ML18142A600
| ML18142A600 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/14/2018 |
| From: | Rochester Gas & Electric Corp |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18142A600 (112) | |
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ME Mil GINNA STATION ANNUAL REPORT 512 JANUARY 1, 1976 TO IDEGEMBER 31, 1976 INGLUSIIVE ACKET 550-244 gp'gag 5o ~
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ANNUAL OPERATING REPORT fP12 OF R. E. GINNA NUCLEAR STATION ROCHESTER GAS & ELECTRIC CORPORATION FOR 1976 DOCKET NO. 50-244 LICENSE NO. DPR-18
TABLE OF CONTENTS PAGE INTRODUCTION HIGHLIGHTS 2- 3
SUMMARY
OF OPERATING EXPERIENCE 4-8 PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY-RELATED MAINTENANCE Amendments to Facility License or Technical Specifications 10 - 11 Facility or Procedure Changes Requiring NRC Approval 12 Tests and Experiments Requiring NRC Approval 13 Other Changes, Tests, and Experiments 14 23
'orrective Maintenance of Safety-Related Equipment 24 Mechanical and Electrical Maintenance 25 56 Instrument and Control Maintenance 57 61 LICENSEE EVENT REPORTS 62 73 OTHER EVENTS OF INTEREST 74 Fuel Performance 75 Fuel Data 76 Core Loading Pattern 77 Steam Generator Inspection(s), 78 82 Summary of Containment Integrated Leak Rate Tests 83- 84 DATA: TABULATIONS 85 Net Electrical Power Generation 86 Unit Shutdowns and Forced. Power Reductions 87 90 Number of Personnel and Man/Rem Exposure by Work and Job Function 91 Plant Organization and Personnel 92 Operating Histogram 93 UNIQUE REPORTING REQUIREMENTS Condenser Tube Maintenance 95 Primary, S/G Blowdown Chemistry 96 Leak Tests on Sources 97 Radioactive Effluents 98- 99 Solid Waste 100 101 Environmental Monitoring 102 105 GLOSSARY 106
INTRODUCTION R. E. Ginna Nuclear Station is a pressurized water reactor of 470 maxi-mum dependable capacity, in MWe net, owned by Rochester Gas 6 Electric Corporation and located in Ontario, New York. The nuclear steam supply system is a Westinghouse PWR. The Architect/Engineer was Gilbert Associ-
~
ates, and the constructor was Bechtel. The condenser cooling method is once-through, and Lake Ontario is the condenser cooling water source. The Plant is subject to license DPR-18, issued September 19, 1969 pursuant to Docket Number 50-244. The date of initial reactor criticality was November 9, 1969, and commercial generation of power began July 1, 1970.
This report was prepared by G. P..Larizza, 716/546-2700, Extension 291-245.
HIGHLIGHTS Robert Emmett Ginna Station began the year in cold shutdown mode, for repair of a tube leak in the "B" steam generator. (See Page 78 of this report). The reactor was brought critical on January 13 of this report period, and the unit was subsequently returned to 100% power. On January 29, the unit was forced out of service due to blade failure in the 82 low pressure turbine. The unit remained shutdown for repair of the 82 LP turbine. During this shutdown, inspection of the "A" and "B" steam generators, annual refueling and other maintenance were also completed.
On April 7 the reactor was taken critical, but inadvertently tripped from less than 1% power by technicians working on:the Overpower AT instrumentation. The unit was returned to power, and on April 12, a turbine runback occurred, due to power range N-44 High Voltage Power Supply Failure (see page 66), causing a reactor power reduction. The power level was increased to 70% on April 14, and reduced to 50% because of vibration of the "A" main feedwater pump. On April 16, the reactor was taken to hot shutdown to retrieve two dropped control rods (Rods G-5 and G-9, see Page 66). The reactor was started again, and power level increased.
On April 18, problems with the E-H Governor caused the unit to be taken off the line. (See Page 5). The unit remained shutdown until April 20, when reactor startup and power escalation began. A tube leak developed in the "B" steam generator, which required a reactor shutdown on April 24.
(See Page 78). The unit remained shutdown until May 7 for inspection and repairs. On May 7, while at 45% power, a gasket in the 1A moisture separator reheat steam line failed, causing the unit to be shutdown.
(See Page 5). The unit was started on May 8, but the reactor tripped due to lo-lo steam generator water level. Recovering from the reactor trip, the unit was brought to 42% power. On May 10, power was increased to 90% power. Reactor power level was reduced to 50% on May 12 due to failure of the "A" main feedwater pump impeller, (see page 5), and remained at 50% power until May 18. The reactor was brought to 100%
power on May 19, but the unit was taken off the line on May 22 due to a fluid leak from the E-H system. (See Page 5). The reactor was started up on May 23, and power was increased to 100% the next day. On June 3, the reactor tripped due to failure of the "B" Inverter (See Page 5). Repairs were made, and the reactor was started up the same day and returned to 100% power, where power remained for the rest of June.
On July 4, 1976, control rods G-5 and G-9 dropped again. (See Page 68);
and the reactor was taken to hot shutdown to retrieve the two rods.
(The same two rods had dropped on April 16). Reactor power was returned to 100% on July 5, decreased to 81% on July 6 to isolate a condensate pump, and returned to 100%. Reactor power was maintained at 100% until July 29, when power was reduced to 50% due to turbine bearing vibration.
(See Page 6). Power maintained at 50% until July 30, then gradually returned to 100% power.
It was suspected that /f2 LP turbine had lost a mass of metal. On August 4, the reactor was shutdown to retrieve two dropped control rods (G-3 and G-ll, See Page 69). The unit remained shutdown until August 6, at which time it was restarted and returned to 100% power.
The following day the unit was manually shutdown due to excessive turbine vibration. Investigation revealed blade failure in i/2 LP turbine.
(See Page 6). The plant remained shutdown for the rest of August for installation of baffles in the Turbine. The unit was returned to power on September 4, and the power level increased to 50%, where it remained until September 7. Then, power was gradually increased to 75%, and then to 90% on. September 8. Power level was decreased to 50% on September 10 due to failure of the "B" main feedwater pump impeller. (See Page 7).
Power was increased to 95% on September 18 after repair of the pump, and remained at that power level for the rest of September. On October 5, power was increased to 100%, and remained there until October 8, at which time the reactor'as placed in the cold shutdown mode due to leakage from piping between the Boric Acid Storage Tanks and the Safety Injection pumps. (See Page 70). Replacement of sections of pipe continued until October 29, when plant heat up commenced. The reactor was brought critical on October 30, and then power level increased to 100% power. Power level remained at 100% throughout November and into December. On December ll, power level was reduced to 45% to inspect 1B-1 condenser for leaks. Power was restored to 100% on December 12, and then returned to 45% due to loss of condenser vacuum caused by failure of the air ejector steam supply valve. Power was restored to 100% power, and on December 13, a turbine runback, caused by Overpower AT, Channel 406B, reduced reactor power to 80%. Power was subsequently returned to 100%.
On December 14, power was reduced to 46% to check the 1B-2 condenser for leaks, and then restored to 100%. Power was decreased to 46% on Decem-ber 17 to again check for leaks in the 1B-2 condenser. While at 46%
power, control rod F-12 dropped (see Page 72). The reactor was shutdown for repairs to the control rod hold circuit".. The reactor was restarted on December 18 and returned to 100% power, wheie it remained for the rest of the year.
SUMMARY
OF OPERATING EXPERIENCE The following is a chronological description of Plant Operations including other pertient items of interest for the twelve month period ending December 31, 1976.
I 1/1 -.1/3 The unit remained shutdown during the period, for repair of a tube leak in the "B" Steam
.,Generator.
1/4 - 1/10 The unit was shutdown during the period.
1/11 - 1/17 The unit was started up on 1/13; the Reactor Power Level was increased to 52% on 1/14 and to 100% on 1/15'and remained at this level for the remainder of the period.
, 1/18 - 1/24 The Reactor Power Level was maintained at 100%
during the period.
1/25 - 1/31 The Reactor Power Level was.maintained at 100%
until 1/29 when the unit was forced out of service due to blade failure in the fj2 Low Pressure Turbine.
2/1 - 2/29 The unit remained shutdown during the period for annual refueling, fP2 Low Pressure Turbine blade repair, "A" and "B" steam generators inspection and repairs, and other maintenance.
3/1 >> 3/31 The unit remained shutdown during the period for annual refueling, fP2 Low Pressure Turbine blade repair, "A" and "B" steam generators inspection and repairs, and other maintenance.
4/1 - 4/3 The annual refueling and maintenance outage continued.
4/4 << 4/10 On 4/7, after the reactor was brought critical, a reactor trip occurred at 0% power while technicians were working on the over power Delta T Instrumentaion.
The annual A,I&0 ended on 4/10 91245; The reactor power was gradually increased to the 20% level.
4/11 - 4/17 The reactor power was increased to the 40% level on 4/ll. On 4/12 a turbine run back occurred which necessitated a reactor power level reduction~27%;
The turbine runback was caused by the failure of Power Range N-44 High Voltage Power Supply. The reactor power was restored to the 40% level after corrective actions were effected. On 4/14 the reactor power was increased to the 70% level; on the same date a power level reduction to 50% due to vibrations noted on the "A" Main Feedwater Pump. On 4/16 the Plant was taken to the hot shutdown. condition to retiieve two dropped Control Rods, G-9 .and G-5. The unit was started up again, on the same date and the reactor power level was increased to 50%. The reactor power..level was maintained at 50% through 4/17.
I
4/18 - 4/24 On 4/18 the unit was taken off line twice, at 0600 and at 1345, due to E.H. Governor problems
'ue to water in oil from oil cooler leak on 4/14/76.
The unit remained shutdown until 4/20 9.0600. The reactor power level was increased to 50% on that date, to the 80% level on 4/20, where it remained through 4/22. On 4/23 the reactor power level was increased to 98%; a further increase to 99% level on 4/24 .was followed by a unit shutdown Q 1745 due to a "B" Steam Generator Tube leak.
4/25 - 4/30 The unit remained shutdown through the end of the report period due to repairs of the "B" Steam Generator.
5/1 The Unit remained shutdown due to a tube leak in the "B" Steam Generator. (This shutdown began on 4/24/76).
5/2 - 5/8 The Unit was started up on 5/7/76 Q 1415; Reactor Power was being increased to the 45% level when a gasket failure in the lA Moisture Separator Reheat Steam Line caused a unit shutdown Q 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. On 5/8, the Unit was started up Q 0545; prior to staxt-up, a Reactor Trip occurred due to low-low Steam Generator level. After startup the Reactor power was increased to the 42% level.
5/9 5/15 The Reactor Power Level was maintained at ~42% until 5/10 when the power level was increased to 90%. On 5/12 the Reactor Power Level was reduced to 50% 9 0715 due to an impeller failure, l>>A Main Feedwater Pump. Reactor Power remianed at this level during the remainder of the period.
5/16 - 5/22 The Reactor Power Level remained at the 50% level until 2215 on 5/18, when impeller repairs were completed. The Reactor Power Level was increased gradually to the 100%
level on 5/19, and remained at the 100% level until 2200 on 5/22 when the unit was removed from service due to an E.H. Fluid leak in the E.H; System.
5/23 >> 5/29 The Unit was started up 9 0245 on 5/23; the Reactor Power Level was increased to 100% on 5/24. On 5/27 the Reactor Power Level was x'educed to 95% due to an E. H. problem. The Reactor Power was restored to the 100% level after corrective action, and remained at this level during the remainder of the report period.
6/1 - 6/5 The reactor power level averaged 100% until 6/3 when a reactor trip and generator trip occurred 9 0715 due to the 1-B Xnverter failure. The trip was due to low Steam Generator water level. The unit was restored to service 9 1530. The reactor power level was grad~
ually increased to 100% where of the period.
it remained for the rest 6/6 - 6/12 The reactor power level averaged 100% during the period.
6/13 >> 6/19 The reactor power level averaged 100% during the period.
6/20 - 6/26 The reactor power level averaged 100% during the period.
6/27 - 6/30 The reactor power level averaged 100% during the period.
7/1 - 7/3 The reactor power level averaged 100% in the period.'/4
- 7/10 The plant was brought to the hot shutdown condition at 0830 on 7/4 to retrieve two dropped control rods. The unit was started up at 1530 after re>>
trieval of dropped control':rods G"5 and G-9. The reactor power was incxeased to the 100% level on 7/5 and remained at this level until 7/6 when the reactor power level was reduced to ~81% to per-form T-5C, Isolation of a condensate pump'; ..he reactor power was increased to 100% level 'after.
restoration of the condensate pump to service, on 7/6. The reactor power remained at this level for the remainder of the period.
7/11 - 7/17 The reactor power level averaged 100% in the period.
7/18 - 7/24 The reactor power level averaged 100% in the period.
7/25 - 7/31 The reactor power level averaged 100% until 7/29.
At approximately 1735 the reactor power level was re-duced to 50%, due to turbine bearing vibration. The reactor power remained at this level until 7/30 when it was gradually increased to the 100% level, on 7/31, after conclusion that the turbine had suffered a mass of metal loss from the fP2 low pressure turbine rotor. The reactor power level was maintained at 100%, for the remainder'of the period.
8/1 - 8/7 The Reactor Power Revel averaged 100% from 8/1'o 8/3.
