ML20041E142
| ML20041E142 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/01/1982 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20041E137 | List: |
| References | |
| NUDOCS 8203100182 | |
| Download: ML20041E142 (59) | |
Text
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WISCONSIN ELECTRIC ANNUAL RESULTS AND
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POWER COMPANY 1981 POINT BE ACH NUCLE AR PL ANT UNIT NOS.1 AND 2 U.S. Nuclear Regulatory Commission Docket Nos. 50-266 and 50-301 i Facility Operating License Nos. -
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- PREFACE This Annual Results and Date Report for 1981 is submitted in accor-dance with Point Beach Nuclear Plant Unit Nos. I and 2, Technical Specifi-cation 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 4 3.3 Tests or Experiments Requiring NRC Approval 4 3.4 Design Changes 5 3.4.1 Unit 1 5 3.4.2 Unit 2 12 3.4.3 Common 16 3.5 Procedure Changes 23 t,
4.0 NUMBER OF PERSONNEL AND MAN-REM EXPOSURE BY WORK AND JOB FUNCTION
, 4.1 1980 27 5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION 5.1 Unit 1 28 5.2 Unit 2 33
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1.0 IllTRODUCT10N The Point Beach fluclear 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 HIGliLIGliTS 2.1 Unit 1 For the period 01-01-81 through 12-31-81, which included a 61-day refueling outage, one steam generator inspection outage, a trip resulting from the inadvertent deenergization of a vital bus inverter due to a contractor opening its breaker, a brief outage for investi-gation of high containment air particulate levels and one trip due to a faulty turbine-generator voltage regulator. Unit 1 operated at an average capacity factor of 60.0% and a net efficiency of 30.6%. The unit and reactor availability were 78.0% and 79.0%, respectively.
Unit 1 generated its 34 billionth kilowatt hour (gross) on 03-21-81; its 35 billionth kilowatt hour on 07-21-81; and its 36 billionth kilowatt hour on 12-29-81.
2.2 Unit 2 For the period 01-01-81 through 12-31-82, which included a 35-day refueling outage, two shutdowns to accommodate license exam startups, two outages for repairs to a high pressure turbine steam leak and an outage to perform maintenance on a turbine stop valve and moisture separator reheater leakage, Unit 2 operated at an average capacity factor of 85.5% and a net efficiency of 32.3%. The unit and reactor availability were 88.6% and 86.2%, respectively. Unit 2 generated its 31 billionth kilowatt hour (gross) on 03-14-81; its 32 billionth kilowatt hour on 07-18-81; its 33 billionth kilowatt hour on 10-08-81; and its 34 billionth kilowatt hour on 12-29-81.
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i 3.0 FACILITY CHANGES, TESTS AND EXPERIMENTS 3.1 Amendments To Facility Operating Licenses During 1981, there were 11 license amendments issued by the U.S.
Nuclear Regulatory Commission to Facility Operating License DPR-24 for Point Beach Unit 1 and 10 amendments issued to Facility Operating License DPR-27 for Point Beach Unit 2. In addition to these license amendments, the U.S. lieclear Regulatory Commission also issued an order for modification of license dated April 20, 1981, which revised the Technical Specification for the Point Beach Nuclear Plant. These license amendments and changes are listed by date of issue and sum-marized as follows.
3.1.1 3-31-81, Amendment 47 to DPR-24, Amendment 52 to DPR-27 These amendments revised the Technical Specifications to reflect modifications to plant equipment to provide addi-tional assurance of the proper functioning of the plant's safety related electrical and emergency power systems.
3.1.2 4-3-81, Amendment 53 to DPR-27 This amendment authorized a one time waiver of the monthly functional test of the turbine stop and governor valves until the start of the Unit 2 seventh refueling outage.
3.1.3 4-20-81, Order For Modification Of Licenses DPR-24 & DPR-27 This order required testing of several prassure isolation check valves associated with Event V loss of coolant acci- 1 dents and imposed limiting conditions for operation and ]
periodic surveillance requirements for these valves by additions to the Technical Specificatons.
3.1.4 Amendment 48 to DPR-24, Amendment 54 to DPR-27 These amendments revised license condition 3.F to include the NRC approved safeguard contingency plan as the revised Chapter 8 to the physical security plan.
3.1.5 5-4-81, Amendment 49 to DPR-24, Amendment 55 to DPR-27 These amendments removed rod bow penalties and requirements related to control rod misalignment and position indication and include several administrative changes to parts of Section 15.3.10 of the Technical Specifications.
3.1.6 6-24-81, Amendment 50 to DPR-24, Amendment 56 to DPR-27 These amendments consisted of a revised definition of oper-ability and several new specifications addressing limiting conditions for operation including inoperability of safety-related system due solely to loss of their normal emergency power supplies.
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> 3.1.7 7-10-81, Amend ent 51 to DPR-24, Amendment 57 to DPR-27, These amen &cnts updated the reactor coolant system tempera-ture and pressure operating curves for Unit 2 and revised the reactor vessel materials surveillance capsule removal schedules for both i tits.
3.1.8 8-20-81, Amendment t to DPR-24, Amenchent 58 to DPk-27 These amendments approved miscellaneous minor achinistrative changer and a revised organization structure for the Point Beach facility.
3.1.9 8-28-81, Amendment 53 to DPR-24, Amendment 59 to DPR-27 The reactor coolant system temperature and pressure oper-ating curves for Unit 1 were updated by this amenchent.
Hinor related administrative changes were also made to the Unit 2 Technical Specifications.
3.1.10 9-25-81, Amendment 54 to DPR-24 This amendment allowed a one-time relaxation of the require-ment for a monthly functional test of the Unit 1 turbine stop and governor valves until the start of the ninth re-fueling outage.
3.1.11 9-30-81, Amendment 55 to DPR-24, Amendment 60 to DPR-27 These amendments approved a series of Technical Specifi-cation additions resulting from THI-2 short term lessons learned Category "A" items.
3.1.12 11-10-81, Amendment 56 to DPR-24 This amendment authorized power operations with up to six tubes in one steam generator having degradation exceeding the plugging limit provided these tubes were repaired by insertion of sleeves inside the tubes to bridge the degraded portion of the tube. It also established a 35% of nominal sleeve wall thickness plugging limit for the repaired tubes.
3.1.13 12-21-81, Amendment 57 to DPR-24, Amendment 61 to DPR-27 These amendments revised the degraded grid undervoltage reJay time delay setpoint for an interim period of 30 days from Dqcember 3, 1981.
3.2 Facility or Procedure Changes Requiring Nuclear Regulatory Lommission <
Approval There were no plant modifications or procedure changes during 1981, 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 1981 which required Nuclear Regulatory Commission approval.
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3.4 Design Changes 3.4.1 Unit 1
- a. Containment Equipment Hatch Power Disconnect (E-150)
The modification installed plug-ins and receptacles on equip-ment hatch wiring to facilitate hatch removal.
Summary of Safety Evaluation: Not nuclear safety related.
- b. Containment Equipment Hatch Outside Lightis.q (E-155)
The modification installed additional lighting outside of the containment equipment hatch.
. Summary of Safety Evaluation: Not nuclear safety related.
- c. Safeguards Power Supply (E-202)
The request installed local / remote transfer and breaker control switches to allow an operator to take local control and thus prevent a potential fire which could occur in con-trol board CO2 from disabling power supplies.
Summary of Safety Evaluation: Local breaker control in addition to remote control is recognized as an improvement in plant safety.
, d. W-2 Switch closure Indication (E-233)
Wiring to the existing breaker green light indicator was put
, in series with the neutral position switch contacts on Wes-tinghouse type W-2 switches in safety-related systems. The change was recommended by a Westinghouse NSD Technical Bulle-tin and NRC IE Bulletin No. 80-20 because of continuity failures of neutral position switch contacts experienced at other nuclear plants.
Summary of Safety Evaluation: The wiring change was similar to that called out in IE Bulletin No. 80-20 and has no effect on the operation of the component the switch is related to.
This modification helps in the evaluaton of the neutral position contacts.
- e. Blowdown Heat Exchanger Pressure Switch Heat Tracing (E-242)
Heat tracing was installed to provide freeze protection for pressure switch 1PS-5955 located in the Unit I lacade.
Summary of Safety Evaluation: Not nuclear safety related.
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- f. Containment Penetration (E-248) l The request installed a new penetration in position E28. The penetration is required to provide leads for approximately 40 new fire detectors to be installed in containment.
Summary of Safety Evaluation: The modification will not have a deleterious effect upon plant safety. The penetration design is equal to or better than original plant design.
- g. Facade 480 V Outlets (E-261)
Additional outlets were installed to support refueling acti-vities.
Summary of Safety Evaluation: Not nuclear safety related.
- h. Bypass Bistable Test Switch on Process Control System's Ana-log Rack's Test Panel (IC-48)
This modification request was proposed by Westinghouse fol-lowing the discovery of a possibly generic problem with bistable output test switches on the test panel associated with the process control system analog racks. A short cir-cuit which shunts the bistable switching element and the switch contact could go undetected using existing equipment. ,
This modification enabled the periodic tests of the logic to j detect this problem. The in-depth review of the design by '
Westinghouse was accepted by the Staff following a detailed explanation of the problem and modification by the I&C Engi-neer.
Summary of Safety Evaluation: The shorting out of one pole of the test switch will better simulate actual operating conditions in the channel under test and make the discovery of previously undetectable grounds possible.
- i. Manipula' tor Hoist Pulley Encoder (IC-144)
A linear displacement encoder was installed on the fuel mani-pulator hoist pulley with readout next to the load cell. The encoder will provide mast height indication accurately and reportably.
Summary of Safe'y t Evaluation: The encoder provides a superior means of determining mast height compared with the present measuring tape method.
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- j. Power-Operated Relief Valve Redundant Pressure Channel (IC-147)
The modification added a redundant pressure signal channel to the controls for the power-operated relief valves when they are used to prevent over-pressurization during low pressure and/or water solid conditions. In addition, a backup gas / air supply was added for the air operators of the PORV's. An interlock through the motor-operated isolation valves was provided for the key-switch indicators.
Summary of Safety Evaluation: The modificaton resolves NRC objections to the original modification (IC-115) and adds redundancy to the existing control channels.
- k. Containment Evacuation Alarm (IC-151)
The modification permanently installed control room wiring for the containment evacuation alarm.
Summary of Safety Evaluation: Not nuclear safety related.
- 1. 5A & 5B Feedwater Heater Drains Valve Trips (IC-164)
The modification changed air supply for SA and 5B heater drain and dump valve controllers to output of the turbine relay dump valve. This causes the drain and dump to open on a turbine trip and lessen the chance of water hammer in the No. 4 heater drain piping.
Summary of Safety Evaluation: Not nuclear safety related.
- m. Light for Steam Generator Feed Pump Two-Minute Timer (IC-172)
The modification installed an indicator light on the main control board to indicate that the main feed pump trip on low suction pressure two-minute time delay relay is timing out.
Light is; needed to alert the operator of eminent main feed pump trip so corrective actions can be initiated.
Summary of Safety Evaluation: Not nuclear safety related.
- n. Feed Pump Trip Reliability Improvement (IC-190)
The modification changed feed pump trips from one out of one low suction pressure or lube oil pressure to two out of three low suction pressure or low lube oil pressure. The change was necessary to enhance feed pump reliability.
Summary of Safety Evaluation: -ot nuclear safety related.
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- o. Primary Sampling System (IC-248)
The modification added trip signals on containment isolation to the following valves in accordance with NUREG-0578, Section 2.1.4: 951, pressurizer steam space sample: 953, pressurizer liquid space sample; 955, reactor coolant system hot leg sample.
Summary of Safety Evaluation: The modification increased assurance that positive isolation of these lines would occur upon a containment isolation.
- p. Steam Generator Blowdown Isolation (IC-250)
The modification request installed a containment isolation signal to valve ICV-2042 and 2045 as required by NUREG-0578, Section 2.1.4.
Summary of Safety Evaluation: The modification request was initiated in accordance with the C. W. Fay to H. Denton May 2, 1980, letter commitment for Category "A" Lessons Learned requirements. No change in the present testing of these valves relative to Appendix J (10 CFR 50) was made as a result of this change.
The modification provided each system with automatic double barrier isolation by diverse signals upon receipt of a valid safety injection signal.
