested on the date of their release into normal service.
(Permanent) 3.5.221 TS-23, Monthly Valve Lineup Verification, Revision 1, 06-01-79 PIV-198 was changed to 238; 199 to 239; 200 to 240; and 201 to 241 in accordance with completed modification request M-535.
(Permanent) 3.5.222 TS-24, Annual Cycling of Flow Path Valves, Revision 1, 07-18-79 Four valve numbers on the test were corrected to agree with the valve numbers contained on the drawing.
(Permanent) 3.5.223 TS-26, Fire Main Flow Test, Revision 2, 07-13-79 Initial Condition 3.3 was added to ensure completion of TS-24 prior to executing TS-26.
In Section 6.0, hydrant #23 was corrected to be hydrant #24.
(Permanent) 3.5.224 TS-27, Sampling of Emergency Fuel Oil Tank, Revision 2, 04-05-79 Step 3.1 was revised to indicate the proper tools required for the job.
In Step 3.2 the replacement gasket type was specified, and in Step 5.7 the requirement to replace the sample connection cap was added to ensure tank closure.
(Permanent) 3.5.225 TS-27, Sampling of Emergency Fuel Oil Tank, Revision 3, 08-21-79 Steps 3.4 through 3.6 were modified to increase the volume of the sample taken per the request of the WE Laboratory Services Group.
(Permanent) 3.5.226 TS-27, Sampling of Emergency Fuel Oil Tank, Revision 4, 10-05-79 Step 3.2 was corrected to designate the correct size of Flexitallic gasket.
Steps 3.5 and 3.6 were corrected to administratively note the packaging and labeling requirements.
(Permanent) 3.5.227 TS-28, Monthly Auxiliary Building Crane Interlock Checks, Revision 0, 05-21-79 This major test replaces PC-26 per new Technical Specification requirements regarding load limits over the spent fuel pit because of spent fuel pool modifications.
(Permanent) 3.5.228 HP 2.7, Radiation Work Permits, Revision 2, 02-14-79 Changes were made to reflect both current practices with respect to RWP's and to detail changes designed to make the use of RWP's more effective in providing radiological safety.
Specific areas outlined were:
(1) ALARA, (2)
Automatic and supervisory termination of the RWP, (3)
Details cf jobs not requiring issuance of RWP's, (4)
Reference to the relationship to the Emergency Plan, (5)
Relationship to the " exposure leveling" policy, (6) Duty Shift Supervisor input to RWP 's, and (7) Revision of the copy distribution.
(Permanent) 3.5.229 HP 13.1.1, Monthly operational Test of the Radiation Monitoring System (Area and Process Subsystems), Revision 3, 07-15-79 The content of the procedure was not changed; however, a new procedure numbering system was established; changing numerical identification.
(Permanent) 3.5.230 HP 13.5.2, Monthly Check of Emergency Monitoring Equipment Located at the Site Boundary Control Center, Revision 2, 07-15-79 In Section 9.0, the title of " Health Physics Supervisor" was inserted in place of " Health Physics Foreman" to conform to the new designation established by the Company. The number of the procedure was changed to reflect the new identification system established.
(Permanent) 3.5.231 PC-26, Auxiliary Building Crane Interlock Checks, Revision 0, 09-30-77 This minor procedure was canceled because it was being replaced with a major Technical Specification-required test.
(Canceled) 3.5.232 PC-41 (Part V), Weekly Swing Periodic Checks, Revision 0, 12-20-77 This part of the minor procedure was canceled because the checks are accomplished by the surveillance computer system.
(Canceled) 3.5.233 WMTP1 4.5, Natural Circulation Test, Cycle 8, Revision 0, 11-27-79 (performed 11-28-79)
Steps 5.9, 5.10 and 5.11 concerning injection of boric acid while monitoring reactivity were omitted because of time limitations.
Performance of the steps was not necessary to properly conduct the test.
(Temporary) 3.5.234 WMTP1 6.1, Low Power Temperature Defect Measurement and Steam Generator Caustic Elution and Neutralization, Revision 0, 08-17-79 Step 4.1 was modified such that a
primary coolant temperature of 400 F was indicated as being an initial condition for performance of the test per Westinghouse and Chemistry recommendations.
(Temporary) 3.5.235 WMTP2 2.1, Core Reloadinq, Revision 0, 02-15-79 (performed 04-02-79) one plug device did not insert properly and was replaced.
(Temporary) 3.5.236 WMTP 9.1, Rod Control Mechanism Timing, Rod Drop and Rod Position Indication, Revision 6, 08-31-79 Precaution 3.4 was added to ensure that more than eight rods are fully withdrawn at any given time, thus providing administrative controls to minimize reactivity insertion.
Precaution 3.5 was added to not withdraw a bank more than 30 steps to ensure the rods remain in the dashpot since
- problems, if they exist, would be identified at this position.
Steps 5.4.4 and 5.4.5 were changed to eliminate needless starts and stops of the oscillograph.
(Permanent) 3.5.237 WHTP 9.2, Power Range Calibration Quarterly Offset Test, Revision 0, 08-31-79 The procedure was completely rewritten to include acceptance
- criteria, precautions on delta flux
- program, more flexibility in sequence, and changing of the addressable points to hose being used.
(Permanent) 3.5.238 WMTP 11.15, Runout Test of PISA, Unit 2, "A" Safety Injection Pump, Revision 0, 02-01-79 (performed 03-31-79)
A typographical error was corrected in Step 5.1.
(Temporary) 3.5.239 WMTP 11.17, 3D Safeguards Diesel Generator 24-Hour Continuous Load Test, Revision 0, 11-05-79 performed 11-09-79)
Step 1.2 was corrected to indicate the formal reply date of the Wisconsin Electric to the NRC letter of 10-26-79 concerning IE Bulletin No. 79-23.
During preparation of the test procedure a draft reply dated 10-10-79 had been used.
Step 4.26.a was added to the procedure to secure the engine.
The step had inadvertently been omitted during typing of the test.
(Temporary) 3.5.240 WMTP 11.18, 4D Safeguards Diesel Generator 24-Hour Continuous Load Test, Revision 0, 11-05-79 (performed 11-07-79),
Step 1.2 was corrected to indicate the formal reply date of the Wisconsin Electric to the NRC letter of 10-26-79 concerning IE Bulletin No. 79-23.
Durig preparation of the test procedure a draft reply dated 10-10-79 had been used.
Step 4.26.a was added to the procedure to secure the engine.
The step had inadvertently been omitted during typing of the test.
(Temporary) 3.5.241 WMTP 11.19, Steam Generator Crevice Cleaning, Revision 0, 11-05-79 In Step 5.5 additional cooling was desired so 500 gallons of DI water per steam generator were added rather than the listed 300 gallons.
Step 5.6 was deleted because the temperature of bulk liquid was increased above that desired.
Step 5.9 was modified to permit operation of only one reactor coolant pump since at the time of test initiation it was uncertain if "B" pump could be utilized because of No.1 seal problems. This concern was later negated.
(Temporary) 3.5.242 WMTP 12.2, Procedure for Verification and Inspection of Concrete Fasteners in Response to NRC IE Bulletin No. 79-02, Revision 2, 08-02-79 In Step 2.3 the frequency for torque wrench calibrations was reduced from weekly to every two weeks.
