ML20081G518

Annual Results & Data Rept 1982

ML20081G518
Person / Time
Site:	Point Beach
Issue date:	12/31/1982
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 8311070111
Download: ML20081G518 (56)

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4 +If.$. ANNUAL RESULTS AND 1 .g g ATA REPORT

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{,.;[. U.S. Nuclear Regulatory Commission W :'r Docket Nos. 50-266 and 50-301 Facility Operating License Nos.

8311070111 821231 N.

. g, -l PDR ADOCK 05000266 R PDR DPR-24 and DPR-27 i

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PREFACE This Annual Results and Data Report for 1982 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 Operating License Nos. DPR-24 and DPR-27, respectively.

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l TABLE OF CONTENTS Page

1.0 INTRODUCTION

1 2.0 HIGHLIGHTS 2.1 Unit 1 1 2.2 Unit 2 1 i .

, 3.0 FACILITY CHANGES, TESTS, AND EXPERIMENTS 3.1 Amendments to Facility Operating Licenses 2

) 3.2 Facility or Procedure Changes Requiring NRC Approval 3 l 3.3 Tests or Experiments Requiring NRC Approval 3

) 3.4 Design Changes 4 3.4.1 Common 4 3.4.2 Previously Omitted Modifications 9 3.5 Procedure Changes 15 4.0 NUMBER OF PERSONNEL AND MAN-REM BY WORK GROUP AND JOB FUNCTION 4.1 1982 18 5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION

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5.1 Unit 1 19 5.2 Unit 2 31 6.0 REACTOR COOLANT SYSTEM RELIEF VALVE CHALLENGES 52 e

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1.0 INTRODUCTION

The Point Beach Nuclear Plant, Units 1 and 2, utilize identical pressurized water reactors rated at 1518 MWt each. Each turbine generator is capable of producing 497 MWe net (524 MWe gross) of electrical power. The plant is located 10 miles north of Two Rivers, Wisconsin, on the west shore of Lake Michigan.

2.0 HIGHLIGHTS 2.1 Unit 1 Highlights for the period 01-01-82 through 12-31-82 included a 49-day refueling outage, a 16-day steam generator inspection outage, a brief outage to repair a ' gasket leak on the "B" loop cold leg RTD manifold orifice flange, and one trip resulting from a heat trace failure causing a steam pressure sensing line to freeze. Unit 1 operated at an average capacity factor of 62.3% and a net efficiency of 31.3%,

with a self-imposed hot leg temperature limitation in an attempt to limit steam generator , tube corrosion. The unit and reactor avail-ability were 81.8% and 82.6%, respectively. Unit 1 generated its 37 billionth kilowatt hour (gross) on 04-24-82; and its 38 billionth kilowatt hour on 08-09-82.

2.2 Unit 2 Highlights for the period 01-01-62 through 12-31-82 included a 40-day refueling outage, a 6-day outage to replace a degraded seal on "B" reactor coolant pump, a brief shutdown resulting from the replacement of an isolation valve on the reactor coolant system's RTD manifold, and a trip resulting from a loose terminal connection causing a spurious safety injection signal. Unit 2 operated at an average capacity factor of 83.1% and a net efficiency of 31.9%. The unit and reactor availability were 86.8% and 87.6%, respectively. Unit 2 generated its 35 billionth kilowatt hour '(gross) en 03-22-82; its 36 billionth kilowatt hour on 07-27-82; and its 37 billionth kilowatt hour on 10-26-82.

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i l 3.0 FACILITY CHANGES, TESTS, AND EXPERIMENTS i

3.1 Amendments to Facility Operating Licenses i

i-During 1982, there were 11 license amendments issued by the U.S.

i Nuclear Regulatory Commission to Facility Operating License DPR-24 1

for Point Beach Unit 1, and 12 license amendments issued to Facility

) Operating License DPR-27 for Point Beach Unit 2. These license amendments and changes are listed by date of issuance and are summar- '

ized below:

3.1.1 1-6-82, Amendment 58 to DPR-24, Amendment 62 to DPR-27

• These amendments revised the Technical Specifications and authorized adjustment of the voltage setpoint and associated time delay of the degraded grid undervoltage relay.

4 3.1.2 3-26-82, Amendment 63 to DPR-27 j This amendment authorized a one time waiver of the monthly functional tests of the turbine stop and governor valves until the start of the Unit 2 eighth - refueling outage.

l 3.1.3 5-25-82 Amendment, 59 to DPR-24, Amendments 64 to DPR-27

The surveillance requirements for hydraulic snubbers were upgraded by these amendments.

Minor administrative changes

, were also made to the Technical. Specifications to reflect j NUREG-0654 Rev. I requirements for checks and tests of

] survey instruments.

! 3.1.4 5-28-82, Amendment 60 to DPR-24, Amendment 65 to DPR-27 i

These amendments upgraded the operability requirements for i

containment fan coolers for Units I and 2.

l 3.1.5 6-25-82, Amendment 61 to DPR-24, Amendment 66 to DPR-27

These amendments revised the Technical Specifications to i

provide primary containment integrated leak test rate require-i ments and schedules consistent in part with the requirements of 10 CFR 50 Appendix J.

3.1.6 7-27-82, Amendment 62 to DPR-24, Amendment 67 to DPR-27 These amendments modified the limiting conditions for opera-tion (LCO) of the Technical Specifications through revisions to the minimum requirements for Auxiliary Feedwater System '

l operability prior to criticality and through revisions to the specified maximum intervals for reduced system availa-bility. .

3.1.7 8-31-82, Amendment 63 to DPR-24, Amendment 68 to DPR-27 These amendments revised the language of the Technical I

i Specifications related to the inservice inspection require- i ments of safety class components.

3.1.8 10-4-82, Amendment 64 to DPR-24, Amendment 69 to DPR-27 These amendments imposed additional restrictions to the operating requirements and surveillance testing requirements j , of the containment purge supply and exhaust isolation valves.

3.1.9 10-15-82, Amendment 65 to DPR-24, Amendment 70 to DPR-27

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{ These amendments allowed a relaxation of the requirements for monthly functional tests of the turbine stop and gover-nor valves during end of cycle operation wnen primary system boron concentration is low.

3.1.10 11-8-82, Amendment 66 to DPR-24, Amendment 71 to DPR-27 These amendments upgraded the existing Technical Specifica-tions in order to provide for redundancy of decay heat removal capability in all modes of operation for Point Beach Units 1 and 2.

3.1.11 11-12-82, Amendment 68 to DPR-24, Amendment 73 to DPR-27 These amendments approved minor administrative changes to radiation area entry control procedures.

3.1.12 11-15-82, Amendment 67 to DPR-24, Amendment 72 to DPR-27 These amendments modified the licenses to include a require-ment to maintain a Security Force Training and Qualification Plan.

3.2 Facility or Procedure Changes Requiring Nuclear Regulatory Commission Approval There were no plant modifications or procedure changes during 1982, 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 1982 which required Nuclear Regulatory Commission approval.

O 3.4 Design Changes 3.4.1 The following design changes were completed during 1982:

a. E-202 (Unit 1) & E-203 (Unit 2), Safeguards System. A local / remote transfer switch and a breaker control switch were installed to permit local control of safe-guards power supplies in the event a fire occurred ,

making main control board CO2 inaccessible.

Summary of Safety Evaluation: Local breaker control in addition to remote control is recognized as an improve-ment in plant safety,

b. E-238 (Common), Sewage Treatment. An improved level control system was installed for operation of P109A&B effluent sump pumps.

Summary of Safety Evaluation: Not nuclear safety related.

c. E-240 (Unit 1) & E-241 (Unit 2), Environmental Qualifi-cation. Envirorimentally qualified solenoid valves, limit switches, electrical conductor seal assemblies, and cable splices were installed as required on nuclear safety related valves to meet NRC IE Bulletin No.79-01B requirements.

Summary of Safety Evaluation: System functions and controls remain unchanged since only- part-for-part replacement is performed.

d. E-265 (Unit 1) & E-266 (Unit 2), Electrical Distribu-tion. Motor operators were installed on the generator breaker disconnect switches to provide for remote operation of the switches from the main control board.

Summary of Safety Evaluation: Not nuclear safety related.

e. IC-271 (Unit 1) & IC-272 (Unit 2), Containment Spray.

The logic for the containment spray pump suction valve -

interlocks was modified such that 870A and 871A are interlocked and 870B and 871B are interlocked. The modification removed the opening permissive from the opposite train 870 valve.

Summary of Safety Evaluation: The modification prevents a single failure (i.e., loss of one diesel) from pre-venting remote operation of both containment spray -

systems during the recirculation phase,

f. H-489 (Unit 2), Main Turbine. A harmonic shroud modifi- .

cation was performed on the L-4 blade group in LP1 per the manufacturer's recommendation.

Summary of Safety Evaluation: Not nuclear safety related.

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g. M-550 (Common), Fire Protection. Sprinklers and asso-ciated equipment were installed in Ready Stores in order to meet NRC and insurance carrier requirements.

Summary of Safety Evaluation: Not nuclear safety related.

, , h. M-586 (Unit 1), Demineralized Water. A demineralized water supply was installed in the service building maintenance shop in order to provide an adequate water supply for steam generator sludge lancing.

Summary of Safety Evaluation: Not nuclear safety related.

i. M-593 (Common), Hydrogen Supply. The hydrogen supply piping in Unit 1 turbine building was relocated to eliminate passage through the turbine lube oil reservoir area. Isolation and excess flow valves were also in-stalled. The modification fulfilled commitrents made to the NRC.

Summary of Safety Evaluation: Not nuclear safety related.

J. M-605 (Common), Fire Protection. Wet pipe sprinkler systems were installed in the vicinity of the safety injection pumps, component cooling water pumps, service water and diesel fire pumps, and in the diesel generator rooms in order to meet NRC fire protection requirements.

Summary of Safety Evaluation: Not nuclear safety related.

k. M-644 (Unit 2), Residual Heat Removal. A lifting beam was installed for moving the containment sump "B" screen to provide more convenient access to the 850A&B valves.

Summary of Safety Evaluation: Not nuclear safety related.

1. M-658 (Unit 2), Buildings & Facilities. The pressurizer compartment access opening was enlarged and platforms were installed to enhance personnel access to the com-partments and improve industrial safety.

Summary of Safety Evaluation: The larger access hole will neither deleteriously weaken the strength of the structure nor increase the risk to the containment liner due to missiles generated in the pressurizer space.

m. M-666 (Unit 2), Condensate. Additional pipe hangers and supports were added on condensate piping to reduce vibration experienced by thermal expansion of the a piping.

Summary of Safety Evaluation: Not nuclear safety related.

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n. M-668 (comon), Waste Disposal. A drain connection with a loop seal was installed in the laundry and hot shower tank (T14) vent to control venting of the tank and to prevent water from entering the ductwork.

Summary of Safety Evaluation: Not nuclear safety related.

o. M-676 (Unit 2), Auxiliary Coolant. Piping and an iso-lation valve were installed downstream of valve 2-958 to facilitate post-accident auxiliary coolant sampling.

Sumary of Safety Evaluation: The design of the piping -

system maintains the isolation and integrity of the system in accordance with acceptable criteria for the systems involved and poses no additional threat or analyzed condition degrading the system.

p. M-693 (Unit 1) & M-694 (Unit 2), Floor Drains. Seal leak-off connections on the containment spray pumps and safety injection pumps were piped to floor drains to ensure that seal leakage was directed to the existing floor drain system without overflowing general floor areas.

Summary of Safety Evaluation: Not nuclear safety related.

q. ,M-707 (Common), Buildings & Facilities. The hydrogen storage facility was relocated to the area of the nitro-i -gen storage tank to accommodate construction of a new maintenance shop in the general area.