- q. Steam Generator Blowdown Sample Valve Isolation (IC-252)
The modificaton request installed a containment isolation j signal to valve 1CV-2083 and 2084 as required by NUREG-0578, Section 2.1.4.
Summary of Safety Evaluation: The modification request was initiated in accordance with the C. W. Fay to H. Denton May 2, 1980, letter commitment for Category "A" Lessons Learned requirements. No change in the present testing of these valves relative to Appendix J (10 CFR 50) was made as a result of this change.
The modification provided each system with automatic double barrier isolation by diverse signals upon receipt of a valid safety injection signal.
- r. Service Water to Generator Bus Coolers Flow Switch (IC-269)
A new type flow switch was installed because of repeated failures of the originally installed flow switch.
Summary of Safety Evaluation: Not nuclear safety related.
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- s. Main Control Board C04 (IC-281)
The modification installed an alarm on IC04 to indicate when-ever reactor water makeup pump discharge valve FCV-111 is not shut. The alarm reads " Potential Dilution in Progress" to alert the operator of the need to shut the valve and to avert a boron dilution during cold shutdown conditions. Completion of the modification fulfilled a commitment made in the C. W. Fay to H. R. Denton letter dated 03/31/81.
Summary of Safety Evaluation: Not nuclear safety related.
- t. Steam Generator Blowdown Tank Remote Level Indicator (IC-283)
The modification added a remote level indicator - for steam generator blowdown tank near blowdown filter pump (IP49).
The indication was needed to aid the operator in correcting blowdown tank level problems.
Summary of Safety Evaluation: Not nuclear safety related,
- u. Gland Steam Condenser Flanges (M-576)
Flanges were installed in the condensate inlet and outlet piping of the gland steam condenser to facilitate access to the tube side.
Summary of Safety Evaluation: Not nuclear safety related.
- v. Fire Protection - Relocate Pull Boxes Turbine Bearing System (M-649)
The pull-boxes were removed from mountings on the main tur-bine generator and installed in panels mounted to the moisture separator reheater structural support steel. This improved accessibility to the pull-boxes and removed them from turbine-generator vibration.
Summary of Safety Evaluation: Not nuclear safety related.
- w. Thimble Plug Removal Tool Storage Bracket (M-651)
A fixture was fabricated and installed in the reactor re-fueling cavity for storage of the thimble plug removal tool.
Summary of Safety Evaluation: Not nuclear safety related.
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- x. Condensate Pipe Supports (M-665)
A snubber support on the condensate piping going from the No. 3 to the No. 4 feedwater heaters was modified and a support added at the emergency bypass valve to reduce piping vibrations.
Summary of Safety Evaluation: Not nuclear safety related.
- y. Post-Accident Coolant Sampling (M-675)
A stainless steel line with associated isolation valves was installed in the residual heat removal piping to obtain the ability to sample coolant when a unit is on residual heat removal at low temperature and pressure.
Summary of Safety Evaluation: The design of the piping systems maintain the isolation and integrity of the system in accordance with the acceptable criteria for the systems involved and poses no additional or unanalyzed conditions degrading the safety of the plant.
- z. Service Water Piping Reroute (M-729)
The modification replaces a seven-foot section of service water return line HB-19 in order to provide adequate clear-ance for replacement electrical penetration Q58.
Summary of Safety Evaluation: The modification had no dele-terious effect upon the existing system but allowed for use of the new electrical penetration to be installed per Modifi- I cation Request E-230. The work was performed during an l outage when service water could be taken out-of-service. >
aa. Fuel Assemblies (M-745)
The request detailed changes to Westinghouse fuel assembly grid corners and affected the new fuel assemblies being utilized during Unit 1 Refueling 9, Unit 2 Refueling 7, and all new assemblies manufactured thereafter.
Summary of Safety Evaluation: The Westinghouse product line modification assisted in reducing the possibility of grid corner damage during fuel handling operations when grid assemblies intsract with grids on adjacent fuel assemblies.
In addition to revising the grid corner taper angle, Westing-house had instituted improved deburring practices to result in a more uniform edge thickness and a smoother ground sur-face at the grid corners.
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, bb. Safety Injection Valve Test Tubing (M-778)
The modification installed high pressure tubing, valves and instruments to perform leak testing on valves 845A-F, 853A-D and 967A&B in accordance with NRC requirements regarding testing of Event V check valves (first and second-off safety injection valves). The leak test system utilizes charging as a supply for backpressure against the valves and is' discon-nected and capped when not in use.
Summary of Safety Evaluation: Not nuclear safety related.
ec. Residual Heat Removal Piping (M-809)
There are two parts to this modification request. Part I required cutting and capping of the 2" pipe connection to each RHR pump discharge line from the refueling water circu-lating pump (P33). This eliminated the need to leak test the piping under post-THI requirements for systems outside con-tainment likely to contain radioactive materials. The piping is not used and its intended function could not be clearly established, although one possibility is that the lines are to be used for adjusting the boron concentration of the RHR loops. This function is performed using other piping con-nections.
Part 2 of the modification request was- the installation of test connections and isolation valves in the spent fuel pit filter discharge connection to the RHR pump suction header.
The change was also required to facilitate leak testing per post-THI requirements for systems outside containment likely to contain radioactive materials.
Summary of Safety Evaluation: Part 1 - discharge lines from P33 to the RHR discharge serve no safety function. Cap in-stallation complied with Westinghouse Pipe Class 601 and other design criteria of the RHR system. Part 2 - not nuclear safety related.
dd. Residual Heat Removal Pump Shaft sleeve "0" Ring (M-812)
An "O" ring was installed between the RHR pumps shaft and sleeve to further reduce leakage.
Summary of Safety Evaluation: Not nuclear safety related.
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l ee Instrument Air Diaphra y-Type Isolation Valves (M-817)
Diaphragm valves _were installed upstream of the instrument air containment isolation valves to provide better isolation when performing Type "C" leak tests.
Summary of Safety Evaluation: Not nuclear safety related.
y 3.4.2 Unit 2
- a. Containment Equipment Hatch outside Lighting (E-154)
The modification installed additional lighting outside of the containment equipment hatch.
Summary of Safety Evaluation: Not nuclear safety related.
- b. Steam-Driven Auxiliary Feedwater Pump Bearing Service Water Solenoid Valve (E-199) i The circuitry for the valve was modified to allow operation of the pump and blowdown at the same time without having to bypass the solenoid valve or run the feedwater pump without bearing service water.
Summary of Safety Evaluation: Improved the integrity of service water cooling to the steam driven auxiliary feed pump bearings.
- c. W-2 Switch Closure Indication (E-234) i i Wiring to the existing breaker green light indicator was put in series with the neutral position switch contacts- on Wes-tinghouse type W-2 switches in safety-related ' systems. The d change was recommended by a Westinghouse NSD Technical Bul-letin and NRC IE Bulletin 80-20 because of continuity failures of neutral position switch contacts experienced at other nuclear plants.
Summary of Safety Evaluation: The wiring change was similar to that called out in IE Bulletin No. 80-20 and has no effect on the operation of the component the switch is related to.
This modification helps in the evaluation of the neutral position contacts.
- d. Blowdown Heat Exchanger Pressure Switch Heat Tracing (E-243) 1 Heat tracing was installed to provide freeze protection for pressure switch 2-PS-5955 which is located in the Unit 2
! facade.
Summary of Safety Evaluation: Not nuclear safety related.
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- e. Power-Operated Relief Valve Redundant Pressure Channel (IC-1ip The r'adification added a redundant pressure signal channel to the contiols for the power-operated relief valves .when they are used to prevent over-pressurization during low pressure and/or water solid conditions. In addition, a backup gas / air supply was added for the air operators of the PORV's. An interlock through the motor-operated isolation valves was provided for the key-switch indicators.
Summary of Safety Evaluation: The modification resolved NRC objections to the original modification (IC-116) and added redundancy to the existing control channels.
- f. Containment Evacuation Alarm (IC-152)
The modification permanently installed control room wiring j for the containment evacuation alarm.
Summary of Safety Evaluation: Not nuclear safety related.
- g. Feed Pump Trip Reliability Improvement (IC 191)
The modification changed feed pump trips from one out of one low suction pressure or lube oil pressure to two out of three low suction pressure or low lube oil pressure. The change was necessary to enhance feed pump reliability.
Summary of Safety Evaluation: Not nuclear safety related.
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- h. Primary Sampling Systems (IC-249)
The modification called for adding trip signals on contain-ment isolation to the following valves in accordance with NUREG-0578, Section 2.1.4: 951, pressurizer steam space sample; 953, pressurizer liquid space sample; 955, reactor i i
coolant system hot leg sample.
Summary of Safety Evaluation: The modification increased assurance that positive isolation of these lines would occur I upon a containment isolation.
- i. Steam Generator Blowdown Isolation (IC-251)
The modification request installed a containment isolation l signal to valves 2CV-2042 and 2045 as required by NUREG-0578, Section 2.1.4.
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Summary of Safety Evaluation: The modification request was i initiated in accordance with the C. W. Fay to H. R. Denton May 2, 1980, letter commitment for Category "A" Lessens Learned requirements. No change in the present testing of these valves relative to Appendix J (10 CFR 50) was n.adt.
The modification provided each system with automatic double barrier isolation by diverse signals upon receipt of a valid safety injection limit.
- j. Steam Generator Blowdown Sample Valve Isolation (IC-253)
The modification request installed a containment isolation signal to valves 2CV-2083 and 2084 as required by NUREG-0578, Section 2.1.4.
Summary of Safety Evaluation: The modification request was initiated in accordance with the C. W. Fay.to H. R. Denton May 2, 1980, letter commitment for Category "A" - Lessons Learned requirements. No change in the present testing of these valves relative to Appendix J (10 CFR 50) was made as a result of this change.
The modification provided each system with automatic double barrier isolation by diverse signals upon receipt of a valid safety injection signal.
- k. Main Control Board C04 (IC-282)
The modification installed an alarm on 2C04 to indicate whenever reactor water makeup pump discharge valve FCV-111 is not shut. The alarm reads " Potential Dilution in Progress" to alert the operator of the need to shut the valve and to avert a boron dilution during cold shutdown conditions.
Completion of the modification fulfilled a commitment made in the C. W. Fay to H. R. Denton letter dated 03/31/81.
Summary of Safety Evaluation: Not nuclear safety related.
- 1. Steam Generator Blowdown Tank Remote Level Indicator (IC-284) i The modification added a remote level indicator for steam generator blowdown tank near blowdown filter pump (2P49).
The indication was needed to aid the operator in correcting f
blowdown tank level problems.
Summary of Safety Evaluation: Not nuclear safety related.
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- m. Steam Dump Header Drains (M-494)
The condenser north steam dump header trap drain header was repiped to discharge to the low pressure trap header instead of directly to the condenser. This change allows the traps to function before establishing condenser vacuum since the low pressure trap header can be aligned to drain to the atmospheric blow-off tank.
Summary of Safety Evaluation: Not nuclear safety related.
- n. Pressurizer Relief Tank Rupture Disk Shield (M-687)
A deflecting shield was installed above the pressurizer relief tank rupture disc to protect cabling located above the rupture disc.
Summary of Safety Evaluation: Not nuclear safety related.
- o. Fuel Assemblies (M-746)
The request detailed changes to Westinghouse fuel assembly grid corners and affected the new fuel assemblies being utilized during Unit 1 Refueling 9, Unit 2 Refueling 7, and all new assemblies manufactured thereafter.
Summary of Safety Evaluation: The Westinghouse product line modification assisted in reducing the possibility of grid corner damage during fuel handling operations when grid assemblies interact with grids on adjacent fuel assemblies.
l In addition to revising the grid corner taper angle, Westing-house had instituted improved deburring practices to result in a more uniform edge thickness and a smoother ground sur-face at the grid corners.
- p. Safety Injection Valve Test Tubing (M-765)
The modification installed high pressure tubing, valves and instruments to perform leek testing on valves 845A-F, 853A-D and 967A&B in accordance with NRC requirements regarding testing of Event V check valves (first and second-off safety injection valves). The leak test system utilizes charging as a supply for backpressure against the valves and is dis-connected and capped when not in use.
Summary of Safety Evaluation: Not nuclear safety related.
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- a. Transmission of R21 Activity Signals to Plant Process Compu-ters (E-131)
This request resulted in more accurate and timely radwaste accounting.