In Section 3.0 a note was added to clarify the proper reporting route for resolution of test problems.
Follcwing Step 3.2.2 a note was added to test all rusted threaded rods on grouted plates in accordance with the pull test procedure.
(Permanent) 3.5.243 WMTP 12.3, WE Procedure for Verification and Inspection of Concrete Fasteners in Response to NRC IE Bulletin No. 79-02, Revision 1, 06-07-79 Step 4.2.1 was modified to require calibration before examination of each base plate and to document the calibration.
Step 4.3 was added to specify personnel training requirements.
(Permanent) 3.5.244 WMTP 12.5, Pull Test Procedure for Verification and Inspection of Concrete Fasteners in Response to IE Bulletin No. 79-02, Revision 1, 06-19-79 in Step 3.1.9 the procedure was revised to measure directly by the use of a rule.
(Permanent) 3.5.245 WMTP 12.5, Pull Test Procedure for Verification and Inspection cf Concrete Fasteners in Response to NRC IE Bulletin No. 79-02, Revision 2, 07-17-79 In Step 2.3, the calibration frequency for the torque wrenches was reduced from weekly to once every two weeks; experience having dictated that the reduction could be made without lessening quality assurance aspects of the job.
(Permanent) 3.5.246 WMTP 12.5, Pull Test Proceaure for Verification and Inspection of Concrete Fasteners in Response to NRC IE Bulletin No. 79-02, Revision 3, 08-02-79 A note was added to Section 3.0 to clarify the reporting route for resolution of testing problems.
A note was added to Step 3.2.3 to permit special test conditions for rusted threaded rods on grouted plates.
(Permanent) 3.5.247 Emergency Plan, Revision 10 (serially reviewed and approved on 03-21-79 The changes consisted of revisions to the Emergency Call List (Section 8.)
(Permanent) 3.5.248 Emergency Plan Manual Section 6, Revision 11, 10-22-79 This revision incorporated requirements to notify the NRC within one hour of the initiation of the on-site evacuation plan; implemented the temporary Technical Support and operational Support Centers; and, implemented the dedicated phone line between the Technical Support Center and NRC Operations Headquarters and updates other communications capabilities.
Responsibilities were clarified and required functions were delineated for:
Technical Advisors (when appointed);
contractor personnel; security personnel (include the Security Supervisor); Office Supervisor; and Energy Information Center personnel.
The implementation sections were re formatted to make it easier to find and follow the responsibilities of different individuals.
Responsibilities between the Duty & Call Superintendents, Coordinator and Duty Shift Supervisor were reorganized.
Many of the changes contained in this revision were in response to the Eisenhut letter of 10-10-79 to all licensees which requires submittal of the Emergency Plan by 01/01/80 for NRC review.
(Permanent) 3.5.249 Emergency Plan, Revision 12, 12-17-79 Reporting of the on-duty operating shift personnel to the control room was clarified; actions of a second Duty & Call Superintendent if on-site were delineated to procedurally ensure his reporting to the site boundary control center; contacting of the "on-call" Technical Advisor by the Coordinator was clarified; and assigned responsibilities of security force personnel were more clearly delineated.
(Permanent) 3.5.250 FEP 2.1, Control Room Fire, Revision 1, 11/05/79.
The procedure was revised to indicate hose reel numbers for specific equipment which has now been installed.
In the previous issue of the procedure asterisks had been used to denote hose reels which had not yet been installed. Specific door numbers were also added to the procedure.
(Permanent) 3.5.251 FEP 2.2, Cable Spreading Room Fire, Revision 1, 11/05/79.
The changes noted above for FEP 2.1, Revision 1, apply.
(Permanent) 3.5.252 FEP 2.3, Vital Switchgear Room Fire, Revision 1, 11/05/79.
The changes noted above for FEP 2.1, Revision 1, apply.
(Permanent) 3.5.253 FEP 2.4, Diesel Generator Room Fire, Revision 1, 11/05/79.
The changes noted above for FEP 2.1, Revision 1, apply.
(Permanent) m 4.0 NUMBER OF PERSONNEL AND MAN-REM BY WORK GROUP AND JOB FUNCTION 4.1 1978 (Corrected)
JOB FUNCTION NUMBER TOTAL REM REACTOR PERSONNEL PER WORK CPERATIONS &
ROUTINE INSPECTION SPECIAL WASTE WORK CROUP
>100 r. Rem CROUP SURVEILLANCE MAINTENANCE ACTIVITIES MAINTENANCE PROCESSING REFLTLING 1.
t wpany Employees Ope rations 43*
77.196 44.651 0.141 8.428 0.895 17.942 5.139 Peak Maintenance 7
and Maintenance 76 92.312 0.49' 37.579 27.718 26.054 0.326 0.147 Chemistry and Health Physics 24*
23.969 21.418 0.318 0.023 0.0 1.677 0.533 Reactor Engineering 3
3.524 1.920 0.0 1.285 0.0 0.0 0.,19 Instrument and Control 7*
3.638 0.0 2.391 1.053 0.194 0.0 0.0 Admir.istration 3
1.103 0.0 0.0
- 1. ? O 3 0.0 0.0 0.0 2.
Contract Workers and Others 128 107.026 0.0 0.0 77.496 29.530 0.0 0.0 TOTALS 284 308.768 68.477 40.429 117.106 56.673 19.945 6.138
- Intercompany transfers people counted only once, not as one person in each group.
e
6 4.0 NUMBER OF PERSONNEL AND MAN-REM BY WORK GROUP AND JOB FLi1CTION 4.2 1979 JOB Pu?iCTICN NUMBER TOTAL RIM REACTOR PE RSONNEL PER WORK OPERATIONS &
ROtfilNE INSPECTION SPECIAL WASTE WrX GROUP
>100 mrem GROUP SURVEILLANCE MAINTENANCE ACTIVITIES MAINTENANCE PROCESSING REFUELING 1.
Company Employees Operations 43*
56.103 35.090 0.260 7.429 0.943 9.641 2.740 Peak Maintenance and Maintenance 77*
91.576 2.012 37.981 25.514 24.725 0.523 0.021 Chemistry and Health Physics 23*
22.906 21.464 0.0 0.017 0.0 1.425 0.0 Reactor Engineering 4*
6.178 2.102 0.0 3.555 0.0 0.0 0.521 Instrument and Control C*
5.499 0.883 3.100 1.177 0.162 0.015 0.162 Administration 3*
4.271 0.0 0.0 4.271 0.0 0.0 0.0 2.
Contract Workers and Others 356 427.774 0.0 0.0 200.493 227.281 0.0 0.0 TOTALS 514 614.307 62.351 41.341 242.456 253.111 11.604 3.444
- Inter-nlant group transfers counted only once, not as one person in each group.
5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION 5.1 Unit 1 03-14-79-Outac3 On 03-14-79 the unit was shutdown with a primary-to-secondary leak rate of 255 gpd.
A static leak test of the "A"
steam generator identified two leaking tubes, which were plugged.
A subsequent 800 psi leak tesc identified an additional six leaking tubes and one wet plug.
A 115 psi leak test of the "B" steam generator disclosed one leaking tube.
All the leaking tubes were explosively plugged; however, the leaking plug was not repaired at the time because the leak was extremely minor.