Sumary of Safety Evaluation: Not nuclear safety related.

r. M-724 (Comon), Buildings & Facilities. The controlled side locker room was extended into the existing mainte-nance shop and laundry room facilities were expanded. A new survey instrument calibration and source room and an enlargement of equipment storage facilities were added.

Sununary of Safety Evaluation: Not nuclear safety related.

s. M-725 (Common), Buildings & Facilities. Security related.

t. M-741 (Comon), Waste Disposal. A hard-piped line from

, the distillate line upstream of BE-71 & 72 to the waste distillate tank overflow line was installed with an associated diaphragm isolation valve. This permitted

• more convenient recirculation of the blowdown evapor-ator.

j Sumary of Safety Evaluation: Not nuclear safety related.

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u. M-766 (Common), Buildings & Facilities. The contractor locker room and offices located on El. 26" of the ser-vice building were enlarged to accommodate greater numbers of contractor personnel engaged in site work activities.

l Summary of Safety Evaluation: Not nuclear safety related.

v. M-770 (Common), Buildings & Facilities. Door locations j in the controlled side locker room were relocated.

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1 Summary of Safety Evaluation: Not nuclear safety related.

l w. M-775 (Common), Buildings & Facilities. An auxiliary i controlled side entry and checkpoint and locker room was installed in the Unit I turbine building at El. 26'.

! Summary of Safety Evaluation: Nat nuclear safety related.

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x. M-811 (Common), Air Conditio.ing. A backup air condi-tioning unit was added to the security computer room.

Summary of Safety Evaluation: Not nuclear safety related.

y. M-813 (Unit 1) & M-814 (Unit 2), Reactor Makeup - Water.

I A valve and test connection were installed to facilitate i testing of the containment isolation valve in the

! reactor makeup water line without. isolating the line to the unaffected unit.

1 Summary of Safety Evaluation: Not nuclear safety related.

z. 82-01 (Common), Heating & Ventilation. Heating in the

} Quality Engineer's office area was upgraded by cross-

connecting Zone 4 supplies to that office and blanking off the Zone 1 supply duct.

Susunary of Safety Evaluation: Not nuclear safety related.

aa. 82-09 (Common), Waste Disposal. Shielding was installed around the front portion of the blowdown evaporator i bottoms filter to reduce personnel exposure.

t Summary of Safety Evaluation: Not nuclear safety related.

) bb. 82-14 (Comunon), Heating & Ventilation. The outside air temperature switch on the service building air condi-tioning unit chiller was permanently jumpered _ to provide-continuous chiller operation.

• Summary of Safety Evaluation: Not nuclear safety related.

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ec. 82-19 (Common), Heating & Ventilation. A new air flow switch was installed in the outlet of each control room fan (W14A&B charcoal filter fans) to increase relia-bility of the fans.

Summary of Safety Evaluation: Not nuclear safety related.

dd. 82-33 (Unit 2) & 87-34 (Unit 1), Electrical Distribu- ,

tion. The main generator breaker "&" auxiliary contact was disconnected from its existing input to the plant computer and rewired as part of 345 KV breaker relay failure scheme to open the main generator breaker if it has failed to open when so called upon. An auxiliary relay was also added via the "b" contact to provide control board indication and input to the computer.

Summary of Safety Evaluation: Not nuclear safety related.

ee. 82-35 (Common), Buildings & Facilities. A new doorway and stairs were installed on the roof of the controlled side locker room to provide access to service building offices on El. 46'.

l Summary of Safety Evaluation: Not nuclear safety related.

ff. 82-43 (Unit 1) & 82-44 (Unit 2), Heating & Ventilation.

The W51 electrical equipment room supply fan speed was increased and the return air damper locked open to in-crease cooling capacity and improve reliability of electrical equipment located in the equipment room.

Summary of Safety Evaluation: Not nuclear safety related.

gg. 82-60 (Common), Communications. ~ A Gai-tronics unit was installed in the heating, ventilating, and air condi-tioning equipment room.

, Summary of Safety Evaluation: Not nuclear safety related.

1 hh. 82-99 (Common), Fire Protection. An isolation valve between the fire water piping accumulator check valve and the header was installed to improve maintainability of the check valve.

Summary of Safety Evaluation: Not nuclear safety related.

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4 3.4.2 The following modifications were completed prior to 1982 but were omitted from previous annual reports. They are dis-cussed herein to complete and update the record.

a. E-100 (Unit 1) & E-101 (Unit 2), Heating & Ventilation (1978). Computer alarming was provided in the control room of the containment cavity cooling fan discharge.

, temperatures.

Summary of Safety Evaluation: Not nuclear safety related.

b. E-131 (Common), Radiation Monitoring (1981). The modifi-cation pr vided for transmitting R21 drumming area vent stack gaseous activity data to the plant process -

computer.

Summary of Safety Evaluation: Not nuclear safety related.

c. E-175 (Unit 2), Security Radio System (1978). Security related. i'

d. E-186 (Common), Lighting System (1980). Security related. \

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e. E-198 (Common), Heating & Ventilation (1979). Air con-ditioning was provided for the telephone equipment. room to accommodate heat loads generated by a new PBX system.-

Summary of Safety Evaluation: Not nuclear safety related.

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f. E-105 (Common), Switchyard Fault Recorder (1981). O Cable was installed from main control board CO2 to a new switchyard fault recorder. This improves the relia- \ s, s bility and quality of information available for fault

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analysis. 1 i^ i Summary of Safety Evaluation: 'Not nuclear safety 'related. '

g. IC-99 (Unit 1), Rod Control System (1981). Rod control -

system logic was rewired per Westinghouse; change recom- *'

'mendation ECR-015 to prevent , unwanted, interaction between rod control and process control W stems; and to avoid injection of line noise , into system logic. The line filter drain wire was relocated from3the zero volt bus to the cabinet chassis area. _3 v

Summary of Safety Evaluation: Not nucleai safety related.

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h. IC-113 (Common), Communications (1981). A tape re-cording system was installed in the control room to monitor conversations within the area as well as the Gai-tronics system channels. Activation of the system i

, is per procedure. l s

i Summary of Safety Evaluation: Not nuclear safety related.

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i. IC-139 (Unit 1), Rod Control System (1981). Feedback

) wires on the sampling resistors were disconnected in all spare circuits and then rewired to the neutral bus to preclude rods dropping as a result of sampling resistor -

deterioration or malfunction.

Summary of Safety Evaluation: Not nuclear safety related.

J. IC-157 (Unit 1) & IC-159 (Unit 2), Waste Disposal System (1981). The control circuit for letdown gas stripper vent cooler gas outlet isolation valve (GW-17A) was modified to bypass the MF-6 interlock and allow cryo-genic gas to backfill the stripper column. This

' alleviated an erroneous low level shut trip caused by a negative <, pressure" in the stripper during cooldown.

Summary of Safety Evaluation

Not nuclear safety related.

l k. IC-173 (Unit 21, Main Control Boards (1981). A light was installed on 2CO3 to indicate the 2-minute timer which trips the main feed pump on low feed suction pressure has been actuated.

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Summary of ' Safety Evaluation: Not nuclear safety related.

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1. IC-180 (Unit 2), Chemical & Volume Control (1980). A

,, vent path with 2-way isolation was installed . in the volume control tank nitrogen and hydrogen supply lines to allow venting of the regulated side of the pressure regulators.

l Summary of Safety Eveluation: Venting .is required to i adjust the regulator setpoint under certain conditions.

Double-valve isolation with a normally capped vent prevents inadvertent depressurization of either gas supply. Small amounts of gas' may be released in the primary auxiliary building, however the gaseous release

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' would be monitored via the vent stack radiation moni-toring system.

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m. IC-195 (Unit 1) & IC-196 (Unit 2), Main Control Boards (1981). Dual bistables for pressurizer low pressure alarms (PC-429/PC-449) were installed in 1&2C159 to replace the single bistable previously installed. The single bistable had been used to change power-operated relief valve interlock setpoints. This had required defeating of the low pressure alarms.

Summary of Safety Evaluation: Not nuclear safety related,

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n. IC-209 (Unit 2), Main Control Boards (1981). An alarm was added to 2C04 to indicate when a power-operated relief valve or Code safety valve on the pressurizer was not shut. The modification was required to meet TMI-short-term commitments.

Summary of Safety Evaluation: Not nuclear safety related.

o. IC-210 (Unit 1) & IC-211 (Unit 2), Auxiliary Feedwater (1981). Flow indication was installed in the auxiliary feedwater supply to each steam generator. The circuitry was powered off the white and yellow instrument bus.

The modifications were necessary to fulfill THI long-term requirements.

Summary of Safety Evaluation: Positive indication of auxiliary feed flow to the steam generators is provided.

p. IC-243 (Unit 2), Heating & Ventilation (1981). Local /

remote control switches for the containment ventilation isolation valves were disabled because if the switches would be in the: local position 3 the valves would be automatically reopened upon reset of containment venti-lation isolation.

Swanary of Safety Evaluation: Disabling of the switch prevents automatic opening following reset, thus re-quiring 2-valve isolation of the ventilation system under all situations. _The subject valves are red tagged shut unless; a unit is in cold shutdown .per Technical 1 Specification requirements.

q. M-369 (Unit 1), Chemical & ' Volume l Control (1980) . The 2" charging pump dampener _ isolation valve was removed

.and the pipe penetration capped because the bladder-type-pulsation dampener had been removed.

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

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r. M-405 (Common), Water Treatment (1979). Hot and cold

' demineralized water was provided to the controlled side laundry and sink to prevent calcium from entering waste liquid systems.

Summary of Safety Evaluation: Not nuclear safety related.

s. M-470 (Common), Heating & Ventilation (1981). Primary ,

auxiliary building ventilation supply for W35 was modi-fled to enable the fan to receive up to 50% recire air from within the building. This would minimize the po-tential for freezing of associated heat exchangers. The

• damper system operates with a manual pneumatic posi-tioner. The coil condensate drain line was improved to facilitate efficient water removal from the unit to minimize the potential for fan room flooding upon a tube rupture in the heat exchanger.

Summary of Safety Evaluation: Not nuclear safety related.

t. M-546 (Common), Chemical & Volume Control (1979). "C" charging pump varidrive was replaced with a DC drive system.

Summary of Safety Evaluation: The DC drive should be more reliable than the varidrive system. Only one drive was changed out pending long-term evaluation of perfor-mance.

u. M-511 (Common), Spent Fuel Pit (1981). New, high density storage racks were procured and installed in the spent fuel pit. A shipping cask anti-tip frame was also installed in the north half of the pit.

Summary of Safety Evaluation: The Staff accepts the thermal, hydrsulic, criticality, seismic and structural analysis completed by the contractor, checked via his QA program and additionally checked by NES and Reactor Engineering. Safety evaluation and Technical Specifi-cations were filed with the NRC.

v. M-587 (Common), Building & Facilities (1980). /. ladder and platform was installed to facilitate access to the radioactive material storage area located on El. 46' of the primary auxiliary building.

Summary of Safety Evaluation: Not nuclear safety related.

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w. M-598 (Common), Fire Protection. Piping was installed in the primary auxiliary building and was tied into existing turbine building headers to provide fire pro-tection capabilities at all levels within the building.

A total of 23 additional hose reels were also added.

The modifications fulfilled commitments made to the NRC on fire protection.

Summary of Safety Evaluation: Not nuclear safety related.

x. M-606 (Common), Service Air. Sampling and analysis of.

service air was performed to verify its suitability for emergency breathing purposes. It further modified breathing piping in the control room to provide breathing connections on the outside of the control boards. The modification was required to fulfill commitments made to the NRC on fire protection.