Summary of Safety Evaluation: Not nuclear safety related.
- b. Atcor System SW-A0V-7 Rewiring (E-217)
The modification changed the control on the subject valve to allow the addition of flush water to the mixer-feeder at all times. Original design restricted operation to the flush cycle. This aided in preventing clogging of the mixer-feeder during processing.
Summary of Safety Evaluation: Not nuclear safety related.
- c. Control Room Heating & Ventilation (E-237) %
The modification replaced the existing selector switch, added another relay and provided indication for control of the control room heating ventilation and air conditioning recir-culation.
Summary of Safety Evaluation: The modification aided system design by preventing control room ventilation from automati-cally returning to normal from the 100% recirculation mode after containment isolation had been reset. Completion of this modificatin fulfilled commitments mady by WE per the 06/06/80 letter to the NRC in response to IE Bulletin No.
80-06.
- d. Lake Thermocouples (E-245)
Four lake thermocouples were moved closer to the intake crib to better detect warm water flow when on ice melt.
Summary of Safety Evaluation: Not nuclear safety related.
- e. Installation of Shutdown Circuit for Blowdown Evaporator &
Letdown Gas Strippers (IC-33)
This modification request was issued to complete a commitment made in an abnormal occurrence report (74-30) and provided for automatic shutdown of the evaporator and strippers in the event high pressure occurs in the evaporator. It also pro-vided for an alarm at a pressure lower than the trip point.
Summary of Safety Evaluation: This modification reduced the potential for radioactive release via the evaporator relief valves by providing automatic alarm and shutdown.
- f. Waste Holdup Tank Pressure (IC-154)
The modification request installed a new wide range instru-ment for PT-155 which increases the existing 0-5 psig range to 0-15 psig. This will enable operating personnel to moni-tor tank pressure during an off-normal condition.
Summary of Safety Evaluation: Not nuclear safety related.
- g. Reboiler Level Control Blowdown Evaporator (IC-159)
The modification permanently installed a level control system which controls condensate level in the blowdown evaporator c reboiler. The system was temporarily installed when problems were encountered with reboiler tube plugging. The modifi-cation provides for permanent installation of electrical cable and level control components including a remote level indicator used by the temporary system.
Summary of Safety Evaluation: Not nuclear safety related.
- h. Addition of Blowdown Evaporator Level Indicator (IC-161)
The modification installed a redundant level indication on the blowdown evaporator and reinstalled level alarms pre-viously disconnected. The modification provides for more reliable level indication and control of the blowdown evapor-ator.
Summary of Safety Evaluation: Not nuclear safety related.
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- i. Installation of Pneumatic Operated Drain Trip Valve in Cryo-genic Drain Piping (M-174)
Addition of the valve prevented a gas stripper pressure increase being made worse by gas feeding back from the cryo-genic unit. The new valve trips and isolates the cryogenic unit from the gas stripper in the event of high pressure in the stripper.
Summary of Safety Evaluation: The addition of the new trip valve reduced the possibility of overpressyre of the gas stripper and, additionally, reduced the magnicude of a radio-active release if the stripper's safety valves operated, the cryogenic unit'being isolated from the stripper at that time.
- j. Boric Acid Evaporators Condensate Drain Traps (M-255)
The bucket-type steam trap on both Units 1 and 2 boric acid evaporator steam heat exchangers were replaced with float-type traps to improve condensate drainage.
Summary of Safety Evaluaton: Not nuclear safety related.
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- k. Radwaste Concrete Mixer - New Atcor System (M-309)
This system was placed in operation several years ago but was inadvertently not included in any previous annual report.
The system improved the capability to solidify liquid rad-vaste and resin using concrete. The system consists pri-marily of .a waste tank, concrete silo and transfer auger, mixer-feeder and associated piping, pumps and controls.
Summary of Safety Evaluation: Not nuclear safety related.
- 1. Ventilation - W35 Complex Recirculation Damper (M-470)
The auxiliary building ventilation supply unit (W35) damper control system was modified to enable h to recirculate up to 50% of its supply air from the auxiliary building. The purpose of increasing recirculation air is to minimize the potential for freezing the heat exchanger coils.
Summary of Safety Evaluation: Not nuclear safety related.
- m. Fire Sprinkler System (M-605)
Wet pipe fire sprinkler systems were installed over the safety injection pumps, component cooling water pumps and service water and diesel fire pumps and the dry pipe system in the emergency diesel generator rooms was converted to a wet pipe system. The modifications were made to comply with
-new NRC regulatory requirements.
Summary of Safety Evaluation: Not nuclear safety related.
- n. Building & Facilities - Compressed Gas Bottle-Storage Building (M-607)
A separate building was constructed to provide safe' storage of compressed gas bottles. The need for the building was identified as an industrial safety improvement item.
Summary of Safety Evaluation: Not nuclear safety related,
- o. Fire Protection (M-620)
The modification installed a smoke and heat vent system for the cable spreading room, control room and computer room; this modification was required to provide suitable smoke removal capability in the e nnt of a potential fire in the subject areas.
Summary of Safety Evaluation: The new system upgraded the fire protection capability of the plant fire protection system which itself is non-nuclear safety related.
i
- p. Fire Protection - Fire Doors & Window Shutters Fan & C'ntrol Room (M-622)
Fire dampers were installed on the control room windows and automatic fire doors were installed on the auxiliary building stack exhaust and carbon filter fan rooms. The fire dampers and doors were installed to meet NRC fire protection upgrade requirements.
Summary of Safety Evaluation: Not nuclear safety related.
- q. Blowdown Evaporator Bottoms Filter Vent & Drain (M-628)
Modifications were made to the filter vent and drain piping to improve filter housing drainage. The changes reduce contamination and personnel exposure when changing filters.
Summary of Safety Evaluation: Not nuclear safety related.
- r. Nitrogen Supply Line From Liquid Storage Tank (M-637)
The supply line from the liquid nitrogen storage tank to the nitrogen distribution header was changed from 51" tubing to 1" pipe to increase its capacity.
Summary of Safety Evaluation: Not nuclear safety related.
- s. Water Treatment Waste Sludge (M-640)
The modification installed new 4" carbon steel piping from the existing desludge line on the clarifier to a new mobile sludge tank such that it could be collected and transported by a contractor to a local landfill for disposal.
Summary of Safety Evaluation: Not nuclear safety related.
- t. Fire Protection - Platforms for Access to Foam Chambers on Fuel Oil Storage Tanks T32A&B (M-641)
Platforms were installed on each fuel oil storage tank to provide access to the fire suppression system foam chamber located on the outside of each tank.
Summary of Safety Evaluation: Not nuclear safety related.
- u. Secondary Sampling, Hard Pipe Secondary Sample Sink Drain (M-642)
The modification hard-piped the subject drain line to the clarifier sump. This modification replaced the arrangement which utilized a temporary hose.
Summary of Safety Evaluation: Not nuclear safety related.
- d. gg jagt,e Compactor (M-670) 1 The modification replaced the existing compactor with a new, more efficient unit which 3 should resuit in an annual waste :
reduction 4f K O ft ; this corresponding to an approximate 1 2-1/2 montJ: burie chace allocation at Barnwell. The existing unit was retained as a backup.
Summary of Safety Evaluation: Not nuclear safety related.
- w. Spent Resin Transfer Cask Service Air Supply (M-686)
The modification installed a removable section of hose in the service air supply to the resin transfer cask. The hose sec-
) tion is removed when the system is not-in use. This modifi-
! cation allayed NRC concerns as identified in .IE Information Notice No. 79-08 entitled, " Interconnection of Contaminated Systems With Service Air Systems Used as a- Source of Breathing Air."
Summary of Safety Evaluation: Not nuclear safety related.
- x. Hydrazine Injection Supply Manifold Valves (M-696)
!' The hydrazine injection pumps' suction piping manifold was modified to provide greater flexibility in aligning the hydrazine injection pumps to the hydrazine tanks.
Summary of Safety Evaluation: Not nuclear safety related.
i y. Radio Base Station Room Cooling (M-715) -
A duct was added off the Unit 2 electrical equipment room
, HVAC supply fan, 2-W51 to supply cooling air to the radio base station room. Air from the radio base station exhausts
- into the electrical equipment room.
I Summary of Safety Evaluation: Not nuclear safety related.
i
- z. RCC Change Tool Storage Bracket (M-716)
A support fixture was fabricated and installed in the spent fuel pit canal for storage of the new RCC change tool.
Summary of Safet'y Evaluation:
. Does not present any addi-tional threat- or unanalyzed problem. The mountings are l
designed to hold the tool in an area of the pool where fuel is not stored.
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, _ - - - - , . - -nn, - - - , - . -- . - - , , . _- .~ - - - , - .
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aa'. Emergency Diesel Generator 3D, Reversal of Flow (M-722) l ..
L " " '
9"
. The modification reversed ventilation air flow through the room for both fans such that outside air rather than turbine
- ~ hall air could be used for ventilating should a fire in the turbine hall preclude access to the diesel room.
Summary of Safety Evaluation: -Air flow reversal was per-formed on a test basis for 4D emergency diesel generator per modification request M-685 and WMTP 11.24. Results of that test were successful and, combined with engineering evalua-tions, verified that this. modification would not reduce the capability of the emergency diesel generator to perform its design function.
bb. Controlled Side Locker Expansion (M-724)
The modification extended the controlled side locker room into the present controlled side maintenance shop and ex-panded existing laundry room facilities. It also provided a new survey instrument calibration and source room as well as providing equipmen storage space on top of the new locker room extension.
Summary of Safety Evaluation: Not nuclear safety related.
ccI Provide Additional Access Ports for Multiple Source Gamma Calibrator (M-726)
W Two additional access ports were added to the gamma calibra-tor assembly to increase the number of calibration points for the Teletector radiation instrument. Plugs were also pro-vided to close the ports when not in use.
Summary of Safety Evaluation: Not nuclear safety related.
dd. Cafeteria E>.pansion (M-743)
Office , space adjacent to the cafeteria was converted to increase the capacity of the cafeteria to support the in-creasing number of plant and contractor personnel.
Summary of Safety Evaluationi Not nuclear safety related.
ee. Install Flexibl'e Boots Around Main Steam & Feedwater Lines in Auxiliary Puilding (Mc760)
Seals wer>. installed in the annular space around the main steam lines and main feedwater lines to prevent air flow from the controlled side of the plant to the clean side.
Summary of Safety Evaluation: Not nuclear safety related.
ff. Drum Positioners for Waste Compactor (M-764) i New positioning brackets were fabricated and installed in the
, waste compactor to eliminate problems with the vendor supplied brackets.
Summary of Safety Evaluation: Not nuclear safety related.
gg. Clean-Side Locker Expansion (M-766)
Offices adjacent to the clean-side locker room were converted to provide additional locker room space for increasing num-bers of plant and contractor personnel.
Summary of Safety Evaluation: Not nuclear safety related.
hh. Female Controlled-Side Locker Room Door' Relocation (M-770)
The entry door to the female controlled side locker room was moved from the south to the west wall to improve access to the locker room.
Summary of Safety Evaluation: Not nuclear safety related.
ii. Resin Sluice Header Waste Tank Bypass (M-772)
The resin sluice header to the waste tank in the waste soli-dification system was. modified to permit bypassing the tank and sluicing directly to a resin shipping container. The purpose of the modification was to minimize resin transfer operations by eliminating the need to transfer resin to and from the waste tank. Flush connections were also added to the resin sluice header.
Summary of Safety Evaluation: Not , nuclear safety related.
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[ 3.5 Procedure Changes These changes were made to start the implementation of NUREG-0737, I.C.I requirements as owners group and staff reviews and input are available. These changes do not represent completion of this project. The changes described below represent a general upgrading of format with additional input and clarification as noted. The acci-dents as analyzed in the FFDSAR remain as the basis for these procedures and are written to aid in the mitigation of the appro-priate conditions.
3.5.1 E0P-1A, Loss of Reactor Coolant, Revision 18, 05-18-81 The procedure was reformatted for consistency with other E0P's and editorially upgraded to provide better clarifi-cation to operating personnel. Guidance for personnel entry into the primary auxiliary building following an accident was added. Containment spray actuation steps were rewritten to reflect modification of spray initiation logic. Pressure criteria for securing safety injection were modified to reflect unit operation at low pressure. NUREG-0578 criteria for containment isolation valve verification were added.