Hand probe eddy current inspection identified defects to be within the tubesheet region.
The unit was returned to line on 03-19-79.
Tubes Plugged "A" Steam Generator "B" Steam Generator R27C49 R22C37 R06C49 R27C44 R24C35 R25C27 R20C28 R20C24 R25C29 Leaking Plugs "A" Steam Generator "B" Steam Generator R31C44 08-05-79 Outage on 08-05-79 approximately two hours after returning to line from a shutdown for strengthening of the auxiliary feed line to the main feed line branch connection saddles, the unit was taken off line with a 1.45 gpm (2,090 gpd) primary-to-secondary leak in the "A"
The leaking tube was visually identified, under static head, as R17C46 in the "A"
steam generator inlet.
Another tube, R17C44, in the "A" inlet was also found to be dripping.
The two leaking tubes were mechanically plugged and a 100% eddy current inspection of both steam generators was performed.
All defective tubes were explosively plugged except for four tubes, including the two leakers, which were mechanically plugged in anticipation of tube pulling during the Refueling 7 outage; i.
e., mechanical plugs being more easily removable.
Final results e f the program disclosed 52 defects in the "A"
steam generator inlet and 45 defects in the "B"
inlet.
All defects were below the top of the tubesheet in the crevice area.
Successful 800 psi hydrostatic leak checks were performed on the steam generators prior to returning the unit to service.
Extent of Eddy Current Inspection "A" SG Inlet 2,944 tubes tested at 400 KHZ through first upport 205 tubes tested at 400 KHZ through U-bend 21 tubes tested at 100 KHZ through first support Outlet 104 tubes tested at 400 KHZ through first support "B" SG Inlet 2,862 tubes tested at 400 KHZ through first support 118 tubes tested at 400 KHZ through U-bend Outlet No tubes were inspected.
"A" Steam Generator Inlet Row Column
% Defect Origin Location Plugged 17 44 100 0.D.
21" above tube end Yes 17 46 100 0.D.
21" above tube end Yes 17 42 95 0.D.
21" above tube end Yes 20 26 95 0.D.
21" above tube end Yes 24 38 87 0.D.
10" above tube end Yes 23 35 84 0.D.
20" above tube end Yes 22 35 76 0.D.
12" above tube end Yes 15 48 85 0.D.
20" above tube end Yes 17 48 91 0.D.
18" above tube end Yes 27 54 90 0.D.
6" above tube end Yes 24 39 87 0.D.
8" above tube end Yes 24 34 89 0.D.
8" above tube end Yes 10 18 69 0.D.
18" above tube end Yes 11 20 90 0.D.
6" above tube end Yes 9
25 96 0.D.
18" above tube end Yes 26 25 85 0.D.
12" above tube end Yes 9
26 89 0.D.
10-18" above tube end Yes 9
27 92 0.D.
14" above tube end Yes 20 27 67 0.D.
20" above tube end Yes 23 30 90 0.D.
6-10" above tube end Yes 21 32 90 0.D.
4" above tube end Yes 27 36 83 0.D.
8" above tube end Yes 25 36 82 0.D.
6-8" above tube end Yes 6
36 73-86 0.D.
10" above tube end Yes 26 37 82 0.D.
8" above tube end Yes 27 41 83 0.D.
4" above tube end Yes 25 38 85 0.D.
10" above tube end Yes 25 42 79 0.D.
8" above tube end les 11 44 76 0.D.
21" above tube end Yes 27 45 81 0.D.
21" above tube end Yes 27 46 78 0.D.
18" above tube end Yes 27 48 89 0.D.
12" above tube end Yes 6
48 87 0.D.
14-16" above tube end Yes 11 49 84 0.D.
14" above tube end Yes 27 50 88 0.D.
14-16" above tube end Yes "A"
iteam Generator Inlet, continued..
Row Column
% Defect Origin Location l'1,tgged 15 50 92 0.D.
21" above tube end Yes 6
51 94 0.D.
18" above tube end Yes 27 51 87 0.D.
20" above tube end Yes 13 52 85 0.D.
20" above tube end Yes 8
52 91 0.D.
21" above tube end Yes 6
52 93 0.D.
12-20" above tube end Yes 27 53 87 0.D.
8" above tube end Yes 5
54 93 0.D.
12-16" above tube end Yes 5
57 89 0.D.
14" above tube end Yes 5
55 89 0.D.
14-18" above tube end Yes 13 60 85 0.D.
20" above tube end Yes 12 60 79 0.D.
20" above tube end Yes 3
67 83 0.D.
10" above tube end Yes 5
49 90 0.D.
18" above tube end Yes 6
50 86 0.D.
18" above tube end Yes 9
52 90 0.D.
21" above tube end Yes 9
53 92 0.D.
21" above tube end Yes 6
81
<20 0.D.
2" above tubesheet No 7
44 31 0.D.
8" above tubesheet No 23 37
<20 0.D.
4" above tubesheet No 23 38
<20 0.D.
4" above tubesheet No 12 29
<20 0.D.
16" above tube end No 14 43
<20 0.D.
2" above tubesheet No 14 44
<20 0.D.
2" above tubesheet No 18 42
<20 0.D.
6" above tubesheet No 23 38
<20 0.D.
2" above tubesheet No 26 20
<20 0.D.
2" above tubesheet No 33 54
<20 0.D.
4" above tubesheet No "A" Steam Generator Outlet No defects detected.
"B" Steam Generator Inlet Row Column
% Defect Origin Location Plugged 31 40 84 0.D.
6" above tube end Yes 22 43 94 0.D.
6" above tube end Yes 23 46 77 0.D.
14" above tube end Yes 23 48 82 0.D.
8-10" above tube end Yes 6
57 94 0.D.
6-8" above tube end Yes 11 14 83 0.D.
6" above tube end Yes 22 40 73 0.D.
14" above tube end Yes 7
79 92 0.D.
4" above tube end Yes 22 49 80 0.D.
8-10" above tube end Yes 24 34 93 0.D.
4" above tube end Yes 21 38 79 0.D.
18" above tube end Yes 23 42 68 0.D.
10" above tube end Yes 11 27 87 0.D.
21" above tube end Yes 11 28 91 0.D.
21" above tube end Yes "B" Steam Generator Inlet, continued..
Row Column
% Defect Oricin Location Plugged 22 32 87 0.D.
10" above tube end Yes 5
33 85 0.D.
14-16" above tube end Yes 23 34 92 0.D.
12-16" cbove tube end Yes 23 36 82 0.D.
6-8" above tube end Yes 8
36 75 0.D.
12" above tuoc end Yes 21 37 86 0.D.
8-20" above tubc end Yes 23 37 79-88 0.D.
8-12" above tube end Yes 23 33 93 0.D.
12" above tube end Yes 26 39 92 0.D.
16" above tube end Yes 25 40 68 0.D.
4" above tube end Yes 21 40 80-87 0.D.
10-18" above tube end Yes 8
40 96 0.D.
12" above tube end Yes 6
40 81 0.D.
12" above tube end Yes 4
40 95 0.D.
10" above tube end Yes 4
41 86 0.D.
10" above tube end Yes 21 41 66 0.D.
16" above tube end Yes 22 41 85 0.D.