Summary of Safety Evaluation: Not nuclear safety related,

y. M-615 (Unit 1), Waste Disposal (1980). The reactor coolant drain tank leakoff was extended to provide sparging and obtain a more even temperature distribution in the tank during pumpdown and to prevent pressure excursions during substantial leakoff flows. A vent hole was installed in the top of the tank to perform an anti-syphon function.

Summary of Safety Evaluation: Not nuclear safet.y related.

z. M-627 (Common), Buildings & Facilities. Security related.

aa. M-625 (Unit 1) & M-646 (Unit 2), Primary Sampling (1981). Sample piping was added to facilitate remote sampling of high activity reactor coolant using a lead shielded sample cylinder.

Summary of Safety Evaluation: A viable means of post-accident primary coolant sampling is now provided.

bb. M-691 (Corraon), Drains. A 2" opening in the bottom of the non-vital switchgear room north doorway was in-stalled to provide a drain path for water if a fire in the room necessitated use of fire equipment.

Summary of Safety Evaluation: Not nuclear safety related.

' cc. M-739 (Common), Buildings & Facilities. The existing turbine building office area was modified to accommodate EQRS personnel by installing a new doorway, partitions, a removing an existing office window and constructing a new office.

Summary of Safety Evaluation: Not nuclear safety related.

1 dd. M-740 (Unit 2), Reactor Core (1981) . Four demonstration optimized fuel assemblies were loaded during Refueling 7 for operation during Cycles 8, 9 and 10 to test and demonstrate improved performance and reliability of the fuel design prior to insertion of a full reload of such fuel.

Summary of Safety Evaluation: Insertion of the assem- ,

blies did not pose an unreviewed safety question and was evaluated to not adversely affect the safety of the plant. No Technical Specification changes were re-quired. The optimized fuel assemblies were evaluated during the overall Manager's Supervisory Staff safety evaluation of the reload core. After the first ;ycle of operation, it was concluded that burnup was a predicted by the vendor and visual inspection disclosed no anom-alies to date.

ee. M-824 (Common), Buildings & Facilities. The existing Operations office area located on turbine building El. 44' was modified to provide increased office space by rearranging walls and adding a doorway.

i Summary of Safety Evaluation: Not nuclear safety related.

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o 3.5 Procedure Changes The following emergency operating procedures were revised during 1982. A 50.59 review was conducted for each change and it was con-cluded that none of the changes increased the probability of an accident as analyzed in the FSAR:

3.5.1 EOP-1A, Loss of Reactor Coolant, Revision 20, 08/31/82. The procedure was editorially upgraded to improve clarity.

Reference to Emergency Plan Implementing Procedures was added as was the requirement to notify the Duty Technical Advisor. A limitation on cooldown rates was added for natural circulation to preclude drawing a bubble in the reactor vessel head. In addition, the WT limitation on natural circulation was increased from 50 to 60*F which is closer to 100% WT and permits a larger thermal driving head to develop. A step was added to ensure fast bus- transfer and energization of safeguards buses. A step was added to remind operators that the primary auxiliary building and

letdown - gas stripper are possible radiation alarm sources j outside containment.

3.5.2 E0P-2A, Loss of Secondary Coolant, Revision 13, 08/31/82.

, The procedure was editorially upgraded to improve clarity. A step was added to remind operators that the primary auxil-

! iary building and letdown gas stripper are possible radia-i tion alarm sources outside containment. A step was added to ensure fast bus transfer and energization of safeguards buses. Emergency Plan Implementing Procedure reference was added. Westinghouse Owners Group recommendations concerning securing of safety injection were incorporated. In addition, the cooldown limitation curve was added.

3.5.3 EOP-3A, Steam Generator Tube Rupture, Revision 14, 08/31/82.

l The procedure was editorially upgraded for clarity. The primary - auxiliary building and letdown gas stripper were added as possible radiation -alarm sources outside contain--

ment. The pressurizer safety valves were added to the pres-surizer integrity checklist. A note was added to ensure cooldown below no-load temperature does not occur until the j faulted steam generator is identified and isolated. Addi-

tional equipment monitoring steps were incorporated. A ,

section cas added to establish conditions required for l

operation of the reactor coolant pumps and depressurization of the primary system while maintaining appropriate-subcool-ing.

a 3.5.4 E0P-4A, Reactor Coolant Leak, Revision 6, 09/13/82. .The l procedure was editorially upgraded for clarity. Equipment l

• tag numbers were corrected. A note . was added to instruct the operator to cool down the primary system as required in order to matt.tain appropriate subcooling limitations.

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, 3.5.5 E0P-5A, Emergency Shutdown, Revision 10, 09/10/82. The

). procedure was editorially upgraded for clarity. Main feed regulating valve automatic operation was clarified. Set-points were corrected as necessary to reflect current settings.

3.5.6 E0P-5B, Inadvertent Safety Injection, Revision 4, 09/10/82. ,

4 The procedure was editorially upgraded for clarity. A step was added to ensure fast bus transfer and energization of safeguards buses. Recirculation from the boric acid storage tanks to the refueling water storage tank was clarified.

3.5.7 E0P-6A, Dropped Rod, Revision 6, 09/10/82. The procedure was editorially upgraded for clarity. Technical Specifi-

cation requirements were added for reduction in power level and maintaininF same.

3.5.8 EOP-6B, Stuck Rod or Malfunctioning Position Indication, Revision 7, 09/10/82. Minor editorial corrections were made, i

3.5.9 E0P-6C, Uncontrolled Withdrawal of RCCA(s), Revision 5, 09/10/82. A step was added to denote charging pump speed control limit high/ low alarm as a symptom of the accident if the reactor is t ot at fall power. Minor editorial correc-l tions were also 1.corporated.

$ 3.5.10 E0P-6D, Uncontrolled Insertion of RCCA(s), Revision 5, 09/10/82. A step was added to denote charging pump speed i

control limit high/ low alarm as a symptom of the accident if the reactor is not at full power. Minor editorial correc-tions were also incorporated.

i .

3.5.11 EOP-7A, Loss of Outside AC Power, Revision 9, 09/10/82.

Equipment tag numbers were corrected. A step was added to limit thermal cooldown to less than 25'F/ hour for prevention of a bubble forming in the reactor vessel head. A note was-added including provisions for emergency bearing cooling via the fire protection system should both safeguards diesels be unavailable coincident with the loss of offsite power.

3.5.12 E0P-7B, Station / Unit Blackout, Revision 1, 09/10/82. .The procedure was upgraded editorially for clarity and was changed from an interim to a permanent procedure. Radio locations were incorporated to assist operating personnel.

References _to local isolation panels were included and backup power supplies noted. A step was added to initiate -

the Emergency Plan as the first subsequent action item.

Steps were also added to isolate the nonenergized safeguards and nonsafeguards buses. A step was _ also added to ensure .

safety injection recire _ line isolation valves are open if

, the safety injection pumps ' are to be operated at shutoff l head. A step .was' clarified to break condenser vacuum to stop 'the' turbine when safeguards buses .are not available.

i 3.5.13 E0P-8D, Spent Fuel Shipping Cask Drop, Revision 2, 09/10/82.

The procedure was editorially upgraded to reference appli-cable operating instructions.

3.5.14 E0P-9A, Service Water System Failure, Revision 3, 09/10/82.

The procedure was rewritten and editorially upgraded to

, achieve consistency in format with other E0P's and to clarify steps for operator assistance.

~

3.5.15 E0P-9B, Loss of Component Cooling, Revision 3, 09/10/82. The procedure was rewritten and editorially upgraded to achieve consistency in format with other E0P's and to clarify steps for operator assistance.

3.5.16 E0P-10A, Control Room Inaccessibility, Revision 8, 09/10/82.

The procedure was rewritten to achieve consistency in format with other E0P's and to provide improved clarity. Steps were added to permit operation of the diesels at the local control station. In addition, the procedure was expanded to include local control of the output breakers to the A05/A06 buses, 4.16 KV and 480 V breakers associated X13/X14 transformers and supply breakers for 480 V safeguards MCC's.

Figures delineating saturation temperature as a function of steam generator wide range level and pressure and for pressurizer level versus cold calibrated level were added for operator information and use.

3.5.17 E0P-12A, Oil & Nonradioactive Hazardous Material Spill, Revision 5, 09/10/82. The procedure was editorially upgraded for clarity.

3.5.18 E0P-13A, Intake Icing / Blockage, Revision 1, 09/10/82. The procedure was rewritten to improve clarity and updated to provide better operator assistance based upon operating experience. It was also reformatted to be consistent with other E0P's.

3.5.19 E0P-13B, High Lake Water Level, Revision 1, 09/10/82. The procedure was renumbered from E0P9C. It was editorially upgraded for clarity and to include installation of a structural modification.

3.5.20 E0P-13C, Emergency Weather Conditions, Revision 3, 09/10/82.

The procedure was rewritten and reformatted to be consictent with other E0P's. It was also upgraded to detail regional naturally occurring phenomenon such as severe cold weather conditions and tornados. The procedure was renumbered from E0P-9D.

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o 5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION The following is a synopsis of the findings resulting from the steam generator tube inspections conducted on both units during 1982. Also included are the results from the steam generator secondary side (annular region) visual inspections.

5.1 Unit 1 03-25-82 Outage On 03-25-82, Unit I was shut down for a scheduled steam generator eddy current inspection. The 2000 psid primary-to-secondary hydro-i static test condition was established during cooldown of the unit. An 800 psid secondary-to-primary leakage check was performed on both steam generators on 03-29-82. The 800 psid seconda ry-to-prima ry 4 leakage check was perfo rmed visually with the aid of remote video I equipment. The visual inspection was initially performed at 0100 j hours on 03-29-82. Due to steam generator humidity conditions, an j additional verification inspection was performed at 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on the

! same day. The secondary side was held at pressure throughout this

inte rval . The specific conditions identified during the leakage checks are noted below.

1 l "A" Steam Generator f Hot Leg 1

t Original Inspection Verification R24C37 Explosive plug .21 drops / min. 16 drops / min.

R21C49 Explosive plug 5 drops / min. Boric acid coated R31C31 Explosive plug 2 drops / min. I drop / min.

R19C33 Explosive. plug Dry F1 drop /2 min.

R03C09 Explosive plug Boric acid coated Boric acid coated i

R01C22 Explosive plug Boric acid coated Boric acid coated R12C25 Tube Wet Wet /1 drop /10 min.

Cold Leg J

i R25C27 Tube 20 drops / min. 20-30 drops / min.

] (Explosive plug removed 10/81)

"B" Steam Generator Hot Leg R23C38 Explosive plug 20 drops / min. 20 drops / min.

• R24C37 Explosive plug Boric acid coated Boric acid coated.

R29C35 Explosive plug 5 drops / min. 5 drops / min.

R29C40 Explosive plug 3 drops / min. 3 drops / min.

. R13C61 Explosive plug Boric acid coated Boric acid coated R29C37 Explosive plug Boric acid coated Boric acid coated R23C63- Tube 2-3 drops / min. D ry -

Cold Leg R24C50 Explosive plug Wet Not verified The eddy current inspection program, performed this outage, consisted of the following:

1. Inspection of all readily remotely accessible tubes to the first '

support plate in the hot legs of both steam generators.

2. Inspection, over the U-bend from the leaking side, of the three tubes which were identified as leakers or potential leakers in the visual tubesheet check.

3. Inspection, to the first tube support from the cold leg, of the tubes in the cold leg of "A" steam generator which had either an explosive or oechanical plug removed during the sleeving demon-stration in 10/81, and were not subsequently sleeved.

Nineteen tubes in the "A" steam generator and 9 tubes in the "B" steam generator were verified to have degradation greater than 40%.