On 8/4, the Plant was manually shutdown to the hot shutdown condition to retrieve two dropped contxol rods (G>>3 and G-ll). The Plant remained shutdown until the generator was placed on line at 0900 on 8/6. The Re-actor Power Level was increased to 100% on 8/6. On 8/7, at 0900 the.:generator was manually tripped due to excessive turbine vibration. Subsequent invest-igation and evaluation revealed the cause to be an L-2 failure~ 82 Low 'Pressure Turbine 8/8 - 8/14 The Unit remained at cold shutdown to repair. turbine blades.
8/15 - 8/21 The repair of turbine blades continues.
8/22 - 8/28 The repair of turbine blades continues.
8/29 - 8/31 Commenced reassembly of ij2 Low Pressure Turbine.
Plant System Lineups commence.
9/1 - 9/4 The turbine generator was placed on the line on 9/4 at 0430, after completion of repairs and mod-ifications to the 82 low pressure turbine. The reactor power level was gradually increased to the 50% level, on 9/4.
9/5 - 9/11 The reactor power remained at the 50% level until 9/7, when it was gradually increased to the 75%
level, On 9/8 the reactor power was increased to the 90% level. A further increase to the 96% level occurred on 9/9, where it remained until 9/10 when
,the power level was reduced to the 50% level due to the impeller failing on the 1>>B Main Feedwater, Pump. The power level remained at 50% during the remainder of the period.
9/12 - 9/18 The reactor power remained at the 50% level from 9/12 to 9/17 at 1830 when repairs were completed on the 1-B Main Feedwater Pump. The reactor power was gradually increased to the 95% level on 9/18.
9/19 - 9/25 The reactor power remained at the 95% level during the period.
9/26 - 9/30 The reactor power was increased to the 97% level on 9/27 and remained there until the end of the period.
10/1 - 10/2 The Reactor, Power level averaged 96% in the period.
10/3 >> 10/9 The Reactor Power was inc'reased to the 100% level on 10/5 and remained there until 10/8. On 10/8 the unit was placed in the Cold Shutdown condition due to a leak'n stainless steel pipe between the boric acid'..
tanks and the safety infection pumps. The unit re-mained shutdown for the remainder of the period.
10/10 - 10/16 The unit remained shutdown while maintenance continued.
10/17 10/23 The unit remained shutdown while maintenance continued.
10/24 - 10/30 The unit remained shutdown from 10/24 to 10/28. On 10/29 plant heat-up commenced; the generator was placed on line 9 0015 on 10/30. The Reactor Power was to the 99% level by the end of the day. in-'reased 10/31 The Reactor Power remained at the 99% level on this date.
11/1 - 11/6 The Reactor Power Level averaged 99% in the period.
ll/7 - 11/13 The ReactorPower Level averaged 99% until 11/10 when the level was raised to 100%, where the remainder of the period.
it remained for 11/14 - 11/20 The Reactor Power Level averaged 100% in the period.
11/21 - 11/27 The Reactor Power Level averaged 100% in the period.
ll/28 - 11/30 The Reactor Power Level averaged 100% in the period.
12/1 << 12/4 The Reactor Power Level was maintained at 100%
during the period.
12/5 - 12/11 The Reactor Power was maintained at the 100% level during the period, with the following exceptions:
On 12/5 the Reactor Power l,evel was reduced to 50%
to perform 1'-1SS, Main Steam Stop Valves Test.
On 12/ll, The Reactor Power Level was reduced to 45%
to check the 1B-1 Condenser for leaks.
12/12 << 12/18 The Reactor Power was maintained at the 100% level during the period with the following exceptions:
On 12/12, the Reactor Power was reduced to the 45%
level due to a loss of condenser vacuum caused by problems with the air ejector steam supply valve.
On 12/13 The Reactor Power was reduced to the 80% level due to a turbine runback that occurred when the over-power Delta T channel 406B failed Hi.
~On 12 14 The Reactor Power was reduced to the 46% level to -check the 1B>>2 condenser for leaks.
O~n12 17, The Reactor Power was reduced to the 46% level to check the 1B-2 condenser for. leaks. At 0114 Control Rod P-12 dropped; the unit was manually shutdown to the hot shutdown condition to retrieve the rod. The unit was started up at 1730; the reactor power was returned to the 100% level on 12/18.
12/19 12/25 The Reactor Power was maintained at the 100% level during the period.
12/26 - 12/31 The Reactor Power Level was maintained at 100% during the period
PLANT or PROCEDURE CHANGES TESTS EXPERIMENTS and SAFETY RELATED MAINTENANCE
AMENDMENTS TO FACILITY LICENSE OR TECHNICAL SPECIFICATIONS On February 4, 1976 the Nuclear Regulatory Commission issued Amendment 89.to the R. E. Ginna Nuclear Power Plant Provisional Operating License No. DPR-18, Paragraphs B.(1), B.(2), B.(3) and C.(2) were revised and new paragraphs B.(4) and B.(5) were added as follows:
"B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses RG&E:
Pursuant to Section 104b of the Act and 10CFR Part 50; "Licensing of Production and Utiliza-tion Facilities", to possess, use, and operate the facility at the designated location in Wayne County; New York, in accordance with the procedures and limitations set forth in this license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as amended through Supplement No. 15; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special :
nuclear material without restriction to chemical or physical form, for sample analysis or instru-ment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclar materials as may be produced by the operation of the facility."
Technical S ecifications The Technical Specifications contained in Appendix A, as revised, are hereby incorporated into the license. The licensee shall operate the facility in accordance with the Technical Specifications.
On March 30, 1976 the Nuclear Regulatory Commission issued Amendment
//10 to the R. E. Ginna Nuclear Power Plant Provisional Operating License No. DPR-18.
The amendment changes the Technical Specifications to revise the core thermal limits curve based on the application of the approved, updated Westinghouse model for fuel densification and clad flattening, a revi-sion in the overpower Delta T and overtemperature Delta T setpoints, a deletion of the allowance for relaxing end-of-life control rod insertion limits, and revised shutdown margin requirements based on a reanalysis of the main steam line break accident.
On November 15, 1976 the Nuclear Regulatory Commission issued Amendment No. 11 to the R. E. Ginna Nuclear Power Plant Provisional Operating License No. DPR-18, Paragraph 2.B.(2). was amended to read:
ItB (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material or reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation as described in the Final Safety Analysis Report, as amended and the Application for License Amendment dated January 30, 1976, supplemented by letters dated May 19, June 3, August 5 and September 29, 1976."
PACILITY OR PROCEDURE CHANGES RE UIRING NRC APPROVAL During the reporting period, Rochester Gas and Electric Corpora-tion received approval from the Nuclear Regulatory Commission to remove the existing spent fuel stacks and replace them with new style spent fuel racks. This change was required to provide additional on-site storage capability for spent fuel. Actual modification began on November 17, 1976.
There were no procedural changes during this reporting period that required prior approval by the Nuclear Regulatory Commission.
TESTS AND EXPERIMENTS RE UIRING NRC APPROVAL There were no tests or experiments during the reporting period which required prior approval by the Nuclear Regulatory Com-mission.
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OTHER CHANGES TESTS AND EXPERIMENTS S stem. Modification SM 75-5 System Modification SM 75-5, Standby Auxiliary Feedwater System was partially completed (Procedure SM 75-5.1, SM 75-5.21, SM 75-5.24, SM 75-5.26, SM 75-5.28, SM 75-5.29 and SM 75-5.32). Piping installation and tie-ins to existing systems were made to feedwater and service water systems in the Auxiliary Building and Containment Vessel.
This modification will provide auxiliary feedwater backup in the event of a high energy pipe break.
A Safety Evaluation was performed and it was determined that the Plant Operating Technical Specifications should be revised to include the additional limiting conditions for operation. This modification does not increase the possibility of an accident as previously evaluated.
This modification will not become operational until approval is received from the NRC.
S stem Modification SM 75-System Modification SM 75-9, Pipe Whip Restraints and. Jet Shields for Chemical Volume and Control Lines was partially completed (Procedures SM 75-9.21 and SM 75-9.23). Relocation of existing piping was made in the Auxiliary Building to accommodate future installations for th'is modification.
This modification will provide protection'or the refueling water storage tank in the event the Chemical Volume and Control Piping ruptures.
A of Safety Evaluation was performed and it was determined that the possibility an accident does not increase as previously evaluated. A revision to the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 75-13 System Modification SM 75-13, Reroute of the Acid, and Caustic Fill Lines was completed.
the end of the New fill valves and associated piping were installed from south Service Building.
This modification will provide a safer route for filling the Acid and Caustic Storage Tank. The old location of the struck by various delivery trucks.
fill lines sub)ected them to being A Safety Evaluation was performed and it was determined.
of an accident is not affected as previously evaluated.
that the possibility A change to the Plant Operating Technical Specifications is not necessary.
14
S stem Modification SM 75-15 System Modification SM 75-15, Auxiliary Building Ventilation Trip was completed. High radiation interlock contacts in the control circuits of Auxiliary Building Exhaust Fans 1A, 1B, lC and 1G and Auxiliary Building Charcoal Filter Fans 1A, lB and Control Access Area Exhaust Fans lA and 1B.
This modification will maintain a continuous air flow from the Auxiliary Building and Control Area.
A Safety Evaluation was performed and it was determined that no safety margin was reduced as previously evaluated. A change to the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 75-18 System Modification SM 75-18, Steam Generator Blowdown Reuse System was completed. A heat exchanger and associated piping and valves were in-stalled and tied to the existing Blowdown System, Condensate System and Makeup Water System.
This modification will recover both the blowdown water and the heat lost which is now being discharged. from the Blowdown System.
A of Safety Evaluation was performed and it was determined that the possibility an accident is not affected as previously evaluated. A change to the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 75-21 System Modification SM 75-21, Air Boosters f'r Diesel Generator Set. An air booster and associated. tubing was installed in each of two emergency diesel generators.
This modification will move the fuel rack (throttle) to the open position before the governor would otherwise be able to, facilitating rapid engine starting.
A of Safety Evaluation was performed.and an accident is not increased.
it was determined that the possibility A change in the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 75-29 System Modification SM 75-29, Replacement of 4kv Undervoltage Relays on Busses llA and llB. Westinghouse SV Undervoltage Relays 273-llA and 274-llA on 11A 4kv bus and 723-11B and. 274-11B on 11B 4kv bus were replaced.
with four new KV-1 relays.
This modification will reduce the probability of relay malfunction.
A Safety Evaluation was performed and it was determined. that no margin of safety is affected. A change in the Plant Operating Technica1 Specifications is not necessary.
S stem Modification SM 75-34 System Modification SM 75-34, Jet Shields and Pipe Whip Restraints in the Intermediate Building. Shielding was installed to protect the Steam Generator Blowdown Containment Isolation Valves.
This modification will protect components and systems required for safe shutdown of the Plant during the postulated transient conditions.
A of Safety Evaluation was performed an accident is not increased.
and. it was determined that the possibility A change in the Plant operating Technical Specifications is not necessary.
S stem Modification SM 75-37 System Modification SM 75-37, Installation of Fisher Controll'er for Main Feedwater Control Valves. A new valve operator with a hydraulic damper was installed on the (2) Copes Vulcan Feedwater Control Valves.
This modification will alleviate instabilities which allow the control valve to act as an amplifier, thereby increasing any forcing function that might exist in the feedwater lines.
A of Safety Evaluation was performed and. it was determined that the possibility an accident was not increased as previously evaluated. A change in the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 75-38 System Modification SM 75-38, Feedwater Pumps Seal Water Control System Modification. Piping valves and, instrumentation were installed for the Main Feedwater Pumps Seal Water System.
This modification will better maintain and indicate proper sea1 water flow in both Main Feedwater Pumps.
A Safety Eva1uation of safety will not was performed and it was determined that the margins be reduced during normal operations or transient conditions. A change in the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 75-43 System Modification SM 75-43, New Plant Sewage System. A domestic sewer line was installed and connected to the Ontario Sewer System.
This modification will provide better means for disposal of nonradioactive wastes from the Service Building, including showers, lavatories, drinking fountains and kitchen.
A Safety Evaluation was performed and it was determined that the margin of safety is not affected and the adequacy of structures, systems and components provided. for the prevention, of and. mitigation of consequences of accidents is not affected. A change in the Plant Operating Technical Specifications is not necessary.
16
S stem Modification SM 75-46 System Modification SM 75-46, Installation of Snubbers on the "B" Feedwater Pump Suction Piping. Four snubbers were installed in the basement level of the Turbine Builhing on the "B" Feedwater Pump Suction Piping.
This modification will add, damping to the piping system and hence reduce the vibration levels and associated stresses.
A Safety Evaluation was performed. and. it was determined that there are no failure modes of the design that would cause or increase the consequences of any of the postulated events. A change to the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 5-47 System Modification SM 75-47, Chemical Volume and Control System Letdown Line Supports. in Containment. Supports MK-CVCS-1 through MK-CVCS-10 were installed to the CVCS Letdown Line on the basement floor of the Containment Vessel.
This modification will provide additional support to the CVCS Letdown Line which will alleviate the overstress condition.
A of Safety Evaluation was performed and it was determined that the an accident was not increased as- evaluated.
possibility A change. in the Plant Operating Technical Specifications's not necessary.
S stem Modification SM 75-48 System Modification SM 75-48, Addition of Four Valves to the Auxiliary Feedwater System. Two check valves and two motor operated crossover valves (less wiring) and associated supports were installed. in the Intermediate Building basement and intermediate level Auxiliary Feedwater Piping.
The addition of two check. valves reduces the amount of auxiliary feedwater piping subject to rupture and.the two crossover valves improve system re-dundancy and increase crossover capability.