4-6" above tube end Yes 14 42 88 0.D.
21" above tube end Yes 13 42 87 0.D.
21" above tube end Yes 22 47 91 0.D.
6-10" above tube end Yes 25 47 86 0.D.
8" above tube end Yes 21 42 74 0.D.
12-14" above tube end Yes 29 49 82 0.D.
8" above tube end Yes 14 50 78-88 0.D.
12-21" above tube end Yes 14 51 88 0.D.
21" above tube end Yes 22 51 74 0.D.
14" above tube end Yes 22 62 88 0.D.
10" above tube end Yes 28 65 91 0.D.
4" above tube end Yes 18 67 81 0.D.
6" above tube end Yes 22 52 91 0.D.
12" above tube end Yes 26 44 90 0.D.
16" above tube end Yes "B" Steam Generator Outlet No eddy current examinations were performed.
08-29-79 Outage s
On August 29,
- 1979, the unit was shutdown with an estimated primary-to-secondary leak rate of 324 gpd. The shutdown occurred 11 days after the unit returned to line following the previous steam generator repair shutdown.
A leak test disclosed tube R20C37 in the "A"
steam generator inlet as the leaker.
A reanalysis of eddy current testing data taken during the previous outage revealed indications of an 88% defect located 20 inches above the tube end buried in a noisy signal.
This tube was mechanically plugged along with another tube at R7C44 with a 31% defect located eight inches above the tube end in the "A" steam generator inlet; the R7C44 tube was plugged as a conservative measure.
No eddy current examination was conducted during the outage.
Refueling 7 Inservice Inspecticn The initial inspection program included only 38 tubes in the inlet of each generator; a post outage review of the August inspection data having raised questions on these tubes.
The progiam was later expanded to a 100% inspection to the first support, however, because of the discovery of additional defects.
A total of 77 tubes were eventually explosively plugged in the "A" steam generator, including 69 tubes with defects greater than 40%; three tubes with defects less than 40%; two good tubes plugged by mistahe; one leaking tube; and two good tubes which were removed for examination.
All except two defects were located in the tubesheet crevice region.
A total of three tubes were removed for examination; one tube with an identified defect; one tube with no identified defect located in the area where deep crevice cracking has been found; and one tube with no identified defect located outside the area where deep crevice cracking has been found.
These tubes were removed in order to perform metallurgical and structural evaluations to learn more about the deep crevice corrosion problem.
The evaluation showed the tubes to be physically acceptable, with even the defective tube exhibiting an acceptable burst pressure of over 5000 psi. The inlet ends of the removed tubes were weld plugged, the cold leg ends being explosively plugged.
A total of 68 tubes were explosively plugged in the "B"
steam generator including 62 tubes with defects greater than 40%; three tubes with defects less than 40%; one good tube plugged by mistake; and two leaking tubes.
All except three defects were located in the tubesheet crevice region.
In addition, four leaking plugs were weld repaired in the "B"
steam generator inlet.
An 800 psi hydrostatic leak check was performed both prior to and after explosive tube plugging in both steam generators.
It should be noted that a review of the magnetic tapes of eddy current inspection of the five tubes with indications at the top of the tubesheet indicate that small signals, undefinable as defects, were evident back to inspections conducted in 1975.
The latest inspection utilizing multi-frequency eddy current techniques rather than the previous 400 KH7 single frequency technique was able to resolve these signals as small volume defects of long standing. These
~
five defects are not considered to have any relationship to the mechanism responsible for deep crevice corrosion and cracking.
Measures taken to inhibit the deep crevice corrosion, include performance of crevice boilout prior to plant startup in an attempt to flush corrosive materials from the crevice, and later plant operation at a reduced core average temperature and power.
In addition, primary system pressure was reduced to 2000 psig to keep Ap across the steam generator tubes to a minimum.
Extent of Eddy Current Inspection "A" SG Inlet 3,009 tubes inspected at 400 KHZ through first support 76 tubes inspected at 100 KHZ through U-bend Outlet No tubes were inspected.
"A" SG Inlet 3,009 tubes inspected at 400 KHZ through first support.
Outlet No tubes were inspected.
"A" Steam Generator Inlet Row Column
% Defect Origin Location Plugged 18 24 84 0.D.
18" above tube end Yes 24 32 91 0.D.
5" above tube end Yes 9
24 93 0.D.
11" above tube end Yes 10 41 94 0.D.
18" above tube end Yes 10 42 95 0.D.
21" above tube end Yes 22 41 87 0.D.
21", 15" and 5" above tube end Yes 15 45 89 0.D.
20" above tube end Yes 23 62 89 0.D.
7" above tube end Yes 19 54 83 0.D.
20" above tube end Yes 19 53 92 0.D.
20" above tube end Yes 12 54 92 0.D.
20" above tube end Yes 12 55 93 0.D.
18" above tube end Yes 8
53 93 0.D.
20" above tube end Yec 12 53 71 0.D.
17" above tube end Yes 13 53 75 0.D.
20" above tube end Yes 9
51 89 0.D.
20" above tube end Yes 13 50 95 0.D.
20" above tube end Yes 5
50 90 0.D.
11" above tube end Yes 13 49 92 0.D.
20" above tube end Yes 15 49 88 0.D.
21" above tube end Yes 19 49 91 0.D.
20" above tube end Yes 15 47 92 0.D.
20" above tube end Yes 10 46 96 0.D.
20" above tube end Yes 10 45 93 0.D.
20" above tube end Yes 17 45 92 0.D.
20" above tube end Yes 19 45 90-81 0.D.
20", 21" above tube end Yes 15 44 97 0.D.
21" above tube end Yes 12 44 84 0.D.
12" above tube end Yes 23 41 80 0.D.
6" above tube end Yes 27 40 81 0.D.
20" above tube end Yes 24 40 93 0.D.
4" to 12" above tube end Yes 23 39 65 0.D.
5" above tube end Yes 27 39 59 0.D.
21" above tube end Yes 23 37 87 0.D.
10" above tube end Yes 24 36 82 0.D.
6" above tube end Yes 22 36 76 0.D.
11" above tube end Yes 27 35 90 0.D.
21" above tube end Yes 23 33 89 0.D.
3-8" above tube end Yes 20 32 42 0.D.
20" above tube end Yes 8
43 82 0.D.
15" above tube end Yes 11 37 88 0.D.
17" above tube end Yes 3
36 91 0.D.
10" above tube end Yes 13 26 95 0.D.
16" above tube end Yes 11 26 92 0.D.
19" above tube end Yes 20 25 92 0.D.
17" above tube end Yes 21 45 79 0.D.
21" above tube end Yes 22 46 55 0.D.
Top of tubesheet Yes "A" Steam Generator Inlet, continued Row Column
% Defect Origin Location Plugged 29 46 84 0.D.
21" above tube end Yes 20 40 99 0.D.
19-21" above tube end Yes 30 53 65 0.D.
14" above tube end Yes 17 53 73 0.D.
15" above tube end Yes 15 51 81 0.D.
20" above tube end Yes 5
56 87 0.D.
10" above tube end Yes 30 57 80 0.D.
Top of tubesheet Yes 17 57 76 0.D.
8" above tube end Yes 9
57 71 0.D.
20" above tube end Yes 15 58 85 0.D.