Of the 2,848 open tubes in the "A" steam generator, 2,835 were in-spected and 2,833 of the 2,850 open tubes in the "B" steam generator were inspected. The tubes that were not inspected are as follows:

Number of Tubes Not Inspected Reason for Not Inspecting "A" "B" 12 14 Contained template plugs 1 2 Restricted tube ends 0 1 Located near environmental ledge (not accessible with remote equipment).

13 17 These tubes were not inspected because of the radiation exposure associated with moving template plugs, manual eddy current probing, and reworking restricted tube ends. The noninspected tubes con-stitute less than 1% of the unplugged tubes, most are not located in the zones where large numbers of defects have occurred, and the overall eddy current 'results did not indicate the necessity to inspect the tubes. Following is a summary of the eddy current indi-cations.

Results of Eddy Current Inspection Indications "A" Steam Generatcr "B" Steam Generator Identified Hot . tg Cold Leg Hot Leg F20% 3 0 1

. 20-29% 0 1 0 30-39% 1 0 2 40-49% 0 0 0 50-59% 2 0 0 60-69% 1 0 1 70-79% 4 0 3 80-89% 3 0 0 90-99% 9 0 5 100% 0 1* O l UDI 18 0 3

• Tube R25C27 was noted leaking during visual inspection. No defect was identified by eddy current ispection.

"A" Steam Generator Hot Leg Side to First Support Plate J

Tube Defect Location Origin Plugged R05C07 97% 15" ATE OD Yes R11C07 UDI 11" ATE OD Yes R18C07 UDI 6" ATE OD Yes R01C13 90/95% 14" & 7 " ATE OD Yes R13C33 UDI 15" ATE OD Yes R18C35 92% 10-15" ATE OD Yes R31C37 UDI 11" ATE OD Yes R18C37 UDI 8" ATE OD Yes R23C38 UDI 20" ATE OD Yes R27C38 UDI 20" ATE OD Yes R10C40 UDI Roll to TTS OD Yes R12C41 79% 12" ATE OD Yes R11C43 87% 18-20" ATE OD Yes R18C44 UDI 20" ATE OD Yes R15C20 UDI 8-14" ATE OD Yes R29C45 79% 7 " ATE OD Yes R29C47 UDI 8-14" ATE OD Yes R09C55 96% 16-19" ATE OD -Yes R12C58 88% 18-20" ATE OD Yes R13C59 UDI 20" ATE GD Yes-R15C59 58% 7-20" ATE OD Yes l R15C60 UDI 14-18" ATE OD Yes I R11C62 52% 21" ATE OD Yes R12C62 93% 18-21" ATE OD ~Yes

, R07C64 89% 12" ATE OD Yes R15C65 UDI 6-19" ATE OD Yes R05C68 F20% \" ATS OD No R15C68 93% 10-15" ATE OD Yes R05C69 -F20% \" ATS OD No R15C71 UDI 8". ATE. OD Yes

Tube Defect Indication Origin Plugged R05C72 78% 7h" ATE OD Yes R08C73 96% 17" ATE OD Yes R09C73 94% 17" ATE OD Yes R08C74 92% 17" ATE OD Yes R06C81 F20% 1" ATS OD No R14C22 UDI 14-17" ATE OD Yes ~

R15C22 UDI 14" ATE OD Yes R18C23 69% 19" ATE OD Yes R15C25 73% 16" ATE OD Yes R15C27 UDI 10-15" ATE OD Yes R36C29 36% TTS OD No Cold Lej Tube Defect Location Origin Plugged R26C53 25% 2" ATS OD No "B" Stea's Generator Hot Leg to First Support Plate Tube Defect Location Origin Plugged R02C15 UDI 6-11" ATE OD Yes R02C17 63% 8" ATE OD Yes R24C25 72% 11" ATE OD Yes R27C30 F20% 1" ATS OD No R21C34 90% 5" ATE OD Yes R08C35 93% 10" ATE OD Yes R14C40 30%, UDI 1" & 20" ATE OD Yes R10C42 73% 21" ATE OD Yes R26C42 30%, UDI 21" & 19-21" ATE OD Yes R13C47 UDI 21" ATE OD Yes R26C50 91% 12" ATE OD Yes R01C47 95% 21" ATE OD .Yes R22C59 72% 17" ATE OD Yes R21C66 UDI 21" ATE OD Yes R03C47 91% 20" ATE OD Yes ATE - Above tube end UDI - Undefinable indication ATS - Above tubesheet TTS - Top of tubesheet I Nineteen tubes in the "A" steam generator and 9 tubes in the "B"

=-

steam generatcr contained indications exceeding . the 40% plugging limit. These tubes and the two leaking tubes in the "A" steam generator were mechanically plugged. Of the 28 indictions exceeding the plugging limit, 11 are new indications in the "A" steam generator -

and six are new indications in the "B" steam generator. The other indications identified were either previously noted as undefinable indications or defects that previously existed, but were not identi-fled in prior inspections. As in the past, all indications were I

l small volume. As a conservative measure, all of the tubes containing undefinable indications were plugged to further ensure the relia-

, bility of the unit. Correct plugging was visually verified via the use of tubesheet photography.

l The explosive plugs verified to be leaking in excess of two drops per i

minute ("A" steam generator hot leg, R24C37; "B" steam generator hot

,, leg, R23C38, R29C35, and R29C40) have been repaired with a welded l plug. Based on the history of the plugs, personnel radiation l exposure encountered during weld repair and future steam generator replacement, the plugs leaking at a very low rate (1 drop per i minute) were not weld repaired this outage.

. Eddy current examinations of the tubes noted to be wet or leaking during the visual leakage check revealed no indications. The tubes verified to be leaking were plugged. Tube R12C25 in the "A" steam generator was plugged with mechanical plugs in both hot and cold legs. Tube R25C27 in the "A" steam generator was plugged with a welded plug in the hot leg since it was sleeved during the 10/81 outage, and mechanically plugged in the cold leg. The type, or

. location, of the defect (s) existing in these tubes is unknown. They

! were both inspected through the U-bend from the leaking side.

i Eddy current examinations of the cold leg ends of tubes which had either mechanical or explosive plugs removed during the sleeving demonstration of 10/81 and which were not subsequently sleeved

revealed one indication in Tube R26C53 of 25% at 2" above the tube-

! sheet from the cold leg side. The indication was previously reported

, during the 1978 refueling outage steam generator inspection. No I other indications were identified.

An eddy current exam of the 12 tubes sleeved during the 10/81 i refueling outage was also performed this outage. The exam consisted-of using the same probe type and eddy current parameters used in.

10/81 and comparing the signals to the 10/81 signals. There were no noticeable changes in the eddy current signals.

To minimize the rate of corrosion, the Unit 1 primary system was returned to power a t. a . reduced hot leg temperature of 557'F. In addition, a crevice flush was performed before the unit was returned 4

to service to remove impurities from the tubesheet crevice.

I j Unit 1 Refueling 10 Inservice Inspection On 10-22-82, Unit I was shutdown for'its tenth refueling outage. The

2000 psid primary-to-secondary hydrostatic test condition was estab-4 lished during cooldown of the unit. An 800 psid secondary-to primary

. 1eakage ' check. was performed on both steam- generators ~ on ' 10-26-82.

I The 800.. paid secondary-to primary leakage . check' was performed i visually with the aid of remote video equipment. -The specific condi-tions identified during the l leakage . checks are noted below. ;(All.

noted leaks were observed from previously plugged tubes.)

L

"A" Steam Generator Hot Leg R03C09 Boric acid coated R12C25 Boric acid coated R14C57 2.0 drops per minute R31C31 1.5 drops per minute "B" Steam Generator Hot Leg R13C63 2.0 drops per minute R28C43 Wet end R29C34 Boric acid coated R29C37 Boric acid coated R31C44 Wet end The eddy current inspection program, performed this outage, consisted of the following:

1. Inspection of essentially all readily remotely accessible tubes to the first support plate in the hot legs of both steam generators.

2. Inspection over the U-bend from the hot leg side of greater than 3% of the tubes in each steam generator.

3. Inspection up to the sixth support of the hot leg tubes con-taining sleeves.

4. A special inspection of all the sleeves in both the hot leg and cold leg.

5. Inspection of tube locations previously identified as containing degradation.

Of the 2,809 open tubes in the "A" steam generator, 2,769 were in-spected and 2,787 of the 2,837 open tubes in the "B" steam generator were inspected. The tubes that were not inspected are as follows:

Number of Tubes Not Inspected Reason for Not Inspecting "A" "B" 19 16 Contained template plugs 1 --

Restricted tube ends-

• 20 34 Under fixture " foot" 40 40 -

l l

l These tubes were not inspected because of the radiation exposure j associated with moving template plugs, moving eddy current equipment, l and the reworking of a restricted tube end. The noninspected tubes l constitute less than 2% of the unplugged tubes, most are not located in the zones where large numbers of defects have occurred, and the overall eddy current results ~did not indicate the necessity to l inspect the tubes. Following is a summary of the eddy current indi-cations.

Results of Eddy Current Inspection Indications "A" Steam Generator "B" Steam Generator Identified Hot Leg Cold Leg Hot Leg F20% 3 1 1 20-29% 0 0 0 30-39% 1 2 0 40-49% 0 0 0 50-59% 0 0 1 60-69% 0 0 1 70-79% 0 0 0 80-89% 3 0 1 90-99% 1 0 0 100% 0 0 0 UDI 0 0 0 "A" Steam Generator Hot Leg Tube Defect Location Origin Plugged R13C48 89% 21" ATE OD Yes R21C48 91% 20" ATE OD Yes R19C56 83% 21" ATE OD Yes R27C58 80% 9 & 11" ATE OD Yes R36C29 34% TTS OD No R05C68 F20% \" ATS OD No R05C69 F20% \" ATS OD , No R06C81 F20% 1" ATS OD No Cold Leg Tube Defect Location Origin Plugged R26C53 34% 2" ATS OD No R20C60 F20% 5" ATS OD No R28C48 36% 2" ATS OD No e

i

)

"B" Steam Generator Hot Leg Tube Defect Location Origin Plugged R27C30 F20% 1" ATS OD No R21C59 56% 8" ATE OD Yes R20C61 80% 20" ATE OD Yes ,

R18C68 69% 20" ATE OD Yes ATE - Above tube end UDI - Undefinable indication

• ATS - Above tubesheet TTS - Top of tubesheet Four tubes in the "A" steam generator and 3 in the "B" steam generator contained indications exceeding the 40% plugging limit. Of the 7 indications exceeding the plugging limit, one is a new indi-cation in the "A" steam generator and 2 are new indications in the "B" steam generator. The other indications identified were either previously noted as undefinable indications or defects that pre-viously existed, but were not identified in prior inspections. As in the past, all indications were small volume and originated on the tube's outside diameter.

The 7 tubes containing indications greater than the plugging limit have been mechanically plugged. Correct plugging was independently verified by visual means.

In addition, the eddy current inspection program identified a total, for both steam generators, of 71 tubes that are restricted to a 0.720" diameter probe at the first support plate on the hot leg side.

Of the 71 restrictions, 20 are in the "A" steam generator and 51 are in the "B" steam generator. The majority of these restrictions are located in the periphery tubes near the " wedge" areas.

All of the restrictions, except for the two in the "B" steam generator, passed 0.650" diameter probe. The two restrictions noted above passed a 0.610" diameter probe. Thirteen of the 20 restric-tions in the "A" steam generator were present during . previous inspections. Forty-two of the 51 restrictions in the "B" steam generator were present during previous inspections. In addition to a slight increase in the total number of restrictions, a slight increase in the extent of the restrictions in some of the- tubes previously noted as containing restrictions was experienced.

All tubes restricted to a 0.700" probe at the first support were probed through the sixth support with a 0.650" or 0.610" probe. Only

• minor denting was noted at the higher supports.

