ML20081B919

Annual Results & Data Rept,1983

ML20081B919
Person / Time
Site:	Point Beach
Issue date:	12/31/1983
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 8403120066
Download: ML20081B919 (59)

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t ANNUAL RESULTS AND DATA REPORT 1983 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2

< LOCKET NOS. 50-266 AND 50-301 WISCONSIN ELECTRIC POWER COMPANY 4

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1.0 INTRODUCTION

The Point Beach Nuclear Plant, Lhits 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 electriccl power. The plant is located 10 miles north of Two Rivers, Wisconsin, on the west shcre of Lake Michigan.

2.0 HIGHLIGHTS

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2.1 Unit 1 Highlights for the period 01-01-83 through 12-31-83 included the shutdown to accommodate refueling and the steam generator replacement.

-On 12-31-83, this outage was in its 91st day. One trip occurred when contractor personnel bumped the "B" train main steam stop valve relay while working on IC02 thus sending a trip signal to the unit. Unit 1 operated at an average capacity factor of 54.8% and an efficiency (electric / thermal) of 33.2% with a self-imposed hot leg temperature 4

limitation in an attempt to limit steam generator tube corrosion. The unit and reactor availability were 74.2% and 70.3%, respectively.

Unit 1 generated its 39 billionth kilowatt hour (gross) on 01-14-83; its 40 billionth kilowatt hour on 05-03-83; and its 41 billionth kilowatt hour on 08-19-83.

2.2 Unit 2 Highlights for the period 01-01-83 through 12-31-83 included a 103-day refueling and steam generator tube sleeving outage, a brief outage to repair the "A" resistance temperature detector bypass manifold isolation valve (the failure of which caused the unit to trip), and a trip caused by the failure of a capacitor in an inverter. Unit 2 operated at an average capacity factor of 69.3% and an efficiency (electric / thermal) of 33.7%. The unit and reactor availability were 71.3% and 72.7%, - respectively. Unit 2 generated its 38 billionth kilowatt hour (gross) on 01-17-83; its 39 billionth kilowatt hour on 07-22-83; and its 40 billionth kilowatt hour on 10-14-83.

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o-3.0 FACILITY CHANGES, TESTS & EXPERIMENTS 3.1 Amenchnents to Facility Operating Licenses During the year 1983, there were 13 license amenchments issued by the

U. S. Nuclear Regulatory Commission to Facility Operating License DPR-24 for Point Beach, Unit 1, and 12 for Point Beach, Unit 2. These license amenchments and changes are listed by date of issuance and are summarized below

' 3.1.1 03-11-83, Amenchnent 69 to DPR-24, Amenchnent 74 to DPR-27 These amenchnents deleted the Appendix "B" Environmental ,

Technical Specification which pertain to the nonradiological water quality related requirements as required by the Federal Water Pollution Control Act Amenchnents of 1972.

3.1.2 04-15-83, Amenchnent 70 to DPR-24, Amenchnent 75 to DPR-27 These amenchnents. addressed Technical Specification testing requirements for the containment airlock doors to achieve compliance with Appendix "J" to 10 CFR 50.

3.1.3 04-04-83, Amenchnent 71 to DPR-24, Amendment 76 DPR-27 These amendments consisted of changes to the Technical Specifications to allow repair of degraded steam generator tubes by sleeving, established primary coolant limits for iodine concentrations and surveillance frequency, and established a plugging limit for sleeved tubes.

3.1.4 04-15-83, Amenchnent 72 to DPR-24, Amenchnent 77 to DPR-27 These amenhents modified the Specification regarding the frequency for conducting independent audits of the emergency preparedness program from once every 24 months to annually in accordance with the requirements of 10 CFR 50.54(f).

3.1.5 05-04-83, Amenchnent 73 to DPR-24, Amenchnent 78 to DPR-27 These amenchnents modified the Technical Specifications to allow temporary . isolation ~f o the shared motor-driven auxiliary feedwater pumps for a unit during periods of

, startup, shutdown and surveillance testing of the other unit provided that the turbine-driven auxiliary feedwater pumps are operable and capable of automatically delivering flow to the steam generators.

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- 3.1.6 09-30-83, Amen &nent 74 to DPR-24, Amendnent 79 to DPR-27 These amendments: revised the degraded grid voltage relay setpoint and associated time delay as presented in Table 15.3.5-1 of the Technical Specifications.

3.1.7 09 30-83, Amendment 75 to DPR-24 This amen &nent approved the steam generator repair program for. Point Beach, Unit 1, and required as a condition of license that the repair operations be conducted in accordance with those comunitments identified in the approved repair report and reflected in the Staff's safety evaluation.

3.1.8 10-06-83, Amendment 76 to DPR-24, Amendnent 80 to DPR-27 These amendnents incorporated various adninistrative changes in the Technical Specifications in order to clarify terminology used -in a limiting condition for operation, clarify language relating to a periodic calibration interval and correct specific portions of the specifications and

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3.1.9 10-05-83, Amendnent 77 to DPR-24, Amendnent 81 to DPR-27 These amen &nents revised the Technical Specifications to permit locating the spent fuel pool neutron absorber surveillance specimens adjacent to the spent fuel divider wall.

3.1.10' 10-31-83, Amendnent 78 to DPR-24, Amendnent 82 to DPR-27 These amendnents revised the loss of voltage relay setpoints and associated time delays in Table 15.3.5-1 of the Technical Specifications.

3.1.11 12-12-83, Amendnent 83 to DPR-27 This amendment allowed a one-time relaxation of the length of time that the backup component cooling water heat exchanger i

I may be out of service for maintenance.

3.1.12' 12-28-83, Amen &nent 79 to DPR-24, Amendnent 84 to DPR-27 These amendnents added an additional reporting requirement to the Technical Specifications to report all challenges to the pressurizer power-operated relief valves and pressurizer safety valves in the annual report.

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M 3.1.13 12-29-83, Amendnent 80 to DPR-24, Amendnent 85 to DPR-27 These revised the Technical Sepcifications surveillance requirements to include monthly testing of the automatic logic circuitry for the auxiliary feedwater pumps.

3.1.14 12-30-83, Amendnent 81 to DPR-24 This amendnent authorized Point Beach, Unit 1, reactor

. operations at either 2250 or 2000 psia after return to power following steam generator replacement.

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. l 3.4 Design Changes 3.4.1 The following design changes were completed during 1983:

a. E-204 (Common), Emergency Diesel Generators. An isolation switch was installed in each Diesel room l

" which disconnects the emergency Diesel generator control circuits from the control room. Automatic start capability is maintained and local initiation of the automatic start sequence is provided.

1 Summary of Safety Evaluation: The modification upgrades the emergency power supply system. It allows the automatic sequencing of the emergency Diesel generator in case of a fire in the main control board.

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E-220 (Common), Radio Power Supply Shift. The radio power source was moved for the F2 frequency from 2DYO2-3 to diesel emergency lighting panel 36E-12, which provides a continuous power supply.

Summary of Safety Evaluation: The small additional load added to emergency Diesel generator G03 is not

' significant and will not reduce the capability of the diesel-to- perform its safety function.

c. E-225 (Unit 2), 2RK-38 Instrement Panel Relocation.

The modification moved the RK-38 panel containing steam generator level and pressure and pressurizer 4: level instrumentation from the west cide of the west wall of the 4160 V safeguards switchgear room to a '

location visible to the operator from the auxiliary feed pump area.

1 Summary of Safety Evaluation: Relocating panel 2RK-38 to be visible from the auxiliary feedwater pump area is necessary to improve local operability' of ' the auxiliary feedwater system, and is consistent with the i

panel ' locations described in the FFDSAR. Components

! and installaticn meet or exceed original' design requirements. The modification is not nuclear safety related since the panels provide indication only.

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.d. E-226 (Unit 1) & E-227 (Unit 2), 4160 V Electrical

. System. {

Three new undervoltage relays were installed l on each 4160 V safeguards bus (1/2A05,1/2A06). Each l of these relays operate an auxiliary relay which is installed on the safeguards bus. Tb auxiliary relas contacts are connected in two trains of 2/3 logic.

Each train of 2/3 logic actuates a time delay relay.

Time delay relays are also located on the safeguards bus. The time delay relay contacts provide the same ,

, functions as presently provided by the single '

undervoltage relay existing on the associated nonsafeguards 4160 V bus (1/2A03, 1/2A04) which is to trip the brealcar that connects the safeguards bus to the nonsafeguards bus and causes annunciator action.

Summary of Safety Evaluation: 2/3 undervoltage logic increases plant safety by ensuring isolation of 4160 V safeguards buses from nonsafeguards power supplies.

Additional unde. voltage protection is required ta meet NRC criteria and a Technical Specification change has been issued requiring the 2/3 logic.

e. E-230 (Unit 1), Electrical Penetrations. Penetration

-Q58 was replaced and new penetrations .were installed in spare positions Q21 and Q22. The penetrations are required to accomplish various post-TMI in-containment system additions.

Summary of Safety Evaluation: The modification does not - have deleterious effect upon plant safety. The penetration design is equal tc or better than original plant design.

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f. E-257 (3D) & E-258 (4D), Emergency Diesel Start Circuitry. Two parallel contacts were installed on A05 and A06 from the two emergency Diesel generator

undervoltage relays that start the diesels on l

undervoltage. This will not allow the output breakers to close until the bus voltage decays even when the supply breakers have been opened.

Summary of Safety Evaluation: This modification will prevent possible damcge to safety-related equipment by i- preventing closure of the emergency diesel generator L output breaker on the safeguards bus before residual

. voltage decays. The parallel contacts in the breaker i

closing circuit provide 1/2 logic for undervoltage protection and protect against single undervoltage relay or contact failure, i.e., failure of one contact to close will not prevent auto closure of the emergency diesel generator output breaker on a loss of AC.

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e C IC-49 (Unit 2), Reactor Protection Circuitry. One of the wires on the bistable test switches in the process control system analog racks was lifted from one terminal and re connected to a different terminal on the same switch. This was done to all bistable test  !

switches in the process control system analog racks <

with the exception of containment spray initiation l from 2/3 containment pressure.

1 Summary of Safety Evaluation: Shorting out of one

. pole of the double pole test switch will better simulate actual operating co.1ditions in the channel under test and m?.he the discovery of previously undatectable grounds possible.

h. IC-107 (Unit 2), Feedwater Heater 4A&B Level Control.

The modification permits alarming of high water level in the feedwater heater but delays the BTV closure to assist in clearing the high level alarm.

Summary of Safety Evaluation: Not nuclear safety related.

i. IC-145 (Unit 2), Manipulator Hoist Pulley Encoder. A linear displacement encoder was installed on the manipulator hoist pulley placing readout next to the load cell.