, 3.5.2 E0P-1A, Loss of Reactor Coolant, Revision 19, 07-15-81 The procedure was editorially upgraded to provide better clarification for operating personnel. A step was added addressing inadequate case cooling considerations.
3.5.3 E0P-2A, Loss of Secondary Coolant, Revision 11, 05-18-81 The procedure was reformatted for consistency with other E0P's and editorially upgraded to provide better clarifi-cation to operating personnel. Guidance was added for personnel entry into the primary auxiliary building if core damage has occurred and recirculation is necessary. Con-tainment spray actuation steps were rewritten to reflect modification of spray initiation logic. Pressure criteria for securing safety injection were modified to reflect unit operation at low pressure. NUREG-0578 criteria for contain-ment isolation valve verification were added.
3.5.4 E0P-2A, Loss of Secondary Coolant, Revision 12, 07-15-81 The procedure was editorially upgraded to provide better clarification to operating personnel. Two steps were revised to specify thermocouple readings to be averaged when monitoring reactor coolant system temperature. A step was added to ensure the letdown orifices are shut if charging pumps are lost to prevent letdown of additional water to the reactor cooalnt system. Steps for monitoring unaffected unit common systems which could be affected were added.
r
3.5.5 EOP 3A, Steam Generator Tube Rupture, Revision 12, 05-18-81 The procedure was reformatted for consistency with other E0P's and editorially upgraded to provide better clarifi-cation to operating personnel. Guidan e was added for personnel entry into the primary auxiliary building if core damage has occurred and recirculation is necessary. Con-tainment spray actuati:n steps were rewritten to reflect modification of spray initiation logic. Criteria for ter-mination of safety injection were changed to more accurately reflect actual conditions of a steam generator tuoe leak.
Per NUREG-0578 verification of containment isolation valves was added.
3.5.6 E0P-3A, Steam Generator Tube Rupture, Revision 13, 07-15-81 The procedure was editorially upgraded to provide better clarification to operating personnel. A step was added to identify under what conditions residual heat removal pumps can be secured. Steps for monitoring unaffected unit common systems were added.
3.5.7 E0P-4A, Reactor Coolant Leak, Revision 4, 05-18-81 The procedure was reformatted for consistency with other E0P's and editorially upgraded to provide better clarifi-cation to operating personnel. A step was added to the "immediate action" section to inform operators on what manual actions to take in the case of a possible reactor coolant pump seal leak.
3.5.8 E0P-4A, Reactor Coolant Leak, Revision 5, 07-15-81 The procedure was editorially upgraded to provide better clarification to operating personnel. Technical Specifi-cation references were added where appropriate as were specific instrument identification numbers.
3.5.9 E0P-5A, Emergency Shutdown, Revision 8, 05-18-81 The procedure was reformatted for consistency with other E0P's and editorially upgraded to provide better clarifi-cation to operating personnel.
3.5.10 E0P-5A, Emergency Shutdown, Rvision 9, 07-15-81 Several steps were arranged for clarification purposes and prioritization. A number of typographical errors were corrected.
1 3.5.11 E0P-58, Inadvertent Safety Injection, Revision 2, 05-18-81 The procedure was editorially upgraded to provide better clarif.' cation to operating personnel. A diagnosis section and an objective section were added. The "immediate action-manual" section was rewritten to enable verification of major parameters. Several action steps were rearranged to permit safety injection reset early to ensure against refueling water storage tank to boric acid storage tank water transfer. Criteria for safety injection termination were modified to reflect unit operation at low pressure conditions.
3.5.12 E0P-58, Inadvertent Safety Injection, Revision 3, 07-15-81 '
Several typographical errors were corrected as well as several steps being upgraded to enhance clarity. Checking pressurizer relief tank level was added to the operator checklists during verification of containment conditions.
Several steps were rearranged to provide better prioriti-zation of activities.
3.5.13 E0P-6A, Dropped Rod, Revision 5, 07-15-81 The procedure was updated to reflect recently approved changes to Technical Specification Section 15.3.10.
3.5.14 E0P-6B, Stuck Rod or Malfunctioning Position Indication, Revision 6, 07-15-81 The procedure was updated to reflect recently approved changes to Technical Specification 15.3.10.
3.5.15 E0P-6C, Uncontrolled Withdrawal of RCCA(s), Revision 4, 07-15-81 The procedure was rewritten to improve clarity and reform-atted to be consistent with other E0P's.
3.5.16 E0P-6D, Uncontrolled Insertion of RCCA's, Revision 4 07-15-81 The procedure was rewritten to improve clarity and was reformatted to be consistent with other E0P's.
3.5.17 E0P-7A, Loss of Outside AC Power, Revision 7, 05-18-81 The procedure was reformatted for consistency with other E0P's and editorially upgraded for clarification to oper-ating personnel. A step was added to reflect the need for operator concern of auxiliary feed pump bearing cooling flow during a loss of AC event. Thermal convection verification was moved from a subsequent action step to an immediate action step.
< 3.5.18 EOP-7A Loss of Outside AC Power, Revision 8, 07-15-81 .
I Numerous editorial changes were made for clarification purposes. Several typographical errors were corrected.
3.5.19 E0P-8A, High Reactor Coolant Activity, Revision 3, 07-15-81 The procedure was reformatted to be consistent with other E0P's. It was also editorially upgraded to provide clari-fication of items to operating personnel.
3.5.20 E0P-8B, Irradiated Fuel Handling Accident In Containment Revision 4, 07-15-81 The procedure was reformatted to be consistent with other E0P's. It was also editorially upgraded to- provide better clarification.
3.5.21 EOP-8C, Irradiated Fuel Handling Accident In Primary Auxiliary Building, Revision 1, 07-15-81 The procedure was reformatted to be consistent with other E0P's. It was also editorially upgraded to provide better clarification.
3.5.22 EOP-8D, Spent Fuel Shipping Cask Drop, Revision 1, 07-15-21 The procedure was reformatted to be consistent with other E0P's. It was also editorially upgraded to provide clari-fication.
3.5.23 E0P-11B, High Airborne Activity, Letdown Gas Stripper Building Ventilation Exhaust, Revision 2, 10-01-81 The procedure was editorially revised for clarification and was reformatted to be consistent with other E0P's.
3.5.24 E0P-12A, Oil & Hazardous Material Spill, Revision 4, 07-15-81 The procedure was editorially revised for clarification and was reformatted to be consistent with other E0P's. A number of steps were changed to reflect new EPA regulations.
4.0 NUMBER OF PERSONNEL AND MAN-REM BY W3RK GROUP AND JOB FUNCTION 4.1 1981 JOB FUNCTION NUMBER TOT AL RE'M REACTOR PERSONNEL PER WORK OPERATIONS & ROUTINE INSPECTION SPECIAL CROUP WhSTE 2100 mPem SURVEILIANCE MAINTENANCE e ACTIVITIES MAINTENANCE PROCESSING REFUELING
- 1. _ Company Employees Operations 65 76.705 52.745 0 12.338 0 8.868 2.754 Peak Maintenance and h=.r.tanance 80 98.294 0 13.688 20.336 37.695 0 26.575 Chemistry & Ifealth Physics 24 29.619 24.360 0 0 0 3.021 2.238 Reactor Engineering 3 1.037 0.239 0 0.305 0 0 0.493
[
Instrument & Control 9 5.265 0 0.917 0.147 2.172 0 2.029 1
Admini s t ra t ic'n , l D. Engineering, Quality [
t & Regulatory Services 12 10.002 1.342 l 0 R.426 0 0 0.234
- 2. Contract Workers and f others 431 346.736 0.703 0 96.364 249.669 0 0
[
TOTALS 624 567.658 79.389 14.605 137.916 289.536 11.889 34.323 l
_ _ - - .. - - = .
5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION 5.1 Unit 1 07-04-81 Outage on 07-04-81, Unit I was shut down for steam generator eddy. current inspection. The 2000 psid primary-to-secondary hydrostatic test condition was established while the unit was being cooled down. An 800 psig secondary-to primary leak check was performed on 07-06-81 and 07-07-81. Detailed inspections of the tubesheets with remote video equipment showed a total of 9 explosive plugs which were either wet, boric acid-coated, or dripping at a slow rate (less than 1 drop every 1 minutes). Of the 9 plugs, 5 had similar observations noted in previous outages. Based upon the low leak rate before shutdown (4 gallons per day) and potential steam generator demonstration sleeving during the Fall 1981 refueling outage, the plugs were not repaired during this outage. The fact that weld repair of an ex-plosive plug involves a relatively high personnel radiation exposure was also a major factor in deciding not to repair the explosive plugs at this time.
The eddy current inspection consisted of remote inspection of all readily accessible tubes to the first tube support plate. Of the 2,853 open tubes in the "A" steam generator, 2,814 were inspected and 2,816 of the 2,861 open tubes in the "B" steam generator were inspec-ted.
Four tubes with indications greater than 40% were plugged on 07-11-81. In addition, tube R06C43 in the "B" steam generator was also plugged. Correct plugging was visually verified the same day.
f Extent of-Eddy Current Inspection' "A" Steam Generator Inlet: 2,814 tubes inspected through first support; multifrequency "B" Steam Generator Inlet: 2,816 tubes inspected through first support:
multifrequency P.esults of Eddy Current Inspection UDI - Undefinable Indications l Indications "A" Steam Generator "B" Steam Generator
! Identified Hot Leg Hot Leg i <20% 2 0 20-29% 0 1 30-39% 1 1 l
40-49% 2 0 50-59% 0 1 60-69% 0 0 70-79% 0 0 80-89% 0 1 90-99% 0 0 i 100% 0 0 UDI 9 0
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"A" Steam Generator Hot Leg UDI - Undefinable Indication ATS - Above Tubesheet ATE - Above Tube End TTS - Top of Tubesheet Row -Column - De fect Origin Location Pluqqed 20 20 <20So OD 5" ATE No 15 29 43So OD 21" ATE Yes 23 36 49So OD 8" ATE Yes 10 54 32So OD TTS No 33 54 <20So OD '5" ATS No 5 7 UDI OD 8" ATE No 11 43 UDI OD 18-20" ATE No 23 43 UDI OD 5-7" ATE No 25 44 UDI OD 13-15" ATE No 25 47 UDI OD _3-18" ATE No 8 55 UDI OD 17" ATE No 15 60 UDI OD 14-18" ATE No 18 68 UDI OD 15-17". ATE No 18 37 UDI OD 8" ATE No "B" Steam Generator Hot Leg Row Column Defect Origin Location Plugged 3 25 80So OD 17" ATE Yes 27 30 29So OD '1" ATS No 6 43 33So OD 21" ATE Yes 16 47 52So OD '5" ATE Yes Unit 1 Refueling 9 Inservice Inspection On 10-09-81 Unit 1 was shut down for refueling. The 2000 psid primary-to-secondary hydrostatic test condition was established during cooldown of the unit. An 800 psig secondary-to-primary leak check was performed in the "B" steam generator on 10-24-81. A.
similar test in the "A" steam generator was performed after sleeving was completed on 11-13-81 and no levels were ..ated. Detailed in-spection of the "B" steam generator tubesheet with remote video equipment showed a total of 7 explosive plugs which were either wet, coated with boric acid, or dripping at a slow rate (2 to 3 drops per minute). Of the 7 plugs, 2 had similar observations noted in previous outages. Based on the low primary-to-secondary leak rate before shutdown (less than 10 gallons per day), the high personnel radiation exposure required 'for weld repair, and potential future sleeving of tubes in the "B" steam generator, the one dripping plug was not repaired during this outage.
l The eddy current inspection programs for the steam generators con-sisted of inspections of all previously degraded tubes through the U-bend in each steam generator, inspection of 3% of the tubes through the U-bend in "A" steam generator and full-length inspection of 3% of the tubes in the "B" steam generator, inspection of essentially all tubes in each steam generator through the first support plate on the hot leg side and full-length inspection of two tubes in the "A" steam generator that exhibited cold leg indications in previous eddy current examinations.
Restrictions with the 0.720" and the 0.700" eddy current probes were encountered in both steam generators. Twenty-seven of the 32 re-stricted tube ends encountered with the 0.720 probe in the "A" steam generator during this inspection were not noted in either the 07-81 or 12-80 inspections. The new restrictions are believed to have been caused by residue from the channelhead decontamination process per-formed on 10-24-81 to 10-26-81. In the "B" steam generator, most of the tubes found restricted at the first support plate with the probes were also noted as restricted in the 07-81 and 12-80 inspections.