18" above tube end Yes 5
59 97 0.D.
10" above tube end Yes 15 61 73 0.D.
17" above tube end Yes 1.,
61 94 0.D.
19" above tube end Yes 13 62 81 0.D.
21" above tube end Yes 14 62 95 0.D.
15" above tube end Yes 13 63 96 0.D.
21" above tube end Yes 12 63 76 0.D.
21" above tube end Yes 13 64 91 0.D.
17" above tube end Yes 23 64 72 0.D.
19" above tube end Yes 13 65 92 0.D.
19" above tube end Yes 5
76 85 0.D.
8" above tube end Yes 15 62 94 0.D.
20" above tube end Yes 19 38 34 0.D.
8" above tube end Yes 28 25 26 0.D.
14" above tube end Yes 24 54 37 0.D.
21" above tube end Yes 15 39
<20 0.D.
5" above tubesheet No 25 43
<20 0.D.
2" above tubesheet No 26 27
<20 0.D.
3" above tubesheet No 26 43
<20 0.D.
2" above tubesheet No 26 45
<20 0.D.
2" above tubesheet No 27 43
<20 0.D.
2" above tubesheet No 28 44
<20 0.D.
2" above tubesheet No Tubes Plugged by Mistake "A" Steam Generator R24C30 R23C63 Tube Plugged - Leaker R29C25 "A" Steam Generator Outlet No eddy current examinations were performed.
"B" Steam Generator Inlet Row Column
% Defect Origin Location Plugged 14 61 86 0.D.
20" above tube end Yes 14 48 86 0.D.
21" above tube end Yes 23 43 63 0.D.
16" above tube end Yes 24 30 90 0.D.
8",
14", 16" above tube end Yes 6
39 90 0.D.
11" above tube end Yes 6
35 87 0.D.
12" above tube end Yes 22 31 90 0.D.
5" above tube end Yes 20 32 83 0.D.
17" above tube end Yes 23 32 94 0.D.
10" above tube end Yes 22 33 78 0.D.
20" above tube end Yes 25 38 90 0.D.
6" above tube end Yes 23 41 87 0.D.
3" to 11" above tube end Yes 12 43 91 0.D.
21" above tube end Yes 24 45 84 0.D.
6" above tube end Yes 23 45 73 0.D.
6" to 8" above tube end Yes 22 45 92 0.D.
15" abcVe tube end Yes 14 46 93 0.D.
21" above tube end Yes 24 46 75 0.D.
13" above tube end Yes 26 54 93 0.D.
20" above tube end Yes 9
16 83 0.D.
3" to 8" above tube end Yes 11 18 83%, 92%
0.D.
15" and 18" above tube end Yes 4
19 77 0.D.
8" above tube end Yes 11 24 73 0.D.
20" above tube end Yes 17 27 78 0.D.
11" above tube end Yes 12 27 72 0.D.
20" above tube end Yes 6
27 64 0.D.
20" above tube end Yes 11 29 85 0.D.
21" above tube end Yes 11 30 92 0.D.
21" above tube end Yes 13 30 53 0.D.
21" above tube end Yes 13 31 92 0.D.
19" above tube end Yes 4
32 76 0.D.
6" above tube end Yes 9
33 86 0.D.
14" above tube end Yes 27 36 95 0.D.
5" above tube end Yes 4
37 90 0.D.
4" above tube end Yes 28 38 45 0.D.
Top of tubesheet Yes s
9 39 62 0.D.
19" above tube end Yes 8
39 95 0.D.
8" to 13" above tube end Yes 32 42 61 0.D.
\\" above tubesheet Yes 29 44 65-50-90 0.D.
20", 10", 5" above tube end Yes 30 44 83 0.D.
Top of tubesheet Yes 31 44 90 0.D.
3" to 13" above tube end Yes 29 45 77 0.D.
J1" above t::be end Yes 10 46 87 0.D.
20" above tube end Yes 27 50 89 0.D.
3" to 15" above tube end Yes 22 56 91 0.D.
10" to 15" above tube end Yes 20 57 82 0.D.
17" above tube end Yes 21 57 83 0.D.
6" above tube end Yes 20 58 93 0.D.
14" to 21" above tube end Yes 3
58 90 0.D.
3" above tube end Yes 6
59 90 0.D.
15" above tube end Yes "A" Steam Generator Inlet continued..
Row Column
% Defect Origin Location Plugged 15 59 84 0.D.
20" above tube end Yes 8
60 89 0.D.
10" above tube end Yes 21 61 94 0.D.
5" to 10" above tube end Yes 23 62 96 0.D.
5" above tube end Yes 13 62 61 0.D.
20" abave tube end Yes 3
62 48 0.D.
15" above tube end Yes 21 64 76 0.D.
7" above tube end Yes 8
65 64 0.D.
12" above tube end Yes 10 66 40 0.D.
13" above tube end Yes 1
68 95 0.D.
11" above tube end Yes 24 71 93 0.D.
5" above tube end Yes 2
77 95 0.D.
3" above tube end Yes 25 36 32 0.D.
21" above tube end Yes 13 39 36 0.D.
15" above tube end Yes 13 59 36 0.D.
20" above tube end Yes Tube Plugged by Mistake "B" Steam Generator R09C32 Tubes Plugged - Leakerr.
R23C47 R20C39 Leaking Plugs Weld Repaired "B" Steam Generator (Inlet)
(RE: These plugs inserted during previous outages.)
R22C41 R23C31 R23C50 R21C26 "B" Steam Generator Outlet No eddy current examinations were performed.
12-11-79 Outtge On 12-11-79, approximately ten days after returning to service after Refueling 7, the unit was shutdown with a primary-to-secondary leak rate of about 250 gpd.
An 800 psi leak check of both steam generators revealed no leakage in the "A"
steam generator and one leaking tube (R2C71) along with one dripping explosive plug and one dripping welded plug in the "B"
The leaking tube was mechanically plugged and the dripping plugs were weld repaired.
Subsequent eddy current inspection of tubes in the area where deep crevice defects have been found, identified 19 defects in the "A"
steam generator and 15 in the "B"
A detailed comparison of the eddy current data collected during Refueling 7, utilizing improved techniques, chowed that 12 of the 19 defects in "A"
and six of the 15 defects in "B"
had had indications not identifiable as defects at the time of the previous inspection.
A total of 20 tubes were mechanically plugged in the "A"
steam generator including 15 tubes with defects greater than 40%; three tubes with defects less than 40%; one tube with an undefineable defect; and one good tube plugged by mistake.
A total of 15 tubes were explosively plugged (except for the leaker which was mechanically plugged) in the "B" steam generator including 11 tubes with defects greater than 40%; one tube with a defect less than 40%;
and three tubes with undefineable indications.
All defects were within the tubesheet crevice region.
A successful 800 psi secondary-to-primary hydrostatic leak check of the "B"
steam generator was performed prior to returning to service.
In addition, a 2000 psig primary-to-secondary hydrostatic test was conducted on both steam generators early in the system heatup phase to give added assurance of tube integrity.
Extent of Eddy Current Inspection "A" SG Inlet 961 tubes tested with multi-frequency through first support Outlet No eddy current inspection performed "B" SG Inlet 862 tubes tested with multi-frequency through first support Outlet No eddy current inspection performed "A" Steam Generator Inlet Row Column
% Defect Origin Location Plugged 8
19 81 0.D.