Summary of Safety Evaluation: The encoder is a superior means of determining mast height compared with the old measuring tape method.

j. IC-184 (Unit 2), Air Ejector Instrumentation Upgrade.

The modification installed Hastings linear mass flow meterc betwen existing flow orifice isolation valves AR-16 and AR-17. The modification is needed in view of more stringent restrictions on condenser air inleakage as it impacts turbine spindle warranty requirements.

Summarf of Safety Evaluation: Not nuclear safety related.

k. IC-240 (Common), Water Treatment. The 0-600 gpm neutralizing tank pump fic,w meter was changed to 0-300 gpm for more accuracy.

Summary of Safety Evaluation: Not nuclear safety related.

1. IC-267 (Unit 1) & IC-268 (Unit 2), Turbine EH Control System Power Supplies. The existing SCI power supplies were removed and Lathbda power supplies were installed. The Lambda power supplies have been demonstrated to be more reliable and are currently being used in Westinghouse turbine EH control systems.

The modifications did not involve a change in ccyonent function.

Summary of Safety Evaluation: Not nuclear safety related.

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m. IC-276 (Common), Water Treatment. Range change kits and new conductivity cells were installed in the anion / mixed bed conductivity recorder (CR-9213) to change the range from 0-50 pmho for anion (0-5 pmho for mixed bed) to a new range of 0-10 pmho for anion (0-1 paho for mixed bed) for more accuracy.

Summary of Safety Evaluation: Not nurlear safety related.

n. IC-285 (Common), Security. The "E" field wiring was modified from a three-wire field to a five-wire field for better sacurity protection.

Summary of 3afety Evaluation: Not nuclear safety related.

o. IC-338 (Common), Security. A terminal and card reader were installed in the Security Supervisor's office.

Summary of Safety Evaluation: Not nuclear safety related.

p. M-130 (Common), Digital Off-Line Noble Gas Monitoring Systems. Procured and installed the subject systems for the main auxiliary building vent, drumming area vent and combined letdown gas stripper vent.

Summary of Safety Evaluation: Besides improving auxiliary vent stack sampling, the modification greatly aproves operator ability to quickly recognize a steam renerator tube failure incident (modification committed to the NRC by letter of 06-26-75, as a direct result of Unit I steam generator tube failure ir.cident of 02-26-75.)

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q. M-326 (Unit 2), Temperature Control Valve for Hydrogen Seal Oil Air-Side Cooler. A temperature control valve )

and associated piping were added to the hydrogen seal oil air-side cooler. This is to maintain the seal oil temperature constant.

Summary of Safety Evaluation: Not nuclear safety related.

r. M-453 (Unit 1) & M-454 (Unit 2), Blu down System. The steam generator blowdown system was modified to include the installation of a blowdown heat recovery system and the connection of the blowdown filter to the waste condensate polishing demineralizers.

Summary of Safety Evaluation: Not nuclear safety related.

s. M-497 (Unit 1) & M-498 (Unit 2), Steam Generator Sampling Isolation Valve. A valve and flushing line were installed to the nearest service water discharge line.

Summary of Safety Evaluation: Not nuclear safety related; however, Chemistry & Health Physics is aware that presumed full flow of this line is to be used in calculating possible discharges.

t. M-577 (Unit 2), Gland Steam Condenser. Flanges were installed on the tube side piping to facilitate removal of the condenser endbells for inspection.

Summary of Safety Evaluation: Not nuclear safety related.

u. M-585 (Unit 2), Demineralized Water. A new demineralized water li.ne was run from the water treatment area to the Unit 2 facade to facilitate the setup of the sludge lance trailer.

Summary of Safety Evaluation: Not nuclear safety

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v. M-597 (Common), Upgrading Fire Barriers. The fire barriers were upgraded through the installation of three-hour fire-rated penetration seals, the design of which has been approved by the NRC. The modifications were accomplished in two phases as follows: Phase 1 was a survey of the location and description of penetrations and preparation of a penetration schedule and wall / floor drawings. Phase 2 consisted of penetration sea) installation. The modification was

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proposed in Section 6.1.14 of the Fire Protection Review and installation fulfills commitments made in response to NRC staff positions PF-32 and PF-40.

Summary of Safety Evaluation: Not nuclear safety related.

w. M-643 (Unit 1), Residual Heat Removal. A safety hoist rig was installed for moving the contaiment sump "B" screen to provide more convenient access to the 850A&B valves.

Summary of Safety Evaluation: Not nuclear safety related.

x. M-655 (Unit 1) & M-656 (Unit 2), Nitrogen Backup to Pressurizer Spray Valves' Inctrument Air Supply. A dedicated nitrogen supply was installed to permit operation of the spray valves if instrument air to the valves is lost. The nitroge.i supply system consists of a nitrogen bottle and associated tubing and valves, and is located outside the pressurizer cubicle.

Summary of Safety Evaluation: Operation of the pressurizer spray valves is not required to prevent or mitigate the consequences of an accident. Howe 7er, adding a . backup nitrogen supply to the valves helps ensure the operator's ability to control reactor coolant system ~ pressure. A relief valve in the nitrogen supply line protects the instrument air supply and associated components from overpressurization,

y. M-664 (Unit 2), Primary Sampling. The modification provided a manual isolation valve, accessible during power operation on the hot leg sample line just upstream and adjacent to 2A0V-955.

Summary of Safety Evaluation: Not nuclear safety related.

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z. M-673 (Common), Instrument Air. A second instrument i air dryer of larger espacity was installed and piped j in parallel to the existing dryer. Dryer e.larms i annunciate in the control room.

Sununaly of Safety Evaluation: Not nuclear safety  !

related.

aa. M-682 (Unit 2), Chemical volume control System. The

• modification installed a containment isolation valve with "T" signal trip in the reactor coolant pump seal wc.*er return line -inside containment. The valve is located downstream of relief valve 2-314.

Sununary of Safety Evaluation: The new valve provides additional assurance of containment isolation. The valve fails closed on loss of air or electric power and is seismically qualified.

. bb. M-684 (Unit 2), Chemical Volume Control System. The modification installed a containment isolation valve on the common letdown line downstream of the orifice block valves inside containment. The isolation valve is equipped with a~"T" signal trip.

Summary of Safety Evaluation: The new valve provides

" added assurance of containment isolation. The valve fails closed on loss of air or electrical power and is seismically qualified.

cc. M-716 (Common), Fuel Handling. A portable coGtrol rod changing tool was obtained for transfer of control rods to the spent fuel pit. This tool allows the transfer of any insert in the spent fuel pit, thus freeing up the cavity for other fuel movements.  :

Summary of Safety Evaluation: The modification does not present any additional threat or unanalyzed problem. The mountings are designed to hold the tool in an area of the pit ehere fuel is not stored.

i dd. M-721 (Common), Sewage Treatment Plant. A new 17,5000 gpd sewage trestment plar.t was constructed to meet the Wisconsin Pollutant Discharge Elimination System (WPDES) permit requirements.

, Summary of Safety Evaluation: Not nuclear safety related.

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ee. M-730 (Unit 1), Steam Generator Elowdown Isolution Valves. Air-operated containment isolation valves were installed inside containment to automatically isolate steam generator blowdown and preserve containment integrity following a seismic event. The modfiication is necessary as a result of NRC IE Bulletin No. 79-02 indicating that the existing valves located outside of containment could be incapacitated by a fallen block wall in the- facade stairway area.

Summary of Safety Evaluation: The new containment isolation valves provide additional assurance of containment integrity. They are seismically supported and in-containment electrical components are environmentally qualified. Valves fail closed on loss of instrument air or electrical power, and will not automatically reopen when the CI signal is cleared.

All safeguards functions on existing steam generator blowdown isolation valves have been added to the new CI valves. . The valves were functionally tested and Type "C" leak-tested.

ff. M-731 (Unit 2), Steam Generator Blowdown. The modification provided air-operated valves for remote automatic isolation of steam generator blowdown inside containment.

Sammary of Safety Evaluation: The new containment isolation valves provide additional assurance of containment integrity. They are seismically supported and, in containment, electrical components will be environmentally qualified. The valves fail closed on loss of instrument air or electrical power and will not_ automatically reopen when the containment isolation signal is cleared. All safeguards functions on existing steam generator blowdown containment isolation valves were added to the new CI valves. The valves were functionally tested and Type "C" leak tested after installation.

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gg. M-742 (Common), Block Wall Upgrading. Certain plant

' concrete block walls were upgraded incorporating

, Bechtel-designed bracing and/or shear transfer connections.

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Summary of Safety Evaluation: This modification was

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performed in response to RC IE Bulletin No. 80-11

'which was issued for review and action by utilities to

.ijnprove nuclear plant safety. Upgrading concrete block ualls improves their structural integrity and

( therefore > enhances plant safety. This modification x

did not create any new unresolved safety issues and

' 'did tiot' decrease the margins of safety assumed in the

, bases for any-Technical Specifications.

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hh. M-750'(Unit 2), Bowser Purification. Connections were

. installed for a portable filter to be used uith the lube, oil Bowser system.

X, Summary of Safety Evaluation: Not nuclear safety

.related ii . M-755 (Unit 1) & M-756 (Unit 2), Primary Sampling.

The 3/4" Grinnel-Saunders diaphragm valves (1&2-284) and associated run of lower pressure piping between the volume control tank and FI-903 were replaced with

, higher pressure pipe and higher pressure valves. This s -was done to' avoid having a run of pipe that is rated f',. - lower than primary system pressure where it is possible to pressurize this piping run to primary system pressure.

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Summary of Safety Evaluation: Not nuclear safety related.

- ' ', - jf. M-759 (Commonl , Seismic Restraints. This modification was in response to the NRC issued IE Bulletin No.

79-14. It required that all safety-related piping greater than 2 " in diameter be in conformance with the original seismic design criteria. The total project included the review of approximately 34,000'

' of safety-related pipe 2\" or larger. A total of 2,842 pipe supports were reviewed and of these, 689

, regt. ired modification.

Summary of Safety Evaluation: This modification ensured that the -:ctsal configuration of the safety-related systems meets the design requirements.

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kk. M-768 (Common), Service Water Piping. Low point flush connections were installed upstream of the motor-operated auxiliary feedwater pump. The flush connections are used to flush sediment from the line to improve motor-operated valve operability and to reduce the potential for sand and silt carryover into the auxiliary feedwater system.

Summary of Safety Evaluation: Piping and installation meet or exceed high pressure service water design requirements. Integrity and operability of service ,

water and auxiliary feedwater systems is not reduced. '

11. M-783 (Common), Auxiliary Building Shielding. The modification installed permanent shielding for MCC 1&2B32 and C59. Temporary, removable shielding recommendations are also provided. The modification was required in order to meet NUREG-0737 requirements.