A Safety Evaluation. was performed and it was determined that the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated, will not be increased. A change in the Plant Operating Technical Specifications is not necessary.
17-
S stem Modification SM 75- 0 System Modification SM 75-50, Jet Shields for Main Steam and Feedwater Lines. Shielding was installed in the Intermediate Building, intermediate level to protect the Main Steam bypass piping and. valves, and. instrument tubing on the feedwater lines.
This modification'proyi'des for'protection from crack breaks or through-wall leakage cracks along the Main Steam and. Feedwater j ines.
A Safety Evaluation in was performed and itinstallation was determined that Plant safety margins are no way diminished by the of get shields. A change in the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 75- 5 System Modification SM 75-55, Removal 'of Guide Stud, from the Reactor Vessel Upper Internals Storage Frame. One of the three guide studs was removed.
from the upper internals storage stand,.
This modification provided clearance to move th'e upper internals without breaking the surface of the water in the reactor cavity, thus minimizing radiation exposure.
A Safety Evaluation was performed and it was determined that plant safety margins are in no way diminished. A change in the Plant Operating Technical Specifications is not necessary.
S stem Modification SM 7 6 System Modification SM 75-56, Modification of Main Feedwater Pump Discharge Piping. Two elbows of new design were:.'installed at the discharge of the two Main Feedwater Pumps.
This modification reduces the turbulence and provides materials with greater resistance to erosion and cavitation.
A Safety Evaluation was performed and it was determined. that the proposed.
modification does not change the consequences previously analyzed. A change in Plant Operating Technical Specifications is not necessary.
S stem Modification SM 76-03 System Modification, SM 76-03, Installation of Perforated Plates on the Steam Generator Moisture Separator Demister End Plates. Twenty-eight plates were installed on the upper tier demisters and twenty-four plates were in-stalled on the lower tier demisters of the "A" and. "B" Steam Generators.
This modification reduces the amount of moisture carry over which was in-creased by previous modifications to the "A" and "B" Steam Generators.
A Safety does not Evaluation was performed and it was determined that this modification affect the adequacy of structures systems or components provided for the prevention of accidents and the mitigation of the consequences of accidents.
A change in Plant Operating Technical Specifications is not necessary .
S stem Modification SM 76-04 System Modification SM 76-04, Speed Control on Fuel Transfer System. The slow speed control was eliminated on the Fuel Transfer System.
This modification was made to improve the performance of the fuel transfer controls since the slow speed control has not performed satisfactorily.
A Safety Evaluation was performed and it was determined. that the margins of safety during normal operation and transient conditions anticipated during the life of the station and the adequacy of structures, systems and components provided for the prevention of accidents and the mitigation of the conse-quences are not changed by this modification. A change in Plant Operating Technical Specifications is not necessary.
S stem Modification SM 76-06 System Modification SM 76-06, Seal Ring Clamp on the Reactor Coolant Pump.
The ring clamp for the "A" Reactor Coolant Pump Seal Assembly was reported..
This modification will force the entire ¹2 seal leakage of approximately 3gph to flow across the ¹3 seal before being directed to the Waste Processing System.
A Safety Evaluation was performed;.and it was determined that the margins of safety during normal operations and transient conditions anticipated during the life of the station have not been affected. It has a1so been determined that the adequacy of structures, systems and components provided. for the prevention of accidents and. the mitigation of the consequences of accidents have not been affected. A change in Plant Operating Technical Specifications is not necessary.
S stem Modification SM 76-07 System Modification SM 76-07, Waste Disposal System Addition. Piping and valves were installed. in the Auxiliary Building to the waste condensate demineralizer, monitor tanks and. waste evaporator.
This modification allows the steam generators to be drained directly to the waste condensate demineralizers and monitor tanks. Additiona11y, simultaneous release of'aundry/shower waste and operation of the waste evaporator is allowed.
A Safety Evaluation was performed. and. it was determined that the modification does not involve any unreviewed safety questions or any Technical Specification Changes.
19
S stem Modification SM 76-09 System Modification SM 76-09, Power Range Current Meter Modification.
A printed circuit card was added and four resistors were changed on the Range Calibrate Assembly.
This modification is being made to facilitate testing and more accurate calibration of the power range detectors.
A Safety Evaluation was performed and it was determined that the modifi-cation does not adversely effect the consequences of'n event nor does any event effect the safety function. A change in Plant Operating Technical Specifications is not necessary.
S stem Modification SM 76-12 System Modification SM 76-12, Air Exhauster Storage Frame. A permanently anchored frame was I
installed for storing the "B" Steam Generator Air Mover.
This modification was performed to provide permanent restraint for the air mover.
A Safety does not Evaluation was performed and it was determined that the modification involve any unreviewed safety questions or any Technical Specification Changes.
S stem Modification SM 76-14 System Modification SM 76-14, Rod Control Capacitor Replacement. Suppression filter capacitors for the Rod Control System power cabinet gripper bridges were replaced..
This modification was,performed since the replacement capacitors'ncreased insulation level prevents excessive capacitor failure.
A Safety Evaluation was performed and does not affect it was determined that the modification the rod withdrawl accident as analyzed. The proposed modifi-cation does not adversely effect the consequences of'an event nor does any event effect the safety function of the modification. A change in the Plant operating Technical Specifications is not necessary.
20
S stem Modification SM 76-15 System Modification SM 76-15, Generator Differential Relay. The zener diode (Z3) across the tripping SCR and the output leads running to station 13A for the main generator differenctia1 relay were removed..
This modification was performed to reduce transient exposure to the SA-1 generator differential relay.
A Safety Evaluation was performed and it was determined that the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated will not be increased by the proposed modification. A change in the Plant Operating Technical Specifi-cations is not necessary.
I S stem Modification SM 76-1 System Modification SM 76-19, Foxboro Bistable Test. The switch rewiring contact in the "neutral" side of the bistable output circuit test switch was removed.
This modification was performed. to reduce the probability of an undetectable failure of the bistable output test circuit.
A Safety Evaluation was performed and it was determined. that the margins of safety during normal operation and anticipated transient conditions, and the adequacy of structures, systems and components provided. for the prevention of accidents and. the consequences of accidents are not significantly changed by this modification. A change in Plant Operating Technical Specifications is not necessary.
S stem Modification SM 76-20 System Modification SM 76-20, Main Feedwater Pump Recixculation Piping Installation. Recix'culation piping was installed and impeller clearances were i,ncreased fax'he "A" and "B" Main Feedwater'umps.
This modificatiqn was performed to pxevent failures of the Main Feedwater Pump ImpelleX s.
A Safety Evaluation gas performed'nd.
does not it was determined that the modification affect Plant safety. A change in Plant Operating Technical Specifications js not necessary.
S stem Modification SM 76-22 System Modification SM 76-22, Installation of Tee and Valve in Laundry and Hot Shower. Tank Transfer Line. A va1ve and. piping were installed. to tie in the neutrqliziang tank to a waste release line.'his
~odifgcation was performed in order to prevent potentially contaminated liquid,ds from being discharged, to the enVironment without being monitored for radar,ogctivity.
A Safety Evaluation was performed. and or unreviewed safety question exists.
itA was change in determined. that no safety hazard Plant Operating Technical Specifications is not necessary.
21-
S stem Modification SM 76-23 System Modification SM 76-23, 25,000 Gallon Holdup Tank Waste Monitoring System. A radiation monitor, strip chart -recorder and miscellaneous piping and controls were installed for the holding tank buried west of the Screen House Building.
The modification was performed to continuously monitor, for radioactivity the waste stream being discharged from the 25,000 ga1lon holding tank.
A Safety Evaluation was performed and will not it was determined that the modification increase the probability of occurrence or the consequences of any accident or malfunction of equipment important to safety, as previously evaluated. A change in Plant Operating Technical Specifications is not necessary.
22
S stem Modification SM 76-25 System Modification SM 76-25, Boric Acid Piping Modification. Piping was replaced. and monitors were installed in the Safety Infection System Suction Piping.
This modification was performed to prevent future leaks and to improve the design of portions of the Safety Infection System Suction Piping.
A Safety Evaluation was performed and it..was determined that the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. A change in Operating Technical Specifications is not necessary.
S ecial Test ST 76-1 Special Test 76-1, Vibration Monitoring Program While Increasing Power from 255 to 405 Nuclear. Vibrations were monitored on the Main Feedwater Piping and valves during power increases for potential degradation;.of systems and equipment.
The test was performed to determine if'scallation of power from 25$ to 40$
presented a potential vibration problem for the Main Feedwater Piping and.
valves.
A Safety Evaluation was perfromed and have an adverse it was determined. that the test will not affect on the safety of the Plant. A change in Plant Operating Technical Specifications is not necessary.
S ecial Test ST 76-2 Special Test ST 76-2, Condenser Back Pressure Control Demonstration. The pressure drop across the L, L-1 baffles installed in the LP-2 Turbine was measured.
The test was performed to determine if the pressure drop across the L, L-1 baffles can 'be reduced. If so, the turbine load can be increased. to the current LP-2 limit accordingly.
A Safety Evaluation was performed and it was determined not have an adverse affect on the safety of the Plant.
that'he test will A change in Plant Operating Technical Specifications is not necessary.
23
CORRECTIVE MAINTENANCE OF SAFETY RELATED UIPMENT 24
MECHANICAL AND ELECTRICAL MAINTENANCE NATURE OF LER OR OUTAGE MALFUNCTION E UIPMENT MAINTENANCE NUMBER CAUSE RESULTS . CORRECTIVE ACTION
'ER Containment Purge Corrective Seats out of Excessive seat, Readjusted the valve seats Supply and Exhaust 76-01 adjustment leakage of. valves Valves 1B Steam Generator Correc'tive LER Tube wall thinn- S team Generator Plugged leaking, tube and 76-08 ing by caus tic tube leakage adjacent tube showing 740%
Shutdown in. corrosion thinning
. '75 Quarterly Lubrication Preventive N/A Scheduled main- Lubricated Rotat- N/A Schedule enance ing Equipment 1B Component, Cooling Preventive N/A Preventive Main- N/A ;Performed inspection of pump Pump tenance Schedule ~'and motor Plant Vent Iodine Corrective N/A Bearing Failure ,
Loss of Sample flo Replaced with spare~and re-Monitor R-10B built. Will increase inspec-tion frequency Waste Gas System Corrective N/A Buildup of corro- Valve stuck open ;Clean valve internals and the Check Valve 1713 sion products on 'disc lapped to the seat seat and plug Charging Valves Preventive N/.A Preventive N/A Repacked valves 392A, 294 and 296 Maintenance lD Traveling Screen reventive N/A Preventive N/A ,Major motor inspection I
Motor Maintenance t.
4160V Bus llB reventive N/A Preventive N/A 'Cleaning & inspection of bus Maintenance MECHANICAL AND ELECTRICAL MAINTENANCE NATURE OF LER OR OUTAGE MALFUNCTION E UIPMENT MAINTENANCE NUMBER CAUSE RESULTS CORRECTIUE ACTION Incore Detector Drive Corrective N/A Internal Inter- Incore path blocked Cleaned and rodded out incore Thimbles B-6, C-3, E-2 ference drive thimbles E-5, G-2) G-6) G-13, H-l, H-10) I"7, J-12, K-4) L"4, L-9 Bus llB Undervoltage Corrective LER N/A Relay failed . Replaced relay with new spare reactor trip relay 76-06 monthly test Reactor Coolant Preventive N/A N/A N/A Replaced filter cartridges Filter 1B Seal injection Preventive N/A N/A N/A Replaced filter cartxidges Filter lA & 1B Steam Preventive N/A Excessive Moist- Mater erosion Installed demister plates at Generators ure Carry over in H.P. Turbine inle t to mo is ture sepax a tor sections of steam generatois 1A & 1B Steam Preventive N/A Phosphate 41 Tubes with wast- 39 tubes with+40% wastage Generators corrosion age greater than- were plugged in 1A steam 40% generator and 2 tubes plugged in 1B steam generatox 1 SS S>WII I ~ ~ ~ ~ II
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MECUANICAL AND ELECTRICAL HAINTENANCE NATURE OF LER OR OUTAGE HALFUNCTION EQUIPNENT 11A INTENANCE NUHBER CAUSE RESULTS CORRECTIVE ACTION Containment iodine Preventive N/A Preventive N/A Inspected and overhauled monitor R-10A Maintenance pump Sample pump Schedule 1B Steam Generator Corrective N/A Undetermined Tube failure Fifteen tubes plugged in the 1B Steam generator in wedge areas Pressurizer manway Corrective N/A Gasket leak Slight leakage Replaced gasket; cleaned seal-at manway during ing surfaces hydro Pressurizer Spray Corrective N/A Packing leak Slight Borate buil Replaced upper packing rings Valve 431A up on stem 33
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MECNANICAL AND ELECTRICAL MAINTENANCE NATURE OF LER OR MALFUNCTION OUTAGE'UMBER E UIPHENT MAINTENANCE CAUSE RESULTS CORRECTIVE ACTION 1B Motor Driven Aux- Corrective Leaking gasket Leakage of bearing Replaced end bearing iliary Feedwater Pump oil cover, gasket Turbine Driven Aux- Corrective N/A Metal .chip in Governor valve Cleaned and inspected iliary Feedwater Pump Relay Valve failed to function hydraulic relay valve properly 35
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MECIIANICAL AND ELECTRICAL MAINTENANCE NATURE OF LER OR OUTAGE MALFUNCTION EQUIPMENT MAINTENANCE NUMBER CAUSE RESULTS CORRECTIVE ACTION Blender System Valve Corrective N/A Valve bonnet Leakage out Replaced valve diaphragm FCW-110B diaphragm fail- valve;bonnet ure Boric acid concentrat Corrective N/A Valve bonnet Restriction Replaced valve diaphragms Valves 1130, 1136 and diaphragms noted in concent<<
1134A stretched rates pump re-circulation line
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41
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HECIIANICAL AND ELECTRICAL MAINTENANCE NATURE OF LER OR OUTAGE MALEUNCT ION E UIFMENT MAINTENANCE NUMBER CAUSE RESUI.TS CORRECTIVE ACTION Plant Fire System Corrective N/A Possible flange Loss of fire Repaired line using welded Supply Main to Auxi:I- leak under trans- system pressure pipe in the inaccessible iary Building. former yard causing frequent area under the transformers makeup to the station fire tank.