8" above tube end (new)
Yes 10 20 42 0.D.
20" above tube end Yes 7
21 95 0.D.
13" above tube end Yes 18 21 85 0.D.
15-21" above tube end Yes 8
25 85 0.D.
15-20" above tube end Yes 9
28 92 0.D.
6" above tube end (new)
Yes 20 31 92 0.D.
15-21" above tube end Yes 18 32 92 0.D.
12" above tube end Yes 19 32 60 0.D.
20" above tube end Yes 18 33 88 0.D.
12" above tube end Yes 7
34 88 0.D.
2%-5" above tube end (new)
Yes 23 40 69 0.D.
10" above tube end Yes 12 57 83 0.D.
20" above tube end (new)
Yes 4
59 94 0.D.
17" above tube end Yes 6
33 84 0.D.
5" above tube end (new)
Yes 27 42
<20 0.D.
12" above tube end Yes 24 47
<20 0.D.
8" above tube end (new)
Yes 19 51 37 0.D.
6" above tube end (new)
Yes 9
54 Undefineable indication; not identified as a defect Tube Plugged by Mistake R18C45 "A" Steam Generator Outlet No eddy current examinations were performed.
. ' ~
"B" Steam Generator Inlet Row Column
% D3fect Origin Location Plugged 11 25 74 0.D.
12" above tube end (new)
Yes 18 28 88 0.D.
5" above tube end (new)
Yes 4
38 92 0.D.
5" above tube end (new)
Yes 12 38 94 0.D.
20" above tube end (new)
Yes 13 46 87 0.D.
21" above tube end Yes 26 46 89 0.D.
19" above tube end Yes 25 54 85 0.D.
7" above tube end (new)
Yes 6
60 82 0.D.
9" above tube end (new)
Yes 3
67 91 0.D.
2\\-11" above tube end Yes 3
68 86 0.D.
2b-5" above tube end (new)
Yes 2
71 100 0.D.
Detected leaking tube; 3" Yes above tube end (new) 12 26
<20 0.D.
20" above tube end Yes 13 44 0.D.
Undefineable indication; 19" Yes above tube end 11 47 0.D.
Undefineable indication; Yes 13-20" above tube end (new) 3 66 0.D.
Undefineable indication; Yes 2 -5" above tube end "B" Steam Generator Weld Repair (Hot Leg) 12 55 Previously explosively plugged (1972) and welded (1977).
20 55 Previously explosively plugged (1972).
"B" Steam Generator Outlet No eddy current examinations were performed.
5.2 Unit 2 Refueling 5 Inservice Inspection Only one tube, R4C57, in the "B" steam generator inlet was identified as having a defect greater than 40% during the Refueling 5 inspection. The defect was a 56% defect located three inches above the tubesheet. The tube was explosively plugged.
Extent of Eddy Current Inspection "A" SG Inlet 133 tubes tested at 400 KHZ through U-bend 347 tubes tested at 400 KHZ through sixth support 90 tubes tested at 100 KHZ through U-bend 98 tubes tested at 7 KHZ through sixth support Outlet 19 tubes tested at 400 KHZ through U-bend "B" SG Inlet 119 tubes rested at 400 KHZ through U-bend 245 tubes tested at 400 KHZ through sixth support 91 tubes tected at 100 KHZ through U-bend Outlet 230 tubes tested it 400 KHZ through first support Results of Eddy Current Inspection "A" SG Inlet 0 tubes with indications greater than 39%
6 tubes with indications from 30-39%
5 tubes with indications from 21-29%
21 tubes with indications less than or equal to 20%
Outlet 0 tubes with indications greater than 39%
1 tube wi'h indications from 30-39%
1 tube with indications from 21-29%
14 tubes with indications less than or equal to 20%
"B" SG Inlet 0 tubes with indications greater than 59%
1 tube with indications from 50-59%
0 tubes with indications from 40-49%
3 tubes with indications from 30-39%
3 tubes with indications from 21-29%
10 tubes with indications less than or equal to 20%
Outlet 0 tubes with indications greater than 39%
5 tubes with indications from 30-39%
92 tubes with indications from 21-29%
133 tubes with indications less than or equal to 20%
"A" Steam Generator Inlet Row Column
% Defect Origin Location Plugged 6
30 22 0.D.
Top of tubesheet No 8
32 30 0.D.
Top of tubesheet No 10 44 Copper No 11 57
<20 0.D.
h" above tubesheet No 14 49
<20 0.D.
%" above tubesheet No 16 40 Copper No 16 41
<20 0.D.
%" above tubesheet No 16 46 Copper No 17 31
<20 0.D.
%" above tubesheet No 17 35 Copper No 17 36
<20 0.D.
" above tubesheet No 17 37 Copper No 17 42 26 0.D.
d above tubesheet No 17 44 Copper No 17 47 Copper No 17 50
<20 0.D.
1" above tubesheet No 18 35
<20 0.D.
1" above tubesheet No 18 36 Copper No 18 37 Copper No 18 38 Copper No "A"
Steam Generator Inlet, continued...
Row Column
% Defect Origin Location Plugged 18 40 Copper No 18 41 Copper No 18 43
<20 0.D.
2" above tubesheet No 18 44 Copper No 18 47 Coppet No 18 51
<20 0.D.
1" above tubesheet No 18 60
<20 0.D.
Top of tubesheet No 19 38
<20 0.D.
\\" above tubesheet No s
19 44 Copper No 19 47 Copper No 19 49
<20 0.D.
\\" above tubesheet No 19 51
<20 0.D.
1" above tubesheet No 19 52
<20 0.D.
\\" above tubesheet No 20 6
37 0.D.
First support No 20 34
<20 0.D.
1" above tubesheet No 20 40
<20 0.D.
\\" above tubesheet No 20 42 21 0.D.
\\" above tubesheet No 20 43 Copper No 20 46 29 0.D.
\\" above tubesheet No 20 47 30 0.D.
\\" above tubesheet No 20 50
<20 0.D.
\\" above tubesheet No 20 52
<20 0.D.
" above tubesheet No 21 34 24 0.D.
" above tubesheet No 21 36
<20 0.D.
\\" above tubesheet No 21 46 30 0.D.
Top of tubesheet No 21 47 37 0.D.
\\" above tubesheet No 21 64 30 0.D.
Top of tubesheet No 22 35
<20 0.D.
\\" above tubesheet No 24 55
<20 0.D.
\\" above tubesheet No 27 34 Copper No 31 72 PV Fourth and fifth support No "A" Steam Generator Outlet Row Column
% Defect Origin Location Plugged 7
29
<20 0.D.
\\" above tubesheet No 8
30 Copper No 8
32
<20 0.D.
\\" above tubesheet No 8
34
<20 0.D.
1\\" above tubesheet No 8
36
<20 0.D.
2" above tubesheet No 9
30 Copper No 9
44
<20 0.D.
1" above tubesheet No 10 39 30 0.D.
2" above tubesheet No 10 40
<20 0.D.
2" above tubesheet No 11 43 Copper No 12 40
<20 0.D.
" above tubesheet No 12 42 26 0.D.
2" above tubesheet No 13 30
<20 0.D.