Summary of Safety Evaluation: The permanent shield walls are designed to seismic criteria and constructed in accordance with appropriate quality control requirements. Protection from toppling is provided for adjacent and nearby safety-related equipment.

mm. M-802, Isokinetic Stack Sampling System. Isokinetic stack sarapling systems were installed on the auxiliary building vent stack and the drumming area vent stack to provide post-accident sampling capabilities required by NUREG-0737.

Summary of Safety Evaluation: Not nuclear safety related.

nn. M-803 (Unit 1) & M-804 (Unit 2), Reactor Coolant Pump Oil Leakage Collection System. An oil collection

• system was added to the reactor coolant pumps capable of collecting lube oil from all potential pressurized and unpressurized leakage sites in the reactor coolant pump lube oil system. The system is required to comply with 10 CFR 50 Appendix R. A " prototype" collection system was installed on one reactor coolant pump to verify fit-up and design before doing the other pumps.

Summary of Safety Evaluation: Not nuclear safety related.

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oo. M-806 (Unit 1) A M-807 (Unit 2), hain Steam Line Drain e

Connection. . A drain connection was installed in each main' steam line's low-point steam trap piping. The purpose of the drain is to. improve the ability to

, drain the steam piping during steam generator leak

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+ f: testing and prevent-flooding the steam lines. This is presently done by removing the steam trap tops.

Summary of SEfety Evaluation: Not nuclear safety related.

-pp. , M-810 (Unit' 2), RHR Piping. There were 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 circulating pump (P33). This eliminated the need to leak test the piping under post-TMI requirements for systems outside containment likely to contain radioactive materials.

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The piping "was not used an'd its intended function could not be clearly established, although one c .

pessibility . is that : the ' lines were to be used for

,~ - '~ b> adjusting the boron concentration of the RHR loops.

This function is p'rformed e using other piping connections.

' ~ Part 2 of the modification tequest was the installation of . test connections and isolation valves

in the spent fuel pit filter discharge connection to the RPlt . pump suctiod . header. These changes also facilita' ec leak testing per post-THI requirements for systems outside containment likely to contain

-radioactive materials.

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. Summary of Safety Evaluation: Part 1 - discharge

, .- '4 lines - from P33 to the RHR discharge serve no safety A > .n function. Cap installation complies with Westinghouse

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Pipe Class 601_ and other. -design criteria of the RHR system. _Part 2 - not' nuclear safety'related.

qq. M-8'?O (Unit 2), Primery Sampling Test Connection, Containment Isolation Valves. The modification adds

- test connections for AOV-951, 953 and 955 and to the s

corresponding root isolation valve for each .

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Summary of Safety Evaluation: Tubing and valve installation meet or exceed original system w j requirements. System integrity is not degraded.

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^ ~, Ccapletion of the modification permits leak testing of the valves in~ accordance with Appendix J of 10 CFR 50.

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rr. M-823 (Commen), Purge Exhaust Sampling. The sample point on the purge exhaust stack was moved to facilitate access for health physics monitoring.

Summary of Safety Evaluation. Not nuclear safety related.

ss. M-826 (Common), Heating Steam. Piping was installed to route condensate from the potable water heater to

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the floor drain. This eliminated the' temporary hose set up. with an open red tag series isolating the condensate return to the condensate system.

Summary of Safety Evaluation: Not nuclear safety related.

tt.82-011 (Common), Service Water Pump Fire Wall. A fire wall was installed between the north and south service water pumps in accordance with 10 CFR 50.48 Appendix R requirements.

Summary of Safety Evaluation: The fire wall is located near safety-related service water pumps and therefore is seismically designed to prevent it from falling on the pumps.

uu. 82-23 (Unit 1) & 82-24 (Unit 2J , Check Valve 862A&B Drain Connection. The modification provides a \"

drain line, isolation valve and pipe cap off of the 1&2CV-826A&B cover flange. The drain connection is required to be able to expose the valve seat to an air environment when performing 10 CFR 50 Appendix J testing.

Summary of Safety Evaluation: Design and installation meets or exceeds original system requirements, and therefore, the ability of tio containment spray system to perform its design function is not degraded.

vv. 82-26-(Unit 2), Primary Sampling. The primary sample sink drain piping was rerouted to the north auxiliary building sump. This routes the sample sink directly to the sump rather than through floor drains.

Summary of Safety Evaluation: Not nuclear safety related. ALARA item.

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ww. 82-45 (Common), Water Treatment. A water connection was installed to facilitate cleaning sludge out of clarifier. This will eliminate or greatly reduce the mess and tripping hazard created and abuse experienced by fire hoses.

Summary of Safety Evaluation: Not nuclear safety related.

xx. 82-49 (Common), Security Fence & "E" Field. Existing fence and "E" field were relocated to provide room for south gatehouse addition.

Summary of Saiety Evaluation: Not nuclear safety related.

yy. 82-51, Fuel Oil Supply. Ninety feet of existing fuel 011 supply piping was rerouted to accommodate modifications made to the plant gatehouse.

Summary of Safety Evaluation: The fuel oil transfer line is classified not nuclear safety related because adequate fuel oil supplies are scintained in the emergency Diesel generator fuel tar.ks and day tanks. >

zz. 82-52 (Common), Fire Doors. Heavy duty closers were installed on the fire doors betwsen the two emergency diesels.

Summary of Safety Evaluation: Not nuclear safety related.

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aaa. 82-53', Auxiliary Feedwater System Inoperable Alarm.

An alarm for each unit was added to the main control board which ar;nunciates any of the following conditions: (1) Steam-driven auxiliary feedwater purap trip valves for that unit closed; (2) either motor-driven auxiliary feedwater pump control switch in pull-out; (3) both main feed pump control switches for that unit in pull-out. The alarms were required to meet a commitment to the NRC resulting from a review of the auxiliary feedwater system.

Summary of Safety Evaluation: There is no change to the existing circuitry because the modification only uses equipment additions or preexisting space equipment. Proper installation of the modification 5as controlled by special maintenance procedurr.

There is no increase in the prceability of a loss of normal feedwater accident. In the event of circuitry failure, there is no accident previously evaluated since only loss of annunciator action is affected.

Post-installation testing includes disabling of an auxiliary feedwater pump (motor-driven pump in pull-out/ turbine-driven pump with trip valve shut) with only one pump disr.nled at a time so an LCO not be violated. The post-installation testing of the main feed pump circuitry was performed with each unit in a shutdown condition.

bbb. 82-54 (Unit 1) & 82-55 (Unit 2), Reactor Protection.

Test switches and lights were installed to reactor protection system racks 1&2c155 and 1&2C165 and associated test jacks to provide for periodic testing ..

of auxiliary feedwater system automatic actuation '

logic.

Summary of Safety Evaluation: There was no change to existing circuitry in the modifications and only equipment additions were considered. Installation and testing was controlled by special maintenance procedure.

The single failure criteria can be supported because independence of the two trains of automatic actuation of auxiliary feedwater is maintained.

ccc. 82-73 (Common), Demineralizer Cubicle Shielding. A new layer of concrete block was installed adjacent to the existing shield walls for the purpose of reducing exposure.

Summary of Safety Evaluation: Not nuclear safety related. ALARA item.

ddd. 82-77 (Unit 1), Main Steam System. The tube bundles were replaced in all moisture separator reheaters with a new design because of excessive tube failures.

Summary of Safety Evaluation: Not nuclear safety related.

eee. 82-83 (Unit 1), Feedwater Systems. The No. 4 feedwater heaters were replaced because extensive corrosion of the tubes has reduced tube wall t.hickness. Tube material was changt.d from Cu-Ni to Type 304 stainless steel.

Summary of Safety Evaluation: Not nuclear safety related.

iff. 82-86 (Unit 2), Containment Personnel Airlock. The modification provides a vacuum ptrap, isolation velve

' and associated hardware necessary to provide a system for leak testing the door seals within 3 days of opening the containment airlocks at El. 66'.

Summary of Safety Evaluation: The modification resulted in a slight change to the airlock door secondary . seal's. pressure boundary. This minor boundary change, past the first "0" ring seal in each

? door, consisted of replacing a pipe plug boundary with a short run of stainless steel tubing, associated fittings and a manual isolation valve. Ackninistrative controls required the proposed isolation valve be 5

normally closed when the opposite airlock door is opened.

4 The modification had no effect upon the structural integrity or other design features of the cont.ainment airlocks. Material and installation will meet or exceed original system specifications and code requirements.

ggg. 82-90 (Common), Security Offices. A new enclosed office was constructed for the Ackninistrative Specialist - Security.

Sununary of Safety Evaluation: Not nuclear safety related.

hhh. 82-93, Circulating Water System. A new intake crib baffle wall which will have an adjustable weir for discharge water bypass was installed.

Summary of Safety Evaluation: Not nuclear safety

related.

iii. 82-95 (Unit 1) & 82-96 (Unit 2), Containment Hatch Labeling & Warning Systems. This modification more

, clearly labels each containment hatch with the appropriate unit number. An audible warning at cach outer hatch indicates the prerence of high radiation fields inside containment. The high radiation i

light / sign has been relocated to a more visible place.

This inodification responds to NRC IE Inspection Report No. 82-12.

Summary of Safety Evaluation: Not nuclear safety related.

jjj.82-116 (Common), New Sewage Treatment Plant. A permanent ladder to the roof of the new sewage treatment plant was installed to facilitate changing the air supply inlet filters.

Summary of Safety Evaluation: Not nuclear safety related.

kkk.82-118 (Conunon), Auxiliary Feedwater System. The seismic adequacy of the auxiliary feedwater system was upgraded to meet the commitments made to the NRC in response to NRC Generic Letter 81-14.

Summary of Safety Evaluation: Supports will be anchored to seismic structures. Drilling will be controlled per existing procedures to ensure rebar is not cut.

• 111. 83-01 (Common), communications. The modification retired existing KRQ-717 facilities and installed 3 new frequencies and control systems with repeater capability. An alternate power supply is provided to increase reliabi 'ty of the radio system. Both the normal and alternate power source are nonsafeguards.

Summary of Safety Evaluation: Not nuclear safety related.

mmm. 83-05 (Unit 2), Primary Sampling. Valve 2-955 was replaced with a valve of higher quality to minimize j leakage.

Summary of Safety Evaluation: The new valve functions  !

L the same as the existing valve and meets or exceeds j original system design and installation requirements.

Seismic evaluations we re performed to ensure seismic qualification is maintained. i L

nnn. 83-12 (Common), Potable Water Remote Alarms. Alarms in the potable water room were wired to annunciat9 in the control room. Potable water alarms can now be acted upon more quickly than before with only a local indicator.