The explosive plugs that were identified to be leaking at a very low -

rate (approximately 2 drops per minute) were not weld repaired this outage. This decision was based on personnel radiation exposure associated with performing weld repairs, the low primary-to-secondary leak rate (less than 10 gallons per day) prior to the outage, 'and the planned steam generator. replacement scheduled in 1983.

An eddy current exam of 11 (Note: One sleeved tube was removed from service in 03/82) tubes sleeved during the 10/81 refueling outage was also performed this outage. An eddy current signal was identified in the hot leg sleeve in tube R28C58. This signal was believed to be an indication of a deposit.on the ID of the sleeve wall. Evaluations of this signal included reprobing of this sleeved tube before and after brushing and honing of the sleeve wall.

~ '

A decision was made to remove the sleeve from this tube for further examination of this indication. The bottom 19-7/8" section of the sleeve was removed for laboratory nondestructive and destructive 0 examination. The hot leg and cold leg ends of tubes R28C58 were subsequently mechanically plugged.

Laboratory eddy current inspection confirmed the indication at 10.5" and studies with a pancake probe showed the indication was on the ID and more pronounced in 0-90-180* circumferential location. A detailed examination of the OD at up to 60X at this location and OD diametral measurements identified no cause for ' the indication and showed a constant OD value of 0.740".

Double wall x-ray radiographs were done at O', 45*, 90', and 135'.

Only a shallow circumferential mark at 10.5" was detected at 45',

90', and 135*. The mark was not found on 0*.

To facilitate further examination, a tube section extending from 9 to 12" was removed and split longitudinally along O' and 180'. At 10.5", a shallow circumferential mark was observed and found to be more pronounced at 0*-90 -180'. Wall thickness measurements indi-cated no localized wall reduction at this location. In addition, the section was searched for localized ferromagnetism because of a possible contribution of a ferromagnetic material to the eddy current signal. No ferromagnetic phase was detected with the instrument used.

The sample was further reduced in size for studies with the scanning electron microscope with the energy dispersive x-ray spectrometer.

The 9-12" tube section from 0-180' was cut transversely one half-inch below and above the 10.5" indiciation and longitudinally at 90'. A scanning electron microscope montage through the indication location shows scratches but otherwise appears featureless. The microstruc-tures at the indication and away from the indication are similar and-were as expected from the in plant wire brushing and honing.

~

The Inconel 600 matrix composition along with Si and Al were found. The latter may have come from the honing.

The above section from 10-11" and 0-90' was cut longitudinally at

• 45'. On the 0-45* section, energy dispersive x-ray spectrometer area analyses were made at 10.4-10.55" (eddy current indication) and at 10.6-10.9" .(away from eddy current indication). The Inconel 600

• matrix composition and Al plus Si were found in both areas.

_ _ _ . .. . _ . ._ _ _ = . .-

Elemental mapping for Ni and Fe was done in the above areas as part of an evaluation of the variation in the concentration of elements that might contribute to magnetic permeability. The concentrations

, of each element in the two areas are similar.

Metallography was performed on the 45* surface of the 45-90* section at 10.5". The wall is uniform in thickness and the structure appears normal.

Additional analyses and examinations are planned as required.

The roll transition and brazed areas of the tube sleeves were also

• inspected using the same eddy current parameters as used during the baseline inspection of 10/81. The data resulting from this in-spection were compared to the baseline data, and no noticeable changes in the eddy current signals were identified. In addition to the sleeve inspections, the hot leg tubes containing sleeves were in-spected through the sixth support from the hot leg side. This inspection was performed with normal eddy current parameters and a 0.650" diameter probe. No indications were identified.

After the normal refueling eddy current testing was completed, sludge lancing was performed as part of normal refueling activities and in preparation for an annular search of the steam generator secondary The annular search is done through the handholes using fiber-side.

optics to scan the region above the tubesheet and between the tubes and the steam generator shell.

On 11/05/82, the annular search of the "A" steam generator identified the following objects, all on the cold leg side:

1

) One 6" "C" clamp with swivel pad missing One "C" clamp swivel pad One 3" stainless steel hose clamp Pieces of lockwire along with residual scale and sludge The "C" clamp was leaning against two tubes which showed definite -

signs of mechanical damage. A third tube in' the area also appeared to have damage although not as severe as the damage to the first two tubes.

The annular search of the "B" steam generator was performed on 11/06/82 and the following items were found on the hot leg side:

One rod bar 1/4" x 3/8" x 58" long _

One piece of metal 1-1/4" x 2-1/4" x 6-1/2" (a second piece identical to the one above was later found as the objects were being removed) .

Pieces of weld rod Pieces of lockwire along with residual scale'and sludge On the cold leg side of the "B" steam generator, the only item found was a piece of slag, which was orignally described as 1/2" in diameter and 1" long.

1 i

l Because of the apparent damage to 3 tubes on the cold leg side of the "A" steam generator, it was decided to perform an eddy current exami-nation of the cold leg peripheral tubes in both steam generators.

All of the hot leg peripheral tubes had been inspected previously through the first tube support plate during the normal eddy current' inspection. To verify that there was no mechanical damage to tubes on the hot leg sides, eddy current tapes of the hot leg peripheral tubes were reexamined with no defects found.

The eddy current inspection was conducted from the cold leg side of both the "A" and "B" steam generators. All outermost peripheral O

tubes were inspected over the U-bend through the top support plate on the hot leg side. Also, all tubes within 2 tube depths of the peri-phery were inspected through the first tube support plate. The results from this inspection showed the existance of mechanical  ;

damage to several tubes. In addition, there were several tubes with signs of degradation which was not related to mechanical damage from foreign objects. Because of these indications, the eddy current inspection program in each steam generator was expanded to include a minimum of 600 additional tubes. This inspection was not required by Technical Specifications, however, the expansion satisfied the Tech-nical Specification requirements for expansion of sample size based upon the number of defect indications found in the initial sample. A total of about 1000 tubes per steam generator were inspected.

Following are the results of the inspection and action taken.

"A" Steam Generator Cold Leg Tube Indication Location Action Comments R42C29 63% 5" ATS Plugged Near "C" clamp R42C31 Bulge \" ATS Plugged Near "C" clamp R43C33 Bulge " ATS Plugged Near "C" clamp R43C34 Bulge \" ATS Plugged Near "C" clamp R43C35 Bulge " ATS Plugged Near "C" clamp R43C36 Bulge \" ATS Plugged Near "C" clamp R35C76 48% 5" ATS Plugged R34C76 Bulge " ATS Plugged R33C77 Bulge " ATS Plugged R34C77 Bulge TTS Plugged R29C82 Dent 2" ATS Plugged R25C85 Dent 1" & 5" ATS Plugged R07C01 41% First support Plugged R44C38 32% TTS ---

Near "C" clamp R45C39 24% \" ATS ---

Near "C" clamp' R31C78 22% '14" ATS ---

ATS - Above tubesheet ATE - Above tube end

. TTS - Top of tubesheet

"B" Steam Generator Cold Leg Tube Indication Location Action Comment R01C02 51% TTS Plugged R01C05 61% First support Plugged z R23C09 51% First support Plugged R36C21 41% First support Plugged R01C92 59% 21" ATE Plugged

• ATS - Above tubesheet ATE - Above tube end

• TTS - Top of tubesheet Mechanical plugging of the above noted tubes was accomplished on i 11/13/82 in the "B" steam generator and on 11/16/82 in the "A" steam i generator.

i Two tubes (R43C32 and R44C35) in the "A" steam generator cold leg had observable damage from the "C" clamp and bad been previously plugged.

R43C32 was plugged in 02/78 after Unit 1 was shut down with a 130 gallon per-day primary-to-secondary tube leak (refer to Licensee Event Report No. 78-001/01T). R44C35 was plugged in 05/78 after Unit I was shut down with a 145 gallon per-day primary-to-secondary leak (refer to Licensee Event Report No. 78-010/01T).

Using fiberoptic photographs to determine defect size, a flow-induced vibration analysis of the two damaged tubes was performed. The defect in the analysis was assumed to be located 5" above the tubesheet, and was assumed to be a notch 1" long axially and 62* wide circumferen-tially. The tube was assumed to be dented and fixed at the first support (conservative for a worst-case analysis).

For anticipated power operating conditions, the maximum resulting fluid-elastic ' stability ratio was 0.2 compared to the theoretical instability threshold of 1.0. The peak turbulent amplitude was calculated to be less than 1.0 mil and assuming a stress concen-i  ; ration factor of 4.0 the maximum calculated alternating stress in the damaged region would be about 1.0 KSI, which is significantly below the high-cycle fatique endurance limit of about 14.0 KSI.

Thus, from the viewpoint of both the fluid-elastic and turbulence excitation, the damaged tubes are judged to be structurally stable.

I Under the influence of normal operating fluid and thermal-mechanical loadings, the damaged tubes are acceptable from .the viewpoint of fatique.

l All of the objects were removed from the "B" steam generator on 11/16/82, except for several pieces of light-weight lockwire which 6 i extended between tube columns and appeared to be fixed in the sludge.

i The 1/4" x '3/8" x 58" rod which was removed was carbon steel and showed no signs of wearing against the. tubes. The rod is estimated .

to have been in the steam generator for at least 5 years and could possibly have been in the steam generator since before the unit was -

placedLin service in 1970. The two metal blocks (carbon steel) were.

identified as items used-in the wrapper support structure (Item 14 on l . _ . _ _

I f

Westinghouse assembly drawing No. 679J446). The blocks did not show any signs of ever being installed, and a remote visual inspection verified that there were no support blocks missing. Thus, the blocks j were extraneous and have been in the steam generator annulus since fabrication.

All of the items were removed from the "A" steam generator on

~

11/24/82. The "C" clamp handle had to be cut before the clamp could be maneuvered out of the steam generator. It is believed that the

. "C" clamp could have fallen into the steam generator in 10/77 when the downcomer flow resistor plate modification was performed. The 7 origin of the hose clamp is unknown. During the retrieval efforts, a pin 1/4" in diameter by 1-1/2" long was lost from one of the re-trieval tools. Attempts to locate the pin were unsuccessful.

Because of the small size of the pin, the potential for tube damage is considered insignificant.

t A tabulation of the personnel exposure for the retrieval efforts resulted in a total whole body exposure of 32.5 man-Rem. Extremity '

exposure totaled 79.8 man-Rem. These doses are based on self-reading pocket dosimeter.

1 After retrieval of the foreign objects was complete, a visual in-4 spection of the tubes 'in the annulus area which could have been affected by the foreign objects path was performed using fiberoptics.

The fiberoptics, which has a magnification capability of 14X, showed only slight scratches and scraping of a limited number of tubes.

a To minimize the rate of corrosion, the Unit 1 primary system was

returned to power at a reduced hot leg temperature of 557*F.

I At the end of this reporting period, the Unit 1 "A" steam generator

, contains 468 plugged tubes and 10 sleeved tubes, and.the "B" steam generator contains-431 plugged tubes.

5.2 Unit 2 Unit 2 Refueling 8 Outage Inservice Inspection 1

, On 04/16/82, Unit 2 was shutdown for ' its eight ~ annual refueling.

I Eddy current examination of the steam generators commenced on 04/24/82. The original eddy current program for each ' steam generator was set up to meet the requirements of the Technical Specifications and Regulatory Guide 1.38. The B" steam generator program consisted of inspecting 575 tubes through the U-bend. This included 276 tubes which were identified as having degradation in the hot leg during the Refueling 7 inspection. The "B" steam generator program ' consisted of 0 inspecting 250 tubes through the U-bend and .31 tubes for the full length. The 250 tubes included.120 which were identified as having degradation in the hot leg during Refueling 7. All 31 tubes in-a spected for the full length had . degradation identified in the cold leg during Refueling-7.

i l

f E

The eddy current program in both the "A" and "B" steam generator hot legs was expanded in accordance with the Technical Specifications as defects were identified. An expansion in excess of 200 tubes in th e hot legs of both steam generators was performed af ter the original program. The results from the expansion in the "A" steam generator required an additional 400 tubes . to be inspected, however, the program was expanded to include essentially all of the tubes in the area of concern. This expansion in "B" was not required by the Technical Specification but was performed for prudent conse rvative engineering reasons. All the tubes in the expansion were inspected through the first tube support plate as the defects found were in the tubesheet region of just above the tubesheet. In excess of 60% of the C tubes in the "A" steam generator and 50% of the tubes in the "B" steam generator were inspected.