1B Main Steam Corrective N/A Gasket leakage Steam leak at head Replaced bonnet gasket Isolation Valve 3516 flange Blender System Corrective N/A Valve Diaphragm Water leakage at Replaced diaphragm Valve 365A Failure Valve bonnet flange Diesel Fire Pump Corrective N/A N/A Tube leak in heat Plugged 1 leaking tube in exchanger engine heat exchanger Safety In)ection Corrective N/A N/A Bonnet Gasket Replaced valve bonnet gaskets Valves 326 C & D Leakage 44
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l NATURE OF LER OR OUTAGE 1 fALFUNCTION E UIPMENT MAINTENANCE NUMBER CAUSE RESULTS CORRECTIVE ACTION i f
. Concentrates Filter PI Corrective N/A Failed -Diaphragm Bonnet Leak Replaced bonnet and diaphragm
.f'I Isolation Valve 1158C 1B Mixed Bed Deminer- Corrective N/A Failed. Diaphragm Bonnet Leak Replaced bonnet and diaphragm lizer Backwash Drain Valve 214 I
MECIIANICAL AND ELECTRICAL MAINTENANCE NATURE OF LER OR OUTAGE MALFUNCTION E UIPMENT MAINTENANCE NUMBER CAUSE RESULTS CORRECTIVE ACTION i
! CVCS Diversion Line Corrective LER N/A Pin hole leak at Repaired leaking weld t
76<<27 socket weld A mixed Bed Deminer- Corrective LER Pin hol'e leak at Repaired weld alizer Piping 76-30 socket weld lA Charging Pump Corrective N/A Worn Packing Leakage at soft Repacked pump packing 1C Service Water Pump Preventive N/A Scheduled Inspec- N/A Performed maj or inspection tion of the pump and motor
- 49
MEC1IANICAL AND ELECTRICAL MAINTENANCE NATURE OF LER OR OUTAGE MALFUNCT ION E UIPMENT MAINTENANCE NUMBER CAUSE RESULTS CORRECTIVE ACTION Control Rod Drive Cables Corrective Outage 76-14 Open circuit Drive Failed Replaced complete plug Drive F-12 LER 76-29 and cable assembly 50
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- 52
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MECllANICAL AND ELECTRICAL MAINTENANCE NATURE OF LER OR OUTAGE- MALFUNCTION E UIPMEHT MAINTENANCE NUMBER CAUSE RESULTS CORRECTIVE ACTION 1A incore neutron Correcgive LER Failed detector Loss of 1A system Replaced detector and cable detector System 76-14 lA & 1B Diesel Preventive N/A Lake sediment Increased dp Cleaned tubes Generators in service water across the lube oil and )acket water coolers Preventive N/A Gasket Leak Borate deposit Replaced gaskets and inspected Pressurizer Spray valves 431A & 431B in 431B at flange of 431B internals Scheduled Maintenance preventive N/A Scheduled N/A Lubricated rotating equip-Quarter+ Lubrication ment Schedule Maintenance 55
MECHANICAL AND ELECTRICAL MAINTENANCE "
NATURE OF LER OR OUTAGE MALFUNCTION E UIPMENT MAINTENANCE NUMBER CAUSE RESULTS CORRECTIVE ACTION A Steam Generator Cut gasket seat North Secondary Hand surface. and excess- Used 'Purmanite process to he3p Hole 'orrective N/A Gasket Leak ive secondary seal gasket seat area system leakage.
1A Charging Pump Corrective N/A Packing Wear Excessive leakage Replaced stuffing box as sem-at stuffing box blies with repacked units.
1B Component Cooling Corrective LER-. Coupling nut Pump to motor Repaired coupling and Pump 76-02 became loose coupling became realigned pump loose & noisy Radiation Monitor .Preventive N/A Preventive . N/A Replaced with spare unit and sample pump for Plant Maintenance rebuilt.
Vent R-13,14 Monitors Schedule Auto Makeup Blender Corrective N/A Diaphragm Bonnet flange Replaced bonnet diaphragm valve FCV-llOC failure leakage B Loop Sample Valve Corrective N/A Bonnet gaske't Release of primary . Replaced bonnet gasket and 955 leak coolant to B loop stem and plug assembly area B Main Steam Line Corrective N/A Second accumu- Leakage of fluid Replaced accumulator unit hydraulic restraint ulator cylinder past accumulator MS-147 piston seal INSTRUMENT AND CONTROL MAINTENANCE NATURE OP LER OR OUTAGE MALFUNCTION E UIPMENT MAINTENANCE NUMBER CAUSE RESULTS CORRECTIVE ACTION Safeguard rack relays Corrective . LER 76-07 Improper factory Relays failed Replaced relays V1 and Cl, A logic train adjustment
'I Source Range detector Corrective N/A Detector faulty Noise in detector Replaced detector N-32 Power range detector N-42 Corrective LER 76-04 High Voltage Suppl Detector failed, Replaced power supply Capacitor failed turbine runback capacitor Boric acid tank level Corrective N/A Amplifier faulty LT-106 unstable Replaced LT-106 during testing amplifier Steam Flow channel 465 Corrective N/A Bistable Setpoint unstable FC465A Repaired bistable failed Control room ventilation Corrective LER 76-11 . Piston mounted System failed flow Reinstalled piston damper actuator improperly test correctly Reactor Coolant Flow Corrective N/A FT411 amplifier Channel spiking Replaced FT411 amplifier channel 411 faulty . hi@h Power Range Detector N-44 Corrective LER 76-14 High voltage power Detector failed; Replaced power supply supply failed turbine runback Primary Loop RTD's 402A, Corrective N/A Ohmic value shifted Replaced RTD's 406A, 407A Control rods G-5 6 G-9 Corrective LER 76-13 Possible printed Dropped rods G-5, Replaced circuit cards circuit cards G-9 failure Accumulator Pressure Channel 941 Corrective N/A plifier faulty Transmitter output Replaced amplifier unstable INSTRUMENT AND CONTROL MAINTENANCE NATURE OF LER OR OUTAGE MALFUNCTION E UIPMENT MAINTENANCE NUMBER CAUSE RESU1 TS CORRECTIVE ACTION Reactor trip logic Corrective N/A Rela'y failed safe Logic train failed Replaced B-train relay 21 ielay . surveillance test Area Monitor R-6 Corrective N/A Transformer in Area monitor failed Repaired drawer drawer Reactor make up water Corrective N/A Integrator failed Spurious operation Replaced integrator flow channel III RMS channel R-15 Corrective N/A Detector failed Channel failed Replaced detector RMS channel R-15 Corrective N/A Detector and Channel failed Replaced detector and preamplifier faile preamplifier RMS channel R-19 Corrective . N/A Steam generator Detector chamber. Cleaned detector chamber tube leak contaminated r Containment spray addi- Corrective LER 76-26 Cold solder joint failed open Repaired joint tive valve 836A controller
'alve in wiring, failed capacitors-Replaced capacitor Boric acid tank 'levels Corrective N'/A Partial plug in LT-106 drifted Performed PT-21 (weekly sensing line high routine surveillance) reamed out all sensing lines.
Boric acid tank levels Corrective N/A artial plug in LT-106 drifted Performed PT-21 (weekly sensing ling high routine~surveillance) reamed .out all sensing lines INSTRUMENT AND CONTROL MAINTENANCE NATURE OF LER OR OUTAGE MALFUNCTION E VIPMENT HAINTENANCE NUMBER CAUS E RESULTS CORRECTIVE ACTION Boric acid tank levels Corrective N/A Partial plug in LT-102 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A Partial plug in LT-171 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A artial plug in LT-172 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A Partial plug in LT-106 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines.
Boric acid tank levels Corrective N/A artial plug in LT-106 and LT-171 Performed PT-21 (weekly sensing line drifted high routine surveillance) reamed out all sensing lines'Boric acid tank levels Corrective N/A artial plug in LT-172 drifted high Performed PT-21 (weekly ensing line routine .surveillance) reamed.out all sensing lines Boric acid tank levels Corrective N/A artial plug in LT-102 and LT-172 Performed PT-21 (weekly
-ensing line drifted high routine surveillance) reamed out all sensing lines INSTRUHENT AND CONTROL MAINTENANCE NATURE OF LER OR OUTAGE MALFUNCTION E UIPMENT . HAINTENANCE NUHBER CAUSE RLSULTS CORRECTIVE ACTION Boric acid tank levels Corrective N/A Partial plug in LT-172 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A Partial plug in LT-106 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A Partial plug in LT-172 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A Partial plug in LT-106 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A, artial plug in LT-102 drifted high Performed PT-21 (weekly ensing line routine.survgillance) reamed out all sensing lines Boric acid tank levels orrective N/A artial plug in LT-171 drifted high Performed PT-21 (weekly ensing line routine. surveillance) reamed out all sensing lines INSTRUMENT AND CONTROL MAINTENANCE NATURE OF LER OR OUTAGE 'ALFUNCTION E VIPMENT MAINTENANCE NUMBER CAUSE RESULTS CORRFCTIVF ACTION Boric acid tank levels Corrective N/A Partial plug of LT-172 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A Partial plug of LT-102 drifted high Performed PT-21 (weekly sensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A artial plug of LT-171 drifted high Performed PT-21 (weekly ensing line routine surveillance) reamed out all sensing lines Boric acid tank levels Corrective N/A artial'plug of LT-102 drifted high Performed PT-21 (weekly ensing line routine surveillance) reamed out all sensing lines LICENSEE EVENT REPORTS During the reporting period of January 1, 1976 through December 31, 1976 a total of 30 events, classified as Reportable Occurrences, took place requiring submittal of Licensee Event Reports to the NRC. The following is a compilation of narrative descriptions of those events:
R.O. 76-01, Excessive Leaka e throu h Containment Vessel Pur e 1-11-76 Su 1 and Exhaust Valves During surveillance testing prior to reactor heatup, the containment vessel purge supply and exhaust valves (Henry Pratt 48" Butterfly Valves) were found to leak excessively., The rubber seat adjustment shoes were realigned and subsequent testing verified return of leak rate to within acceptable limits. This event is similar to events which occurred on 2-14-74 and 11-19-74. The cause of the excessive leakage may be low ambient temperature during reactor shutdown, when these events have occurred; the valves have not failed testing during power operation. An engineering evaluation of the valve seat material was performed and determined that seat material should be adequate. Vendor representative inspected valves and indicated that the leakage may have been caused simply by maladjustment of adjusting shqes.
This event has not recurred since reported under Reportable Occurrence 76-01. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-02, "B" Com onent Coolin Pum Cou lin Failure 1-12-76 During normal rounds while the reactor was at cold shut-down, an operator noticed that the "B" component cooling pump was running noisily; both pumps were running at the time. The "B" pump was stopped and an inspection by station mechanics disclosed that the pump-to-motor coupling (Thomas Coupling Division Dry Type Model DBZ226 for 2 1/8" to 2 3/8" shaft) was damaged, with two of the coupling bolts failed. It appeared that two of the bolts initially became loose which then led to further deterioration of the coupling. After repairing the coupling, the motor was realigned and the pump returned to service. This was considered to be a random failure and no further action to prevent recurrence was necessary.
No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-03$ Reactor Coolant Drain Tank Pum Suction Flan e Leaka e 1-15-76 Durin H drostatic Test During routine 105 psig hydrostatic test of residual heat removal suction piping from containment sump B to reactor coolant drain tank pumps, the "A" pump suction flange was found to be leaking at a rate of approximately 2.85 gal/hr. It was determined that maintenance personnel"who had previously done some work on the pump had tightened the flange bolts, but not tight enough to withstand the test pressure. The flange bolts were subsequently properly tightened and the leakage reduced to zero. To prevent recurrence of this event, the procedure used to perform maintenance on the pump was modified to assure proper tightening of the..'flange. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-04, Failure of, Power Ran e Channel N42 Hi h Volta e Power 1-29-76 ~Su i~1 During load decrease for shutdown, both control fuses blew for power range channel N42. Resultant de-energizing (trip mode) of the bistable relay drivers automatically initiated a turbine runback to 30% power. Channels N41, N43 and N44 continued to function properly. A capacitor in the auxiliary power supply board for the high voltage power supply (Power Design Company 300 V to 1500 V positive polarity 0-15 ma Model UPMDX54), found to be open, was replaced. Pith the replacement capacitor, the power su'pply was tested satisfactorily and returned to service. This occurrence was deemed a random component failure, and hence, no action to prevent recurrence was necessary. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-05, Abnormal De radation of Primar Containment Com onents 2-8-76 Discovered Durin Containment Inte rated Leak Rate Test During performance of the containment integrated leak rate test, while holding at a pressure of 35 psig, it became apparent that there was excessive leakage through several containment components as follows: containment purge supply valve inside containment, containment purge exhaust valves, MOV's,813 and 814, check valve 1713. The components were repaired and leakage brought back within allowable limits to permit continuation of the leak rate test. No danger to the health and welfare of the general public was pos'ed by this occurrence.