Sunnary of Safety Evaluation: Not nuclear safety related.

ooo. 83-13 (Unit 1) & 83-14 (Unit 2), Steam Jet Itr Ejector Drain Loop-Seal. A loop-seal was installed on the steam jet air ejector to prevent the air ejector from

. sucking air through the drain line.

Summary of Safety Evaluation: Not nuclear safety related.

ppp. '83-34 (Unit 2), Reactor Coolant System. Calcium silicate box insulation jacketed with stainless steel was installed on the piping between the pressurizer and the safety valves.

Summary of Safety Evaluation: Insulating the loop seals helps to ensure maintaining the integrity of the reactor coolant system by reducing the - dynamic loads on the safety valves and piping. The increased temperature of the loop seal piping is within its design-limits.

qqq. 83-35 (Unit 2), Reactor Coolant System. Pressurizer i

safety valve discharge piping was upgraded by deleting some existing supports and adding new supports.

Summary'of Safety Evaluation: Design of the new supports is based apon analysis to ensure stress limits are not exceeded. The new supports provide additional assurance that the Integrity of the subject piping is not degraded by relief valve operation. Per Technical Specification 15.3.13.5, there is no requirement to inform NRC prior to installation.

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. . 1 rrr. 83-38 (Unit 2) & 83-39 (Unit 1), Waste Disposal. The lap joint flange on the inlet of relief valve GW-82B  ;

was replaced with a socket welded flange. The pipe support on the inlet to relief valve GW-82A was removed at El. 29'7".

Summary of Safety Evaluation: The modification  !

improves the integrity of the gas stripper piping and thereby reduces the potential for leakage from a

. system containing radioactive materials. &

sss. 83-45 (Common), New R&Jiation Monitoring System.

Heating steam piping was moved to facilitate calibrating radiation monitoring system equipment.

Summary of Safety Evaluation: Not nuclear safety related.

ttt. 83-50 (Unit 2), Residual Heat Removal Pumps. An "0" ring seal was added between the shaft and the shaft sleeve to stop leakage allowed by the existing retal gasket.

Summary of Safety Evaluation: Not nuclear safety ;

related.

uuu. 83-62 (Unit 1), Polar Crane Limit Switches. Upper limit switches were installed on the Unit 1 polar crane to prevent two-blocking. There are two lir.it switches of different types and' they deenergize the main power supply and hoist drive motor.

Summary of Safety Evaluation: The two limit switches  !

deenergize both the main power supply and the hoist  !

drive motor. Before only the hoist drive motor was deenergized on the upper limit.

vvv. 83-64 (Common), Turbine Building Crane Limit Switches.

Upper limit switches were installed on the turbine building crane to prevent two-blocking. There are two limit switches of different types and they deenergize the main power supply and hoist drive motor.

Summary of Safety Evaluation: The two limit switches deenergize both the main power supply and the hoist.

drive motor. Before only the hoist drive motor was deenergized on the upper limit.

4

\

. i www. 83-75 (Unit 2), Generator Potential Transformer. The *

, cable running from the grounding transformer to the potential transformer neutral bus was disconnected at each end. A 4/0 cable was then connected to the potential transformer neutral bus to the station ground located in the potential transformer compartment.

Summary of Safety Evaluation: Not nuclear safety related.

xxx.83-077 (Common), Radwaste Compactor. A new radwaste compactcr was installed to facilitate the extra waste generated during the steam generator outage.

Summary of Safety Evaluation: Not nuclear safety related.

yyy. 83-78 (Common), Female Controlled Side Locker Room.

The female controlled side locker room was moved from the area between the Unit 1 Nos. 4 & 5 feedwater heaters to its present location near the turbine building elevator. It was moved to facilitate the No.

4 feedwater heater replacement on Unit 1.

Summary of Safety Evaluation: Not nuclear safety related.

zzz. 83-80 (Unit 2), Switchyard. Instantaneous overcurrent relays were connected to separate current transformers in the ground leads from the free-standing CT structures to the ground grid. The overcurrent relays are connected to initiate the 345 KV Nos. 2 and 4 (Units 1 & 2, respectively), bus differential lockout.

The modification detects a failed 345 KV CT and initiates fault clearing.

Summary of Safety Evaluation: Not nuclear safety related.

aaaa. 83-83 (Common), Security Facilities. White crushed rock was placed along "E" field instrusion detection zones to enhance closed circuit television observation and prevent weed growth.

Summary of Safety Evaluation: Not nuclear safety related.

4 e

bbbb. 83-96 (Unit 1), Generator Potential Transformer. The neutra,1 cable from the potential transformer compartment was disconnected at the potential transformer neutral bus and at the grounding neutral transformer and grornded at both ends. A ground wire was installed from the neutral bus (potential trr.nsformer) to the ground bus at the potential transformer compartment.

, Summary of Safety Evaluation: Not nuclear safety related.

cccc.83-136 (Common), Cranes & Hoists. An I-beam trolley and chain hoist arrangement was installed in the G02 emergency Diesel generator room. It enables Maintenance personnel to more efficiently remove the cylinder banks of the emergency Diesel generator so the seal between the banks and the crankcase can be inspected.

Summary of Safety Evaluation: An evaluation of the lifting beams was done to ensure that they will remain intact during a seismic event. Use of the beams will only be for emergency Diesel generator maintenance.

dddd.83-143 (Commonl, Water Heater for Primary Auxiliary Building El. 26' Decon Pad. A water heater was installed in the vicinity of the decon pad to aid in decontamination operations.

Summary of Safety Evaluation: Not nuclear safety related.

a 3.4.2 The following modification was completed prior to 1983 but

' was omitted from the previous annual report. It is discussed herein to complete and update the record.

M-772 (Common), Resin Transfer Line. A 2" stainless steel sluice pipe extension and spoolpiece were installed downstream of A0V-14 and will extend through the truck access area wall. A flush connection was installed in the existing resin sluice line near the transfer cask setdown area. The

• purpose of the modification is to permit slulcing resin directly to a shipping cask without going through the Atcor system.

Summary of Safety Evaluation: Not nuclear safety related.

3.5 Procedure Changes The following emergency operating procedures were revised during 1983.

3.5.1 EOP-8B, Irradiated Fuel Handling Accident in Containment, Revision 5, 03-23-83. The procedure was upgraded to delete the use of containment check-in cards for personnel inventory.

3.5.2 EOP-8C, Irradiated Fuel Handling Accident in Primary Auxiliary Building, Revision 2, 12-05-83. The procedure was editorially upgraded to improve clarity. Reference to the Emergency Plan Implementing Procedures was added as well as the radiation monitors used for analysis of primary auxiliary building atmosphere. The titles of " Shift Supervisor" and

" Duty & Call Technical Advisor" were upgraded to " Shift Superintendent" and " Duty Technical Advisor," respectively.

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o 5.0 ST'AM GENERATOR TUBE INSERVICE INSFECTION The following is a synopsis of findings resulting from steam generator tube inspections conducted during 1983.

5.1 Unit 1 During the Refueling 11 outage, the Unit I steam generators were replaced. Results of the preservice inspection will be contained in

, the 1984 operating report.

5.2 Unit 2 Refueling 9 Inservice Inspection Prior to sleeving of tubes during this ninth refueling outage, eddy current inspection of both steam generator inlets was performed.

In "A" steam generator inlet, 2,983 tubes were tested through the first support and 182 tubes were inspected full-length. Four tubes were not inspected because of their location under the eddy current fixture foot.

In "B" steam generator inlet, 2,991 tubes were tested through the first support, 192 tubes were inspected through the U-bend and 33 tubes were examined full-length. Four tubes were not inspected because of their location under the eddy current fixture foot.

A Results of Eddy Current Inspection "A" SG Inlet "B" SG Inlet

<20% 366 159 20-29% 225 156 30-39% 33 47 40-49% 2 2

. 50-59% ---

1 60-69% 1 ---

70-79% --- ---

80-89% 9 1 90-100% 4 ---

e "A" Steam Generator Examination Results The following tubes were plugged as a result of eddy current examinations in "A" steam generator:

Tube Defect Location Origin Side R19C05 47% ITSP OD Inlet R05C21 43% TTS OD Inlet R18C24 '899, 10-12" ATE OD Inlet R26C27 UDI/81% 5\-17\" ATE OD Inlet R16C30 100% 5\-16" ATE OD Inlet R17C30 UDI/81% 6-14" ATE /12 " ATE OD Inlet R16C31 UDI 9-15" ATE OD Inlet R17C31 UDI 12\-16" ATE OD Inlet RICC33 UDI/84% 6-15" ATE /6" ATE OD Inlet R17C34 UDI/86% 7-16" ATE /7\" ATE OD Inlet

-R20C36 UDI 13-18" ATE OD Inlet R17C37 UDI 3\-7\" ATE OD Inlet R19C37 UDI/88% 4\-17 " ATE /5" ATE OD Inlet R20C37 100% 6-15" ATE OD Inlet R17C38 UDI 6-18" ATE OD Inlet R17C39 UDI/89%/41% 10-17\" ATE /14" ATE /ITS OD Inlet R17C42 100% 3-16" ATE OD Inlet R18C42 68% 16" ATE OD Inlet R16C43 91% 5\-18h" ATE OD Inlet R17C44 UDI 8-13" ATE OD Inlet R18C44 UDI 4\-14" ATE OD Inlet R14C45 UDI 13-19h" ATE OD Inlet R15C46 UDI 8\-16" ATE OD Inlet R09C47 UDI 11\-17 ATE OD Inlet R13C47 UDI 11 -16" ATE OD Inlet R14C47 UDI 14-18" ATE OD Inlet R19C47 UDI 7-16" ATE OD Inlet R12C48 UDI 8-8\" ATE OD Inlet R14C48 89% 12-15" ATE OD Inlet "B" Steam Generator Examination Results R01C90 83% 4\-8" ATE OD Inlet R17C28 51% TTS OD Inlet R09C48 43% 1" ATS OD Outlet ATE - Above tube end UDI - Undefinable indication ATS - Above tubesheet TTS - Top of tubesheet

"A" Steam Generator Hot Leg Row Column Indication Location Origin Plugged 18 5 34% 1 TSP OD No 19 5 47% 1 TSP OD Yes 20 5 31%/38% 1 TSP /2 TSP OD No 23 8 29% 1 TSP OD No 4