" above tubesheet No 14 40
<20 0.D.
" above tubesheet No "A" Steam Generator Outlet continued...
Row Column
% Defect Origin Location Plugged 18 35
<20 0.D.
1" above tubesheet No 19 33
<20 0.D.
h" above tubesheet No 20 34
<20 0.D.
1" above tubesheet No 21 35
<20 0.D.
1" above tubesheet No 21 46
<20 0.D.
1" above tubesheet No "B" Steam Generator Inlet Row Column
% Defect Origin Location Plugged 13 52
<20 0.D.
1" above tubesheet No 14 39
<20 0.D.
Top of tubesheet No 16 37
<20 0.D.
" above tubesheet No 18 41
<20 0.D.
1" above tubesheet No 19 40 21 0.D.
" above tubesheet No 19 42 31 0.D.
" above tubesheet No 21 34
<20 0.D.
Top of tubesheet No 23 58
<20 0.D.
Top of tubesheet No 23 59
<20 0.D.
1" above tubesheet No 25 51
<20 0.D.
1" above tubesheet No 26 41 24 0.D.
Top of tubesheet No 27 27 31 0.D.
Top of tubesheet No 29 54 Ccpper No 31 42
<20 0.D.
\\" above tubesheet No 32 40 23 0.D.
" above tubesheet No 32 42 32 0.D.
Top of tubesheet No 33 41 (20 0.D.
" above tubesheet No 43 57 56 0.D.
3" above tubesheet Yes "B" Steam Generator Outlet Row Column
% Defect Origin Location Plugged 4
32
<20 0.D.
1" above tubesheet No 4
39
<20 0.D.
Top of tubesheet No 5
27
<20 0.D.
2" above tubesheet No 5
28 26 0.D.
1\\" above tubesheet No 5
32
<20 0.D.
1" above tubesheet No 5
33
<20 0.D.
2" above tubesheet No 6
27
<20 0.D.
1" above tubesheet No 6
32
<20 0.D.
2" above tubesheet No 6
43 21 0.D.
Id above tubesheet No 6
51 21 0.D.
1" above tubesheet No 6
52 21 0.D.
1" above tubesheet No 6
53
<20 0.D.
\\" above tubesheet No 6
54
<20 0.D.
1" above tubesheet No 6
57 21 0.D.
Top of tubesheet No 7
26
<20 0.D.
1" above tubesheet No 7
28 24 0.D.
1" above tubesheet No 7
30 26 0.D.
2" above tubesheet No "B" Steam Generator Outlet, continued Row Column
% Defect Origin Location Pluqqed 7
32 23 0.D.
2" above tubesheet No 7
34
<20 0.D.
2" above tubesheet No 7
36 27 0.D.
2" above tubesheet No 7
39 21 0.D.
2" above tubesheet No 7
40 21 G.D.
2" above tubesheet No 7
41 27 0.D.
1" above tubesheet No 7
43
<20 0.D.
2" above tubesheet No 7
49 21 0.D.
2" above tubesheet No 7
50 24 0.D.
2" above tubesheet No 7
51 22 0.D.
2" above tubesheet No 7
52 21 0.D.
2" above tubesheet No 7
53 21 0.D.
3" above tubesheet No 7
54
<20 0.D.
1" above tubesheet No 7
55
<20 0.D.
2" above tubesheet No 7
56 27 0.D.
1" above tubesheet No 7
57
<20 0.D.
1" above tubesheet No 7
59 23 0.D.
2" above tubesheet No 7
63
<20 0.D.
2" above tubesheet No 7
64
<20 0.D.
2" ab,ve tubesheet No 7
65
<20 0.D.
2" above tubesheet No 7
67 29 0.D.
1" above tubesheet No 7
69 24 0.D.
" above tubesheet No 8
28
<20 0.D.
1" above tubesheet No 8
32
<20 0.D.
1\\" above tubesheet No 8
34
<20 0.D.
2" above tubesheet No 8
38 30 0.D.
2" above tubesheet No 8
52
<20 0.D.
2" above tubesheet No 8
54
<20 0.D.
2" above tubesheet No 8
55 21 0.D.
2" above tubesheet No 8
64
<20 0.D.
2" above tubesheet No 9
26 21 0.D.
1" above tubesheet No 9
28
<20 0.D.
1" above tubesheet No 9
31
<20 0.D.
2" above tubesheet No 9
32
<20 0.D.
1 " above tubesheet No 9
33 21 0.D.
2" above tubesheet No 9
34
<20 0.D.
1\\" above tubesheet No 9
36
<20 0.D.
2" above tubesheet No 9
41
<20 0.D.
2" above tubesheet No 9
44 23 0.D.
2" above tubesheet No 9
47
<20 0.D.
2" above tubesheet No 9
48 26 0.D.
1\\" above tubesheet No 9
49 21 0.D.
2" above tubesheet No 9
50 21 0.D.
2" above tubesheet No 9
51
<20 0.D.
2" above tubesheet No 9
52 32 0.D.
2" above tubesheet No 9
53
<20 0.D.
2" above tubesheet No 9
54 24 0.D.
2" above tubesheet No 9
55 21 0.D.
2" above tubesheet No 9
56
<20 0.D.
2" above tubesheet No 9
60 24 0.D.
2" above tubesheet No "B" Steam Generator Outlet, continued...
Row Column
% Defect Origin Location Plugged 9
61 21 0.D.
2" above tubesheet No 9
62 23 0.D.
2" above tubesheet No 9
65
<20 0.D.
1\\" above tubenheet No 9
66 21 0.D.
1" above tubesheet No 9
67
<20 0.D.
1 " above tubesheet No 9
72 23 0.D.
1" above tubesheet No 10 26 23 0.D.
1 " above tubecheet No 10 27
<20 0.D.
1" above tubesheet No 10 34
<20 0.D.
2" above tubesheet No 10 40 21 0.D.
1" above tubesheet No 10 42
<20 0.D.
2" above tubesheet No 10 43 27 0.D.
1 " above tubesheet No 10 44
<20 0.D.
1" above tubesheet No 10 45
<20 0.D.
1 " above tubesheet No 10 46
<20 0.D.
1" above tubesheet No 10 47
<20 0.D.
Ih" above tubesheet No 10 49
<20 0.D.
1 " above tubesheet No 10 52
<20 0.D.
2" above tubesheet No 10 54
<20 0.D.
2" above tubesheet No 10 56 29 0.D.
2" above tubesheet No 10 60 21 0.D.
1" above tubesheet No 10 62 21 0.D.
1\\" above tubesheet No 10 65
<20 0.D.
1" above tubesheet No 10 67
<20 0.D.
1" above tubesheet No 10 70
<20 0.D.
1" above tubesheet No 11 30
<20 0.D.
1!
Sove tubesheet No 11 32
<20 0.D.
15 aove tubesheet No 11 36
<20 0.D.
2" bove tubesheet No 11 39 24 0.D.
1" above tubesheet No 11 43 21 0.D.
1 " above tubesheet No 11 44 21 0.D.
1" above tubesheet No 11 49
<20 0.D.
l' above tubesheet No 11 50
<20 0.D.
1-above tubesheet No 11 52
<20 0.D.
1\\" \\bove tubesheet No 11 54
<20 0.D.