R.O. 76-06, Failure of Undervolta e Rela on Bus 11B 2-6-76 During surveillance testing of undervoltage relays, U.V. device 274/11B (Westinghouse type SV 75-150V) failed to drop to de-energized position after under-voltage signal was simulated. Redundant device 273/11B operated properly. All four U.V. relays associated with busses llA and llB were field stripped, the plungers cleaned, and retested sati'sfactorily.
Three previous occurrences of this type were reported on 6/8/73, 12/13/74 and 8/5/75. In order to prevent recurrences of this type event, new style U.V. relays were installed under the direction of System Modifica-tion SM 75-29. Since this modification, there has been no recurrence of this event. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-07, Failure of Safe uards Train "A" MG-6 'Rela s to 2-7-76 0 crate Pro erl During valve alignment verification following manual containment -isolation (CI) and containment ventilation isolation (CVI) in.preparation for containment ILRT, Sump A discharge AOV had not closed. Investigation revealed that MG-6 relays (Westinghouse type MG-6 relays, style 289B363All, 125 VDC) in the "A" safe-guards train for CI and CVI would not latch and would drop back on release of pushbutton. Train "B" functioned properly. There has been no previous failure of this type. Inspection of the relays disclosed improper latch screw factory adjustment. The relays were replaced with one from spare parts stock which were adjusted, tested, installed and retested satisfactorily. No further action is needed to prevent recurrence, but the latch screw adjustment will be checked during the 1977 refueling outage. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-08, . Steam Generator Tube Wasta e Greater Than 40/ of 2-27-76 Wall Thickness During planned eddy current examination of steam generator (S/G).tubes, 39 tubes in the "A" S/G inlet and 2 tubes in';the "B" S/G inlet showed defects above the 40/ limit. These tubes, and those corresponding on the outlet side, were plugged explosively. Local
.corrosion of tube OD surface has been caused by residual phosphates from phosphate water treatment. In November, 1975, an~all-volatile water treatment program was started which is expected to eliminate further aggravation of this problem. Secondary-side S/G tube lancing has been utilized to remove the residual phosphates. The all-volatile water treatment program, plus the finalization of the full-flow condensate polishing demineralizer system, should help eliminate tube wastage problems.
No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-09, Premature Tri in of "A" Diesel Generator Breaker 2-27-76 to Bus 18-During preventive maintenance circuit breaker testing while in refueling mode, the "A" Diesel Generator breaker to Bus 18 tripped prematurely. Investigation revealed that the "C" phase overcurrent trip relay (Westinghouse style 24Y4712BA8, 1600 amp 60 Hz) actuated 6 seconds earlier than minimum time allowed by station procedures. An identical relay was bench tested, installed, and field tested satisfactorily.
Station Electricians dismantled the defective relav, cleaned some internal contamination from the relay and found that the relay now operated properly. This appears to be an isolated instance and not a generic problem with relays of this type. No dangers to the health and welfare of the general public was posed by this occurrence.
R.O. 76-10, Failure of CV Pur e Su 1 Valves to Close Pro erl 2-28-76 During surveillance testing of the air volume between the purge supply valves, excessive leakage was observed from the valve inside the containment. Leakage through the redundant valve was minimal. Operation of the leaking valve revealed erratic valve motion and the valve would not close fully each time. The actuator was dismantled and it was found that excessive friction between the spring cylinder 'assembly piston shaft and its guide bushing resulted from the shaft being excessively scored. The shaft.:.was smoothed, polished and dry lubricated, and the valve subsequently-retested satisfactorily. The vendor has been contacted concerning recommended spare parts, and an order has been placed for those parts. If received in time, the parts will be used to rebuild the actuator during the Spring 1977 refueling outage. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-11 Control Room Ventilation Dam er Mounted Im ro er1 3-8-76 During flow test of control room ventilation system in preparation for charcoal filter efficiency test there was zero flow indicated for cleanup filtering path, designed for 25% total system flow. Investigation revealed the damper. drive piston was mounted improperly during construction, the deficiency had not been identified by the commercial tester during preoperational testing, and following corrective action had not taken place. Previous testing of this system had included observance of damper operation, which assumed that actuatora had been properly installed. The actuator was reinstalled properly. Presently existing Q.A.
Program requirements prevent this from happening in future modifications at the plant, by assuring their acceptability prior to turnover for use. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-12, Personnel Air Lock Door Handwheel Shaft Packin 3-31-76 ~Leaka e During routine 60 psig pressure decay test of personnel hatch air lock, excessive leakage was observed from an outer door handwheel shaft packing. Leakage from'an air lock shaft packing was previously reported 3/30/70.
Investigation revealed that the packing nut had backed off due to repeated usage of the air lock during the refueling outage. The packing nut was threaded back in place and tightened, and the airlock volume retested satisfactorily. Action to prevent recurrence was two-fold. Procedures for testing the volume includes a prerequisite to verify proper installation of the packing gland nut after extended usage of the airlock. A modifi-cation is to be implemented during the Spring 1977;:.;
refueling which would install "keepers" on the packing nut to prevent it from backing off. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-13, Dro ed Rods G-5 and G-9 .
4-16-76 During a power escalation, a turbine runback occurred as a result of 2 control rods, G-5 end G-9 in Bank B, dropping. With unit at hot shutdown, an alarm indicated stationary "A" regulation failure which in turn indicates 4 printed circuit cards as the possible cause. The urgent failure alarm was reset to operate the control rods to find the cause and the alarm stayed clear. The 4 cards were replaced, and the rod operated properly.
The cards were sent to Westinghouse for investigation which did not disclose any defects. No further corrective action was considered necessary at this time. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-14, Power Ran e Channel N44 Hi h Volta e Power Su 1 Failure 4-12-76 During steady state power operation N44 power range high
'oltage power supply (Power Design Inc. 300 V to 1500 V positive polarity 0-15 ma Model UPMD-X54) failed to 57 VDC initiating a false dropped rod bistable actuation, resulting in a turbine runback. - Three. redundant channels continued-'to function properly. The'faulty'unit was removed and during troubleshooting in the shop to function properly "again. It was it started returned to the manufacturer for further investigation.
Meanwhile, the unit had been replaced with one from spare parts stock, tested in place and the channel returned to service. The manufacturer inspected the power supply, found the defective part and replaced it.
They could not explain the intermittent operation noted during troubleshooting. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-15, Steam Generator (S/G) Tube Leak 4-24-76 During normal power operation, the air ejector and blowdown radiation monitors alarmed. Analysis of the "B" S/G blowdown indicated a calculated leak rate of 0.06 gpm. Unit was shutdown and subsequent eddy current (EC) testing verified the leaking tube. Twelve other tubes in ad)acent areas were found to have similar ID type indications at about 2 1/2 to 3 inches above the tubesheet. All 13 tubes were explosively plugged. See R.O. 76-08 for corrective actions taken to prevent recurrence. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-16, Ino erabilit of Rod Position Deviation Channel 5-20-76 During a check of the computer rod position program, the rod position deviation channel was found to be inoperable.
A hand log was started as specified by the plant Technical Specifications. Computer technician re-initialized the program and the rod position program check was completed satisfactorily. The channel was deemed operable and hand logs terminated. Operator verification of Tech.
Specs. related programs in the computer will be continued, and a Technical Specification change is being considered to eliminate further reports of this event when a suitable alternative is in effect. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-17, "A" Accumulator Level Below Technical S ecifications 6-18-76 Limit During the performance of S-16.11, "Monitoring and Enrichment of Accumulator Boron Concentration," the "A" Accumulator drain valve was opened with the controls at the rear of the control board. The operator then proceeded to the front of the control board to monitor the level on indicators reading more conservative than those on the back of the board, and noted that the level had dropped to 46X. The level was returned to within Technical Specifications limits within 11 minutes. This occurrence was caused by the fact that S-16.1C did not provide precautions that the operator remain at the valve controls during the draining operation. The procedure has since been changed to incorporate these requirements.
No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-18, Dro ed Rods G-5 and G-9 7-4-76 During steady state power operation turbine runback occurred, caused by dropping of two,control rods, G-5 and G-9, in Bank "B". These two rods had dropped on three previous occasions, two of which involved water leaks over the rod control cabinet. The alarm indication was a stationary "A" regulation failure.
The reactor was brought to hot shutdown, the alarm was reset and could not be made to recur. The unit was brought back to power without any further control rod problems. Based on similar incidents at other Westing-house facilities, certain electronic components in the rod control circuitry have been replaced with upgraded components, at Westinghouse's suggestion (This action was taken subsequent to a similar incident, R.O. 76-21 reported later in this section). These replacements have included changing the power supply fuses and sending the ones which had been in service back to Westinghouse for X-ray examination. The examination revealed no flaws which could have caused the failure. Investigations are presently underway for further modifications to the cabinet wiring as suggested by Westinghouse. The cabinet is Westinghouse model 916E646G03. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-19, Unmonitored Release of Radioactive Water 7-8-76 During reclaiming of steam generator blowdown water by determined that, from 6-17-76 to 7-6-76, ll it processing in the main water treatment plant, was releases were made from the neutra)izing tank with activity ranging from less )han 2.25 x 10 pCi/cc to less than 3.96 x 10 >Ci/cc. Although the activity was determined by sampling prior to each release, the releases were made without the use of a continuous gross activity monitor. The recently installed blowdown recovery system processes blowdown water as described above. The demineralizer resins concentrate the small amount of activity, the resins are regenerated and the spent regenerants are collected in the neutralizing tank. This tank is then sampled and the contents analyzed for pH and activity prior to release. The maximum concentration in the cagal due to release of this effluent was 2.7 x 10 pCi/cc which posed no hazard to the public.
To prevent unmonitored release from recurring a temporary line was run to allow releasing the neutralizing tank through radiation monitor R-18. Subsequently, a radiation monitor was installed to monitor the gross activity of the effluent being released from the neutralizing tank. This monitor installation is described under SM 76-23.
R.O.76-207 Failure of 0 erator to Start "B" Diesel Generator 7-9-76 When "A" Diesel Generator Was Ino erable Upon completion of the monthly surveillance test of the "A"
.Diesel Generator (DG), the operator opened Bus 14 breaker from the DG and received a DG supply breaker overcurrent trip alarm. Knowing this alarm was invalid, he tried to reset the breaker electrically with the control switch; the alarm did not clear. The "B" DG was not started. The Electrician Foreman was notified and he immediately went to the breaker location to investigate, and he reset the breaker manually. It was determined that wear had occurred on the shunt trip alarm switch linkage preventing the breaker from resetting electrically. With the "B" DG running the shunt trip assembly spring tension was increased and the breaker operated successfully several times. The breaker was then returned to service and both DG's returned to standby. At PORC's direction, the applicable alarm response cards were updated to assure that the redundant DG is placed in service immediately upon notification that one DG or its bus ties is inoperable. No danger to the health and welfare of the general public was posed by this occurrence.
R.O; 76-21, Dro ed Rods G-3 and G-ll 8-4-76 During steady state power operation, a turbine runback occurred, caused by control rods G-3 and G-11 in Bank "D" partially dropping into the core. This is ."the first occurrence concerning these two rods, but the fifth such occurrence associated with the 2 BD power cabinet.
The reactor was brought to hot shutdown, the urgent failure alarm was reset and could not be made to recur.
Wires and connections in the Logic Cabinet and 2 BD Power Cabinet were checked with no abnormalities found.
See R.O. 76-18 for further action taken. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-22, Chan e in Westin house Safet ,Anal sis 8-5-76 Westinghouse notified RG&E of a change in the assumptions to be used by Westinghouse in the safety analysis methods for the LOCA. The original analysis assumed the upper head region fluid temperature was at T-cold. It has been shown that for Westinghouse PWR's the actual upper head fluid temperature was somewhat greater. Westinghouse has reanalyzed the LOCA with T-cold equal to T-hot on a generic basis and found that the peak cladding temperature increases by 40 F. Westinghouse has advised RG&E that other accident analyses for Ginna are not affected. No further corrective action is necessary, and no danger to the health and welfare was posed by this occurrence.
R.O. 76-23$ Use of Im ro er Materials in Safet In ection Test Line 8-19-76 During review by contractor's Q.C. personnel of work documentation package, it was determined that piping installed in S.I. Test Line during Spring 1976 refueling shutdown was 3/4" Sch 40 S stainless steel rather than 3/4" Sch 80 S stainless steel specified. The PORC has determined that with the Sch 40 pipe section S.I. pumps would have been capable of fulfilling their intended function. With unit shutdown contractor immediately replaced piping with correct material. This occcurrence was caused by inadequate administrative controls to assure proper installation of materials. Contractor procedures and plant procedures have been changed to provide more stringent controls for the verification of acceptability of installed modifications prior to plant operation.