• 16 <20% TTS OD No 7 16 <20% TTS OD No 8 16 <20% TTS OD No 2 17 <20% TTS OD No 3 17 <20% TTS OD No 4 17 28% TTS OD No 5 17 <20% TTS OD No 6 17 <20% TTS OD No 7 17 <20% TTS OD No 8 17 <20% TTS OD No 6 18 <20S, TTS OD No 7 18 <20% TTS OD No 8 18 <20% TTS OD No 9 18 22% TTS OD No 10 18 <20% TTS OD No 11 18 <20% TTS OD No 3 19 <20% TTS OD No 4 19 <20% TTS OD No

'5 19 <20% TTS OD Na 6 19 <20% TTS OD No 7 19 <20% TTS OD No 9 19 22% TTS OD No 10 19 28% TTS OD No 11 19 <20% TTS OD No 12 19 <20% TTS OD No 15 19 <20% TTS OD No 4 20 <20% TTS OD No 5 20 <20% TTS OD No 6 20 <20% TTS OD No i 7 20 <20% TTS OD No 8 20 <20% TTS OD No 9 20 <20% TTS OD No 10 20 <20% TTS OD No 12 20 <20% TTS OD No 13 20 <20% TTS OD No 14 20 <20% TTS OD No

' 15 20 <20% TTS OD No 17 20 <20% TTS OD No 5 21 43% TTS OD Yes 6 21 <20% TTS OD No 8 21 30% TTS OD No 10 21 <20% TTS OD No 12 21 <20% TTS OD No 13 21 <20% TTS OD No t

Pow Colunin Indication Location Origin Pluqqed 14 21 <20% TTS OD No 15 21 <20% TTS OD No 16 21 <20% TTS OD No 17 21 <20% TTS OD No 18 21 <20% TTS OD No 3 22 39% TTS OD No 7 22 <20% TTS OD No 8

• 22 <20% TTS OD No 9 22 <20% TTS OD No

-10 22 <20% TTS OD No

. 13 22 21% TTS OD No 17 22 <20% TTS OD No

'18 22 <20% TTS OD No 20 22 <20% TTS OD No 23 22 <20% TTS OD No 26 22 <20% TTS OD No 3 23 34% TTS OD No 4 23 <20% TTS OD No 5 23 <20% TTS OD No 6 23 <20% TTS OD No 7 23 <20% TTS OD No 8 23 23% TTS OD No 9 23 <20% TTS OD No 10 23 <20% TTS CD No 12 23 23% TTS OD No 13 23 <20% TTS OD No 16 23 <20% TTS OD No 17 23 <20% TTS OD No 18 23 21% TTS OD No 19 23 22% TTS OD No 20 23 24% TTS OD No 21 23 <20% TTS OD No 23 23 <20% TTS OD No 25 23 <20% TTS OD No 26 23 <20% 1TS OD No 5 24 <20% TTS OD No 6- 24 ~ <20% TTS OD No 7 24 <20% TTS OD No 8 24 <20% TTS OD No 9 24 <20% TTS OD No 11 24 <20% TTS OD No 12 24 <29% TTS OD No 13 24 <20% TTS OD No i 16 24 31% TrS OD No

-17 24 <20% TTS OD No 18 24 89% 10-12" ATE OD. Yes 19 24 38% TTS OD No 21 24 <20% TTS OD No 24 24 f20% TTS OD No 6_ 25 28% TTS OD No

-7 25 < 20*6 TTS OD No

O Row Column Indication Location Origin Plugged 8 25 <20% TTS OD No 9- 25 <20% TTS OD No 10 25 <20% TTS OD No 11 25 <20% TTS OD No 12 25 <20% TTS OD No 13 25 21% TTS OD No 14 25 <20% TTS OD No 15 . 25 <20% TTS OD No la 25 35% TTS OD No 19 25 35% TTS OD No 20 25 39% TTS OD No 21 25 <20% TTS OD No 22 25 <20% TTS OD No 4 26 < 20*; TTS OD No 6 26 <20% TTS OD No 7 26 <20% TTS OD Nc-

, 8 26 <20% TTS OD No 9 26 <20% TTS OD No

. 10 26 <20% TTS OD No 11 26 <20% TTS OD No 12 26 <20%/26% TTS/%"ATS OD No 13 26 <20% TTS C3 No 14 26 <20% TTS OD No 15 26 <20% TTS OD No 16 26 <20% TTS OD No 17 26 28% TTS OD No 18 26 38% TTS OD No 19 26 24% TTS OC No 20 26 32% TTS OD No 21 26 <20% TTS OD No 4 27 <20% TTS OD No 5 27 <20% TTS OD No 6 27 24% TTS OD No

~7 27 27% TTS OD No 8 27 <20% TTS OD No 9 27 <20% TTS OD No 11 27 <20% TTS OD No 12 27 <20% TTS OD No 13 27 <20% TTS OD No 14 27 <20% TTS OD No j 16 27 <20% TTS OD No 17 27 <20% TTS OD No 15 27 39% TTS OD No 19 27 37% TTS OD No 20 27 27% TTS GD Nc

~ 21 27 <20% TTS OD No 26 27 81%/UDI 5b-17% ATE OD Yes 5 28 31% TTS OD No 6 28 36% TTS OD No 7 28 34% TTS OD No 8 28 23% TTS OD No

Row Column Indication Location Origin _Pluqqed 11 28 <20% TTS OD No 12 28 <20% TTS OD No 13 28 <20% TTS OD No 14 28 <20% TTS OD No 18 28 <20% TTS OD No 19 28 39% -TTS OD No 20 28 26% TTS OD No 21 12 8 <20% TTS OD No 23 -

28 28% TTS OD No 5 29 <20%- TTS OD No 6' 29 <20% TTS OD No 8 29 <20% TTS OD No 9 29 <20% TTS OD No 10 29 <20% TTS OD No 11 29 <20% TTS OD No 12 29 <20% TTS OD No 4 30 <20% TTS OD No

, 6 30 <20% TTS OD No 7 30 <20% TTS OD No 8 30 <20% TTS OD No 11 -30 <20% TTS OD No 12 30 <20% TTS- OD No 13 30 <20% TTS OD No 16: 30 100% 5%-16" ATE OD Yes 17' 30 81%/UDI 12%" ATE /6-14" ATE OD Yes 4 31 <20% TTS OD No 8 31 <20% TTS OD No 13 31 <20% TTS OD No 14 31 . <30% TTS OD No 16 31- UDI 9-15" ATE OD Yes 17 .31 UDI ~12%-16" ATE OD Yes 8 32 <20% TT" OD No 12 32 <20% TTS OD No 13 32 <20% TTS OD No 14 32 <20% TTS OD No 27 32 ' <20% TTS OD No

6 33 <20% TTS OD No 11 33 <20% TTS OD No 12 33 ~<20% TTS OD No 13 33 <20% TTS OD No 18 33 84%/UDI 6" ATE /6-15" ATE OD Yes

.21 33 <20% TTS OD No

.10- 3<4 <20% TTS OD No 11 34 <20% TTS OD No 12 .34 32%- TTS OD No 171 34 86%/UDI 7%" ATE /7-16" ATE OD Yes 20 -- 34 <20%/21% TTS/%" ATS OD No 21 34 <20% TTS OD No 27~ 34 23% TTS OD No 9 35 <20% TTS OD No 12- .35 <20% -TTS OD No

f.

Row Column Indication Location Origin Pluqqed 13 35 <20% TTS OD No 21 35 34% TTS OD No 22 35 <20% \" ATS OD No 26 35 22% TTS OD No 7 36 <20% TTS OD No 9 36 <20% TTS OD No 10 36 <20% TTS OD No 11 36 <20% TTS OD No 12 36 <20% TTS OD No 20 36 UDI 13-18" ATE OD Yes 7 37 <20% TTS OD No 8 37 <20% TTS OD No 9 37 <20% TTS OD No 10 37 31% TTS OD No 11 37 <20% TTS OD No 12 37 <20% TTS OD No 13 37 26% TTS OD No 14 37 <20% TTS OD No 17 37 UDI 3\-7\" ATE OD Yes 19 37 88%/UDI 5" ATE /4\-17h" ATE OD Yes 20 37 100% 6-15" ATE OD Yes 9 38 <:20% TTS OD No 10 38 <20% TTS OD No 17 38 UDI 6-18" ATE OD Yes IC 38 <20% 1" ATS OD No 19 38 <20% 1" ATS OD No 8 39 <20% TTS OD No 9 39 35% TTS OD No 11 39 .<20% TTS OD No 12 39 23% \" ATS OD No 17 39 41%/89%/UDI TTS/14" ATE /10-17\" ATE OD Yes 18 39 <20% 1" ATS OD No 7 40 <20% TTS OD No 8 40 <20% TTS OD No 11 40 <20% TTS OD No 12 40 <20% TTS OD No 16 40 37% TTS OD No 25 40 22% TTS OD No 6 41 <20% TTS OD No 8 41 <20% TTS OD No 16 41 <20% 1" ATS OD No 25 41 <20% TTS OD No 33 -41 <20% TTS OD No 6 42 <20% TTS OD No 8 42 <20% TTS OD No 13 42 <20% TTS OD No 17 42 100% 3-16" ATE OD Yes 18 42 68% 16" ATE OD Yes 20 42 25% \" ATS OD No 24 42 <20% TTS OD No 25 42 <20% TTS OD No

Row Column Indication Location Origin Pluqqed 26 42 <20% TTS OD No 6 43 <20% TTS OD No 8 43 <20% TTS OD No 16 43 91% 5\-18\" ATE OD Yes 19 43 33% TTS OD No 23 43 <20% TTS OD No 25 43 <20% TTS OD No 26 43 22% TTS OD No 27 -

43 <20% TTS OD No 28 43 30% TTS OD No 5 44 <20% TTS OD No 6 44 <20% TTS OD No 10 44 <20% TTS OD No 17 44 UDI 8\-13" ATE OD Yes 18 44 UDI 4\-14" ATE OD Yes 20 44 <20% TTS OD No 24 44 <20% TTS OD No 25 44 <20% TTS OD No 27 44 26% TTS OD No 28 44 <20% TTS OD No 29 44 29% TTS OD No 31 44 24% TTS OD No 6' 45 <20% TTS OD No 7 45 <20% TTS CD No 9 45 <20% TTS OD No 14 45 UDI 13-19\" ATE OD Yes 21 45 <20% TTS OD No 22 45 <20% TTS OD No 29 45 <20% TTS OD No 31 45 <20% TTS OD No 5 46 <20% TTS OD No 6 46 <20% TTS OD No 9 46 <20% TTS OD No 10 46 25% TTS OD No 14 46 <20% TTS OD No 15 46 UDI 8\-16" ATE OD Yes 18 46 24% 3" ATS OD No 20 46 31% TTS OD No