1" amove tubesheet No 11 56 27 0.D.
2" above tubesheet No 11 57
<20 0.D.
2" above tubesheet No 11 58
<20 0.D.
1" above tubesheet No 11 59 21 0.D.
2" above tubesheet No 11 61 26 0.D.
2" above tubesheet No 12 27
<20 0.D.
1" above tubesheet No 12 30 21 0.D.
1" above tubesheet No 12 32
<20 0.D.
1\\" above tubtsheet No 12 33
<20 0.D.
1 " above tubesheet No 12 34
<20 0.D.
1\\" above tubesheet No 12 35 21 0.D.
1\\" above tubesheet No 12 37
<20 0.D.
1 " above tubesheet No 12 42
<20 0.D.
\\" above tubesheet No 12 47
<20 0.D.
" above tubesheet No 12 49
<20 0.D.
1" above tubesheet No "B" Steam Generator Outlet, continued...
Row Column
% Defect Origin Location Plugged 12 50
<20 0.D.
1" above tubesheet No 12 56
<20 0.D.
1" above tubesheet No 13 31 21 0.D.
\\" above tubesheet No 13 32
<20 0.D.
1" above tdbesheet No 13 33
<20 0.D.
1" above tubesheet No 13 35
<20 0 D.
1" above tubesheet No 13 37 21 0.D.
1" above tubesheet No 13 39 24 0.D.
1" above tdaesheet No 13 40 21 0.D.
1" above tubesheet No 13 41
<20 0.D.
1" above tubesheet No 13 42
<20 0.D.
\\" above tubesheet No 13 43 21 0.D.
\\" above tubesheet No 13 44
<20 0.D.
" above tubesheet No 13 49
<20 0.D.
1" above tdaesheet No 13 50
<20 0.D.
1" above tubesheet No 13 51
<20 0.D.
1" above tubesheet No 13 54 21 0.D.
1" above tubesheet No 13 56
<20 0.D.
1" above tubesheet No 13 61
<20 0.D.
1\\" above tubesheet No 13 62
<20 0.D.
1 " above tubesheet No 13 65
<20 0.D.
1" above tubesheet No 14 33
<20 0.D.
1 " above tubesheet No 14 36
<20 0.D.
1" above tubesheet No 14 39 21 0.D.
" above tubesheet No 14 47
<20 0.D.
\\" above tubesheet No 14 49
<20 0.D.
\\" above tdoesheet No 14 50
<20 0.D.
1" above tubesheet No 14 51 21 0.D.
" above tubesheet No 14 52
<20 0.D.
1" above tubesheet No 14 53
<20 0.D.
1" above tubesheet No 14 54
<20 0.D.
1" above tubesheet No 14 55
<20 0.D.
1" above tubesheet No 14 57
<20 0.D.
1" above tubesheet No 14 58
<20 0.D.
1" above tubesheet No 14 62 21 0.D.
1" above tubesheet No 14 67
<20 0.D.
" above tubesheet No 14 68
<20 0.D.
\\" above tubesheet No 15 33
<20 0.D.
1" above tubesheet No 15 37
<20 0.D.
" above tubesheet No 15 47
<20 0.D.
" above tubesheet No 15 48
<20 0.D.
\\" above tubesheet No 15 49
<20 0.D.
\\" above tubesheet No 15 50
<20 0.D.
b" above tubesheet No 15 51
<20 0.D.
1" above tubesheet No 15 52
<20 0.D.
1" above tubesheet No 15 53
<20 0.D.
1" above tubesheet No 15 54 32 0.D.
1" above tubesheet No 15 55
<20 0.D.
1" above tubesheet No 15 56
<20 0.D.
1" above tubesheet No 15 59
<20 0.D.
1\\" above tdaesheet No "B" Steam Generator Outlet, continued...
Row Column
% Defect Origin Location Plugged 15 63 24 0.D.
1" above tubesheet No 15 64
<20 0.D.
" above tubesheet No 15 65 21 0.D.
" above tubesheet No 15 67 21 0.D.
\\" above tubesheet No 16 35 (20 0.D.
\\" above tubesheet No 16 36
<20 0.D.
\\" above tubesheet No 16 49
<20 0.D.
" above tubesheet No 16 51
<20 0.D.
\\" above tubesheet No 16 52
<20 0.D.
1" above tubesheet No 16 53
<20 0.D.
1" above tubesheet No 16 55
<20 0.D.
1" above tubesheet No 16 60 21 0.D.
1" above tubesheet No 17 50
<20 0.D.
" above tubesheet No 17 62
<20 0.D.
" above tubesheet No 18 29 23 0.D.
1" above tubesheet No 18 30
<20 0.D.
" above tubesheet No 18 32 30 0.D.
1" above tubesheet No 18 33
<20 0.D.
\\" above tubesheet No 18 49 22 0.D.
" above tubesheet No 18 50 74 0.D.
1" above tubesheet No 18 51 i9 0.D.
1" above tubesheet No 18 52
<20 0.D.
1" above tubesheet No 18 53
<20 0.D.
" above tubesheet No 18 54 21 0.D.
1" above tubesheet No 18 55
<20 0.D.
1" above tubesheet No 18 56
<20 0.D.
1" above tubesheet No 18 57 21 0.D.
1" above tubesheet No 18 58 21 0.D.
1" above tubesheet No 18 61 21 0.D.
\\" above tubesheet No 18 63 22 0.D.
" above tubesheet No 19 50 35 0.D.
1" above tubesheet No 19 52
<20 0.D.
1" above tubesheet No 19 53 24 0.D.
\\" above tubesheet No 19 55
<20 0.D.
1" above tubesheet No 19 57 21 0.D.
1" above tubesheet No 19 58 21 0.D.
1" above tubesheet No 19 59 23 0.D.
1" above tubesheet No 19 62
<20 0.D.
" above tubesheet No y
19 67
<20 0.D.
Ih" above tubesheet No 19 68 21 0.D.
2" above tubesheet No 21 38
<20 0.D.
" above tubesheet No 21 52
<20 0.D.
\\" above tubesheet No 21 54
<20 0.D.
\\" above tubesheet No 21 58 21 0.D.
1" above tubesheet No 22 51
<20 0.D.
\\ above tubesheet No 22 52 29 0.D.
\\" above tubesheet No 22 53 23 0.D.
" above tubesheet No 22 55
<20 0.D.
1 " above tubesheet No 22 56 21 0.D.
" above tubesheet No 22 57 27 0.D.
1" above tubesheet No "B" Steam Generator Outlet, continued...
Mow Column
% Defect Origin Location Plugged 23 52 21 0.D.
\\" above tubesheet No 23 54 23 0.D.
1" above tubesheet No 23 55 24 0.D.
1" above tubesheet No 23 56 21 0.D.
" above tubesheet No 23 57 27 0.D.
b" above tubesheet No 23 58 21 0.D.
" above tubesheet No 23 65 21 0.D.
1" above tubesheet No 24 54 21 0.D.
\\" above tubesheet No i
24 56
<20 0.D.
" above tubesheet No 24 57 21 0.D.
\\" above tubesheet No 24 58 23 0.D.
\\" above tubesheet No 25 56 24 0.D.
\\" above tubesheet No 27 65
<20 0.D.
" above tubesheet No O
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