No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-24, Leaks in Safet In ection Suction Pi in 10-8-76 During normal operations an auxiliary operator found an accumulation of borated water near a valve in the Safety Injection System piping between the boric acid storage tanks and the SIS pumps. Further investigation revealed a leak in a section of 8" schedule 10S stainless steel piping between valves 826A and 826B. The reactor was immediately taken to cold shutdown in an orderly manner and further investigation initiated to determine the cause of the leakage. Subsequent examination of all welds, and piping in the safety injection suction piping which was heat traced, revealed five other sections of piping which required replacement. Examination method consisted of ultrasonic and liquid penetrant examination of welds plus hydrostatic testing of the piping to 265 psig. One section of piping which was inaccessible due to high radiation was subjected to the hydrostatic test only, as was one weld which was in a concrete wall penetration. The repair to this system consisted of the following: replacement of defective schedule 10 piping with schedule 40 piping, installation of additional seismic pipe supports, installation of two vents in the piping to assure complete purging of all air from the system, and installation by Westinghouse of acoustic leak detectors near the inaccessible welds to provide early warning of leaks. The mechanism of the piping deteriora-tion was identified as transgranular chloride stress corrosion, caused most probably by the concurrence of residual construction contamination of the piping ID, stagnation of borated water in this section of line thus preventing periodic cleaning of the system, and continuous supply of heat to the piping by the heat tracing system. Further corrective action to be performed during the 1977 refueling outage consists of: system hydrostatic testing of the safety injection and containment spray systems, random ultrasonic examination of selected welds in the SIS pump suction line from the BAS Tanks, and rerouting of the line as it passes through the high radiation areas to make that section of piping accessible for Inservice Inspection.
No danger to the health and welfare of the general public was posed by this occurrence. The repairs made to the piping are described in System Modification SM 76-25.
R.Q. 76-25, Leak in RHR Return Line 10-10-76 During operation of the RHR system while shutdown, an operator noticed a puddle of water on the floor and identified the source as a leak at a weldolet to nipple on a 3/4" vent on the RHR return line. This is similar to Reportable Occurrence 74-5. Station personnel replaced the nipple and rewelded in accordance with approved procedures. The cause of this occurrence was improper fitup of the nipple to the socket during original construction which resulted in stresses induced in the weld. Welding procedures presently in use address proper fitup to assure this does not recur. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-26$ Failure of NaOH Valve 836A Controller 11-16-76 During routine control board surveillance of status of safeguards valves, it was noticed that the controller output for CV spray. NaOH valve 836A was less than 10ma.
The redundant controller was operable. The controller is a Foxboro model 62H-2E (special), 118V60Hz, 14W, 10-50 ma output proportional controller. Controller YIC 836A was removed to replace an overheated bias board.
Two capacitors consisting of 2 banks of 3 capacitors each had broken down and a resistor had opened due to acid leakage from the failed capacitors. The next day after return of the controller to service, it was again found inoperable. Investigation revealed a cold solder joint on a wire to the proportional band .potentiometer. After resoldering this joint, the controller was bench tested for an hour before reinstalling. The controller action was checked in MANUAL and AUTO modes and found to respond properly.'his failure is considered to be an isolated component malfunction. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-27, Leak in Letdown Divert Line to CVCS Holdu Tanks 12-16-76 During normal power operation, operator noticed leaking water from overhead in "Valve Alley" leading to the Gas Decay Tank room. Further investigation identified a pinhole leak in a socket weld on the letdown divert line.to the CVCS Holdup Tanks. The weld was built up in the area of the leak and a visual*.examination performed up return of system to service to verify the integrity of the repair. This section of line is to be replaced during the Spring 1977 Refueling shutdown. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-28 Dro ed Rod F-12 Bank "A" 12-17-76 During steady state power operation at a reduced load, rod F-12 in Bank "A" dropped into the core. Operators, with the assistance of I and C Technicians, attempted to retrieve rod unsuccessfully. Investigation showed the problem to be an open stationary hold coil circuitry, unrelated to the three dropped rod occurrences reported earlier this year. The unit was shutdown due to excessive power tilt (see R.O. 76-29) and inspections of the mechanism coils and cable connectors was made. Inspections identified the problem to be an open circuit point at the reactor head assembly plug connection. The plug connection consisted of a Grouse Hinds B-pin male plug and receptacle Cat. No.
RPE-133-153-SO-INT-107 and RPE-233-014-PO-INT-107, with pins for $!8 and f/10 wire. The faulty equipment was replaced, readings were taken at the rod control cabinets to verify the integrity of the mechanism power supply, and the rod was exercised to verify operability. The unit was .subsequently returned to power. No further corrective action to prevent recurrence was considered necessary. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-29, adrant to Avera e Power Tilt Ratio Exceeded 1.12 12-17-76 During a dropped rod condition (see R.O. 76-28) core quadrant to average power tilt ratio calculation showed 1.125, in excess of Technical Specifications limit 1.12.
Unit was placed in:.hot shutdown and returned to service upon repair of the faulty control rod mechanism described in R.O. 76-28. No danger to the health and welfare of the general public was posed by this occurrence.
R.O. 76-30, Leak in Letdown Line to Mixed Bed Demineralizer 12-17-76 During normal rounds an operator discovered liquid leaking from the "A" mixed bed demineralizer inlet piping. This was determined to be a pinhole leak in a socket weld between a pipe and an elbow. The hole was ground out and repaired using proper welding proceduere. The piping was visually:inspected for leakage after return to service.
No further corrective action was considered necessary.
No danger to the health and welfare of the general public was posed by this occurrence.
OTHER EVENTS of INTEREST FUEL PERFORMANCE Primary system sampling from January 1, 1976 through December 31, 1976 showed no increases indicative of significant fuel failures.
Periodic flux maps data indicated normal flux distribution in the core, and confirmed that hot channel factors were below design limits.
Predicted and measured boron concentration versus burn-up of core were within 15 ppm boron, well below the limit of 1% reactivity anomolies.
During this reporting period, the reactor was refueled. Cycle V was terminated January 29, 1976. Thirty-six new fuel assemblies were introduced in the core. Cycle VI operation began April 7, 1976.
For major events of interest for the year 1976, see'Highlights" section of this report. Major events during Cycle Vl include the following:
- 1. 4-18-76 Water in EH oil 2 Days
- 2. 4-24-76 B Steam Generator Tube Leak 13 Days
- 3. 8-4<<76 Dropped Two Control Rods 2 Days
- 4. 8-7-76 Turbine Blade Failure 28 Days
- 5. 10-8-76 Boric Acid Line Leak 21 Days Fuel sipping was performed during the refueling shutdown of 1976.
A total of forty-two fuel assemblies were sippod, .two of which were classified as l~akers and two as possible,leakers.
FUEL DATA The following outline addresses topics in which the Nuclear Regulatory Commission has expressed interest.
Fuel Vendor: Westinghouse Electric Corp.
Exxon Nuclear Co.
Babcock and Wilcox Co.
Fuel Loading Data: Core consists of 121 fuel assemblies.
Each fuel assembly consists of 14 x 14 array of 179 fuel rods, 16 RCC guide thimbles, and one instrument tube.
Core Loading Map of Cycle VE: see Figure 1 Rochester Gas & Electric Corp.
Ginna Station CYCLE VI Core Loading Pattern a c.
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8 C 3 X X = Fuel Assembly Ident ific'a t ion F.IGURE $ , >?
I
UNIT 81 STEAM GENERATOR INSPECTION(S) 1976 During this period Eddy Current testing was performed three times.
- .A 100/ Eddy Current Inspection was made in February and on the other two occasion (January and April), the testing was in localized areas
- surrounding a known leaking tube. Both occasions were involving the B Steam Generator. The nature of the tube failure in April, being unique, made it appear prudent to examine all tubes in the wedge areas.
The failure was due to inside wall defects and was located in one of the "hard wedge" areas. All the tubes of the B Generator were in-spected to the first support. Both inlet and outlet sides of the B Generator were examined, as well as those in the A Generator inlet side.
Tubes examined in the area of the leaking tube in January 1976 (200"tubes) were tested at 400 KHZ to the first support. Besides the leaking tube, one additional tube showed wastage >40/.
Eddy Current Test Results Steam Generator B (Inlet)
Site R. E. Ginna - Unit 81 Test Frequency 400 KHZ Dated January 1976 ROM COLVHH TYPE OF INDICATION LOCATION 40 68 0 D Defect 4" below Tube Sheet 39 : 69 0 D Defect 1" above Tube Sheet The tubes tested in the B Generator Xn the area of the leaking tube (April 1976) exhibited inside wall defects and the examination scope was extended (when brought to our attention) to include all wedge areas in both inlet and outlet sides. Further indications of inside wall de-fects promted a decision to open the A Generator to begin a study of the tubes in wedge areas of"that Generator. In all, there were 12 tubes in the B Generator inlet that exhibited inside wall defects (exclusive of the faild tube - also inside defect), one tube found to have >40/
outside wall wastage, and one tube .plugged by improper placement of a repair thimble. No tubes in the B Generator outlet and no tubes in the A Generator inlet were found to have defects.
Current Test Results Steam Generator B (Inlet)
Site R. E. Ginna - 'ddy Unit /31 Test Frequency 400 KHZ Dated April 1976 ROW "'COLUMN TYPE OF INDICATION LOCATION
'l 92 1 D Defect 4" above Tube Sheet 91 ID Defect + 4" above Tube Sheet TYPE OF INDICATION IQCATION 10 91 I D Defect - 4" above Tube Sheet 91 I D Defect 4" above Tube Sheet 12 91 I D Defect ~" 4" above Tube Sheet 90 I D Defect - 4" above Tube Sheet 14 90 I D Defect 4" above Tube Sheet 15 90 89 I D Defect .- 4" above Tube Sheet 16 89 I D Defect - 4" above Tube Sheet 17 89 I D Defect -'" above Tube Sheet 30 0 D Defect at Tube Sheet 31 15 I D Defect 4" above Tribe Sheet 32 15 I D Defect - 4" above Tribe Sheet 33 15 I D Defect - 4" above Tube Sheet
>>79-
In February, 1976, decision had been made to perform 100/ Eddy Current inspection of both Generators (both inlet and outlet sides). This type of m~dnation would supplement information for baseline data and would also assure that we would include those tubes in the areas of greatest concern. Testing was perforaed at 400 KHZ (all tubes to at least the first support plus 264 in the A Generator inlet over the U-bend, 218 tubes in the B Generator over the U-bend, and an additional 864 tubes in each Generator acaazined.to the 6th support). Additional testing was performed at 25 KHZ (to the first support) for sludge profiles in both Generators.
The same probe was used for both 400 KHZ and 25 KHZ testing with brush recorder speed at 5mn/sec and 25 nm/sec respectively. All data was recorded on magnetic tape and recorder strip. charts with data recorded during each shift.. A dent evaluation study was also, performed in both Generators. The dents were reported. < 5 mils, probably a result of original fabrication, and were considered to be of no iaaediate concern.
The total number of tubes inspected at 400 KHZ was 3192 in the A Generator and 3247 in the B Generator. They are categorized as follows:
A Generator B Generator Inlet Outlet Inlet Outlet Required Sample 437 201 200 Bcamined First Time 1019 2627 1315 2731 Baseline without E.C. 897 2312 . 1115 1630 Indications Baseline with E.C.
Indications 122 315 (Dent Only) ~ (Dent Only)* '1 20020/)*
1001 (all < 20/)*
The extra auger of tubes tested provides a history for ccmparison in future inspections.
Thirty-nine tubes showed defects greater than 4F/ indicating only a 5K growth rate over the past year, considering. tubes which had indicated > 2F/ over for the previous two inspections.
Eddy Current Test Results Site R. E. Ginna Qu.t g1 Steam Generator A (Inlet)
Test Frequency 400 KHZ Date February 1976
- OF THOSE EXAMINED THE FIRST TIME TYPE OF INDICATION 25 0 D Defect 1" above Tube Sheet 15 27 0 D Defect 2" above TUbe Sheet 15 28 0 D Defect 2" above Tube Sheet 18 0 D Defect 2" above Tube Sheet 20 0 D Defect 1" above TUbe Sheet 0 D Defect 2" above Tube Sheet 20 52 0 D Defect 3P,'bove Tube Sheet 60 0 D Defect 2" above Tube Sheet 14 68 0 D Defect 2" above Tube Sheet 29 0 D Defect 1" above Tube Sheet 19 30 0 D Defect 2" above Tube Sheet 19 31 0 D Defect 4" above Tube Sheet 20 0 D Defect 3", above Tube Sheet 21 31 0 D Defect 2" above Tube Sheet 24 0 D Defect 1>" above TUbe Sheet 25 36 0 D Defect lg'bove Tube Sheet 46 0 D Defect 2g'bave Tube Sheet 18 42 0 D Defect 2" above Tube Sheet 29 0 D Defect >" above Tube Sheet 29 47 0 D Defect Q" above Tube Sheet 29 0 D Defect Q" above Tube Sheet 29 0 D Defect 1" above Tube Sheet 23 0 D Defect 1" above Tube Sheet 26 45 0 D Defect 1" above Tube Sheet 28 0 D Defect 2" above Tube Sheet 27 0 D Defect 2" above Tube Sheet 27 0 D Defect 2" above Tube Sheet
COLURE TYPE OF INDICATION 28 47 0 D Defect 2" above Tube Sheet 27 48 0 D Defect 2" above Tube Sheet 68 0 D Defect 2" above Tube Sheet 10 68 0 D Defect 2" above Tube Sheet 12 68 0 D Defect 2" above Tube Sheet 12 65 0 D Defect- >" above Tube Sheet 75 0 D Defect:=' above TUbe Sheet 73 D Defect 1" above TUbe Sheet 71 0 D Defect 2" above Tube Sheet 70 0 D Defect 2" above Tube Sheet 70 0 D Defect 1" above Tube Sheet 70 0 D Defect 1" above Tube Sheet Eddy Current Test Results Site R. E. Ginna Unit gl Steam Generator 8 (Inlet)
Test Frequency 400 KHZ Date February 1976 TYPE OF INDICATION IDCATION 18 28 0 D Defect lg'bove Tube Sheet 51 0 D Defect 4" above Tube Sheet I
SUMMARY
OF CONTAINMENT INTEGRATED LEAK RATE TEST A Reactor Containment Building Integrated Leak Rate Test (Type A Test) was performed during February, 1976.