A two-fold evaluation of the "B" steam generator cold leg indications were done to determine if it was necessary to expand the eddy current program in the cold leg. First, the history of the tubes with indi-cations was looked at. The percentage size of the indication was compared with what was reported for the previous 5 years. Second, the Level IIA evaluator did a direct comparison of this year's eddy current signal with last year's signal. Both comparisons indicated that there was not a significant change in the condition in the cold leg and coupled with the exposure associated with setting up in the cold leg an expansion was not conducted.

The results of the eddy current inspection identified a total of 13 tubes which required plugging. The following table lists the tubes which required plugging along with three other tubes in the."A" steam generator which were plugged as a conservative measure.

"A" Steam Generator Hot Leg Tube Indication Location R16C34 96% 5" above tube end R17C36 90% 5" above tube end R20C38 77% 6-9" above tube end R18C41 82% 8" above tube end R18C43 Undefinable 9-13" above tube end R17C45 Undefinable 12" above tube end R13C46 89% 14" above tube end R21C47 87% 14" above tube end R12C63 48% .5" above tubesheet R19C63 27% 14" above tube end e

e 1

1 "B" Steam Generator Hot Leg Tube Indication Location R23C26 55% Top of tubesheet R23C27 43% Top of tubesheet R27C27 41% Top of tubesheet R23C28 53% Top of tubesheet R21C34 43% Top of tubesheet R06C44 55% .5" above tubesheet v All of the above-listed tubes were archanically plugged on 04/29/82.

All of the tubes with indications at the top of the tubesheet or above had identifiable indications during the 1981 refueling outage.

Only one of the tubes with indications in the tubesheet area had an identified indication during the 1981 refueling outage. Tube R19C63

! had a 25% indication at 14" above the tube end in 1981.

The indications within the tubesheet area are believed to be the result of intergranular attack caused by caustic corrosion. The indications at the top of the tubesheet or above are beleved to be remnants of phosphate wastage as evidenced by the fact that they were noted during previous outages. The called indications are not signi-ficantly greater than the plugging limit and for the most part the difference in comparison to previous outages is within the expected range of scatter for small volume indications which are masked by a tubesheet signal.

Following the normal refueling eddy current inspection, an annular search of the steam generator secondary side was performed.

The search of the Unit 2 steam generators revealed one piece of wire in the "A" steam generator and nothing in the "B" steam generator.

The wire in the "A" steam generator was on the hot leg side about 4' around from the handhole on the nozzle side. The wire was initially spotted leaning up against the tube bundle.

Following photographic documentation, the wire was removed from the steam generator. The area where the wire had been leaning against the tubes was slightly discolored from the corroding wire. An extensive fiberoptics inspection of the area was undertaken to check for possible tube damage. No signs of damage to the steam generator was apparent.

The wire measured 14" in length and 1/8" in diameter. The wire was identified to be a weld rod, e

It is not known how long the wire was inside the generator, but it is assumed that it was probably dropped in during the secondary side mechanical modifications effort in 04/78.

a In an attempt to reduce corrosion, the steam generators have been sludge lanced. Also, a crevice flush was conducted prior to returning the unit to service.

At the end of this reporting period, the Unit 2 "A" steam generator

contains 92 plugged tubes, and the "B" steam generator contains 41 plugged tubes.

Results of Eddy Current Examination Indication "A" SG "A" SG "B" SG "B" SG Identified Hot Leg Cold Leg Hot Leg Cold Leg F20% 361 1 117 2 20-29% 44 0 27 13 30-39% 13 0 28 17 '*

40-49% 1 0 3 0 50-59% 0 0 3 0 60-69% 0 0 0 0 70-79% 1 0 0 0 80-89% 3 0 0 0 90-99% 2 0 0 0 100% 0 0 0 0

, UDI 102 0 49 0 ATE - Above tube end ATS - Above tubesheet TTS - Top of tubesheet UDI - Undefinable indication TSP - Tube support plate 1

l t

I l

l l

I

a e

"A" Steam Generator Hot Leg Row Column Indication Location Origin Plugged 18 5 27% ITSP OD No 19 5 36% ITSP OD No 20 6 32-25% ITSP-2 TSP OD No 23 8 25% ITSP OD No 11 18 24% TTS OD No 4 19 F20% TTS OD No 9 19 24% TTS OD No 10 19 26% TTS OD No 11 19 21% TTS OD No 12 19 F20% TTS OD No 5 20 F20% TTS OD No 13 20 F20% TTS OD No 5 21 F20% TTS OD No 6 21 F20% TTS OD No 8 21 F20% TTS OD No 3 22 36% TTS OD No 4 22 UDI TTS OD No 5 22 UDI TTS OD No 6 22 UDI TTS OD No 7 22 F20% TTS OD No 8 22 F20% TTS OD No 9 22 F20% TTS OD No 13 22 26% TTS OD No 17 22 F20% TTS OD No 18 22 F20% TTS OD No 20 22 F20% TTS OD No 23 22 F20% TTS OD No 26 22 F20% TTS OD No 3 23 37% TTS OD No 4 23 31% TTS OD No 5 23 F20% TTS OD No 6 23 UDI TTS OD No 7 23 F20% TTS OD No 8 23 F20% TTS OD No 9 23 UDI TTS OD No 10 23 F20% TTS OD No 11 23 UDI TTS OD No 12 23 28% TTS OD No 13 23 F20% TTS OD No 14 23 UDI TTS OD No 16 23 F20% TTS OD No

' 17 23 UDI TTS OD No 18 23 31% TTS OD No 19 23 21% TTS OD No

• 20 23 TTS OD No 21%

21 23 F20% TTS OD No 22 23 F20% TTS OD No 23 23 F20% TTS OD No 24 23 F20% TTS OD No

- _ _ _ _ . . . . _ _ _ . - . __ m i

Row Column Indication Location Origin Plugged 25 23 F20% TTS OD No 26 23 F20% TTS OD No 5 24 F20% TTS OD No 6 24 F20% TTS OD No 7 24 F20% TTS OD No 1 8 24 UDI TTS OD No I

9 24 UDI TTS OD No 11 24 21% TTS OD No 12 24 F20% TTS OD No -

13 24 F20% TTS OD No 16 24 F20% TTS OD No i 17 24 F20% TTS OD No 1

18 24 F20% TTS OD No

19 24 30% TTS OD No 21 24 F20% TTS OD No

! 22 24 F20% TTS OD No

! 24 2 ** F20% TTS OD No 6 25 31% TTS OD No 7 25 F20% TTS OD No j 8 25 F20% TTS OD No 9 25 24% TTS OD No 10 25 F20% TTS OD No i

11 25 F20% TTS OD No 12 25 F20% TTS OD No 13 25 21% TTS OD No 14 25 34% TTS OD No 15 25 F20% TTS OD .No 16 25 UDI TTS OD No 17 25 F20% TTS OD No 18 25 F20% TTS OD No 19 25 23% TTS OD No 20 25 F20% TTS OD No i 21 25 F20% TTS OD No I

4 26 F20% TTS OD No 6 26 F20% TTS OD No 7 26 F20% TTS OD No

! 8 26 F20% TTS OD No

9 26 F20% TTS OD No 10 26 F20% TTS OD No 11 26 F20% TTS OD No 12 26 F20% TTS OD No j 13 26 F20% TTS OD No l 14 26 F20% TTS OD No l 15 26 F20% TTS OD No l 16 26 F20% TTS OD No. a j 17 26 24% TTS OD No

! 18 26 29% TTS OD No l 19 26 33% TTS OD No .

20 26 25% TTS OD No 21 26 F20% TTS OD No 22 26 F20% TTS OD No-23 26 F20% TTS OD No -

.s Row Column Indication Location Origin Plugged 4 27 24% TTS OD No 5 27 F20% TTS OD No 6 27 F20% TTS OD No s 7 27 F20% TTS OD No -'

, 8 27 F20% TTS OD No ,

9 27 F20% TTS OD No 11 27 F20% TTS OD No j 12 27 F20% TTS OD No- i

• l 13 27 F20% TTS OD ,' No 14 27 F20% TTS OD. No ,

15 , 27 F20% TTS OD '-

No 16 27 F20% TTS OD No 17 27 F20% TTS OD No 18 27 27% TTS OD No 19 27 27% TTS OD No 20 27 27% TTS OD No 21 27' F20% TTS OD No q 26 27 F20% TTS OD No -1 5 28 F20% TTS OD No il 6 28 27% TTS OD No 'l 7 28 24% TTS OD No 8 28 F20% TTS OD No l 9 28 F20% TTS OD No l 10 28 F20% TTS OD No i 11 28 F20% TTS OD No 12 28 F20% TTS OD No 13 28 F20% TTS OD No 14 28 F20% TTS OD No .,

16 28 F20% TTS OD No / 3 18 28 F20% TTS OD No s i

g' i

19 28 F20% TTS OD No s 20 28 F20% TTS OD No 21 28 UDI N :TTS OD No .O 26 28 UDI  ; TTS. OD No 5 29 UDI '7TS- OD No 6 29 UDI ,

TTS OD _'No.

9 29 UDI TTS OD No 10 29 F20% TTS OD No 11 29 F20% TTS OD No ,

20 29 .F20%\ '

TIS OD No 6 30 F20% \ TTS OD. No 7 30 UDI TTS OD No 8 30 F20% .;TTS OD No 4 31 F20% (TTS OD No

' 8 31 F20% N ~TTS OD No 14 31 F20% 'i

• h,TTS OD No 4

17 31 F20% ' 'TTS OD -No O 8 32 F20% TTS s. OD No 12 32 F20% TTS OD No.

13 32 F20%- TTS , OD No 'g 14 32 F20% TTS \,- OD No 5

• 4 21 32 UDI TTS -t OD .No ..

27 32 F20% TTS ,, 0D No

[

t

% a

Row Column Indication Location Origin Plugged l

6 33 F20% TTS OD No 11 33 F20% TTS OD No 12 33 F20% TTS OD No 13 33 F20% TTS OD No 21 33 26% TTS OD No 10 34 F20% TTS OD No '

11 34 F20% TTS OD No 12 34 26% TTS OD No 16 34 77% 6 to 9" ATE OD Yes

• 17 34 F20% TTS OD No 20 34 23% TTS OD No 21 34 F20% TTS OD No 27 34 25% (TS OD No 11 35 F20% TTS OD No 12 35 F20% TTS OD No 13 35 F20% TTS OD No 21 35 F20% TTS OD No 26 35 F20% TTS OD No 9 36 F20% TTS OD No

10 36 F20% TTS 0D No 11 36 F20% TTS OD No 12 36 F20% TTS OD No

, 17 36 90% 1/2" ATS OD Yes 9 37 F20% TTS OD No 10 37 F20% TTS OD No 11 37 F20% TTS OD No 12 37 F20% , TTS OD No s 13 '

37 F20% TTS OD No 14 37' F20% TTS OD No 9 38 , F20% TTS OD No 10 38 -

'F20% TTS OD No 12' (' 38 UDI TTS OD No 13' 38 F20% . TTS OD No

, . 16 38 3

, F20% TTS OD No 18 - 38 / F20% TTS OD No 19 38 i F20% TTS OD No 20 38 96% 5" ATE OD Yes 8 39 F20% 'TTS OD No 9 39 22% TTS OD No 11 39 F20% TTS OD No 12 39 F20%

f l TTS OD No 13 39 UDI '

! TTS OD No 17 39 F20%

TTS OD No-7 40 F20% g' TTS' OD No 8 40 F20% TTS OD No a 11 40 F20% TTS OD No 12 40 F20% \ TTS OD No 13 40 F20%

• TTS OD No a 15 40 UDI TTS OD No l

h .