Plugging of 16 of the 17 tubes with indications greater than 40% was completed by 11-14-81. One tube in the "A" steam generator, R18C68, which was found to have an indication greater than 40% was sleeved as part of a sleeving demonstration program. A total of 12 tubes were sleeved in the "A" steam generator during refueling 9, 6 of which had indications greater than 40%.
Extent of Eddy Current Inspection
, i "A" Steam Generator Inle': 2,641 tubes inspected through first support; multifrequency 88 tubes inspected through U-bend; multifrequency 33 tubes inspected full-length; multifrequency "A" Steam Generator Outlet: 33 tubes inspected from hot leg side "B" Steam Generator Inlet: 2,692 tubes inspected through first support; multifrequency
, 3 tubes inspected through U-bend; i
multifrequency 112 tubes inspected full-length; multifrequency "B" Steam Generator Outlet: 112 tubes inspected from hot leg side Results of Eddy Current Inspection Indications "A" SG "A" SG "B" SG Identified Hot Leg Cold Leg Hot Leg
<20% 3 0 0 20-29% 1 1- 3 30-39% 2 1 0 40-49% 0 0 1 50-59% .3 0 0 60-69%- 1 0 1 70-79% 4 0 3 80-89% 1 0 1 90-99% 1 0 1 100% 0 ,- 0 0 UDI 15 0 1 "A" Steam Generator Hot Leg Row Column Defect Origin Location Plugged 5 7 UDI OD 15" ATE No 36 29 38% OD TTS No 31 37' UDI OD 11". ATE No 18 37 UDI OD 8" ATE No 23 38 UDI OD 20" ATE No 27 38 UDI OD 20" ATE No 30 39 56% OD 11" ATE Yes-12 41 tDI OD 12" ATE No 23 42 70% OD 20" ATE Yes 23 43 67% OD 5-7" ATE Yes 11 '43 UDI OD 18-20" ATE No 25 44 UDI OD 150 ATE Sleeved 10 40 UDI OD Roll to TTS No 10 21 90% OD 15" ATE Yes 20 20 50% OD 5" ATE Yes 15 20 UDI OD 8-14" ATE No 25 47 57% OD .18" ATE Yes 29 47 UDI OD 8-14" ATE No 33 54 38% OD " ATS Sleeved 10 54 <20% OD TTS No 8 55 UDI OD 17" ATE Sleeved 15 59 UDI OD 7-20" ATE No 15 60 UDI OD .14-18" ATE No 8 64 77% OD 12" ATE Yes 23 67 73% OD 8" ATE Yes 18 68 80% OD 15-17" ATE Sleeved 15 68 UDI OD 10-15" ATE No 5 68 21% OD " ATS No 5 69 <20% OD %" ATS No 11 74 77% OD 15-18" ATE Yes 6 81 <20% OD 1" ATS No t .,
l "A" Steam Generator Cold Leg l Row Column Defect Origin Location Plugged l'
, 28 48 35% OD 2" ATS No 2 13 80% OD 6" ATE Yes 2 15 UDI OD 6-11" ATE No 18 27 64% OD Top of Roll Yes 20 28 77% OD 3-17" ATE Yes 27 30 25% OD 1" ATS No 14 40 29% OD 1" ATS No 26 42 29% OD 1" ATS No 25 46 95% OD 8" ATE Yes 26 47 44% OD 8-20" ATE Yes 27 52 70% OD 20" \TE Yes 11 78 75% OD 8" ATE Yes "B" Steam Generator Hot Lej Row Column Defect Origin Location Plugged 2 13 80% OD 6" ATE Yes f UDI OD 6-11" ATE No
! 2 15 18 27 64% OD Top of Roll Yes 20 -28 77% OD 3-17" ATE Yes 27 30 25% OD 1" ATS No 14 40 29% OD 1" ATS No 26 42 29% OD 1" ATS No 25 46 95% OD 8" ATE Yes 26 47 44% OD 8-20" ATE Yes 27 52 70% OD 20" ATE Yes 11 78 75% OD 8" ATE Yes "A" Steam Generator Sleeved Tubes Row Column Indication Date Plugged Type 25 27 None 03/79 Explosive 8 75 89%, 20" ATE 07/80 Mechanical 9 72 76%, 20" ATE 07/80 Mechanical-18 68 47%, 18-20" ATE 07/80 Mechanical l Not Plugged 28 58 None N/A i 10 66 None Not Plugged N/A 26 53 83%, 21" ATE 07/80 Mechanical 27 64 51%, 3" ATE 07/80 Mechanical-
! 8 55 UDI Not Plugged N/A 25 44 UDI Not Plugged N/A 33 54 38%, " ATS Not Plugged N/A 27 49* Leaker 03/79 Explosive
- Sleeved in hot and cold leg. All other tubes were sleeved in the cold leg only.
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5.2 Unit 2 Unit 2 Refueling 7 Outage Inservice Inspection On 04-20-81 an 800 psig secondary-to-primary leak check was performed in each steam generator. Detailed inspections of the tubesheets with remote television equipment showed leakage from the explosive plug in tube R32C15 in the "A" steam generator. The leakage rate was about
, 2 drops per minute. Another plug in the "A" steam generator, R31C52, was heavily coated with boric acid but no water was present. 'After considering the location of the leaking plug, which is in the tube-sheet periphery, and the effect that repair of the plug would have on exposure, critical' path, and problems associated with repairs in the area, the decision was made not to repair the plug during this outage. An additional consideration was the fact that the primary-to-secondary leakage rate in the steam generator was only 1 gallon per day before the outage.
The initial eddy current inspection programs for the "A" and "B" steam generators consisted of inspection through the U-bend of 3% of the tubes in each steam generator plus all previously degraded tubes, in accordance with Technical Specification requirements. Addi-tionally, the "A" steam generator program included a full-length inspection for a previous indication in the cold leg and inspections through the U-bend of about 190 tubes in connection with tube degra-dation at contact with anti-vibration bars (AVB) reported by other plants. The program for the "B" steam generator inlet included inspecting 33 previously degraded tubes through the U-bend and 172 randomly located tubes to meet Technical Specification requirements and for AVB tube degradation. The program for the "B" steam gener-ator outlet consisted of inspection through the first support plate of all previously degraded tubes, inspection through the first support plate of about 200 tubes in the problem areas determined by previous inspections, and inspection through the third support plate of about 170 tubes around the periphery. The programs in the inlets of both steam generators were later expanded in accordance with Technical Specifications, resulting in ar. inspection of essentially all tubes in the "A" steam generator through the first support plate and approximately 75% of the tubes in the "B" steam generator through the first support.
Results of the eddy current inspections showed 25 pluggable tubes in the "A" steam generator and 16 pluggable tubes in the "B" steam generator. One of the. tubes in the "A" steam generator, R15C73, was pulled for detailed analysis and the hole was weld plugged on 04-30-81. A degraded tube in the "B" steam generator, R24C25, had interference preventing insertion of a mechanical plug. The tube entrance area was rerolled and then successfully plugged. Plugging of all tubes was completed on 04-30-81. Photographs of the tube-sheets taken later the same day verified plugging of the proper tubes.
4 Extent of Eddy Current Inspection "A" Steam Generator Inlet: 491 tubes inspected through U-bend; multifrequency 2693 tubes inspected through first support; multifrequency 4
1 tube inspected for full length; multifrequency "A" Steam Generator Outlet: 1 tube inspected from hot leg side i "B" Steam Generator Inlet: 208 tubes inspected through U-bend; multi-frequency 2061 tubes inspected through first support; I
multifrequency
- "B" Steam Generator Outlet
- 307 tubes inspected through first support;
- multifrequency 163 tubes inspected through thirds support; multifrequency Results of Eddy Current Inspection Indications "A" SG "A" SG "B" SG "B" SG Identified Hot Leg Cold Leg Hot Leg Cold Leg
<20% 309 1 81 253 20-29% 150 0 62 30 30-39% 123 0 60 1 40-49% 20 0 11 0 50-59% 5 0 3 0 l, 0 0 0 60-69% 2
, 70-79% 0 0 0 0 l 80-89% 0 0 0 0
! 90-99% 2 0 0 0
! 100% 0 0 0 0 i
l' 1
I l
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"A" Steam Generator Hot Leg ATE - Above Tube End ATS - Above Tubesheet TTS - Top of Tubesheet Row Column Defect Origin Location Plugged 18 5 21% OD 1st Support No 19 5 35% OD 1st Support No 20 6 32%/38% OD 1st Support /-
2nd Support No 23 8 36 OD 1st Support No 4 16 <20% OD TTS No 7 16 <20% OD TTS No 8 16 <20% OD TTS No 2 17 <20% OD TTS No 3 17 <20% OD TTS- No 4 17 <20% OD TTS No 5 17 <20% OD TTS No 6 17 <20% OD TTS No 7 17 <20% OD TTS No 8 17 <20% OD TTS No 9 17 <20% OD TTS No 3 18 <20% OD !s"ATS No 6 18 <20% OD- TTS No 8 -18 <20% OD TTS No 11 18 35% OD TTS No 3 19 <20% OD TTS No 4 19 22% OD TTS No 6 19 <20% OD TTS No 7 19 <20% OD TTS No 9 19 21% OD TTS No 10 19 21% OD TTS No 11 19 35% OD TTS No 12 19 <20% OD TTS No-4 20 <20% OD TTS No l 5 20 31% OD TTS No
- 6 20 OD TTS No
<20%
- 7 20 <20% OD TTS No
11 20 <20% OD TTS No j 12 20 <20% OD TTS No
} 13 20 22% OD TTS No i 14 20 <20% OD TTS No j 5 21 39% OD TTS No
- 6 21 21% OD TTS No
! 8 21 36% OD TTS No 10 21 <20% OD TTS No j 12 21 <20% OD TTS No 1 14 21 <20% OD TTS No 16 21 <20% OD TTS No i
Row Column Defect Origin Location Plugged 17 21 <20%_ OD TTS No
- 18 21 <20% OD TTS No 3 22 33% OD TTS No 4 22 <20% OD TTS No 8 22 33% OD TTS No 9 22 <20% OD TTS No 12 22 52% OD TTS Yes 13 22 32% OD .TTS No 17 22 <20% OD TTS No 18 ' 22 <20% OD- TTS No 20 ' 22 <20% OD TTS No 23 22 <20% OD TTS No 26 22 <20% OD TTS No i 3 23 39% OD TTS No 4 23 32% OD TTS No 5 23 21% OD TTS No 7 23 <20% OD TTS No 8 23 39% OD TTS No 10 23 <20% OD TTS No 11 23 <20% OD TTS No 12 23 21% OD TTS No 13 23 <20% OD TTS No-14 23' <20% OD TTS No 16 23 <20% OD TTS No
) 17 23 <20% OD TTS No 18 23 27% OD TTS No 19 23 38% OD TTS No 20 23 22% OD TTS No 21 23 30% OD TTS No 22 23 <20% OD TTS No 23 23 21% OD TTS No 24 23 <20% OD TTS No 25 23 <20% OD TTS No 26 23 <20% OD TTS No