The maximum aU.owable leakage rate (Lt) for reduced pressure testing, at pressure Pt, is determined by the relation Lt=La fPtg 1/2 and is .153 percent per de., Paj Where: La = maximum allowable leakage rate at pressure Pa and pressure Pa
~
is 60 psig.
The acceptance criteria for the type A test was set at 75%%u of the Lt value or ,115 percent per day for testing at 35 psig.
The initial class A test was aborted when the plot of containment air weight Vs. time continued to exhibit a weight loss, seven hours after start of the stabilization period, in excess of the. acceptance criteria.
During, the stabilization period two areas of positive leakage from the con-tainment building atmosphere to the adjacent building atmosphere were ident-ified. One leakage path was from the maintained open vent valve, downstream of the containment isolation check valve in the Nitrogen supply line to the Reactor Coolant Drain Tahk. The other leakage path was from the containment ventilation purge exhaust volume. This .volume is bounded by two 48 inch isolation dampers in series configuration with one located inside the con-tainment building and the other located on the outside.
Although the measured and observed leakage from the two i:solation boundaries did not seem to be of sufficient magnitude to account for the total weight loss as exhibited by the mass weight plot the decision'was made to depress-urize the reactor containment building and effect repairs of the leakage paths. Leakage from the two identified areas was evaluated at a rate of
.056 percent per day of the containment test atmosphere.
Prior to the initi,al pressurization of the reactor containment building, a visual inspection of accessible containment interior and exterior surfaces was conducted to uncover any evidence of structural deterioration which would effect either containment structure integrity or leak tightness. There was no evidence of structural deterioration, however it was noted that inside the containment building numerous stainless steel liner sheets (approximately 4 ft. X 7 ft.) which cover the liner insulation were-separated at their
)oining edges. The 'insulation cover serves only to protect the insulation from mechanical damage and to provide a smooth washdown surface during maint-enance.
In retrospect it would seem that the Reactor Building contents were not corn-pletely stabilized prior to the first attempted type A test. All indications point to absorption of air into the porous containment liner insulation. The continuation of ingassing to the insulation free volume, after the desired pressure is attained in the containment; creates the illusion of substantive reactor building leakage.
Repair of the two leakage paths, previously mentioned, were completed on ..
February 10, 1976, and preparations were begun to repressurize the Reactor Containment Building.
The type A retest was concluded sucessfully, the leakage rate determined by applying a linear least squares fit to the weight of containment air at each hour of the test was .044 percent per day.
A report titled "Reactor Containment Building Integrated Leak Rate Test" was submitted to the Director of Nuclear Reactor Regulation, Branch No. 1 on May 19, 1976. The enclosure also included the summary results of type B and C test that were performed since the last type A test (1972). This report was submitted in accordance with the requirements of Appendix J to 10 CFR Part 50.
DATA TABULATION NET ELECTRICAL POWER GENERATION YR TO DATE CUl HJLATIVE NUMBER OF HOURS REACTOR WAS CRITICAL......... 5 300 81 ~46 469.
REACTOR RESERVE SHUTDOWN HOURS ~ ~ ~ ~ ~ ~ ~ ~ . ~ . ~ 759.19 1 051.32*
HOURS GENERATOR ON LINE,... ~ .. ~ . 5115 45 027.63 UNIT RESERVE SHUTDOWN HOURS.... ~ . ~ ~ ~ ~ ~ ~, 0 8~5
- GROSS THE)0(AL ENERGY GENERATED ()0)H)......... 6 984 360 57 355 234 GROSS ELECTRICAL ENERGY GENERATED (MWH) . . . . . . . 2 179 615 18 985 476 NET ELECTRICAL ENERGY GENERATED (blWH) 2 060 941 17 953 80644 REACTOR SERVICE FACTOR ~ . ~..... ~ .. ~ 60% 74%
REACTOR AVAILABILITYFACTOR, . . . ~ . ~ ~ ~ ~ ~ ~ ~ 69% 76%
L UNIT SERVICE FACTOR . ~ . . . ~ , ~ ~ . . . ~ . ~ ~ ~ 58% 72%
UNIT AVAILAVILITYFACTOR. . . , ~ ~ ~ . . .'
~ ~ 58% 72%
UNIT CAPACITY FACTOR (using:.'DC). 50% 65%
UNIT CAPACITY FACTOR (using Design MWe) 50% 65%
UNIT FORCED OUTAGE RATE ~ . ~ ~ , ~ ~ . . . ~ ~ ~ ~ 28% 12%
- Cumulative Data Commencing January 1, 1975
- + Corrected value. Error in June, 1971 calcuations RA
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r STANDARD FORMAT FOR REPORTING NUMBER OF PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION Number of Personnel (> I00 mrcm) Total hlan.Rcm Contract Workers Contract 1Vorkers IVork &, Job Function Station Employccs UtilityLrmployccs and Others Station L'mployccs f Utility mployces and Others Itcactor Operations &, Survcillancc hlaintenancc I'crsonncl 4 0 0 4.670 0 0 Operating I'crsonncl I lcalth Physics I'ers>>nncl 25 0 0 21.840 . 0,* 0 0 0 0 0 0 0 Su per vie>r y Personnel I'.nginccring I'ersunncl 0..
1 0 0 0.
0.580 0 0 0
. 0 0 0 Itoutinc Maintcnancc hlaintenanec I'crsonncl, 36 180, 224 44.610 172,860,. - 89.340 Operating I'crsonnct .-0 0 0 0 0 0 l ical >h I'hysics I cru>nncI .8 1 '5 7. 000 0. 340 4.000
--= .0=
Sut)cfv>u>ly Personnel 8 Q 1. 740 -0 0
,8. 0 =
lin}',h>ccrh>g Personnel 4 0. 480: . 7. 920 0 Inscrviec inspection ihlaintenancc Pcrsonncl 0.250 2.550 1 ..5 f 2.450 Operating Peruu>nel 0 0 0 0 0 0 lleahh I'hysies I'crwnncl Supervisory I'crsonncl I'.nginecring Personnel 0,=
0
- 0. == =- -
'0
.0
-0
=
0 0
0 Q..
0 Q.
0 0
0 0 . ~
Special hlah>tcnancc hlaintcnancc Pcrsonncl 6, 147 .136 Operating
.2,,1QO 64..600 86.390 I'ersonnc)'lc>>lth I'hysks Personnel 0 ., -0 .=0. 0." 0 0
!iupcrvisory I'crsi>nneI 6 0 --=9 3. 230 . 0 0.. 860
)inginccring Personnel 4 1 0. 930 0.820 IVastc Processing 2 0. 81Q. -0.230 hlaintcnancc I'crsonncl. 15 10. 000 Operating I'crsonncl 32 0 9.990 10 8.715 0 llcalth Physics Peru>nncl'
!iupcrvisory Personnel .)0 6 - .0 5
0..
~
3.850 0 0
- 1. 060 nginccring Pcrsonncl Itcfucting 0
0- 0
,0.
0
.Q...=.
0 0
hlaintcn>>ncc Pcrsonncl -'. ' 0 27.430 0 =-21 .2 1. 210 Operating I crsnnncl Ilealth I'hysics I'crsonncl 6 0 0 4.350 0 0
- 3. '0 3.. Qr600 0 0.600 S>lpervisury Pcrsonncl I'.ngh>ecring Personnel 1 .. 0 0
0
.0--
. 1.300
- 3. 270 --0-----
0.
0 0
TOTAL htain tenancc I crsonn et 36 260 252 61.630 267.440 189;380 Operating Peru>nnei 2$ 0 34.905 .0..
I leal th I'hysics Personnel Supervisory Pcrsonncl
- 8. 0.340 0.930
- 6. 520 I'.ngineering Personnel 8 0.820 Orand Total 264 123. 745 277.440 196. 50 Plant Or anization and Personnel The retirement of C.E. Platt, Superintendent of Ginna Station, was announced in December and became effective January 1, 1977.
The following announcements were also made in December and became effective January 1, 1977 L.S. Lang Superintendent B.A. Snow Assistant Superintendent C.H. Peck Operations Engineer G.F. Larizza Nuclear Engineer t ~ ~
C l ~" A' a
5 C
- ta ~ (
a ~
QQlcca ~A Nalt II Opetattag NC ~ Catcaa ltlc JANUARY FEBRUARY MARCH JUNE JULY AUGUST SEPTEMBER OCTOBER NOVEMBER OECEMBER W
- E9
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Figure 2 UNX UE REPORTS 1976 CONDENSER TUBE MAINTENANCE Since changing the secondary chemistry control from phosphate to all volatile treatment, the integrity of the condenser has become much more important. During 1976 we had several leaks in the condenser.
During February with the plant shutdown, a flourescein dye hydro was performed. No leaks were found. An eddy current examination of all the admiralty tubes in the "B" condenser was performed during the spring shutdown. Two tubes had indications of O.D. thinning greater than 50%. One tube was in the 1Bl and one in the 1B2 and both were plugged.
With the plant shutdown in August, another flourescein dye hydro indicated a leaking tube in the 1A2 water box in an unusual area.
Eddy current examination of tubes in this area indicated two more severely damaged tubes. Four tube sections were pulled for closer examination. The attack. seemed to be erosion from the bottom side of the tubes. The probable cause of the attack was ascertained to be steam impingement from the steam dump lines located below the affected tubes. This attack could be seen in both condensers over the most frequently used steam dump lines.
In August, one tube was ruptured by some ob)ect falling into the condenser. A turbine repair )ob had just been'completed, therefore, it was quite possible something was dropped during that gob.
In September power was reduced to 50% and a tube leak was found and plugged in the 1B2 condenser. The tube was located in the bottom periphery where erosion has been a problem.
In December, two power reductions had to be made to locate and repair tube leaks. One leak occurred in the interior of the 1Bl tube bun-dle. The other leak was located in a typical peripheral area in the 1B2 bundle.
During 1976, a total of ten tubes were plugged. All except one of the tubes were admiralty. The one stainless tube was ruptured by a falling object.
With the use of on-line sodium analysers, leaks in the 100-200 cc/min.
can be detected and located to a specific water box. These leaks can then be found using a combination of plastic film and foam. From the experience gained this year with the use of foam and plastic film, any leak which is indicated by sodium level in a specifc water box can be found as long as enough care is taken in performing the check.
~
Primary Coolaht SG BlawQown Report 'nnual Total ECfLQ.vs 1976 Radioactivity Tritium Ci/
>Ci/cc Max F85 E-4 January 13-28 Avg. 15.50 .563 .04 .005 .311 783 5.44 E-5 Min. 4.24 .243 ('.05 ~ 01. .151 236 4. 8 E-7
.03 February Avg. .435 .05 <. 01 2216 Min .35 .05 .01 2107 Max .37 .11 .01 2307 Avg .34 .06 (.01 2234 Min .32 .05 .01 2081 Max 10. 85 .189 .05 . (.01 <5 .156 2189 April 10-24 Avg 3.97 .174 (.05 .112 1487 Min 0.28 .15 (.05 .019 1069 Max 19.00 .221 .13 <. 01 5-10 .331 1575 1.97 E-4 May 7-31 Avg 9. 54 .158 .09 (5 .194 1170 2.3 E-5 Min l. 65 .081 .05 <5 .075 987 < MD@
Max 25.40 .24 .05 (.01 <5 .279 1041 June 1-30 Avg. 22.50 .19 <.05 .361 976 Min 18.44 .1 Max 26.10 .52 .06 <.01 <5 .436 944 July 1-31 Avg 23.68 .48 ( .05 .366 856 Min 20.33 .43 August 1-7 Max Avg Min
- 23. 78 20.40 16.62
.45
.45
.45
'.05 .06
.05
(: 01* (5 .551
~ 551
.551 1043 930 799 Maz 16:71 .50 .15 <.Ol 5-10 .271 1201 September 4-31 Avg 9.54 .41 .07 < 5 .183 859 2 5
.084 14 Max 22.50 .54 g.05 (.Ol <5 .283 1078 2.7 E-6 October 1-8 Avg 18.93 .46 .283 814 5.1 E-7 Min 15.2 1 .283 688 < MBA.
Max 23. 06 .63 '-.! .10 $.01 <5 .415 680 November 1-31 Avg. 17. 12 .51 ~ <.05 .290 631 Min 11.89 .39 <.05 .197 572 Maz .33 .54 - <.05 <. Ol <5 . 530 740 3.04 E-6 December 1-31 Avg 24. 63 44 .358 549 4. 0 E-7 Min 21. 26 .36 .290 489 < MM.
t4pg~
mna I~~= 2.1E-7 Leak Tests on Sources Tech. Spec.
6.9.3 c Leak tests on sources containing greater than 10CFR 30.71 Schedule B quantities wel e made at 6 month intervals.
No tests performed on the sources revealed the presence of
.005 microcurie or more of removable contamination.
F I
KQIESKR CAS 63 ELECTRIC IEE%41ICN 1976 APPENDIX A REPORT OF RADIOACTIVE EFFLVENTS FACILITY DOCKET YEAR I'IOIIID RELEASES AP
' "'V'"""33~ OV, AI 101AL PCLTASC CVRI CS 89195 'gl M/- &=0 . 78-9 IHTQ 817 80 m5K:
Wh.-iI. NSJQ.