'21 46 30% TTS OD No 23 -46 <20% TTS OD No 6 47 22% TTS OD No 7 47 <20% TTS OD No 8 47 <20% TTS OD No 9 47 UDI 11\-17\" ATE OD Yes 13 47 UDI 11\-16" ATE OD Yes 14 47 <20%/UDI \" ATS/14-18" ATE OD Yes 15 47 <20% TTS OD No 19 47 <20%/UDI \" ATS/7-16" ATE OD Yes 32 .47 <20% TTS OD No 2 48 <20% TTS OD No 5 48 <20% TTS OD No

e -

Row Column Indication Location Origin Plugged 7 48 <20% TTS OD No 12 48 <20%/UDI TTS/8h-16" ATE OD Yes 13 48 (20% TTS OD No 14 48 89% 12-15" ATE OD Yes 32 48 24% TTS OD No 4 49 <20% TTS OD No 6 49 <20% TTS OD No 7 ,

49 21% TTS OD No 8 49 <20% TTS OD No 12 49 24% TTS OD No 13 49 24% TTS OD No 32 49 27% TTS OD No 5 50 <20% TTS OD No 6 50 <20% TTS OD No 7 50 22% TTS OD No 8 50 <20% TTS OD No 9 50 <20% TTS OD No 11 50 35% TTS OD No 12 50 <20% TTS OD No 13 50 <20% TTS OD No 20 00 38% 1" ATS OD No 28 50 <20% TTS OD No 7 51 <20% -TTS OD No 8 51 <20% TTS OD No 10 51 <20% TTS OD No 11 51 <20% TTS OD No 12 51 <20% TTS OD No 18 51 27% " ATS OD No 21 51 <20% TTS OD No 22 51 22% TTS OD No 23 51 27% TTS OD No 24 51 <20% TTS OD No 27 51 <20% TTS OD No 9 52 <20% TTS OD No 10 52 <20% TTS OD No 11 52 <20% TTS OD No 14 52 <20% TTS OD No 21 52 29% TTS OD No 22 52 <20% TTS OD No 23 52 23% TTS OD No 24 52 <20% TTS OD No 38 52 <20% 1" ATS OD No 5 53 <20% TTS OD No 6 53 <20% TTS OD No 9 53 <20% TTS OD No 11 53 <20% TTS OD No 21 53 22% TTS OD No 22 53 <20% TTS OD No 23 53 28% TTS OD No 26 53 <20% TTS OD No 30 53 <20% TTS OD No

e a Row Column Indication Location Origin Plugged 31 53 <20% TTS OD No 32 53 <20% TTS OD No 6 54 <20% TTS OD Wo 18 54 <20% TTS OD No 20 54 <20% TTS OD No 21 54 <20% TTS OD No 22 54 22% TTS OD No 24 , 54 28% TTS OD No 30 54 <20% TTS OD No 31 54 <20% TTS OD No 7 55~ <20% TTS OD No 8 55 <20% TTS OD No 20 55 33% TTS OD No 21 55 25% TTS OD No 22 55 22% TTS OD No

- 24 55 22% TTS OD No 29 55 <20% TTS OD No

.30 55 <20% TTS OD No 9 56 <20% TTS OD No 10 56 <20% TTS OD No 18 56 <20% TTS OD No 19 56 <20% TTS OD No 20 56 <20% TTS OD No 21 56 <20% TTS OD No 24 56 <20% TTS- OD No 29 56 <20% TTS OD No 9 57 28% TTS OD No

. 11 57 22% TTS OD No 12 57 <20% TTS OD No 19 57 4

<20% TTS OD ho 20 57 <20% TTS OD No 23 57 <20% TTS OD. No 24 57 <20% TTS OD No 9 58 <20% TTS OD No 12 58 <20% TTS OD No 17 58 <20% TTS OD No 19 58 <20% TTS OD No 20 58 <20% TTS OD No 23 58 <20% TTS OD No 26 58 29%- TTS OD No 36 58 30% 12" ATE OD No 8 59 <20% TTS OD No 9 59 <20% TTS OD No 12 59 34%  %" ATS OD No 19 59 <20% TTS OD No 27 59 <20% TTS OD No

'6 60 <20% 2%" ATS OD No 12 60' <20% TTS OD No 26 60 <20% TTS OD No

, 27- 60 <20% TTS OD No 10 61 <20% TTS OD No s

Row Column Indication Location Origin Plugged 11 61 <20% TTS OD No 17 61 <20% TTS OD No 21 61 <20% TTS OD No 22 61 <20% TTS OD No 24 61 <20% TTS OD No 25 61 <20% TTS OD No 26 61 <20% TTS OD No 6 ,

62 21% TTS OD No 7 62 <20% TTS OD No 9 62 <20% TTS OD No 10 62 21% TTS OD No 11 62 <20% TTS OD No 13 62 <20% TTS OD No 14 62 26% TTS OD No 17 62 <20% TTS OD No 18 62 <20% TTS OD Ho 19 62 <20% TTS OD No 20 62 24% TTS OD No 23 62 <20% TTS OD No 24 62 <20% TTS OD No 25 62 <20% TTS OD No o 10 63 <20% TTS OD No 17 63 26% TTS OD No 13 63 <20% TTS OD No 22 63 25% TTS OD No 23 63 <20% TTS OD No 24 63 22% TTS OD No 5 64 <20% TTS OD No 9 64 26% TTS OD No 12 64 <20% TTS OD No 19 64 21% TTS OD No 20 64 23% TTS OD No 22 64 <20% TTS OD No 10 65 <20% TTS OD No 11 65 <20% TTS OD No 12 65 <20% TTS , OD No

-15 65 <20% TTS OD No 20 65 <20% TTS OD No 21 65 <20%/26% 12" 1 TSP /TTS OD No 22 65 <20% TTS OD No 8 66 <20% TTS OD No 9 66 <20% T'IS OD No 12 66 <20% TTS OD No 14 66 <20% TCS OD No 16 66- <20% TTS OD No 17 66 <20% TTS OD No 20 66 <20%/22% TTS/1 TSP OD No

- 21 66 25% TTS OD No 8 67 <20%~ TTS OD No 12 67 <20% TTS OD No 17' 67 <20% TTS OD No

s Row Column Indication Location Origin Pluqqed 19 67 <20% TTS OD No 20 67 24% ITSP-TTS OD No 21 67 <20% TTS OD No 6 68 <20% TTS OD No 7 68 <20% TTS OD No 16 68 <20% TTS OD No 17 68 <20% TTS OD No 19 68 29% TTS OD No 20 68 <20% TTS OD No 6 69 <20% TTS OD No 7 69 <20% TTS OD No 12 69 <20% TTS OD No 17 69 <20% TTS OD No 18 69 <20% TTS OD No 19 69 <20% TTS OD No 20 69 <20% TTS OD No 12 70 <20% TTS OD No 13 70 <20% TTS OD No 16 70 <20% TTS OD No 17 70 <20% TTS OD No 18 70 <20% TTS OD No 19 70 <20% TTS OD No 13 71 <20% TTS OD No 16 71 <20% TTS OD No 18 71 <20% " ATS OD No 12 72 <20% ITSP-TTS OD No 13 72 <20% TTS OD No 14 72 <20% TTS OD No 16 72 <20% TTS OD No 17 72 <20% TTS OD No 4 73 21%  %" ATS OD No 13 73 <20% TTS OD No 14 73 <20% TTS OD No 17- '73 <20% ITSP-TTS OD No 12 74 23% TTS OD No

~13 74 <20%/28% TTS OD No 14 74 38%/39%  %" ATS/ " ATS OD No 6 75 <20% TTS OD No 10 75 <20% TTS OD No 11 75 <20% TTS OD No 12 75 <20%/23% TTS/TTS OD No 3 76 <20% TTS OD No

,. 6 7e <20% TTS OD No 7 76 22% TTS OD No 10 76 <20% TTS OD No 11 76 21% TTS OD No 13 76 <20% ITSP-TTS OD No 3 77 <20% TTS OD No 4 77 24% TTS OD No 8 77 <20% TTS OD No 9 77 <20% TTS OD No

s Row Column Indication Location Origin Pluqqed 10 77 <20% TTS OD No 10 78 <20% TTS OD No 4 79 <20% TTS -

OD No 10 79 <20% TTS OD No 9 80 21% TTS OD No 1 90 83% 4\-8" ATE OD Yes e

,, --,7 , ., e , , - . ..--e - , , - - - - . - ~..e-.,,,,.v. , - ,.- ,n,,,.ee

4

. "B . Steam - Generator

, c liot Leg r Rov Column Indication Location Origin Pluqqed 6 16 <20% ~ TTS OD No

~

5 ' 17 <20% TTS -

OD No 5 18 25% _ \" ATS OD No 7 18 <20% TTS OD No

/ 10 -

18 <20% TTS OD No 11 18 - <20% '

TTS .. OD No 7' 19 <20% TTS OD No (6 19 <20% TTS OD No 6 21 38% TTS OD No 17 21 <20% TTS OD No 7 , 2?q <20%  %" ATS , 00- No

'6' 23 ~ <20% TTS. OD No 8 23 ~ ~ _ <20% 1" ATS OD No 14 23 <20% TTS OD No 1G' _ 23 .<20% TTS OD No 31 24 ' 23 f<20% TTS OD No 9 24 <20%  %" ATS OD No 14 i 24 <20% TTS OD No

. 317 24 ~

23% TTS OD No 19 24 <20% TTS OD No 22 24 <20% TTS OD No

'24 24 <20% TTS OD No 9 25 '- -22%' TTS OD No 12 '25 ', <20% TTS OD No 13 25 14 25

', ' <20%

'20%

TTS TTS OD OD No No 23 25 <20% TTS OD No

. 25 25 27% 7TS OD No

9. '

26 31%/28% 1" ATS/%" ATS OD No

,10 . 26' 23% . 1" ATS OD No

.. 21 -26 <20%

~TTS ~ OD No 24 26 ,

35%  %" ATS OD No 25 26 35% TTS ' OD No 12  : 27 ~ <20% TTS OD No 18 27 . . <20% TTS OD No 19 27 ' <20% " ATS OD No 21 27' <20% TTS OD No 22 27- <20% TTS OD No c9 28 25% '1"'ATS OD No 17 ' 28 51% TTS OD Yes 21- 28  ;<20%. TTS OD No 27 , 28 <20% TTS OD No

' 6~ ' 29 <20% .TTS OD No 13 - 29 <20% TTS OD No 18 29 ~ <20% *S OD No 23 ; 29 23% '14. OD No 24 29 ~ <20% , TTS OD No

'6 .

30 . -26% TTS OD No

, < v .