- 5 24 21% OD TTS No
, 6 24 <20% OD TTS No i 7 24 31% OD TTS No 10 24 44% OD TTS Yes
[ 11 24 36% OD TTS No l
. 12 24 21% _OD TTS No 13 24 36% OD TTS No 16 24 25% OD TTS No 17 24 <20% OD TTS No 18 24 32% OD TTS No 19 24 38% OD TTE No 20 24 41% OD TTS Yes 21 24 29% OD TTS No 22 24 <20% OD TTS No l
j 24 24 <20% OD TTS No 6 25 35% OD TTS No 7 25 <20% OD TTS No l TTS No
/ 8 25 25% OD i 9 25 35% OD TTS No i
i i
,. - ,--n - w,--,-ra- ,-..-,.~-,_r,..- -
Row Column Defect Origin Location Plugged 10 25 38% OD TTS No 11 25 37% OD TTS No 12 25 21% OD TTS No 13 25 38% OD TTS No 14 25 22% OD TTS No 15 25 <20% OD TTS No 16 25 <20% OD TTS No 17 25 24% OD TTS No 18 25 35% OD TTS No 19 25 35% OD TTS No 20 25 38% OD TTS No 21 25 21% OD TTS No 22 25 <20% OD TTS No 4 26 <20% OD TTS No 6 26 22% OD TTS No 7 26 <20% OD TTS No 8 26 <20% OD TTS No 9 26 25% OD TTS No 10 26 25% OD TTS No 11 26 24% OD TTS No 12 26 22% OD TTS No 13 26 25% OD TTS No 14 26 <20% OD TTS No 15 26 <20% OD TTS No 16 26 <20% OD TTS No 17 26 40% OD TTS No 18 26 40% OD TTS No 19 26 38% OD TTS No l 20 26 34% OD TTS No 21 26 <20% OD TTS No f 22 26 <20% OD TTS No
! 23 26 <20% OD TTS No 4 27 39% OD TTS No 5 27 25% OD TTS No 6 27 30% OD TTS No 7 27 32% OD TTS No 8 27 26% OD TTS No 9 27 <20% OD TTS No 10 27 <20% OD TTS No 11 27 <20% OD TTS No 12 27 32% OD TTS No 13 27 <20% OD TTS No 14 27 22% OD TTS No 15 27 25% OD TTS No 16 27 36% OD TTS No 17 27 36% OD TTS No 18 27 35% OD TTS No 19 27 36% OD TTS No 20 27 36% OD TTS No 21 27 29% OD TTS No 22 27 <20% OD TTS No 26 27 29% OD TTS No Row Column Defect Origin Location Pluqqed 5 28 <20% OD TTS No 6 -28 33% OD TTS No I
7 28 32% OD TTS No1 8 28 25% OD TTS No 10- 28 27% OD TTS No 11 28 <20% OD TTS No
'12 28 No
<20% OD TTS 13 28 <20%. OD- TTS No 14 28 <20% OD TTS No
- j. 15 28 <20% OD TTS No t
16 28 24%- OD TTS No '
17 28 28% OD TTS No 18 28 22% OD TTS No 19 28 33% OD TTS No 20 28 29% OD TTS No
8 29' 21% OD TTS No 10 29 <20% OD TTS No 11 29 <20% OD TTS No
- 12 29 <20% OD TTS No
- 14 29 <20% OD TTS No 18 29 35% OD TTS No 19 29 55% OD TTS Yes 4 30 <20% OD TTS No 4 6 30 <20% OD TTS No 8 30 32% OD TTS No 11 30 <20%. OD TTS No 12 30 33% OD TTS ENo 13 30 38% OD TTS No 21 30 21% OD TTS No 22 30 <20% OD TTS No 23 30 <20% OD TTS 'No
> 24 30 <20% OD TTS No i 25 30 <20% OD TTS No
! 4 31 <20% OD TTS No 8 31 <20% OD TTS No
- 13 31 22% OD TTS No
. 17 31 21% OD TTS No-l 26 31 59% OD b" ATS Yes 8 32 <20% OD TTS No 12 32 21% OD TTS No j 13 32 <20% OD TTS No l 14 32 <20% OD TTS No-
! 27 32 38% OD TTS No 6 33 <20% OD ' TIS No 11 33 <20% OD TTS No l
l 12 33 38% OD TTS No
! 13 33 <20% OD TTS No
! 17 33 95%/26% OD 6" ATE /TTS Yes ,
i l
L N h
I j Row Column Defect Origin ' Location Pluqqed 24 33 -<20% OD TTS No i 10 34 <20% OD TTS No i 12 34 39% OD TTS No 20 34 28% OD TTS No j 21 34 <20%/38% OD '1" ATS/TTS No 27 34 23% OD TTS No
. 11 35 <20% OD TTS No
! 12 35 31% OD TTS No 13 35 <20% OD TTS No 21 35 28% OD TTS No 23 35 <20% OD TTS No 26 35 25% OD TTS No 9 36 <20% OD TTS No 10 36 <20% OD TTS No l 11 36 <20% OD TTS No l 12 36 25% OD TTS No 16 36 <20%' OD TTS No 21 36 <20% OD TTS No-8 37 <20% OD TTS No 9 37 <20% OD TTS No
! 10 37 21% OD TTS No 11 37 26% OD TTS No 12 37 21% OD TTS No
, .13 37 23% OD TTS No i 14 37 26%' OD TTS No
- 15 37 <20% OD TTS No
! 26 37 <20% OD TTS No
! 8 38 <20% OD TTS No I
9 38 <20% OD TTS No 10 38 <20% OD TTS No 13 38 <20% OD TTS No 16 38 36% OD TTS No 20 38 <20% OD '1" ATS No 8 39 <20% OD TTS No 9 39 32% OD TTS No 11 39 <20% OD TTS No 12 39 <20% OD TTS No 16 39 <20% OD TTS No 17 39 39% OD TTS No 19 39 92% OD 9-13" ATE Yes 22 39 <20% OD TTS No 29 39 <20% OD TTS No 31 39 <20% OD TTS No 7 40 <20% OD TTS No 8 40 <20% OD TTS No 9 40 <20% OD TTS No 10 40 <20% OD TTS No 11 40 26% OD TTS No 12 40 <20% OD TTS No 13 40 <20% OD TTS No 16 40 <20% OD TTS No 20 40 <20% OD TTS No 21 40 <20% OD TTS No 25 40 <20% OD TTS No
Row Column Defect Origin Location Plugged 31 40 30% OD TTS No 32 40 <20% OD TTS No 6 41 <20% OD TTS No 8 41 <20% OD TTS No 10 41 <20% OD TTS No 11 41 <20% OD TTS No 12 41 46% OD TTS Yes 16 41 <20% OD TTS No 19 41 33% OD TTS No 20 41 41% OD. TTS Yes 23 41 45% OD TTS Yes 24 41 25% OD TTS No 25 41 25% OD TTS No 31 41 <20% OD TTS No 32 41 <20% OD TTS No 6 42 <20% OD TTS No 8 42 <20% OD TTS No 12 42 36% OD TTS No 13 42 38% OD TTS No 17 42 29% OD TTS No 20 42 39% OD TTS No 22 42 <20% OD TTS No 23 42 27% OD TTS No 24 42 21% OD TIS No 25 42 <20% OD TTS No 26 42 <20% OD TTS No 6 43 <20% OD TTS No 7 43 <20% OD TTS No 8 43 <20% OD TTS No ,
12 43 42% OD TTS Yes 19 43 36% OD TTS No 22 43 28% OD TTS No 23 43 (20% OD TTS No 25 43 <20% OD TTS No 26 43 <20% OD TTS No 27 43 28% OD TTS No 28 43 <20% OD TTS No 6 44 <20% OD TTS No 8 44 <20% OD TTS No 9 44 <20% OD TTS No 10 44 22% OD TTS No 13 44 45% OD TTS Yes 19 44 51% OD TTS Yes-20 44 33% OD TTS No 21 44 51% OD TTS Yes 22 44 49% OD TTS Yes 24 44 <20% OD TTS No 25 44 <20% OD TTS No 26 44 <20% OD TTS No 27 44 <20% OD TTS No 28 44 25% OD TTS No 29 44 39% OD TTS No i
l Row Column Defect Origin Location Pluqqed 31 44 36% OD TTS No 6 45 <20% OD TTS No 7 45 <20% OD TTS No 8 45 21% OD TTS No 9 45 32% OD TTS No 10 45 43% OD TTS Yes 11 45 41% OD TTS Yes 12 45 28% OD TTS No 23 45 47% OD TTS Yes 24 45 32% OD TTS No 25 45 <20% OD TTS No 27 45 <20% OD TTS No 29 45 <20% OD TTS No 31 45 <20% OD TTS No 5 46 25% OD TTS No 6 46 <20% OD TTS No i 9 46 <20% OD TTS No
< 10 46 36% OD TTS No I
20 46 36% OD TTS No 21 46 36% OD TTS No 23 46 <20% OD TTS No 24 46 <20% OD TTS No 29 46 <20%- OD TTS No 6 47 22% OD TTS No 8 47 <20% OD TTS No 10 47 <20% OD TTS No 15 47 <20% OD '1" ATS No 19 47 <20% OD '5" ATS No 21 47 26% OD TTS No 32 47 <20% OD TTS No 5 48 <20% OD TTS No 7 48 <20% OD TTS No 9 48 39% OD TTS No 13 48 <20% OD. TTS No 31 48 <20% OD TTS No 32 48 25% OD TTS No 6 49 <20% OD TTS No 7 49 <20% OD TTS No 8 49 <20% OD TTS No 11 49 22% OD TTS No 12 49 25% OD TTS No 13 49 35% OD TTS No 14 49 <20% OD TTS No 32 49 36% OD TTS No 33 49 43% OD TTS Yes 34 49 <20% OD TTS No 5 50 <20% OD TTS No 6 50 <20% OD TTS No 7 50 <20% OD TTS No 9 50 <20% OD TTS No 10 50 26% OD TTS No 11 50 25% OD TTS No
[. .- -
j
- t Row Column Defect Origin Location Pluqqed 12 50 25% OD TTS No 13 50 <20% OD TTS No 29 50 <20% OD TTS No 8 51 32% OD TTS No i 10 51 27% OD TTS No
! 20 51 28% OD TTS No 22 51 26% OD TTS No i -23 51 26% OD TTS No 24 51 <20% OD TTS No 9 52 <20% OD TTS No i 11 52 <20% OD TTS No
. 14 52 <20% OD TTS No 19 52 22% OD TTS No 20 52 22% OD TTS No j 21 52 32% OD TTS No 22 52 21% OD TTS No 23 52 28% OD TTS No 24 52 22% OD TTS No-6 53 <20% OD TTS No 10 53 <20% OD TTS No 11 53 <20% OD TTS No 12 53 -<20% OD TTS No 19 53' 26% OD .TTS No 20 53 21% OD TTS No j 21 53 30% OD TTS No 22 53 26% OD TTS No 23 53 28% OD TTS No 24 53 35% OD TTS No 26 53 <20% OD TTS No 29 53 <20% OD TTS No 30 53 ' <20% OD TTS No 31 53 <20% OD TTS No 32 53 <20% OD TTS No 6 54 <20% OD TTS No 11 54 <20% OD TTS No 12 54 36% OD TTS No 13 54 <20% OD TTS No 18 54 22%' OD TTS No 19 54 25% OD TTS No 20 54 35% OD TTS No 21 54 28% OD TTS No 22 54 28% OD TTS No 24 54 36% OD TTS No 30 54 <20% OD TTS No 7 55 <20% OD TIS No 8 55 <20% OD TTS No 12 55 <20% OD TTS No 13 55 <20% OD TTS No f
l l
l l
Row Column Defect Origin Location Plugged 19 55 25% OD TTS No 20 55 39% OD TTS No 21 55 37% OD TTS No 22 55 31% OD TTS No 24 .55 31% OD TTS No 25 55 42% OD TTS Yes 30 55 25% OD \" ATS No 8 56 <20% OD TTS No 9 56 <20% OD TTS No 10 56 28% OD TTS No 12 56 25% OD TTS No 18 56 25% OD TTS No I 19 56 22% OD TTS No 20 56 25% OD TTS No
( 21 56 28% OD TTS No 22 56 30% OD TTS No 23 56 30% OD TTS No 24 56 36% OD TTS No 25 56 <20% OD TTS No 29 56 <20% OD TTS No 9 57 35% OD TTS No 11 57 <20% OD \" ATS No 12 57 32% OD TTS No 3
- . 18 57 <20% OD TTS No 19 57 26% OD TTS' No 20 57 32% OD TTS No 23 57 28% OD TTS No 24 57 <20% OD TTS No 9 58 <20% OD TTS No 11 58 <20% OD \" ATS -No 12 58 <20% OD TTS No 17 58 <20% OD TTS No 18 58 <20% OD TTS No 19 58 32% OD TTS No 20 58 25% OD TTS No 22 58 25% OD TTS No 23 58 26% OD TTS No 26 58 <20% OD TTS No 8 59 <20% OD TTS No 9 59 <20% OD TTS No 12 59 <20% OD TTS No 17 59 <20% OD TTS No 18 59 <20% OD TTS No 19 59 22% OD TTS No 22 59 28% OD TTS No 23 59 28% OD TTS No 24 Ro 39% OD TTS No 27 59 <20% OD TTS No 7 60 <20% OD TTS No 16 60 <20% OD TTS No 19 60 31% OD TTS No 20 60 33% OD TTS No 21 60 33% OD TTS No 22 60 36% OD TTS No Row Column- Defect Origin Location Plugged 23 60 <20% OD TTS No 25 60 <20% OD TTS No 26 60 <20% OD TTS No 9 61 <20% OD TTS No 10 61 36% OD TTS No 11 61 33% OD TTS No 12 61 21% OD TTS No 17 61 ' <20% OD TTS No 18 61 <20% OD TTS No 19 61 26% OD TTS No 21 61 30% OD TTS No 22 61 34% OD TTS No 23' 61 22% OD TTS No 24 61 <20% OD TTS No 25 61 <20% OD TTS No 26 61- <20% OD TTS No 6 62 <20% OD TTS No 7 62 <20% OD TTS No 9 62 <20% OD TTS No 10 62 39% OD TTS No 11 62 39% OD TTS No 12 62 <20% OD TTS No 14 62 <20% OD TTS No 16 62 <20% OD TTS No 17 62 25% OD TTS No 18 62 28% OD TTS No 19 62 35% OD TTS No.