0 AVLRACL CONCCNTRATION RILCAStD vCI sa.
Vciiia L11E=It l.llEK 83E-9 9&&9 5VIEM L.168-A9E-1T .OOE- .1 E-CI NAXIIRAI CONCCNIRA'I ION RtlCA510 L16E=8 9 5%=9.. ~7~E- OE-Ial I IVII AI 101AL RTLLA5C Its L15 ~7.537 S.ill 0 AVfaA OSN NIRAIION R LCASCO CUR VCI 04.
12L34 2 .6m'lE~34E~ K79K:
ZQQ 15 2422EGR X32k2 DIS lVTO NOSIC I A$ 5 AI TOTAL RTLTASC CUR I CS TL I 0 AVIRACC CONCfNIRATICN RtltA510 VCI/NL CR05S ALPNA ~ ADIOACIIVITV A IOTA R A RI . 50E-5 L2- LARES L26EJI .5E.
.0 LI3 Ofd 93 - / .19ETlli BJ'11ENlli
- 5. VIRIAK Of LIOVID VATIC 10 DISCIIAROC CANAL Liltas 1.65387 7.5185 6.76E5 5.70E5 3.51E6 2.42E 6 8.19E5 .55 Z28E7 I 1.289E7 2.658E6 i.52285 6
7.
VOlINC Of Dl lUII ON TAITR 15010PCS RtlfASLD Liltas CURT CS 5.55281 . AS 55kl K338EIl DA LA.140 SR.89 L26 73IOOOD{-
Xt 135 Il .00027 cs.137 CS 135
%5I'I 22
.Q162~09ll
.Q0599 TJXOPi
.QQ JXQ
~0 102 Q
CO. .Q1638 'Q Q106/59 .051 CR 51 J$ 126 .I12 5 .00016
.OQL341 0042 QX02 .QQQQ6 3533 00003 0 ~2
- 8. PTRCINI Of ItCNNICAL $ ftCIT ICATION LINIT TOR TOTAL ACIIVIIVRTLTASLD 1.65 .017 .012 .003 .12 c. 1 c ((X'I < .001 <.001 .001 0 05It .007 .00039 .00008 00017 .00005 .0 3 Agll@f .002717 00002 .00001 .00011 .001227 .00012 NI .00001 .00002 .00004 000010 .00002 .00430 811103 .002115 00291 .00937 .00032 .000056 .00006 .00002 .00002, .000004 .00002 000016 tID .01491 Ccl41 .001685 00026 .00156 .00009 .000686 HD HD HD .000007 000116 ND .00440 Nocbcster Oss and Electric Corporation 1976 II AIRBORNE RELEAStS Utl I TS FEB APR IIAY JUIIE JULY SEPT. OCT IIOV BE~ TOTA IOIAl RoslC GASCS IX I 2~Ii . 1 9 . 252 391 1M 6t)S 2ltb12 3.74.852 ~.6 l&L87? Ll&831 122.235
~0073 Ms?..
3 IOIAL HALOGINS CVRICS ~01 2 3 INL~!L DD29 J)O O01 013 .DD125 JXXSS
- 3. feral tARIlcvlAIc GRB5$ RAolo*cllvllY Is,v) CVR I 15 RI 4.52&-6
.5685 6.3E 5 l.2&-6 1 1 2.2&-6 2~22 3.75E-6 Im 1.58-6 M2~ ~.015. - .SSE-6 2.32E-6 .57E-6
.1%
1.33E-6
.582 8.95E-5 tAiilIll!LAIC. Mbss ALAFIIA RAOIBAcllvi'lv S. 121IAL 1 2l$ 8:6 3&ED 3.38=7 "4&=7 . 7 6E~ l 65F 7 2~s7 JI.3~ J)SE~ 5 L&2 . 211~E-9.0E-9 3210 5.50E-6 59;200 6.
7.
110XIMiii IIO14LL GAS RClA51 RAIC PIRCL22I Of AttlICABLC lliillIORI I SCC 50. 000 ~00 L4go 7UD mL ~2&0 S 84 A~ 20OBLC GAS$ 15 c~o8 ~24 011 .D 5 6t),OM R. HAliCCRS J+ 6 34 1J$ 3 C. FARIICVlAICS C. 001 ooL <N
- 8. ISOIttt RtllASloi CVRICS
~ AR'I I C IA AlC S CS.)37 B A. l A. 140 .32m-8 h53-
.28&-8 ND 14~ 1.51F-T 6J2L2. 4F 7 8 1)D
.67'R . 4a7IE-7 lo
.05E-o F
-8 2:24E=6 CS-134 5R.89 NQ LJ3E3
~? 63%1 02-7
.HD
~%=7 3 K%E 22Aliv4205 01 1 Z~= 002D JKZH .00065 0011 .DO12 0 ZOOM +31
.20%
I~ I 0011 ~2.0M AIg .00016 DOD22 Moo J)ODL I ~ 135 GASFS aa 8A 33Ji16 91&
xt.133 Do J16D03 Idol ld XR-87
~4l XR a00 IID +34 1 132 ~ 18 .2962 .?91 13 020 Xt-135M xt. I 35 AR. ~ I onila5 As Attaotelalc ISttclfY 49 JRl Tsl 20 I ttD 8~8 8
15 4432 9 l8606
~
~ 3on J)6lh 0102 K) el A!48 .I!a J)0&& tlQ k.
529 F7 la01E~ ID 80 5.03E-Z .3. 2- IID 2.NFd l QSN-6 .6~r~ 2N+s t.lss&w.83m>> a.92S 7 45'LE=FAN OF~ 1.20E-T 5',058-6
~000!
7 9253 FFJ) I'3,73&r6 Nn= ~a 2.65EW L.53EJL )0 HD 6..87E-
- 5. 56.$ -.7 3,55F 8. ID o DSE57 Nn R 1.66E-92$ FG.S SO In 9~ Jlk8 LOOF=
226E-1.8!m7 SD ln L&5EYS D 9.5&-9 8)
Nb 95 4.6 1&-7 ND 6.64&-8 2.00E-T 3.44E-S 1.24E-6 Te 132 2.77e-T )iD ND 2.77E-7 CO141 2.14E-S ki) 2.14E-S 8)54 2.8E-9 t&3 2.&OE-9 ND - None Detected JANUARY 1, 1976 throu h JUNE 30, 1976 During the six @anth period, 8223 cubic feet of solid waste conta~
Curies of radioactive material was shipped offsite for burial. All the 85.82 solid waste was buried at the Horehead, Kentucky burial ground operated by Nuclear Engineering Gcxqmny Inc.
DATE SHIPPH) COBIC FEET DISPOSITION 12 January 3.50 345 NECO, Morehead, Kentucky 6 February 6.79 360 NEQO, Morehead, Kentucky 19 Fdzxexy 0.12 1008 NECO, forehead, Kentucky 21 February 24 100 NECO, Morehead, Kentucky 23 February 5.13 338 NECO, Morehead, Kentucky 4 March 3.79 456 NECO, Morehead, Kentucky 8 March 4.09 426 NHCO, Horehead, Kentucky 12 March 2.94 499 NKD, Mor~, Kentucky 17 March 12.7 384 NECO, forehead, Kentucky 22 March 2.35 492 NECO, Morehead, Kentucky
~
25 March 0.08 406 NKO, Morehead, Kentucky
-30 March 3.87 331 NECO, Morehead, Kentucky 5 April 1.88 299 NECO, behead, Kentucky 9 April 0.84 494 NECO, forehead, ~cky 14 April 2.54 331 NECO, Horehead, Kentucky 20 April 3;30 NEO, Morehead, Kentucky 20 May 0.82 654 NECO, Morehead, Kentucky 26 May 5.39 385 NECO, Horehead, Kentucky 9 June 1.69 603 NKO, Morehead, Kentucky TOILS 85.82 8223
-100-,
SOLID PASTE JULY 1 1976 throu h DECEMBER 31 1976 During the six month period, 1671 cubic feet of solid waste containing 12.01 Curies of radioactive material was shipped offsite for burial. All the solid waste was sent to the Barnwell, South Carolina burial ground operated by Chem-Nuclear Systems, Inc.
DATE SHIPPED UUEIEE CUBIC FEET 30 August 3.36 348 20 October 4.72 419 23 November 1.71 404 13 December 2.22 500 July >> Dec. Total 12.01 1671 Jan << June Total 85.82 8223 1976 Total 97.83 9894 23 Shipments
-101-
ANNUAL REPORT fk 12
( JANUARY 1 1976 TO DECEHBER 31 1976 INCLUSIVE 6.9.3.a ENVIRONEHNTAL MONITORING 1975
'c)+
A summary of the survey results is presented below:
(1) (ab)+ (d)+
NO. OF NO. OF RESULTS TYPE LOCATION SAMPLES ~FRE UENCY AVG. MAX.
Atomspheric Dust 556 Weekly .053 1.112*pcs/m~
Fall Out 60 Monthly 25.0 103 pci/m2 Radiation Film 12 144 Monthly ~
<10 <10 mRem TLD 23 92 Quarterly 16.9 23.0 mRem Lake Ontario Water 180 Weekly 6.00 12.2 pCi/1 Deer Creek Water 22 Monthly 7.15 13.3 pCi/1 Well Water Monthly 9.56 18.3 pCi/1 Milk Monthly 1.24** 9.48*pCi/1 Fruit Annual .013 ~Cf Csl37 gm iRarine Organisms Fish 1 28 Quarterly 2.84 3.45 pCi/gm Algae 11.3 pCi/gm Lake Bottom Annual 8.98 pCi/gm
- Sample values high due to Chinese Nuclear Bomb Test fallout.
- Avg. milk result would be 0.42 pCi/R if one high value (indicated in *) is excluded.
6.9.3.a (2) No sample or measurement indicated statistically significant levels of radioactivity above the background.
+ Technical. Specification Paragraph
-102-
Loki Onlori p tntahe Tunnel GINNA ITE Loyout and To ttgraphy
~ n sn Weather Tower creen I
s ouse 0 Plant lwcatini WA
~~ g~
or'nW). nn t ~~'.
5e. ~s mao r~
m~
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IV
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~
P't ~ ss Nearest Otf.she
" Residents ~
~+A t ~w y
11%0 AAe
~ ll ~ OQH&tO OI ~~
I
, il ~erst Ltd Control Subatatton Kouae L Ji Q Sampling Station FIGURE LOCATION N MONITORING AND SAMPLING STATIONS (Inner Ring)
-103-
~zp~~>~.ekt H lin Braddock Bay Manitou Beach LAKE ON TAB'10
})RADDOCK Braddock Hts rr i r GINNA SITE 3
O / per( neyyj// e PAIR BARKS PT us L k Bluff h 3'e 2 P
~ 22 ~ Peeper ea Win sor Par g Cen
'59 rrt: h~
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nion Latteside 2)' 7 OdUS P 0 crs rc~ ce ~~Lartin 2t)
~
i o E' e Wall trgto dams }04 REF(E ~ Wg b!i> 0 ta00 itriamsop g~ }p4
'f
. Basin ~ Cen Ontai)o Nilli )son~ Resorts,}
see, Alton
')8 roy 3$
c fPoft l9 QU@0 enfiel 3 g Cen J.' S Sodus
.t. 0 a4')IEST t, %en g I wath.o
~A ee g 11 $ -") field pen field rion way i)e Chili 4'st0 afe Roc ester
~t4 t '-'ats th Fa'i e 8 Cen acedo ili ~)3 je dhurct) y' 15:sg ., qs Cen E Lock erlin ~
P 'Pittsf 'irport f
't ( i acedon aim yr $
ro M OON stare} Clillon
'l r enn< a l6 Bushnell <<gyPt Palmyra 6 Lyons
'- '8
.a e W I b Henrielta L3 one
. NettIfa k 3
Marengo Gaibvttt coll ill endon Mormon Gibs >> 6 ee>> +ay '~ Mumford
~
l9 e ~ Cen k Mob
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~ e Rush M ndon 3 7
CrtltrA DA I)3~ 5 Honeoye Victor 6 tw a/I nius l9 Manchesler [" Clrfton 8 Sr;orlstrille 'prs eanepling Station Scale Miles IO )5 FIGURE '-~. LOCATIONS OF RADIATIONMONITORING -104-AND SAMPLING STATIONS (Outer Ring)
Lake k Onf gr i'o GINNA TV/0 ONE SITE THREE ~ MILE MILE FOUR MILE MILE C. ROAD O LAKE Q JJ: SHEPHERD RO. 0 PUTNAM Q Q ROAD JJJ eeG'OSTON Q Q BRICK CHURCH RD JJJ V 45 ROAD Q z Q Q WSW GEELEY RO 4700m X O Q 0 Q JJ: BERG ROAD d Z O z l JJJ z Q z KENYON ROAD
$ 00 8000m z
J 30 Q 0 N T A R I 0 SSE 5500m PENN CENTRAL HIGHWAY, ROUTE 10 STATE RD RIDGE 45 SSW 7000m
~ADDY LANE LEGEND:
WHITNEY ROAD (20) NUMBER OF COWS IN HERD (SSW) SECTOR INVOLVED 7 n NUMBER OF COWS PER HERD SUPPLIED BY THE WAYNE COUNTY AGENT,
- Fig.. 5 Locations of dairy herds within 5 miles of 'the Ginna reactor.
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GLOSSARY The terms, words, and phrases used in this report are commonly used throughout the nuclear industry, and no special definitions or explanations are necessary.
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