N ,

4 #

, ' 3 ,

g '<

^"

.- ,c,.-  %, . ~ , . . . - . , , , . , . - , , . - . , - - - - -. - . . -

Es Row Column Indication Location Origin Pluqqed 22 30 26% 1" ATS OD No 23 30 26% TTS OD No 6 31 <20% TTS OD No 7 31 <20% TTS OD No 21 31 <20% TTS OD No 23 31 <20% 2" ATS OD No 18 32 34% 1" ATS OD No 19 32 <20% TTS OD No 21 -

32 <20% TTS OD No 23 32- 34% 2\" ATS OD No 6 35 <20% TTS OD No 7 35 21% \" ATS OD No 9 35 <20%/28% TTS/1" ATS OD No 5 36 <20% TTS OD No 6 36 <20% TTS OD No 9 36 <20% TTS OD No 11 36 24% 1" ATS OD No 28 36 <20% TTS OD No 5 37 <20% TTS OD No 6 37 <20% TTS OD No 7 37 29% 1" ATS OD No 5 38 <20% TTS OD No 7 38 23% 1" ATS OD No 8 38 29% 1" ATS OD No 14 38 35% \" ATS OD No 28 38 <20% TTS OD No 30 38 32% TTS OD No 5 39 <20% TTS OD No 6 39 <20% TTS OD No 7 39 , 30% 1" ATS OD No 14 39 <20% \" ATS OD No 28 39 <20% TTS OD No 29 39 25% TTS OD No 32 39 25% \" ATS OD Mo 33 39 211 \" ATS OD No 5 40. 32%/37% TTS/TTS OD No 7 40 30% TTS OD No 9 40 33% \" ATS OD No 13 40 <20% TTS OD No 28 40 <20% TTS OD No 29 40 32% TTS OD No 32 40 33% \" ATS OD No 6 41 <20% TTS OD No 7 41 31% TTS OD No 9 41 <20% TTS OD No 12 41 <20% TTS OD No 26 41 22% TTS OD No 28 41 <20% TTS OD No 33 41 29% \" ATS OD No 5 42 <20% TTS OD No 6 42 <20% TTS OD No

_m _ _ _

Row- Column IndicatiEn Location Origin Pluqqed 9 42 27% TTS OD No 28 42 <20% TTS OD No 31 42 <20%  %" ATS OD No 32 42 24% TI S - OD No 7 43 32%' TTSf OD No 13 43 <20% -\" ATS OD No 19 43 30% , TTS OD No 24 , 43 <20% TTS OD No 5 44 29% TTS OD No 7 44 33% TTS OD No 10 44 <20% TTS OD No 11 44 30% 1" ATS CD No 12 44 21% 1" ATS OD No 19 44 26% TTS OD No 21 44 21% TTS OD No 22 44 <20%/23% TIS /4\" ATS OD No 23 44 36% TTS OD No

.24 44 <20% TTS OD No 28 44 <20% TTS- OD No 29 44 <20% TTS OD No 30 44 33% TTS OD No 31 44 25% .TTS OD No 5 45- <20% TTS OD No 6 45 33% TTS OD No 10 45 <20% TTS- OD No 11 45 <20% TTS OD No 15 45 _< 20% TTS OD No 22 45 35% TTS OD No 23 45 <20% TTS OD No 29 45 <20% TTS OD No 9 46 -39% 1" ATS OD No 10 46 <20%~ TTS OD No 11 46 <20%. TTS OD No 13 46 -<20% TTS OD No 22 46 29% TTS OD No 32 46 38% TTS OD No 7 47 40% 1\" ATS OD No 10 47 <20% TTS OD No 11 47 1 <20% TTS OD No 13 47 <20% - TTS -

OD No 14 47 i 25% TTS OD No

'16- 47 <20% " ATS OD No 17 47 <20% \" ATS OD No 18 47 <20% \" ATS OD No 21 47, 32% TTS OD No 22 ' 47

<20% \" ATS OD No

-23 47 <

~20% \" ATS OD No a . 24 47. <20% \" ATS OD No 26 47 <20% TTS OD No 6 48 34% 1" ATS OD No 7 ,

48 .35% 1" ATS OD No a

+ -

Row Column Indication Location Origin Pluqqed 9 48 43% 1" ATS OD Yes 11 48 <20% TTS OD No 12 48 <20% TTS OD No ,

13 48 <20% T75 OD No 14 48 <20% TTS OD No 17 48 <20%  %" ATS OD Nc 21 48 33% TTS OD No 22 48 <20%/32% 2%" ATS/TTS OD No 23 48 <20% 1" ATS OD No 24 48 <20% TTS OD No 25 48 <20% TTS OD Nc 6 49 28% 1" ATS OD No 9 49 27% 1" ATS OD No 12 49 32% TTS OD No 13 49 <20% TTS OD i!o 14 49 <20% TTS OD No 17 49 <20% TTS OD No 18 49 <20% TTS OD No 21 49 25% TTS OD No 22 49 <20% TTS OD No 3 23 49 <20% TTS OD No 24 49 <20% " ATS OD No 25 49 25% TTS OD No 26 49 35% TTS OD No 27 49 <20% TTS OD lio 28 49 <20% TTS OD No 29 49 <20% TTS OD No 11 50 <20% TTS OD No 13 50 <20% TTS OD No 18 50 28%/32% 1" ATS/TTS OD No 19 50 <20%/36% TTS/TTS OD No 21 50 <20% TTS CD No 22 50 36% TTS OD No 23 50 33% TTS OD No 25 50 31% 1%" ATS OD No 26 50 <20% TTS OD No 27 50 <20% TTS OD No 28 50 34% TTS OD No 29 50 35% TTS OD No 4 51 <20% TTS OD No 6' 51 <20%/27% TTS/1" ATS OD No 9 51 <20% TTS OD No 18 51 <20% TTS OD No 22 51 <20% TTS OD No 23 51 <20% TTS OD No 25 51 <20%/25% TTS/l" ATS OD No 28 51 33% TTS OD No 9 52 33% 1" ATS OD No 11 52 <20% TTS OD No 24 52 27% TTS OD No 25 52 <20%  %" ATS OD No 44

e Row Column Indication Location Origin Pluqqed 28 52 <20%  %" ATS OD Ne 29 52 23*, TTS OD No 11 $3 <20% TTS OD No 23 53 <20% 1'- ATS OD No 25 53 <20% TT3 OD No

~6 53 <20% TTS OD No 28 53 23% TTS OD No 11 , 54 <20% TTS OD No 17 54 <20% TTS OD No 21 f4 <20% TTS OD No 22 54 21% 1" ATS OD No 25 54 <20% TTS OD No 27- 54 <20% TTS OD No 28 54 <20% TTS OD No 29 54 <20% TTL OD No 22 SS <20% TTS OD No 26 55 <20% TTS OD No

.23 56 <20% TIS OD No 24 56 <20% TTS OD No 27 56 <20% TTS OD No 10 57 (20% TTS OD No 21 57 <20% TTS OD No 17 SB <20% TTS OD No 21 56 33% TTS OD No 23 58 30%  %" ATS OD No 7 59 39%  %" ATS OD No i

9 59 <20% TTS OD No 21 ~59 <20% TTS OD No 22 59 <20% TTS OD No 23 59 35%  %" ATS OD No 16 60 <20% TTS OD No

17. 60 <20% TTS OD No 23 60 32%  %" ATS OD No s23 61 <20% TTS OD No 14 64 <20% TTS OD No 7 65 <20% TTS OD No 9 65 <20% TTS OD No 23 65 <20% TTS OD No 7 66 <20% TTS OD No 15 66 <20% TTS OD No 23 66 <20% TTS OD No 7 67 <20% TTS OD No 10 .68 <20% TTS OD No 11 68 <20% TTS OD No 5 73 26% TTS OD No 4 74 <20%' TTS OD No 5 74 <20% TTS OD No 7 74 <20% TTS OD No 9 .74 <20% TTS OD No 9 75 25% TTS OD No i .-

Row Column Indication Location Origin Plugged 10 76 23% TTS OD No 6 77 <20% 'iTS OD No During the outage, a total of 1,501 sleeves were installed in the "A" steam generator and 1,500 sleeves were installed in the "B" steam generator. Inprocess and baseline eddy current examinations were performed for all tubes sleeved. No indications other than nondetectable and minor ID variations (dent ripples) were identified &tring the course of these inspections.

i E

List of Sleeved Tubes "A" Steam Generator.

R02C20 R03C20 R04C20 R05C17 R05C68 R06C64 R02C21 R03C21 R04C21 R05C18 R05C69 R06C65 R02C22 R03C22 R04C22 R05C19 R05C70 R06C66 R02C23 R03C23 R04C23 R05C20 R06C16 R06C67 R02C24 R03C24 R04C24 R05C21 R06C17 R06C68 R02C25 R03C25 R04C25 R05C22 R06CIC R06C69

.R0ic26 R03C26 R04C26 R05C23 R06C19 R06C70 R02C27 R03C27 R04C27 R05C24 R06C20 R07C16 R02C28 R03C28 R04C28 R05C25 R06C21 R07C17 R02C29 R03C29 R04C29 R05C26 R06C22 R07C18 R02C30 R03C30 R04C33 R05C27 R06C23 R07C19 R02C31 R03C31 R04C31 R05C28 R06C24 R07C20 R02C32 R03C32 R04C32 R05C29 R06C25 R07C22 202:33 R03C33 R04C33 RC5030 R06C26 R07C23 R02C34 R03C34 R04C34 R05C31 R06C27 R07C24 R02C35 R03C35 R04C35 R05C32 R06C28 R07C25 R02C36 R03C36 R04C36 R05C33 R06C29 R07C26 R02C37 R03C37 R04C37 R05C34 R06C30 R07C27 R02C38 R03C38 R04C38 R05C35 R06C31 R07C28 R02C39 R03C39 R04C39 R05C36 R06C32 R07C29 R02C40 R03C40 R04C40 R05C37 R06C33 R07C30 R02C41 R03C41 R04C41 R05C38 R06C34 R07C31 R02C42 R03C42 R04C42 R05C39 R06C35 R07C32 R02C43 R03C43 R04C43 R05C40 R06C36 R07C33 R02C44 R03C44 R04C44 R05C41 R06C37 R07C34 R02C45 R03C45 R04C45 R05C42 206C38 R07C35 R02C46 R03C46 R04C46 R05C43 R06C39 R07C36 R02C47 R03C47 R04C47 R05C44 R06C40 R07C37 R02C48 R03C48 R04C48 R05C45 R06C41 R07C38 R02C49 R03C49 R04C49 R05C46 R06C42 R07C39 ,