Row Column Indication Location Origin Plugged 16 40 F20% TTS OD No 20 40 F20% TTS OD No 24 40 UDI TTS OD No 25 40 UDI TTS OD No

, 31 40 l'DI TTS OD No 6 41 P20% TTS OD No 16 41 F20% TTS OD No 18 41 82% 8" ATE OD Yes 19 41 F20% TTS OD No 24 41 UDI TTS OD No 25 41 UDI TTS OD No 32 41 UDI TTS OD No 6 42 F20%- Tr3 OD No 8 42 F20% TTS 00 No 10 42 UDI TTS OD No 12 42 F20% TTS OD No 13 42 F20% TTS OD No 17 42 UDI TTS OD No 19 42 UDI TTS OD No 23 42 UDI TTS OD No 24 42 F20% TTS OD No 25 42 F20% TTS OD No 26 42 F20% TTS OD No 27 42 F20% TTS OD No 20 42 21% .5" ATS OD No 6 43 F20% TTS OD No 8 43 F20% TTS OD No 18 43 UDI 9" to 13" ATE OD Yes 19 43 27% TTS OD No 22 43 UDI TTS OD No 25 43 F20% TTS OD No 26 43 F20% TTS OD No 27 43 F20% TTS OD No 28 43 F20% TTS OD No 6 44 UDI TTS OD No 7 44 UDI TTS OD No 8 44 UDI TTS OD No i 10 44 F20% TTS OD No 11 44 UDI TTS OD No 20 44 F20% TTS OD No 24 44 F20% TTS OD No 25 44 F20% TTS OD No 26 44 F20% TTS OD No 27 44 F20% TTS OD No

' 28 44 F20% TTS OD No 29 44 F20% TTS OD No 31 44 22% TTS OD No O 6 45 F20% TTS '0D No 7 45 F20% TTS OD No 8 45 UDI TTS OD No 9 45 UDI TTS OD No 12 45 UDI TTS OD No i l l

1 Row Column Indication Location Origin Plugged 17 45 UDI 12" ATE OD Yes 19 45 UDI TTS OD No 21 45 F20% TTS OD No 22 45 F20% TTS OD No 27 45 F20% TTS OD No 29 45 F20% TTS OD No

• 5 46 F20% TTS OD No 6 46 F20% TTS OD No 9 46 F20% TTS OD No .

10 46 F20% TTS OD No 11 46 UDI TTS OD No 13 46 89% 14" ATE OD Yes 14 46 F20% TTS OD No 20 46 F20% TTS OD No 21 46 32% TTS OD No 23 46 F20% TTS OD No j 24 26 F20% TTS OD No 29 46 UDI TTS OD No 6 47 F20% TTS OD No 7 47 UDI TTS OD No 8 47 UDI TTS OD No 12 47 UDI TTS OD No 13 47 UDI TTS OD No 15 47 F20% \" ATS OD No 19 47 F20% " ATS OD No 21 47 F20%-89% TTS-14" ATE OD Yes 32 47 F20% TTS OD No 2 48 F20% TTS OD No j 5 48 F20% TTS OD No 7 48 F20% TTS OD No 9 48 F20% TTS OD No 10 48 UDI TTS OD No 13 48 F20% TTS OD No 20 48 UDI TTS OD No 21 48 UDI TTS OD No 32 48 F20% TTS OD No 4 49 UDI TTS OD No 6 49 F20% TTS OD No 7 49 F20% TTS OD No 8 49 F20% TTS OD No l 11 49 UDI TTS OD No

' 12 49 F20% TTS OD No 13 49 F20% TTS OD No 14 49 UDI TTS OD No 19 49 UDI TTS OD No 3 20 49 UDI TTS OD No 32 49 26% TTS OD No 5 50 F20% TTS OD No .

6 50 F20% TTS OD No 7 50 F20% TTS OD No 8 50 F20% TTS OD No l l

Row Column Indication Location Origin Pluated 8 50 F20% ITS OD No 9 50 F20% TTS OD No 10 50 F20% TTS OD No 11 50 F20% TTS OD No

, 12 50 F20% TTS OD No 13 50 F20% TTS OD No 20 50 UDI TTS OD No 22 50 UDI TTS OD No 28 50 F20% TTS OD No 8 51 F20% TTS OD No 10 51 F20% TTS OD No 11 51 F20% TTS OD No 12 51 F20% TTS OD No 13 51 F20% TTS OD No 20 51 22% TTS OD No

22 51 F20% TTS OD No l 23 51 F20% TTS OD No

9 52 F20% TTS OD No

10 52 UDI TTS OD No 11 52 F20% TTS OD No 12 52 F20% TTS OD No 20 52 F20% TTS OD No 21 52 F20% TTS OD No 22 52 F20% TTS OD No 23 52 F20% TTS OD No 24 52 F20% TTS OD No 6 53 F20% TTS OD No 11 53 F20% TTS OD No 12 53 F20% TTS OD No 29 53 F20% TTS OD No 20 53 F20% TTS OD No 21 53 F20% TTS OD No 22 53 F20% TTS OD No 23 53 F20% TTS OD No 24 53 F20% TTS OD No 29 53 F20% TTS OD No 30 53 F20% TTS OD No 31 53 F20% TTS OD No 32 53 F20% TTS OD No 6 54 F20% TTS OD No 10 54 UDI TTS OD No 12 54 F20% TTS OD No 18 54 UDI TTS OD No 19 54 UDI TTS OD No

' 20 54 F20% TTS OD No 21 54 F20% TTS OD No.

22 54 F20% TTS OD No c>

24 54 F20% TTS OD . No 30 54 ~F20% TTS OD No 8 55 F20% TTS OD No 12 55 F20% TTS OD No

_ ~ _ _ - _ _ . . _ _

4 d

il Row Column Indication Locatiori Origin Plugged

. 13 55 UDI TTS OD No i

19 55 UDI TTS OD No 20 55 31% TTS OD No i 21 55 21% TTS OD No 22 55 UDI TTS OD No 23 55 UDI TTS OD No '

24 55 21% TTS OD No 26 55 UDI TTS OD No 30 55 21% " ATS OD No

• I 9 56 F20% TTS OD No

, 10 56 F20% TTS OD No 12 56 UDI TTS OD No 17 56 F20% TTS OD No i 18 SG UDI TTS OD No 19 56 UDI TTS OD No 20 56 F20% TTS OD No 21 56 F20% TTS OD No 22 56 UDI TTS OD No 23 56 F20% iTS OD No 24 56 F20% TTS OD No 9 57 F20% TTS OD No 10 57 UDI l

TTS OD No 11 57 F20% " ATS OD No 12 57 F20% TTS OD No 19 57 F20% TTS OD No 20 57 F20% TTS OD No 23 57 F20% TTS OD No 24 57 F20% TTS OD No

, 9 58 F20% TTS OD No 12 58 F20% TTS OD No 17 58 F20% TTS OD No 18 58 UDI TTS OD No 19 58 F20% TTS OD No 20 58 F20% TTS OD No 22 58 UDI TTS OD No 23 58 F20% TTS OD No

, 24 58 UDI TTS OD No l 26 58 F20% TTS OD No 9 59 F20% TTS OD No 12 59 F20% TTS OD No 17 59 F20% TTS OD No 18 59 F20% TTS OD No 19 59 F20% TTS OD No 23 59 UDI TTS OD No 24 59 UDI TTS OD No )

11 60 UDI TTS OD No 17 60 UDI TTS OD No 18 60 UDI TTS OD No e 19 60 UDI TTS OD No 20 60 F20% TTS OD No 21 60 UDI TTS OD No l 22 60 UDI TTS OD - No l 26 60 F20% TTS OD No

- _1

Row Column Indication Location Origin Pluated 4

10 61 F20% TTS OD No 11 61 30% TTS OD No 12 61 F20% TTS OD No 17 61 F20% TTS OD No 18 61 UDI TTS OD No 8'

19 61 UDI TTS OD No 21 61 F20% TTS OD No 22 61 UDI TTS OD No 6 23 61 F20% TTS OD No 24 61 F20% TTS OD No 25 61 F20% TTS OD No 26 61 F20% TTS OD No 6 62 F20% TTS OD No 9 62 F20% TTS OD No 10 62 29% TTS OD No 11 62 F20% TTS OD No 12 62 F20% TTS OD No 17 62 F20% TTS OD No 18 62 F20% TTS OD No 19 62 F20% TTS OD No 20 62 F20% TTS OD No 23 62 F20% TTS OD No 24 62 F20% TTS OD No 25 62 F20% TTS OD No 10 53 F20% TTS OD No 12 61 48% .5" ATS OD Yes 17 61 F20% TTS OD No 18 63 F20% TTS OD No 19 63 F20%-27% TTS-14" ATE OD Yes 22 63 21% TTS OD No 23 63 F20% TTS OD No 24 63 F20% TTS OD No 25 63 F20% TTS OD No 9 64 F20% TTS OD No 12 64 F20% TTS OD No 13 64 F20% TTS' OD No 17 64 UDI TTS OD No 19 64 F20% TTS OD No 20 64 F20% TTS OD No 22 64 F20% TTS OD No 10 65 F20% TTS OD No 11 65 F20% TTS OD No 12 65 F20% TTS OD No 20 65 21% TTS OD No

( 21 65 F20% TTS OD No 22 65 F20% TTS OD No 9 66 F20% TTS OD No n, 11 66 UDI- TTS OD No 12 66 F20% TTS OD No 13 66 F20% TTS OD No

Row Column Indication Location Origin Plugged 16 66 UDI TTS OD No 17 66 F20% TTS OD No 20 66 F20% TTS OD No 21 66 F20% TTS OD No 8 67 UDI TTS OD No 12 67 F20% TTS OD No '$

13 67 F20% TTS OD No 17 67 F20% TTS OD No 18 67 UDI TTS OD No a 19 67 F20% TTS OD No 20 67 F20% TTS OD No 21 67 F20% TTS OD No 6 68 F20% TTS OD No 12 68 F20% TTS OD No 16 68 F20% TTS OD No 17 68 F20% TTS OD No 19 68 F20% TTS OD No 20 68 F20% TTS OD No 7 69 F20% TTS OD No 8 69 F20% TTS OD No 14 69 UDI TTS OD No 16 69 UDI TTS OD No 17 69 F20% TTS OD No 18 69 F20% TTS OD No 19 69 F20% TTS OD No 20 69 F20% TTS OD No 12 70 F20% TTS OD No 17 70 F20% TTS OD No 18 70 F20% TTS OD No 19 70 F20% TTS OD No 20 70 UDI TTS OD No 13 71 F20% TTS OD No 16 71 F20% TTS OD No 18 71 F20% TTS OD No 10 72 F20% TTS OD No 11 72 F20% TTS OD No 12 72 F20% TTS OD No i 13 72 F20% TTS OD No 14 72 F20% TTS OD No 17 72 F20% TTS OD No 4 73 UDI TTS OD No 13 73 F20% TTS OD No 14 73 F20% TTS OD No 17 73 F20% TTS OD No 10 74 UDI TTS OD No 3 12 74 F20% TTS OD No l 13 74 21% TTS OD No l 14 74 25% TTS OD No a l 9 75 UDI TTS OD No 10 75 F20% TTS OD No 11 75 F20% TTS OD No l

)

i Row Column Indication Location Origin Pluaaed 12 75 25% TTS OD No 13 75 UDI TTS OD No 4 76 UDI TTS OD No 6 76 F20% TTS OD No 7 76 F20% TTS OD No 8 76 F20% TTS OD No 9 76 UDI TTS OD No 10 76 F20% TTS OD No G 11 76 TTS F20% OD No 12 76 UDI TTS OD No 13 76 F20% TTS OD No 3 77 UDI TTS OD No 4 77 29% TTS OD No 8 77 UDI TTS OD No 9 77 F20% TTS OD No 10 77 F20% TTS OD No 4 79 F20% TTS OD No 10 79 F20% TTS OD No 9 80 F20% TTS OD No.