20 62 39% OD TTS No 21 62 47% OD TTS Yes 23 62 <20% OD TTS No 24 62 <20% OD TTS No 25 62 <20% OD ITS No 10 63 25% OD TTS No 12 63 <20% OD TTS _No 16 63 <20% OD TTS No 17 63 26% OD TTS No 18 63 21% OD TTS No 19 63 21%/25% OD 17" ATE /TTS No 22 63 28% OD TTS No 23 63 <20% OD TTS No 24 63 <20% OD TTS No 9 64 28% OD TTS No 12 64 <20%' OD TTS No 13 64 <20% OD TTS No 17 64 <20% OD TTS No 19 64 OD TTS No (20%
20 64 36% OD TTS No 22 64 <20% OD I-IS No 10 65 <20% OD TTS No 11 65 <20% OD TTS No 12 65 21% OD TTS No 13 65 <20% OD TTS No 15 65 <20% OD TTS No 20 65 39% OD TTS No l
l I l l
l l
l Row Column Defect Origin Location Plugged 21 65 31% OD TTS No
- 22 65 <20% OD TTS No l 8 66 <20% OD TTS No i 9 66 <20% OD TTS No i 11 66 <20% OD TTS No i 12 66 31% OD TTS No 13 66 21% OD TTS No 14 66 <20% OD TTS No 15 66 <20% OD TTS No 16 66 24% OD TTS No 17 66 <20% OD TTS No 19 66 46% OD TTS Yes
. 20 66 36% OD TTS No 21 66 24% OD TTS No t
8 67 21% OD TTS No 12 67 21% OD TTS No 13 67 <20% OD TTS No 17 67 25% OD TTS No 19 67 30% OD TTS No 20 67 21% OD TTS No 21 67 <20% OD TTS No 6 68 <20% OD TTS No 7 68 <20% OD TTS No 8 68 <20% OD TTS No 12 68 21% OD TTS No 13 68 25% OD TTS No 15 68 <20% OD TTS No-16 68 <20% OD TTS No 17 68 <20% OD TTS No 19 68 28% OD TTS No 20 68 33% OD TTS No 6 69 <20% OD TTS No 7 69 <20% OD TTS No 12 69 24% OD TTS No 14 69 25% OD TTS No 16 69 <20% OD TTS No 17 69 26% OD TTS No 18 69 <20% OD TTS No 19 69 27% OD TTS No 12 70 26% OD TTS No 13 70 <20% OD TTS No i
14 70 <20% OD TTS No 16 70 <20% OD TTS No 17 70 33% OD TTS No 18 70 30% OD TTS No 19 70 <20% OD TTS No 10 71 <20% OD TTS No 11 71 <20% OD TTS No 12 71 41% OD TTS Yes 13 71 <20% OD TTS No 16 71 <20% OD TTS No 17 71 41% OD TTS Yes 18 71 22% OD TTS No 11 72 37% OD TTS No
Row Column Defect Origin Location Pluqqed 12 72 25% OD TTS No 13 72 <20% OD TTS No 14 72 <20% OD TTS No 17 72 <20% OD- TTS No 4 73 <20% OD %" ATS No 8 73 <20% OD TTS No 9 73 <20% OD TTS No 13 73 36% OD TTS No 14 73 31% OD TTS No 15 73 41% OD TTS Yes 17 73 <20% OD TTS No 9 74 <20% OD TTS No 12 74 21% OD TTS No 13 74 31% OD TTS No
'4 75 <20% OD TTS No 5 75 <20% OD TTS No 10 75 38% OD TTS No 11 75 <20% OD TTS No 12 75 26% OD TTS' No 3 76 <20% OD TTS No 4 76 28% OD TTS No 5 76 <20%. OD TTS_ No 6 76 32% OD TTS No 7 76 38% OD TTS No 8 76 26% OD TTS No 10 76 25% OD TTS No
-11 76 25% OD TTS No-13 76 26% OD TTS ~No 3 77 28% OD 12S No 4 77 24% OD TTS No 8 77 35% OD TTS No 9 77 30% OD TTS No 10 77 <20% OD TTS No 10 78 27% OD TTS No 4 79 <20% OD TTS- No 10 79 <20% OD TTS No 9 80 25% OD TTS No "A" Steam Generator Cold Leg Row Column Defect Origin Location Plugged 12 42 <20% OD 2" ATS No 4
"B" Steam Generator Hot Leg Row Column Defect Origin Location Pluqqed l 6 16 22% OD TTS No 5 17 <20% OD TTS No 6 17 60% OD TTS 'Yes 7 17 66% OD TTS Yes i 5 18 25% OD TTS No 6 18 41% OD TTS .Yes 7 18 24% OD TTS No I 10 18 <20% OD TTS No 11 18 22% OD TTS No 6 19 41% OD TTS Yes
! 7 19 <20% OD TTS No 8 19 <20% OD TTS Yes
! 6 20 46% OD TTS No 6 21 28% OD TTS No 17 21 <20% OD TTS No 6 22 46% OD TTS Yes 10 22 <20% OD TTS No 14 22 46% OD TTS Yes 6 23 36% OD TTS No 14 23 '37% OD TTS No 18 23 <20% OD TTS No 14 24 30% OD TTS No 17 24 30% OD TTS No 18 24 <20% OD TTS No 19 24 21% OD TTS No 24 24 21% OD TTS No l
9 25 24% OD TTS No 14 25 22% OD TTS No 17 25 22% OD TTS No j 22 25 38%/46% OD TTS Yes 23 25 39% OD TTS No 24 25 46% OD TTS Yes 25 25 21% OD TTS No 26 25 41% OD TTS Yes 9 26 29% OD %" ATS No 21 26 <20% OD TTS No 23 26 36% OD TTS No 24 26 33% OD TTS No 25 26 33% OD TTS No 26 26 41% OD TTS Yes 7 27 <20% OD " ATS No 18 27 30% OD TTS No 19 27 26% OD TTS No Row Column Defect Origin Location- Plugged ,
- j 21 27 30% OD TTS No 22 27 28% OD TTS No 23 27 21% OD TTS No 26 27 21% OD 'TTS No 27 27 31% OD TTS No-13 28 <20% OD TTS No 19 28 32% OD TTS No 20 28 25% OD TTS No 21 28 <20% OD TTS No l 26 28 38% OD TTS No 27 28 37% OD TTS No 6 29 22% OD TTS No 10 29 <20% OD TTS No 13 29 <20% OD TTS No 18 29 <20% OD TTS No 22 29 44%' OD TTS Yes 23 29 28% OD TTS . No 24 29 23% OD .TTS No i 26 29 36% OD TTS No ,
27 29 39% OD TTS No i 6 30 28% OD TTS No 11 30 <20% OD TTS No 12 30 <20% OD TTS No '
16 30 37% OD TTS No .u 22 30 34%/28% OD 1" ATS/TTS No 27 30 39% OD TTS No 6 31 <20% OD TTS No 7 31 <20% OD TTS No 21 31 21% OD TTS No l 15 32 54% OD 'TTS Yes
- 20 33 <20% OD TTS No i
21 34 35% OD \" ATS -No 5 36 <20% OD 'TTS No 5 37 <20% OD TTS No 12 37 <20% OD TTS No 14 38 34% OD TTS No 28 38 22% OD TTS No
- 29 39 22% OD TTS No
- 32 39 22% OD TTS No
32 40 21% OD TTS No 26 41 <20% OD TTS No i 27 41 <20% OD TTS No 28 41 25% OD TTS No
~l
I
~t A f Ro'w' Sholumn Defect Origin Location Pluqqed 6 42 <20% OD TTS No 9 42 22% OD TTS No 19 42 27% OD " ATS No
.- 28 42 <20% OD TTS No
- )" 31 42 27% OD %" ATS No 32 42 34% OD TTS No 19' 43 31% OD TTS No 21 43 38% OD TTS No 24 43 27% OD TTS No
, ' 2 6' 43 <20% OD TrS No 5 44 <20% OD TTS No 6 44 <20% OD TTS No 19 44 30% OD TTS No 21 44 34% OD TTS No 22 ~ 44 30% OD TTS No
, ; ,c I -L23 44 23% OD TTS No i 24 44- <20% OD TTS No j 28 44 <20% OD TTS No 31 44 21% OD TTS No 5 45 <20% OD TTS No 6 45 25% OD TTS No 10 45 <20% OD TTS No 15 45 <20% OD %" ATS No 22 45 36% OD TTS No 23 45 <20% OD' TTS No
' 10 46 <20% OD TTS No 11 46 <20% OD TTS No.
12 46 <20% OD TTS No 22 46 28% OD TTS No 32 46 34% OD TTS No
+ i i -10 47 <20% OD TTS No
- 11 47 <20% OD TTS No 13 47 <20% OD TTS No 14 47 <20% OD TTS No 2' 47 29% OD TTS No 22 47 21% OD %" ATS No 23 47 <20% OD %" ATS No 26 47 <20% OD TTS No 12 48 28% OD TTS No 13 48 <20% OD TTS No 17 48 <20% OD " ATS No 21 48 33% OD TTS No 22 48 38%~ OD TTS No j 23 48 <20% OD %" ATS No 24 48 <20% OD TTS ~ No 25 48 <20% OD TTS No 9 49 23% OD TTS No 10 49 21% OD TTS No 12 49 30% OD TTS No 13 49 <20% OD TTS No l 14 49 <20% OD TTS No 21 49 '
,23% OD TTS No
,, 22 49 21% OD TTS No l'
i l
Row Column Defect Origin Location Pluqqed 23 49 <20% OD TTS No 24 49 <20% OD \" ATS No 25 49 23% OD TTS No 26 49 30% OD TTS No 27 49 <20% OD TTS No 28 49 <20% OD TTS No 8 50 <20% OD TTS No 11 50 38% OD TTS No 13 50 <20% OD %" ATS No 14 50 <20% OD %" ATS No 19 50 <206 OD TTS No 21 50 <20% OD TTS No 22 50 39% OD TTS No 23 50 32% OD TTS No 26 50 25% OD TTS No 28 50 38% OD TTS No 29 50 31% OD TTS Po 6 51 <20% OD TTS No 9 51 <20% OD TTS No 10 51 <20% OD TTS No 22 51 27% OD TTS No 25 51 30%/24% OD 1" ATS/TTS No 28 51 32% OD TTS No 11 52 30% OD TTS No 22 52 <20% OD TTS No 24 52 28% OD TTS No 29 52 28% OD TTS No 11 53 32% OD TTS No 21 53 24% OD TTS No 25 53 25% OD TTS No 26 53 <20% OD TTS No 27 53 35% OD TTS No 28 53 31% OD TTS No 29 53 38% OD TTS No I 30 53 26% OD TTS No 9 54 <20% OD TTS No 11 54 24% OD T'fS No 25 54 27% OD TTS No 27 54 28% OD TTS No 28 54 35% OD TTS No 25 55 <20% OD TTS No 26 55 36s OD TTS No 10 57 25% OD TTS No 17 57 <20% OD TTS No 21 57 27% OD TTS No 17 58 <20% OD TTS No 21 58 36% OD TTS No 23 58 23% GD %" ATS No i 43 58 <20% OD 2" ATS No 9 59 <20% OD TTS No 15 59 <20% OD TTS No 21 59 32% OD TTS No 23 59 <20% OD \" ATS No Row Column Defect Origin Location Plugged 13 60 <20% OD TTS No 9 64 53% OD TTS Yes 14 64 27% OD TTS No 7 65 31% OD TTS No 9 65 31% OD TTS No 23 65 26% OD TTS No 7 66 30% OD TTS No s 66 37% OD TTS No 15 66 26% OD TTS No 7 67 32% OD TTS No 4 73 35% OD TTS No 5 73 31% OD TTS No 5 74 38% OD TTS No 6 74 46% OD TTS Yes 7 74 31% OD TTS No 9 74 30% OD TTS No 8 76 50% OD TTS Yes 10 76 32% OD TTS No 6 77 36% OD TTS No I
i
"B" Steam Generator Cold M Row Column Defect Origin Location Plugged 6 26 <20% OD 1" ATS No 7 26 <20% OD 1" ATS No 8 26 <20% OD 1" ATS No 9 26 27% OD 1" ATS No 10 26 21% OD 1" ATS No 6 27 <20% OD \" ATS No 7 27 <20% OD \" ATS No 8 27 <20% OD \" ATS No 9 27 <20% OD \" ATS No 10 27 <20% OD 1" ATS No 5 28 <20% OD \" ATS No 7 28 <20% OD \" ATS No 8 28 <20% OD \" ATS No 9 28 24% OD \" ATS No 10 28 <20% OD " ATS No 6 29 <20% OD \" ATS No
! 7 29 <20% OD \" ATS No I 8 29 <20% OD \" ATS No 9 29 <20% OD \" ATS No 10 29 <20% OD \" ATS No 18 29 <20% OD \" ATS No 6 30 <20% OD h" ATS No 7 30 <20% OD \" ATS No
/ 8 30 <20% OD \" ATS No 9 30 <20% OD \" ATS No 10 30 <20% OD \" ATS No 11 30 <20% OD \" ATS No 12 30 <20% OD " ATS No .