R02C50 R03C50 R04C50 R05C47 R06C43 R07C40 R02C51 R03C51 R04C51 R05C48 R06C44 R07C41 R02C52 R03C52 R04C52 R05C49 R06C45 R07C42 R02C53 R03C53 R04C53 R05C50 R06C46 R07C43 R02C54 R03C54. R04C54 R05C51 R06C47 R07C44 R02C55 R03C55- R04C55 R05C52 R06C48 R07C45 R02C56 R03C56 R04C56 R05C53 R06C49 R07C46 R02C57 .R03C57 R04C57 R05C54 R06C50 R07C47 R02C58 R03C58 R04C58 R05C55 R06C51 R07C48 R02C59 R03C59 R04C59 R05C56 R06C52 R07C49

.R02C60 R03C60 R04C60 R05C57 R06C53 R07C50 R02C61 R03CC1 R04C61 R05C58 R06C54 R07C51 R02C62 R03C62 R04C62 R05C59 R06C55 R07C52 R02C63 R03C63 R04C63 R05C60 R06C56 R07C53 R02C64 R03C64 R04C64 R05C61 R06C57 R07C54 R02C65 R03C65 R04C65 R05C62 R06C58 R07C55 R02C66 R03C66 R04C66 R05C63 R06C59 R07C56 R02C67 R03CG7 R04C67 R05C64 R06C60 R07C57

R02C68 R03C68 R04C68 R05C65 R06C61 R07C58 R02C69 R03C69 R04C69 R05C66 R06C62 R07C59 R02C70 R03C70 R04C70 R05C67 R06C63 R07C60

.. i

m _ _ _ _ . _ . _ .. . _ ._ . . _ _

. t

, R07C61 R08C54 R09C47 R10C42 R11C40 R12C39 R07C62 R08C55 R09C48 R10C43 R11C41 R12C40 R07C63 R08C56 R09C49 R10C44 R11C42 R12C42 i R07C64 R08C57 R09C50 R10C46 R11C43 R12C45 j R07C65 R08C58 R09C51 R10C47 R11C44 R12C46 -

R07C66 R08C59 R09C52 R10C48 R11C46 R12C47 4

'R07C67 R08C60 R09C53 R10C49 R11C47 R12C48 R07C68 R08C61 R09C55 R10C50 R11C48 R12C49 '

R07C69- R08C62 R09C56 R10C51 R11C49 R12C50

.R07C70 R08C63 R09C57 R10C52 R11C50 R12C51 R07C71. R08C64 R09C58 R10C53 R11C51 R12C52 R07C72 R08C65 R09C59 R10C54 R11C52 R12C53 R07C73 R08C66 R09C60 R10C55 R11C53 R12C54 R07C74 'ROSC67 R09C61 R10C56 R11C54 R12C55 R07C75 R08C68 R09C62 R10C57- R11C55 R12C56 s 'R07C76 R08C69 R09C63 R10C60 R11C56 R12C57 R08C16 R08C70 R09064 R10C61 R11C57 R12C58 ,

.R08C17 R08C71 R09C65 R10C62 R11C58 R12C59 R08C18 R08C72 R09C66 R10C63 R11C59 R12C60 R08C19 R08C73- R09C67 R10C65 R11C60 R12C61 R08C20 R08C74 R09C68 R10C66 R11C61 R12C62 R08C21 R08C75 R09C69 R10C67 R11C62 R12C63

' R08C22 - R08C76 R09C70 R10C68 R11C63 R12C64 R08C23' R09C16 R09C71 R10C69 R11C64 R12C65 R08C24 R09C17 R09C72 R10C70 R11C65 R12C66 R08C25- R09C18 R09C73 -R10C71 R11C66 R12C67

-R08C26 R09C19 R09C74* R10C72 R11C67' R12C68 R08C27 .R09C20 R09C75 R10C73 R11C68 R12C69 R08C28 R09C21 R09C76' R10C74 R11C69 R12C70 R08C29 R09C22 R09C77 R10C75 R11C70 R12C72 R08C30 R09C23 R10C16 R10C76 R11C71 R12C74 R08C31 R09C24~ R10C17 R10C77 R11C72 R12C75

.R08C32 R09C25 R10C18 R11C18 R11C73 R12C76 R08c33 R09C26 R10C19 R11C19- R11C75 R13C18 R08C34 R09C27 R10C20 R11C20 R11C76 R13C20 R02C35- R09C28- R10C21 R11C21 R12C18 R13C21 4 R08C36 R09C29 R10C22 R11C22 R12C19 R13C22

i. R08C37 .R09C30 .R10C23 R11C23 R12C20 R13C23

'R08C38 R09C31 R10C25 R11C24 R12C21 R13C24 R08C39 R09C32 R10C26 R11C25 R12C13 R13C25 R08C40-R09C33 R10C27 E11C26 R12C24 R13C26

.R08C41 'R09C34 R10C28 R11C27 R12C25 R13C27

. R08C42 R09C35- R10C29 R11C28 R12C26 R13C28

-R08C43 R09C36 _R10C30 R11C29 R12C27 R13C29

'R08C44 R09C37 R10C31 R11C30 R12C28 R13C30

.R08C45 R09C38 R10C32 R11C31 R12C29 R13C31 R08C46' R09C39 R10C33 R11C32 R12C30 R13C32 R08C47. R09C40 R10C34 R11C33 R12C32 R13C33 R08C48 R09C41 R10C35 R11C34 R12C33 R13C35 >

R08C49 R09C42 RICC3G R11C35 R12C34 R13C37 R08C50 R39C43 R10C37 R11C36 R12C35 R13C38 ROSC51 R09C44' R10C38 R11C37 R12C36 R13C39 R08C52 R09C45 -R10C40 R11C38 R12C37 R13C40 LR08C53 R09C46 R10C41 R11C39 R12C38 R13C42 1

'

• sleeved tube subsequently plugged.

o

.i

• e R13C43 R14J41 R15C36 R16C31 R17C26 R18C27 R13C45 R14C42 R15C37 R16C32 R17C27 R18C28 R13C47 R14C43 R15C38 R16C33 R17C28 R18C29 R13C48 R14C44 R15C39 R16C35 R17C29 R18C30 R13C49 R14C45 R15C40 R16C36 R17C30 R18C31 R13C50 R14C46 R15C41 K16C37 R17C31 R18C32 R13C51 R14C47 R15C42 R16C38 R17C32 R18C33 R13C52 R14C48 R15C43 R16C39 R17C34 R18C34 R13C53 R14C49 R15C44 R16C40 R17C35 R18C35 R13C54 R14C50 R15C45 R16C41 R17C37 R18C36

• R13C55 R14C51 R15C46 R16C42 R17C38 R18C38 R13C56 R14C52 R15C47 R16C43 R17C39 R18C39 R13C57 R14C53 R15C48 R16C44 R17C42 R18C40 R13C58 R14C54 R15C49 R16C45 R17C43 R18C42 R13C59 R14C55 R15C50 R16C46 R17C44 R18C44 R13C60 R14C56 R15C51 R16C47 R17C46 R18C45 R13C61 R14C57 R15C52 R16C48 R17C47 R18C46 R13C62 R14C58 R15C53 R16C49 R17C48 R18C47 R13C63 R14C59 R15C54 R16C50 R17C49 R18C48 R13C64 R14C60 R15C55 R16C51 R17C50 R18C49 R13C65 R14C61 R15C56 R15C52 R17C51 R18C50 R13C66 R14C62 R15C57 R16C53 R17C52 R18C51 R13C67 R14C63 R15C58 R16C54 R17C53 R18C52 R13C68 R14C64 R15C59 R16C55 R17C54 R18C53 R13C69 R14C65 R15C60 R16C56 R17C55 R18C54 R13C70 R14C66 R15C61 R16C57 R17C56 R18C55 R13C71 R14C67 R15C62 R16C58 R17C57 R18C56 R13C72 R14C68 R15C63 R16C59 R17C58 R18C57 R13C73 R14C69 R15C64 R16C60 R17C59 - R18C58 R13C74 R14C70 R15C65 R16C61 R17C60 R18C59 R13C75 R14C71 R15C66 R16C62 R17C61 R18C60 R13C76 R14C72 R15C67 R16C63 R17C62 R18C61 R14C18 R14C73 R15C68 R16C64 R17C63 R18C62 R14C19 R14C74 R15C69 R16C65 R17C64 R18C63 R14C20 R14C75 R15C70 R16C66 R17C65 R18C64 R14C21 R14C76 R15C71 R16C67 R17C66 R18C65 R14C22 R15C18 R15C72 R16C68 R17C67 R18C66 R14C23 R15C19 R15C74 R16C69 R17C68 R18C67 R14C24 R15C20 R15C75 R16C70 R17C69 R18C69 l

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10 6.0 REACTOR COOLANT SYSTEM RELIEF VALVE CHALLENGES There was one challenge to the Unit 2 reactor coolant system power-operated relief valves during 1983. With the reacter coolant system " solid" and stable at 180*F and 375 psig, an attempt was made to equalize the levels of the two safety injection accumulators. Immediately after opening the isolation valves between the two accumulators and the cold leg injection header for equalization, an "overtemperature/ overpressure" alarm was received. The two valves were promptly closed. It was not realized that two other isolation valves between the cold leg injection line header and the primary system had been reopened for plant startup. These two open valves created a direct connection between the two accumulators and the primary system. Between the time that the valves were opened to equalize the accumulators and the time that they were closed after the alarm, the primary system pressure reached 410 psig and lifted at least one power-operated relief valve under the control of the low temperature overpressurization system.

There were no challenges to the Unit I reactor coolant system power-operated relief or safety valves during 1983.

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ntsconsin Electnc mencower 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 March 1, 1984 Mr. J. G. Keppler, Regional Administrator PRif1CIPAt_ STAFF / -

Office of Inspection and Enforcement, 7A P*P" V h'l/M f Region III 3/RA GE U U. S. NUCLEAR REGULATORY COMMISSION E/RA DHS" 799 Roosevelt Road 3C 99A_ ,

Glen Ellyn, Illinois 60137 3Cs IAo_

SGA, 4 5sI- File, h

Dear Mr. Keppler:

DOCKET NOS. 50-266 AND 50-301 ANNUAL RESULTS AND DATA REPORT POIN'I BEACH NUCLEAR PLANT, UUITS 1 AND 2 Enclosed herewith are two copies of the Annual Results and Data Report for the Point Beach Nuclear Plant, Units 1 and 2, for the year 1983. This report is submitted in accordance with Technical Specification 15.6.9.1.B and contains information regarding steam generator inservice inspections, personnel occupational exposures, and descriptions of facility changes, tests, and experiments as required pursuant to 10 CFR Section

50. 59 (b) .

Very truly yours,

,h day y Vice Preside'nt-Nuclear Power C.-W. Fay Enclosure Copies to NRC Resident Inspector Director, Office of Inspection and Enforcement (40 copies)

WAR 5y

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