"A" Steam Generator Cold Leg Row Column Indication Location Origin Pluated 12 42 F20% 2" ATS l

l 5

i l 4 L.

s

. _ - - . . _ _ = __ . . _ _ . - _ _ - -- -

"B" Steam Generator

!, Hot Leg f

Row Column Indication Location Origin Plugged l 5 18 29% .5" ATS OD No l 7 18 F20% TTS OD No l 11 18 F20% TTS OD No j 6 21 F20% TTS OD No '

6 23 21% TTS OD No 8 23 24% .5" ATS OD No 14 23 F20% TTS OD No * '

18 23 F20% TTS OD No 19 23 28% 1" ATS OD No 9 24 F20% 1" ATS OD No 14 24 F20% TTS OD No 17 24 39% TTS OD No 18 24 F20% TTS OD No 19 24 F20% TTS OD No 24 24 F20% TTS OD No 9 25 21% TTS OD No 12 25 F20% TTS OD No t

13 25 F20% TTS OD No 14 25 F20% TTS OD No 15 25 F20% TTJ OD No 17 25 F20% TTS OD No 18 25 F20% TTS OD No 23 25 UDI TTS OD No 25 25 36% TTS OD No 9 26 23% " ATS OD No 21 26 F20% TTS OD No 23 26 55% TTS OD Yes 24 26 39% .5" ATS OD No 25 26 37% TTS OD No 18 27 F20% TTS OD No 21 27 26% TTS OD No 22 27 F20% TTS OD No 23 27 43% TTS OD Yes 27 27 41% TTS OD Yes 17 28 UDI TTS OD No 20 28 F20% TTS OD No 21 28 F20% TTS OD No 23 28 53% TTS OD Yes 26 28 26% TTS OD No 27 28 F20% TTS (D No 6 29 22% TTS. GD No I

13 29 F20% TTS OD No 18 29 F20% TTS OD No )

23 29 34% TTS OD No 24 29 29% TTS OD No 6 30 25% TTS OD No 2 22 30 21% 1" ATS OD No 23 30 28% TTS OD No

, 6 31 F20% TTS OD No i

7 31 F20% TTS OD No l

I

Row Column Indication Location Origin Plugged 21 31 F20% TTS OD No 18 32 UDI TTS OD No 21 32 F20% TTS OD No 16 33 F20% ITS OD No 20 33 F20% TTS OD No

, 13 34 F20% TTS OD No 21 34 43% " ATS OD Yes 6 35 UDI TTS OD No 9 35 UDI TTS OD No 5 36 F20% TTS OD No 4 37 F20% TTS OD No 5 37 F20% TTS OD No 7 37 UDI TTS OD No 6 38 UDI TTS OD No 14 38 28% .5" ATS OD No 28 38 26% TTS OD No 30 38 39% TTS OD No 5 39 F20% TTS OD No 6 39 F20% TTS OD No 9 39 UDI TTS OD No 12 39 UDI TTS OD No 28 39 F20% TTS OD No 29 39 23% TTS OD No 32 39 30% .5" ATS OD No 33 39 34% .5" ATS OD No 5 40 39% TTS OD No 13 40 F20% TTS OD No 28 40 F20% TTS OD No 29 40 35% TTS OD No 32 40 29% .5" ATS OD No 6 41 F20% TTS OD No 12 41 F20% TTS OD No 26 41 F20% TTS OD No 28 41 F20% TTS OD No 33 41 32% .5" ATS OD No 5 42 UDI TTS OD No 6 42 F20% TTS OD No 9 42 22% TTS OD No 19 42 32% .5" ATS OD No 28 42 F20% TTS OD No 31 42 F20% .5" ATS OD No 32 42 23% TTS OD No 6 43 UDI TTS OD No 7 43 UDI TTS OD No 13 43 F20% TTS OD No A

19 43 UDI TTS OD No 26 43 UDI TTS OD No 6 44 55% \" ATS OD Yes 0

10 44 F20% TTS OD No 11 44 UDI TTS OD No 12 44 UDI TTS OD No 19 44 27% TTS OD No 21 44 UDI TTS OD Ka 1

'l Row Column Indication Location Origin Plugged 21 44 UDI TTS OD No 22 44 F20% TTS OD No 23 44 38% TTS OD No 24 44 F20% TTS OD No 31 44 22% TTS OD No 5 45 F20% TTS OD No 6 45 F20% TTS OD No  ?

10 45 F20% TTS OD No 11 45 F20% TTS OD No 22 45 39% TTS OD No

• 23 45 F20% TTS OD No 10 46 F20% TTS OD No 11 46 F20% TTS OD No 13 46 UDI TTS OD No 22 46 39% TTS OD No 32 46 32% TTS OD No 10 47 F20% TTS OD No i 11 47 F20% TTS OD No 13 47 F20% TTS OD No 14 47 F20% TTS OD No 21 47 39% TTS OD No 22 47 F20% \" ATS OD No 23 47 F20% " ATS OD No 26 47 F20% TTS OD No 9 48 F20% TTS OD No 12 48 F20% TTS OD No 13 48 F20% TTS OD No 14 48 F20% TTS OD No 17 48 F20% " ATS OD No 21 48 39% TTS OD No 22 48 39% TTS OD No 23 48 F20% " ATS OD No 24 48 F20% TTS OD No 25 48 UDI TTS OD No 9 49 F20% TTS OD No 12 49 21% TTS OD No 21 49 23% TTS OD No 22 49 39% TTS OD No 23 49 F20% TTS OD No 24 49 F20% \" ATS OD No 25 49 34% TTS OD No 26 49 29% TTS OD No 27 49 22% TTS OD No 1 28 49 F20% TTS OD No I

29 49 UDI TTS OD No 30 49 UDI TTS '0D No )

11 50 F20% TTS OD No 13 50 F20% " ATS OD No 18 50 F20% TTS OD No .)

19 50 F20% TTS OD No l 21 50 F20% TTS OD No l

I Row Column Indication Location Origin Plugged 21 50 F20% TTS OD No 22 50 34% TTS OD No 23 50 39% TTS OD No 26 50 F20% TTS OD No 27 50 UDI TTS OD No 28 50 39% TTS OD No 29 50 39% TTS OD No 4 51 F20% TTS OD No 6 51 F20% TTS OD No O 9 51 F20% TTS OD No 10 51 F20% TTS OD No 11 51 F20% TTS OD No 18 51 F20% \" ATS OD No 23 51 UDI TTS OD No 25 51 21"-F20% 1" ATS-TTS OD No 26 51 UDI TTS OD No 28 51 31% TTS OD No 11 52 F20% TTS OD No 22 52 UDI TTS OD No 23 52 UDI TTS OD No 24 52 F20% TTS OD No 25 52 F20% " ATS OD No 28 52 F20% \" ATS OD No 29 52 F20% TTS OD No 9 53 UDI TTS OD No 11 53 F20% TTS OD No 26 53 UDI TTS OD No 28 53 21% TTS OD No 29 53 UDI TTS OD No 10 54 UDI TTS OD No 11 54 F20% TTS OD No 25 54 F20% TTS OD No 26 54 UD1 TTS OD No 27 54 F20% TTS- OD No 28 54 F20% TTS OD No 29 54 F20% TTS OD No 23 55 UDI TTS OD No 25 55 F20% TTS OD No 26 55 F20% TTS OD No 10 56 UDI TTS OD No 21 56 UDI TTS OD No 22 56 UDI TTS OD No 23 56 UDI TTS OD No 10 57 UDI TTS OD- No 13 57 UDI TTS OD No l -14 57 UDI TTS OD No 21 57 F20% TTS OD No 22 57 UDI TTS OD No L 11 58 UDI TTS OD No 14 58 UDI -TTS OD -No 15 58 UDI TTS OD No 17 58 F20% TIS OD No 21 58 UDI TTS OD No e

l Row Column Indication Location Origin Pluased 23 58 33% " ATS OD No 24 58 UDI TTS OD No 9 59 F20% TTS OD No 15 59 F20% TTS OD No 21 59 F20% TTS OD No 22 59 F20% TTS OD No 23 59 36% TTS OD No  ?

24 59 UDI TTS OD No 11 60 UDI TTS OD No 17 60 UDI TTS OD No D j 23 60 UDI TTS OD No 64 14 F20% TTS OD No 7 65 F20% TTS OD No 9 65 F20% TTS OD No 23 65 22% TTS OD No 7 66 F20% TTS OD No 15 66 F20% TTS OD No 4 7 67 F20% TTS OD No 4

4 73 F20% TTS OD No 5 73 22% TTS OD No 5 74 28% TTS OD No 8 74 F20% TTS OD No 10 74 F20% TTS OD No 9 75 UDI TTS OD No 6 77 F20% TTS OD No i

l J

l

I l l

l "B" Steam Generator Cold Leg Row Column Indication Location Origin Plugged 9 26 33% 1" ATS OD No 10 26 24% 1" ATS OD No g 9 28 36% \" ATS OD No 18 32 28% \" ATS OD No 7 35 36% " ATS OD No 4 9 35 31% \" ATS OD No 11 36 27% " ATS OD No 7 37 24% \" ATS OD No 7 38 F20% \" ATS OD No 8 38 29% \" ATS OD No 7 39 35% 1" ATS OD No 5 40 29% " ATS OD No 7 40 32% \" ATS OD No 7 41 30% \" ATS OD No 7 43 23% " ATS OD No 7 44 31% " ATS OD No 11 44 31% " ATS OD No 12 44 F20% " ATS OD No 9 46 31% \" ATS OD No 7 47 33% 1.5" ATS OD No 6 48 31% 1" ATS OD No 7 48 30% 1" ATS OD No 9 48 31% 1.5" ATS OD No 9 49 24% 1.5" ATS OD No 18 50 29% " ATS OD No 19 50 36% " ATS OD No 6 51 24% 1" ATS OD No 9 51 27% \" ATS OD No 9 52 24% 1" ATS OD No 23 53 21% \" ATS OD No 7 59 31% 1.5" ATS OD No 23 54 33% " ATS OD No l

L f'

l

6.0 REACTOR COOLANT SYSTEM RELIEF VALVE CHALLENGES There were no challenges to either the reactor coolant system power-operated relief or safety valves during the year 1982.

1

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'I F

Wisconsin Electnc ecua cour>ur 231 W. MICHIGAN. P.O. BOX ?O46. MILWAUKEE, WI 53201 February 25, 1983 Mr. J. G. Keppler, Regional Administrator Office of Inspection and Enforcement, Region III U. S. NUCLEAR PIGULATORY COMMISSION 799 Roosevelt Road Glen Ellyn, Illinois 60137

Dear Mr. Keppler:

DOCKET NOS. 50-266 AND 50-301 ANNUAL RESULTS AND DATA REPORT POINT BEACH NUCLEAR PLANT, UNITS 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 1982. 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, "6 -

Vice President - Nuclear Power C. W. Fay k Enclosure f/ g l%

Copies to NRC Resident Inspector g '

Director, Office of Inspection and /

Enforcement (40 copies) d r