I 7 31 <20% OD \" ATS No 8 31 <20% OD \" ATS No 9 31 <20% OD \" ATS No !
10 31 <20% OD \" ATS No l 12 31 <20% OD \" ATS No 13 31 <20% OD \" ATS No 6 32 <20% OD " ATS No 7 32 <20% OD \" ATS No 8 32 <20% OD \" ATS No 9 32 <20% OD \" ATS No 10 32 <20% OD \" ATS No ,
11 32 <20% OD \" ATS No 12 32 <20% OD \" ATS No 13 32 <20% OD \" ATS No 14 32 <20% OD \" ATS No 15 32 <20% OD \" ATS No 18 32 22% OD " ATS No 6 33 <20% OD \" ATS No 7 33 <20% OD \" ATS No 8 33 <20% OD \" ATS No 9 33 <20% OD \" ATS No 10 33 <20% OD \" ATS No 12 33 <20% OD \" AIS No
Roy Column Defect Origin Location Plugged 13 33 <20% OD \" ATS No 14 33 <20% OD \" ATS No 6 34 <20% OD b" ATS No 7 34 (20% OD \" ATS No 8 34 <20% OD \" ATS No 9 34 <20% OD \" ATS No 10 34 <20% OD \" ATS No 11 34 <20% OD \" ATS No 12 34 <20% OD b" ATS No 13 34 <20% OD \" ATS No 14 34 <20% OD \" ATS No 6 35 <20% OD \" ATS No 7 35 25% OD \" ATS No 8 35 <20% OD \" ATS No 9 35 21% OD \" ATS No 10 35 <20% OD \" ATS No 11 35 <20% OD \" ATS No 12 35 <20% OD \" ATS No 13 35 <20% OD b" ATS No 14 35 <20% OD \" ATS No 6 36 <20% OD \" ATS No 8 36 <20% OD \" ATS No 9 36 <20% OD \" ATS No 10 36 <20% OD \" ATS No 11 36 21% OD \" ATS No 12 36 <20% OD \" ATS No 13 36 <20% OD " ATS No 14 36 <20% OD " ATS No 6 37 <20% OD \" ATS No 7 37 21% OD \" ATS No 8 37 <20% OD \" ATS No 9 37 <20% OD \" ATS No 10 37 <20% OD \" ATS No 11 37 <20% OD \" ATS No 12 37 <20% OD \" ATS No 13 37 <20% OD \" ATS No 14 37 <20% OD \" ATS No 6 38 <20% OD " ATS No 7 38 21% OD \" ATS No 8 38 21% OD \" ATS No 9 38 (20% OD \" ATS No 10 38 <20% OD \" ATS No 11 38 <20% OD \" ATS No 12 38 (20% OD \" ATS No 13 38 <20% OD \" ATS No 14 38 <20% OD \" ATS No 6 39 <20% OD \" ATS No 7 39 24% OD 1" ATS No 8 39 <20% OD \" ATS No 9 39 <20% OD \" ATS No 10 39 <20% OD \" ATS No 11 39 <20% OD \" ATS No 12 39 <20% OD \" ATS No
Row Column Defect Origin Location Plugged 13 39 <20% OD \" ATS No 14 39 <20% OD \" ATS No 5 40 24% OD \" ATS No 7 40 21% OD \" ATS No 8 40 <20% OD \" ATS No 9 40 <20% OD \" ATS No 10 40 <20% OD \" ATS No 11 40 <20% OD \" ATS No 12 40 <20% OD \" ATS No 13 40 <20% OD \" ATS No 14 40 <20% OD \" ATS No 6 41 <20% OD \" ATS No 7 41 22% OD \" ATS No 8 41 <20% OD \" ATS No 9 41 <20% OD 1" ATS No 10 41 <20% OD \" ATS No 11 41 <20% OD \" ATS No 12 41 <20% OD \" ATS No 13 41 <20% OD \" ATS No 14 41 <20% OD \" ATS No 6 42 <20% OD \" ATS No 7 42 <20% OD \" ATS No 8 42 <20% OD " ATS No 9 42 <20% OD \" ATS No 10 42 <20% OD \" ATS No 11 42 <20% OD \" ATS No 12 42 <20% OD \" ATS No 13 42 <20% OD \" ATS No 14 42 <20% OD \" ATS No 6 43 <20% OD \" ATS No 7 43 25% OD \" ATS No 8 43 (20% OD \" ATS No 9 43 (20% OD h" ATS No 10 43 <20% OD " ATS No 11 43 <20% OD \" ATS No 12 43 <20% OD \" ATS No 13 43 <20% OD " ATS No 14 43 <20% OD \" ATS No 6 44 <20% OD \" ATS No 7 44 23% OD " ATS No 8 44 <20% OD \" ATS No 9 44 <20% OD \" ATS No 10 44 <20% OD b" ATS No 11 44 21% OD \" ATS No 12 44 21% OD h" ATS No 13 44 <20% OD \" ATS No 14 44 <20% OD \" ATS No 6 45 <20% OD \" ATS No 7 45 <20% OD \" ATS No 8 45 <20% OD \" ATS No 9 45 <20% OD \" ATS No 6 46 <20% OD \" ATS No
Column Defect Origin Location Plugged R_ow_
7 46 <20% OD \" ATS No 8 46 <20% OD \" ATS No 9 46 22% OD \" ATS No 6 47 <20% OD 1" ATS No 7 47 27% OD 1\" ATS No 8 47 <20% OD 1" ATS No 9 47 <20% OD 1" ATS No 6 48 21% OD 1" ATS No 7 48 29% OD 1" ATS No 8 48 <20% OD 1" ATS No 9 48 31% OD 1\" ATS No 6 49 <20% OD 1" ATS No 7 49 <20% OD 1\" ATS No 8 49 <20% OD 1" ATS No 9 49 22% OD 1h" ATS No 18 49 <20% OD 1" ATS No 6 50 <20% OD 1\" ATS No 7 50 <20% OD 1\" ATS No 8 50 <20% OD 1" ATS No 9 50 <20% OD 1" ATS No 18 50 21% OD \" ATS No 19 50 24% OD \" ATS No 6 51 24% OD 1" ATS No 7 51 <20% OD 1\" ATS No 8 51 <20% OD 1" ATS No 9 51 23% OD \" ATS No 14 51 <20% OD \" ATS No 18 51 <20% OD 1" ATS No 19 51 <20% OD 1" ATS No 6 52 <20% OD 1" ATS No 7 52 <20% OD 1\" ATS No 8 52 <20% OD 1\" ATS No 9 52 25% OD 1" ATS No 18 52 (20% OD 1" ATS No 19 52 <20% OD \" ATS No 21 52 <20% OD \" ATS No 22 52 <20% OD " ATS No 23 52 <20% OD \" ATS No 6 53 <20% OD 1" ATS No 7 53 <20% OD 1\" ATS No 8 53 <20% OD 1" ATS No 9 53 <20% OD 1" ATS No 18 53 <20% OD \" ATS No 19 53 <20% OD \" ATS No 22 53 <20% OD 1" ATS No 23 53 <20% OD " ATS No 6 54 <20% OD 1" ATS No 7 54 <20% OD 1" ATS No 8 54 <20% OD 1" ATS No 9 54 <20% OD 1\" ATS No 13 54 <20% OD 1" ATS No 15 54 <20% OD 1" ATS No 18 54 <20% OD 1" ATS No
Row . Column Defect Origin Locaticn Plugged 19 54 <20% OD- \" ATS No 21 54 <20% OD \" ATS No 23 54 21% OD h" ATS No 24 54 <20% OD \" ATS No 6 55 <20% OD 1" ATS No 7 55 <20% OD 1\" ATS No 8 55 <20% OD 1" ATS No 9 55 <20% OD 1\" ATS No 18 55 <20% OD 1" ATS No 19 55 <20% OD 1\" ATS No 22 55 <20% OD 1" ATS No 23 55 <20% OD \" ATS No 24 55 <20% OD \" ATS No 6 56 <20% OD 1" ATS No 7 56 <20% OD 1" ATS No 9 56 <20% OD 1" ATS No 10 56 <20% OD 1\" ATS No 11 56 <20% OD 1" ATS No 18 56 <20% OD 1" ATS No 19 56 <20% OD \" ATS No 20 56 <20% OD 1" ATS No 21 56 <20% OD 1" ATS No 22 56 <20% OD 1" ATS No 23 56 <20% OD \" ATS No 24 56 <20% OD \" ATS No 25 56 <20% OD \" ATS No 6 57 <20% OD 1" ATS No 7 57 <20% OD 1" ATS No 8 57 <20% OD 1\" ATS No 9 57 <20% OD 1\" ATS No 10 57 <20% OD 1\" ATS No 11 57 <20% OD 1\" ATS No 7 58 <20% OD 1" ATS No 8 58 <20% OD \" ATS No 11 58 <20% OD 1" ATS No 18 58 <20% OD \" ATS No 19 58 <20% OD \" ATS No 21 58 <20% OD \" ATS No l
22 58 <20% OD \" ATS No l 23 58 <20% OD \" ATS No 24 58 <20% OD \" ATS No 6 59 <20% OD \" ATS No 7 59 22% OD 1\" ATS No 8 59 <20% OD 1" ATS No 11 59 <20% OD " ATS No 19 59 <20% OD " ATS No 9 60 <20% OD 1" ATS No 10 60 <20% OD 1" ATS No 11 60 <20% OD 1" ATS No 16 60 <20% OD 1" ATS No 9 61 <20% OD 1" ATS No 10 61 <20% OD 1" ATS No 11 61 <20% OD 1" ATS No I
Row Column Defect Origin Location Plugged 18 61 <20% OD \" ATS No 9 62 <20% OD 1" ATS No 10 62 <20% OD 1" ATS No 11 62 <20% OD 1" ATS No 14 62 <20% OD 1" ATS No 15 63 <20% OD \" ATS No 18 63 <20% OD \" ATS No 15 65 <20% OD \" ATS No 23 65 <20% OD \" ATS No 9 66 <20% OD \" ATS No 7 67 <20% OD 1" ATS No 15 67 <20% OD 1" ATS No 19 68 <20% OD 1" ATS No 7 69 <20% OD " ATS No 8 69 <20% OD 1" ATS No 9 72 <20% OD %" ATS No
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ .