ML20151A837

Annual Operating Rept to NRC,870101-1231

ML20151A837
Person / Time
Site:	Browns Ferry
Issue date:	12/31/1987
From:	Gridley R, Jacqwan Walker TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8804070254
Download: ML20151A837 (108)

Text

{{#Wiki_filter:i 1 l 1 TENNESSEE VALLEY AUTHORITY OFFICE OF NUCLEAR POWER BROWNS FERRY NUCLEAR PLANT ANNUAL OPERATING REPORT TO THE NU' CLEAR REGULATORY COMMISSION January 1,1987 - December 31,1987 DOCKET NUMBER 50-259,50-260, and 50-296 LICENSE NUMBER DPR-33, DPR-52, and DPR 68 i wa- . y.--JL3'/7 A XtB $/10/88 pgRans!A85&886 r3 DCD

o T TABLE OF C')NTENT8 Summary of Plant Conditions . . . . . . . . . . . . . . . 1 Summary of Plant Modification Completions . . . . . . . . 6 Summary of Plant Modifications Partially Complete . . . 15 New Temporary Alterations . . . . . . . . . . . . . . . . 20 Summary of Special Teots Completed . . . . . . . . . . . . 31 New Procedures Issued . . . . . . . . . . . . . . . . . . 42 ; Plant Instruction Revisions . . . . . . . . . . . . . . . 54 Restart Test Program Test Instrictions It. sued . . . . . . 65 i Change Notices for Restart Test Procedures Issued . . . . 80 1987 Release Summary . . . . . . . . . . . . . . . . . . . 89 Transmission Line Corridor Herbicide Usage .- . . . . . . 90 Reactor Vessel Fatigue Usage Evsluation . . . . . . . . . 91 Challenges To or Failures of Main Steam Relief Valve . . . 92 Occupatior,al Exposure Data . . . . . . . . . . . . . . . . 93 i i i i i I l l l l w.-- . , , _ , ------...c.---x---w,,__, ... , _ _

SUMMARY

OF PLANT CONDITIONS JANUARY 1,1987 - DECFMBER 31, 1987 Unit 1 Unit I was placed on administrative hold in March 1985 to resolve TVA and NRC concerns and has been in outage for 1027 days. The unit also began its sixth refueling on June 1, 1985, with a scheduled restart date to be determined. The sixth refueling will involvo loading 8x8R (retrofit) fuel assemblies into the core. The prior-to-startup unit 1 items are environmental qualification (EQ) of electrical equipment (10CFR50.49), torus modification (Nuclear Regulatory

• Commission Rest'lation [KUREG] 0661), containment modifications (EUREC 0737), electrical changes (Appendix R 10CFR50) (all) main steam isolation valve (MSIV) modifications, modification of masonry ,

walls (IEB 80-11), evaluation of the vent drain and test connections (VDTC) (Licensee Event heport (L'dR) 82020), valve modification (Appendix J), high pressure coolant injection (HPCI) concerns, modification of primary containment isolation system (PCIS) logic (LER 259/85009), replacement of plant process computers, seismic qualifications of piping (IEB 79-02/14), postaccident evaluation (KUREG 0737), resctor protection system (RPS) modifications (IE notice 78-45) H 02 2 sample line modification (LER 81050), radiation monitors modification (LER 80033), emergency equipment cooling water (EECW) carbon to stainless pipe change out, and all NRC commitment items except anticipated transients withcut scram (ATWS) modifications which are sr.heduled for next outage. There are zero assemblies in the reactor vessel. The opent fuel storage pool presently containw 284 new assemblics, 764 end of cycle (EOC)-6, 252 EOC-5, 260 EOC-4, 232 EOC-3, 156 EOC-2, and 168 EOC-1 assemblics. At the end of this reporting period capacity of thn fuel pool was 1355 locations. Unit 2 Unit 2 was placed on administrative hold on September 3, 1985, to resolve TVA and URC safeta concerns. The unit also began its fifth cefueling on S2ptember 15, 1984, with a scheduled restart date to be determined. The fifth refueling intolves loading 8x8R (retrofit) fuel assemblies into the core. The prior-to-startup unit 2 items are control rod drive (CRD) scram discharge instrument volume (SDIV) piping modification (IEB d0-17), EQ of electrical equipment (10CFR50.49), torus modifications (NUREG 0661), containment modification (MUREG 073?). electrical changes (Appendix R 10CFR50) (partial), MSIV modifications, modification of masonry walls (IEB 80-11), addition of feedwater nozzle temperature monitoring (NUREG 0619), evaluation of the VDTC (LER 82020), valve modification (Appendix J) (partial), diesel generator (DG) speed sensor 1

e installation (LER 81004), high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) testable check valve change out, modification of PCIS logic (LER 259/85009), HPCI concerns, seismic program review, and EECW carbon to stainless pipe change out. There are zero assemblies in the reactor vessel. At the end of 1987 there were 304 new assemblies, 764 EOC-5, 248 EOC-4, 352 EOC-3, 156 EOC-2, and 132 EOC-1 assemblies in the spent fuel storage pool. At the end of the year, available capacity of the spent fuel pool was 1481 locations. All high density racks (HDRs) have been installed in the pool with the exception of two. Unit 3 Unit 3 was shut down on March 9, 1985, and placed on administrative hold to resolve various TVA and NRC concerns. The restart date has not been determined. It has been shutdown for 1020 days. On January 30, 1987, fuel unloading was started. This task was completed on March 2,1987. The sixth refueling outage has been scheduled to start on September 21, 1988, and involves loading 8x6R (retrofit) assemblies into the core and ATWS modifications. The prior-to-startup unit 3 items are EQ of electrical equipment (10CFR50.49), containment modifications (NUREG 0737), electrical changes (Appendix a 10CFR50) (all), MSIV modifications, modification of masonry walls (IEB 80-11), evaluation of the VDTC (LER 82020), valve modifications (Appendix J), HPCI concerns, replacement of plant process computer, seismic qualifications of piping (IEB 79-02/14), postaccident evaluation (NUREG 0737), addition of redundant drywell control air supply, RPS modification (IE Notico 78-45), H 022 sample line modification (LER 31050), radiation monitor modification (LER 80033), replacement of jet pump holddown beams assemblies (IEB 80-07), change out of switches in standby gas treatment (SBGT) (LER 83018), EFCW carbon to stainless pipe change out, and plant design upgrade to seismic qualification. There are zero assemblies in the reactor vessel. There are 764 assemblies to finish EOC-6, 248 EOC-5, 280 EOC-4, 124 EOC-3, 144 EOC-2, and 208 EOC-1 assemblies in the spent fuel storage pool. At the end of this reporting period, available capacity of the fuel pool was 585 locations. All HDRs have been installed in the pool with the exception of six, s - 2

o OPERATING DATA Operatino Status - Unit 1

1. Unit Name: Browns Ferry Unit One

2. Reporting Period: Decenber 1987

3. Licensed Thennal Power (MWt): 3293

4. Nameplate Rating (Gross MWe): 1152

5. Design Electrical Rating (Net MWe): 1065

6. Maximum Dependable Capacity (Gross MWe): 1098.4

7. Maximum Dependable Capacity (Net MWe): 1065

8. If Changes Occur in Capacity Rt ting (Items Nunber 3 Through 7) Since Last Report , Git , reasons: N/A

9. Power Lever To Which Restricted, if Any (Not MWe): N/A

10. Reasons for Restrictions, In Any: N/A 1987 Cumulative

11. Hours in Reporting Period 8760 117.680

12. Nunter of Hours Reactor Was Critical 0 59.521.38

13. Reactor Reserve Shutdown Hours 0 6.997.44

14. Hours Generator On-Line 0 58.267.26

15. Unit Reserve Shutdown Hours 0 0

16. Gross Thennal Energy Generated (MWH) 0 168.066.787

17. Gross Electrical Energy Generated (MWH) 0 55.398.130

18. Net Electrical Energy Generated (MWH) -14.233 53.706.402

19. Unit Service Factor 0 49.51

20. Unit Availability Factor 0 49.51

21. Unit Capactty Factor (Using MDC Net) 0 42.85

22. Unit Capacity Factor (Using DER Net) 0 42.85

23. Unit Forced Outage Rate 100 41.12

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

N/A

25. If Shutdown at end of report period, estimated date of startup:

To be determined i 3

e OPERATING DA7A Operatina Status - Unit 2

1. Unit Name: Browns Ferry Unit Two

2. Reporting Period: pecenber 1987

3. Licensed Thennal Power (MWt): 3fi3

4. Nameplate Rating (Gross MWe): .1152

5. Design Electrical Rating (Net MWe): 1065

6. Maximum Dependable Capacity (Gross MWe): 1098.4

7. Maximura Dependable Capacity (Net MWe): 1065

8. If Changes occur in Capacity Rating (!tems NJrber 3 Through 7) Since Last Report Give Reasons: N/A

9. Power Lever To Which Restricted, If Any (Net MWe): N/A

10. Reasons for Restrictions, in Any: N/A __

1987 Cumulative

11. Hours in Reporting Period 8760 112.567 _

12. Nunber of Hours Reactor Was Critical 0 55.860.03

13. Reactor Reserve Shutdown Hours 0 14.200.44

14. Hours Generator on-Line n 54.338.36

15. Unit Reserve Shutdown Hours 0 0

16. Gross Thennal Energy Generated (MWHi 0 153.245.167

17. Gross Electrical Energy Generated (MW ) 0 50.771.798

18. Net Electrical Energy Generated (MW) -34.470 49.183.932

19. Unit Service Factor 0 48.27

20. Unit Availability Factor 0 48.27

21. Unit Capacity Factor (Using MDC Net) 0 41.03

22. Unit Capacity Factor (Using DER Net) 0 41,03

23. Unit Forced Outage Rate 100 40.25

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Aration of Each):

N/A

                                                                                                                                                                                                                                    ~

25. If Sh W own at eno of report period, estimated date of starten:

To be d.d.eg ined l 4 l

e OPERATING DATA Oceratino Status - Unit 3

1. Unit Name: Browns Ferry Unit Three

2. Reporting Period: Decenter 1987

3. Licensed Thermal Power (MWt): 3293 _

4. Nameplate Rating (Gross MWe): 1152

5. Design Electrical Rating (Net MWe): 1065

6. Maximum Dependable Capacity (Gross MWe): 1098.4

7. Maximum Dependable Capacity (Net MWe): 1065

8. If Changes Occur in Capacity Rating (Items Nunter 3 Through ?) Since Last Report , Give Reasons: N/A

9. Power Lever To Which Restricted, If Any (Net MWe): N/A

10. Reasons for Restrictions, In Any: N/A 1987 Cumulative

11. Hours in Reporting Period _ 8760 94.992

12. Number of Hours Reactor Was Critical 0 45.306.08

13. Reactor Reserve Shutdown Hours 0 5.149.55

14. Hours Generator on-Line 0 ,,,,_

44.194.76

15. Unit Reserve Shutdown Hours 0 _

0

16. Gross Thermal Energy Generated (MWH) 0 , 131.868.267

17. Gross Electrical Energy Generated (Mul) 0 , 43.473.760

18. Net Electrical Energy Generated (MWH) , -50.980 42.041.148

19. Unit Service Factor 0 46.52

20. Unit Availability Factor 0 46.52

21. Unit Capacity Factor (Using MOC Net) 0 41.56

22. Unit Capacity Factor (Using DER Not) 0 41.56

23. Unit Forced Outage Rate 100 13J

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

N/A

25. If Shutdown at end of report period, estimated date of startup:

To be determined 5

F~ n

SUMMARY

OF PLANT MODIFICATION COMPLETIONS JANUARY 1, 1987 - DECEMBER 31, 1987 Engineering Change Notice (ECM) P0898 1.%difications of Fire Protection System - Unit 2 Description The sprinklers in the aqueous film-forming foam (AFFF) system were raised or lowered as required by this modification in the unit 2 reactor building on elevation 639*. Certain sprinklers were also rearranged slightly to percit adequate dispersion or water density over the reactor recirculatlon MG sets. Safety Evaluation The modification increased the effectiveness of the AFFF system by improving the spacing, coverage, and discharge of the sprinkler heads over the reactor recirculation motor generator (MG) sets. The AFFF system does not perform a safety-related function. However, proper design considerations were taken to ensure safety-related equipment in the immediate vicinity of sprinklers was not adversely affected. Thus, the margin of safety was not reduced. ECN P0896 Modifications of Fire Protection System - Unit 2 Description Installed automatic sprinklers in the unit 2 reactor building abovo elevation 593'. The modification included hangers to stpport the piping. Safety,_Ey uuation The modification minimizes the pr.tential spretni of a fire through the equipment hatch openings and stairwelic between elevations 593' and 621' of the unit 2 reactor building. Also, sprinklers were provided for areas on elevation 593' that were not covered by the existing sprinklo system. Therefore, the fire - protection capabilities of the plant were improsed. The modification did not degrade the reactor building integrity. The marrin of safety was not reduced. i ECN P0897 Modifications of Fire Protection System "n'.t 2 Description Installed automatic sprinklers in the unit 2 reactor building above elevation 621'. The modification included hangers to support the piping. j 4 4 6

           , ._yu     . _ _ _ , _ - - . . - ,

Safety Evaluation . The change minimized the potential spread of a fire through the equipment hatch openings and stairwells between elevations 621' and 639' of the unit 2 reactor building. The modification provided sprinklers for areas on elevation 621' that were not previously covered by the sprinkler system, thus improving the fire protection capabilities of the plant. The piping and supports meet the Class II requirements and do not degrade the Class I structures. Since fire protection was improved in the unit 2 reactor building, the margin of safety was not reduced. ECN P0895 Modifications of Fire Protection System - Unit 2

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i Description Installed automatic sprinklers in the unit 2 reactor building above elevation 565'. The modification included hangers to support the piping. The ECN was totally completed as it only covered unit 2. Safety Evaluation The modification minimizes the potential spread of a fire, thus improving the fire protection capabilities of the plant. The additional piping and supports meet the Class II requirements and do not degrade class I structures. Since the change improved fire protection in the unit 2 reactor building, the margin of safety was not reduced. ECN P0894 Modifications of Fire Protection System - Unit 2 l Description Installed automatic sprinklers that formed a water curtain below the unit 2 reactor building elevation 565' floor slab in the residual j heat removal (RHR) corner rooms. The modifications included hangers to support the piping. Safety Evaluation i The modification minimizes the potential spread of a fire, thus improving the fire protection capabilities of the plant. The additional piping and rapports meet the Class II requirements and do not degrade Class I structures. Since the change improved fire f protection in the unit 2 reactor building the margin of safety was j not reduced. I I i 7 6

F O ECN P0828 Modifications of Reactor Recirculation System - Unit 2 Description Installed fuse blocks, associated wiring, and fuses in the protective relay circuits of the 4KV reactor recirculation pump MG sets 2A and 2B. Safety Evaluation The modification was performed to bring BFN into compliance with 10CFR50 Appendix R. The added coordinating fuses did not degrade the seismic qualifications of the associated circuitry. The margin of safety was not reduced. ECN P5500 Modifications of Reactor Building Roof - Common Description Inspected and repaired the damaged areas of the reactor building roof. Safety Evaluation The repairs to the roof enhanced the margin of safety and improved secondary containment integrity. ECN P5420 Modifications of Test Ports (System 30, 31, 64) - Common Description The ECN was written for documentation only to "as-construct" drawings. The drawings provided for the installation of ventlock test holes in the heating, , ventilation, and air conditioning systems. Safety Evaluation The modification will not affect the seismic qualification or designed function of the heating, ventilating, and air-condi'loning c (HVAC) systems. It will facilitate the required postmodification testing for ECNs P3138, P3139, P0956, and P0870 through P0875. ECN P5303 Modifications of EECW System - Common Description The modification was for documentation only to "as construct" drawings to depict the actual plant configuration. TACF 0-85-029-067 was removed, which j had previously implemented the modification. The modification removed the l 8

i raised face from flange joints of the temperature control valve 67-62. Safety Evaluation The modification only altered the flange configuration used for mating valve 67-62 to its associated process piping. The modified joint will perform the same function as the original flange joint and its modified design meets the EECW system design requirements. Based on this, the margin of safety was not reduced. ECN PS309 Modifications of Cable Tray Support System - Unit 2 Description Performed modifications required to seismically qualify affected cable tray systems for unit 2 startup and interim operation. Work involved the cable tray supports at the intake pumping station and the unit 2 cable spreading room and the cable tray tunnel (rectangular portion of tunnel from the intake structure to the plant). The ECN was completed. Safety Evcluation The modifications assure that the affected cable trays meet the seismic qualifications of seismic Class I cable tray / supports for unit 2 startup and interim operation. The margin of safety was not reduced. ECN PS295 Modifications of Reactor Building Closed Cooling System - Unit 2 Description Revised "as designed" drawings to the 1/2-inch cleanup recirculating pump test connection isolation valves 70-666A, 666B, 667A, and 667B to conform to the actual unit 2 configuration. Safety Evaluation The valves were not required for system operation and were not required for any nuclear safety function. The modification did not affect the RBCCW system's ability to provide any required design basis function. Based on this, the ' margin of safety as defined in the basis for any technical specification (TS) was not reduced. ECN PS277 Modifications of Cable Tray Support #21 - Unit 2 Description Fabricated and installed the cable tray cupport baseplate (support #21) in the unit 2 reactor building. 9

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Safety Evaluation The structural integcity of the seismic cable tray support was evaluated by the Division of Nuclear Engineering (DNE). It was determined that the function, operation, and qualification of the support and other associated equipment was not adversely affected. Thus, the margin of safety was not reduced. ECN PS228 Modifications of Main Steam System - Unit 2 Description Installed tees in instrument lines for PT-1-79, 81, 90, 93, and PI-1, 94, 97, 100, 103, 109, 112 which are located on panels 25-110, 25-111, a.nd 25-112. Safety Evaluation The modification was designed so that it did not degrade the qualification of the safety-related equipmerit which provides input to the RPS. Based on this, the margin of safety was not reduced. ECNs P5224 and P5433 Modifications of Refuel Floor Components - Units 1, 2, and 3 Description Revised the cribbing requirements for all components set on the refuel floor so as not to exceed 1000 PSF loading on the floor. Provided third laydown area for the unit ' reactor pressure vessel head during disassembly. Safety Evaluation Compliance with these cribbing requirements will ensure the refuel floor loading limits will not be exceeded. Therefore, the margin of safety was not reduced. ECN P3137 Modifications of Main Steam System - Unit 2 Descriptiqn The ECN was for documentation only to "as-construct" drawing 47W600-11 R17. The ECN added note 5 to the drawing. The piping classification was changed from P to M for CV-1-14D, 15D, 26D 27D 37D, 38D, 51D, and 52D. Safety Evaluation The valves are only needed for testing. They are not required for any safety-related function associated with closing the MSIV. Thus the Class P portion of piping associated with the test valves could be reclassified without adversely affecting the safety of the plant. 10

ECN P0877 Modifications Fire Protection Draft Stops - Unit 2 Description Installed draft stops around vertical openings between different reactor building floor elevations. Safety Evaluation The modification improved the effectiveness of the fire protection system. The operations of other systems were not affected. Therefore, the change did not reduce the margin of safety. ECW P0672 Modifications of Reactor Feedwater System - Unit 2 Description Removed pressure switches PS-3-57A, B, C, and D, and their associated wiring and replaced them with a wire connecting the mode switch (SA-S1) and relays 5A-K11A through D. Safety Evaluation The modification did not adversely impact plant safety. A seismic analysis showed the changes did not adversely affect qualification of any seismic Class I equipment. The margin of safety was not reduced. 4 ECN P0585 Modifications of Standby DGs - Units 1 and 2 Description Replaced speed sensing syctem consisting of frequency generator and Electro-Motive Division of General Motors Corporation (EMD) speed switch panels with magnetic pickups and associated speed switches. Safety Evaluation The speed sensing panels were replaced with more reliable, solid state panels.

The function of the panels was not changed. The new panels were seismically qualified Class 1E. Postmodification testing proved the acceptability of the i modification. The margin of safety was not reduced.

! ECN P0244 Modifications of Reactor Water Recirculation System - Unit 2 Description Replaced Foxboro transmitters, PDT-68-65 and PDT-68-82, with Rosemount. 11

Safety Evaluation The Rosemount transmitters serve the same functions as the Foxboro models, have proved to be more reliable, and their use will enhance the function served by the transmitters. Hence, the margin of safety has been enhanced. ECN P0093 Modifications of Primary Containment - Unit 2 Description Fabricated and installed torus access ladders inside the unit 2 torus. Modified the torus vent header collar supports. Fabricated and installed MK1 plates and increased existing fillet welds. The major portion of the work covered by the ECW was implemented during unit 2 cycle 4. It will be completed during the current unit 2 cycle 5 outage. Safety Evaluation ECN P0093 covers the modifications for the long-term torus integrity programs. The modifications covered by the ECN were tested and analyzed by CE. The data gathered by GE showed that the modifications, when performed, would be a great improvement over the present condition. Therefore, the probability of occurrence or the consequences of an accident or the malfunction of equipment important to safety was not increased. ECN P0570 Modifications of Emergency DGs - Unit 3 peseription Replaced viscous type vibration damper with gear type vibration damper on the i unit 3 DG engines. The ECN was completed for unit 3. The modification had been previously implemented on units 1 and 2. Safety Evaluation ! The exchange of DG vibrational dampers (a gear-type for a viscous type) did not reduce the ability of the DGs to perform their function or increase the probability of their failure. The margin of safety was not reduced. ECN L1896/L1916 Modifications of Control Air System - Unit 2 Description Revised power sources for FSV-32-28A and FSV-32-29B and made final electrical tie-in for FCV-32-91. Power source for FSV-32-28A was changed from the plant , nonpteferred AC system B to an I&C AC system bus. FSV-32-29B power was changed from the I&C bus to the plant nonpreferred bus. The ECNs were field completed but postmodificattor testing is remaining, i 12

Safety Evaluation . The modification increased the reliability for keeping FSV-32-28A and -29B open during normal plant operation. The implementation of the ECNs will prevent spurious scram due to failure of one of the buses and, thereby, add to the safe operation of the plant. ECW P0974 Modifications of Cooling Tower 4KV Switch Gear - Common Description This ECW provides for the connection of the Environmental Qualification warehouse substation and the stoplight at the north nuclear plant road to the newly constructed AKV loop line/ cooling tower 4KV switchgear. The ECN is not complete. Safety Evaluation The addition of this modification and/or possible failure of the equipment will not create any different type of accident than previously evaluated. Also, loss of these additional loads will not impact safety of the plant operation or safe shutdown. Therefore, the margin of safety is not affected. ECN P0957 Modifications of Recirculation System - Unit 2 Description Replaced and/or modified part of the existing recirculation system piping susceptible to intergranular stress corrosion cracking (IGSCC) with an improved stainless steel material that has increased resistance to IGSCC. The major portion of the field work covered by the ECN has been completed. Some minor support work and hydrostatic testing are all that remains to complete the ECN. Safety Evaluation All structures and components affected by this modification were replaced / repaired with equivalent or better materials which meet or exceed current design requirements. The modification does not adversely affect the system operation. Therefore, the margin of safety will not be reduced. ECNs L1692/P0857 Modifications of RHR/RHRSW Systems - Unit 1 Description Installed flange connections on RHR/RHRSW heat exchangers 1B per ECH L1692. This represented only a partial completion of the ECN for unit 1 and then the remainder of the work was cancelled. No work on this ECN was completed for units 2 and 3 and the ECN for these units has been cancelled. ECN P0857 13

                                                                                               .                                                                                                                                          i pertained to field change requests to revise drawing discrepancies found as the work was being perfonned on ECM L1692.                       ECN P0857 was completed on unit 1.

i Safety Evaluation - The function, operation, and qualification of the RHR heat exchanger and the

RHR service water inlet line were not adversely affected. Based on this, the

, margin of safety as defined in the basis for any TS is not reduced, a . i i i 1 i 14

i

SUMMARY

OF PLANT MODIFICATIONS PARTIALLY COMPLETE JANUARY 1, 1987 - DECEMBER 31, 1987 P ECW P0879 Partial Modifications of Fire Protection (Doors) - Common Description Replaced or modified various compartmentation boundary doors which did not meet UL requirements and/or did not provide the fire barrier ratings required to  : meet the 10CFR50 Appendix R requirements. Modifications have been completed i for doors 455, 482, 531, 600, 643, 644, 656, 810, 811, 824, and'827. Various other doors are being modified but modifications are not completed. Safety Evaluation The door modifications improve the fire protection system design without changing or adversely affecting the operation of the secondary containment system. Therefore, the margin of safety is not reduced. l ECN P0381 Partial Modifications of Various System - Units 2 and 3 Description CEMAC transmitters are being replaced with Rosemount transmitters on an as-needed basis on designated systems of all three units. The ECN has not been completed on any unit. Safety Evaluation 3 The Rosemount transmitters are compatible with existing equipment, perform the same functions as the CEKAC transmitters, and meet the requirements specified for the functions performed. The margin of safety margin required by the TS has not been reduced. ECN PS213 Partial Modifications of RCIC/HPCI j Description Modified the bonnet flange joints on valves HCV-71-32 and HVC-73-24. The ECN was completed for unit 2 but not for 1 and 3. l t 15

Safety Evaluation . The modification will allow the bonnet flange seal to be leak rate tested under the local leak rate test program. The modified valves perform the same function and meet the same requirements but facilitate testing. The margin of safety was not reduced by this modification. ECN P5016 - Partial Modifications of Reactor - Recirculation System peggription Documented the removal of the 4-inch bypass line and the associated snubbers, SS-9, on loops A and B which were removed in accordanc'e with ECN L1633. Also, documented was the removal, by maintenance requests, of the snubber support brackets in order to facilitate weld overlay repairs. The ECN was completed for unit 2, was previously completed on unit 3, but was only partially implemented on unit 1. Safety Evaluation

  ~

Based on the results of DNE calculations, the pipe break analysis is still valid, and the seismic analysis was not adversely affected by the modification. ECN p0207 Partial Modifications HPCI/RCIC Systems Unit 2 Description Fabricated and installed access platforms for the HPCI and RCIC

>        temperature switches. The ECW was not completed for unit 2.

Inspection and repair is being performed during the unit 2, cycle 5 outage. Safety Evaluation l The platform does not have a safety-related function. It is seismic Class I so that it will not fall during an earthquake and possibly damage safety-related equipment. The platform is anchored to the reactor building concrete structure. The additional loading of the concrete structure is negligible. Therefore, the probability of an accident or malfunction of equipment important to safety per;viously evaluated was not increased. 16

ECN L1970 Partial Modifications of EECW System . Description Replaced existing valves and made piping tie-in to the new stainless steel components (EECW supply for RHR pump seal heat exchangers). A small portion of the work covered by the ECM has been partially implemented on all 3 units. f Safety Evaluation The replacement valves and piping tie-ins do not change the operation or affect the ability of the EECW system to respond to an accident situation. The integrity of the EECW piping system was not reduced by the new material. The margin of safety was not reduced. ECN P0027 Modifications of Spent Fuel Storage - Unit 3 Description Removed the remaining old style fuel storage racks, control rod racks, and seismic bracing. New base plates, fuel racks, and control rod racks are remaining to be installed on unit 3. The ECN has been completed on unit 1 but only partially implemented on unit 2.

                                                                                                                                                                                                                                                                                                        ,e Safety Evaluation Seismic qualification of the racks and their arrangements were evaluated and approved.                                                                                                      NRC approval was obtalned prior to implementation of the ECN. The margin of safety will be enhanced with the complete implementation of this ECN.

ECN P0085 Modifications of Primary Containment - Unit 2 Description Performed modifications for upgrading the existing drywell temperature and pressures instrumentation and annunciator used for the manual initiation ' of containment spray by the unit cperator. The major portion of the work covered by the ECN has been completed. Rerouting of some cables and performing postmodification testing will be performed prior to unit 2 startup. Safety Evaluation Implementation of the ECN serves to ensure mitigation of excessive drywell pressures and temperature. Therefore, the consequences of design basis accidents can be held within certain acceptable limits and maintain the containment integrity. The margin of satety will be improved through implementation of the ECN. ,

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17

                                                                                                                                                                                               ..,--_<_r                                                        - _ _ - - _ . _ _    . . --_ - - _   _-

ECM P0126 Modifications of Various Systems - Unit 1, 2, & 3 Description The ECW provides for the replacement of the currently installed pressure, level, and temperature switches with analog transmitter / trip unit combinations (which provide continual monitoring of critical parameters in addition to performing basic logic trip operations) with the proposed analog transmitter / trip unJt system for engineered safeguard sensor trip inputs. Work that has been completed includes the installed conduits, cables, junction boxes, panels, and pressure tested instrument pensing lines on unit 2. Safety Evaluation The instrument functions will not be changed by the modification. The modification package was reviewed and approved by NRC prior to implementation. The likelihood of failure will be reduced by implementing the ECN. ECW P0533 Modifications of Temperature Monitoring System Description The ECN provides for the design, procurement, and installation of an improved temperature monitoring system for the torus. A major portion of the work has been completed during the unit 2, cycle 5 outage. The modification is still in progress. The physical work for this ECN was previously completed on unit 1. Postmodification test, however, remains to ba performed. The major portion of the ECN has been completed on unit 3. Safety Evaluation i The new monitoring system will provide a more accurate indication of torus water bulk temperature and the local temperature at each quencher than the old system could provide. The modification provides assurance that the torus temperature is within the prescribed limits set forth in the TSs; therefore, the margin of safety was not reduced. j ECN P0768 Modifications of RHRSW/EECW/RCW - Common Description i Revised drawings to reflect the required design pressure and temperatures for EECW and RHRSW systems, and the interfacing portion of the RCW system. Physical work remains to be implemented, hence, the ECN was not l complete. I 18

Safety Evaluation . The documentation change will not provide additional modes of failure. Documenting the "as-is" status of the equipment will provide an accurate base of technical knowledge on which to base future design or operations. No physical work was done to any system which would impact any actual I existing margin of safety already built into plant equipment. 4 ) I i i t

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NEW TEMPORARY ALTERATIONS JANUARY 1, 1987 - DECEMBER 31, 1987 0-87-001-26 Installation of Alternating Current (ac) Recording Voltmeter Purpose This temporary alteration was installed to troubleshoot false starts on fire pump "C" and detect whether these false starts were initiated by the auto start circuitry of fire protection equipment. This was accomplished by attaching a recording ac voltmeter to the coil of the auto start relay for fire pump "C" motor (FPXC) and to the coil of the fire pump ' push button auxiliary relay (FPSL). Safety Evaluation Neither the probability of the occV"Jence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the Final Safety Analysis Report (FSAR) was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the temporary - alteration did not affect fire pump "C" operability, the recording voltmeter was fused so that in the event of an accident or malfunction the meter would have tecome shorted and would have been cleared from the circuit on auto initiation, and work performed was within the scope of the current TS. 0-87-005-90 Change Continuous Air Monitor (CAM) Detectors to Meet New Sensitivity Requirements Pu rpose This alteration was accomplished in order to comply with the radiological environmental technical specification (RETS) and was issued to make I changes to CAM RM 90-252. The change included the replacement of the 4 particulate and iodine channel collectors with filter collectors and the replacement of the gaseous channel detectors from CM tubes to scintillation detectors. This alteration will remain installed until permanent changes are implemented by Design Change Requesc 3421. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the building effluents will be continuously monitored by a noble gas monitor (instead of by noble gas, particulate, and lodino 20

monitors) and significant increases in particulate and lodine activity are always accompanied by increases in noble gas activity. The alteration does change the type of detectors referenced in FSAR Section 7.12.6 but the change is to the type required by the new TS (RETS ) .

                                 )

0-87-008-57 Eliminate Loss of power and Undervoltage Load Shed Initiations Purpose This temporary alteration was issued to eliminate the loss of power and undervoltage load shed initiation on breaker 2B on 480V shutdown board 1A. The 480V shutdown board 1A is the power source for the control bay ventilation board A which feeds the control room emergency ventilation (CREV) system train A. This alteration will ensure the CREVS train A is operable for any design basis accident. Safety Evaluation Welther the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. The plant standby AC power supply feeds only one control bay ventila'.lon board which is in conflict with the FSAR. This temporary alteration resulted in supplying both control bay ventilation boards with standby AC power. The loading of the DG will not be increased by the adding of this extra load becauso RHR pump 1A will be removed from service. The power requirements for the other engineering safeguard systems will be satisfied. Also, during any design basis accident, two trains for the CREV system will be automatically available. 1-87-003-26 Temporary Repair of Fire Protection System piping Pu rpose This temporary alteration was issued to install a temporary patch on an 8-inch high pressure fire protection system distribution pipe to stop a pinhole spray leak. The location of the, patch is elevation 565' unit 1 SC penetration between unit 1 and unit 2. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR 21

was increased or created, and the margin of safety.as defined in the basis of any TS was not reduced because the installation of the temporary patches on the high pressure system distribution piping does not render the system or portion of the system inoperable and does not invalidate piping design per FSAR 10.11.5.1. d 1-87-004-90 Change CAMS Detectors to Meet Sensitivity Requiraaents Purpose This alteration was accomplished in orda , comply with the radiological environmental TS and war , sued to make changes to unit 1 CAMS RM 90-249. -250, and . These changes included the replacement of the particu1** ca iodine channel collectors with filter collectors and , placement of the gaseous channel datec' ors from ... to scintillation detectors. This alteration

                                                                                                                    ..w 11ed until permanent changes are implemented by Desian Change Request 3421.

Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the l basis of any TS was not reduced because the building effluents will be continuously monitored by a noble gas monitor (instead of by noble gas, particulate, and iodine monitors) and significant increases in particulate and lodine activity are always accompanied by increases in noble gas activity. The alteration does change the type of detectors referenced in FSAR section 7.12.6 but the change is to the type required by the new TS (RETS). 1-87-005-26 Temporary Repair of Fire Protection System Piping Pu rpose This temporary alteration was issued to 'all a temporary patch on i a 4-inch high pressure fire f,rotection sye .m distribution piping to allow this section of the system to remain in service and to stop the ! existing leak. The location of this patch is in unit I reactor building elevation $65' in overhead R7 Q-line. 4 Safety Evaluation Neither the probability of the occurrence or the consequences of an i accident or malfunction nor the possibility for an accident or 4 malfunction of a different typo than previously evaluated in the FSAR was incressed or created, and the margin of safety as defined in the i 22 1

basis of any TS was not reduced. Installation of temporary patches on the high pressure fire protection system distribution piping does not render the system or portion of the system inoperable and does not invalidate piping design per FSAR 10.11.5.1. The increased loading on the pipe and pipe supports is negligible. 1-87-006-26 Temporary Repair of Fire protection System piping Purpose This temporary alteration was issued to install a temporary patch on a 3-inch high pressure fire protection system distribution pipe to stop a pinholo leak. The location of the patch is in unit I reactor building on elevation 593' at the north wall. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. Installation of temporary patches on the high pressure fire protection system distribution piping does not render the system or portion of the system inoperable and does not invalidate piping design per FSAR 10.11.5.1. The increased loading on the pipe and pipe supports is negligible. 1-87-008-26 Temporary Repair of Fire protection System Piping Purpose This temporary alteration was issued to install a temporary patch on a 2.5-inch high pressure fire per .ection system distribution pipo to stop a leak located at R2.5 U-line elevation 565' in the unit i reactor building. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. Installation of temporary patches on the high pressure fire protection system distribution piping does not render the system or portion of the system inoperable and does not invalidate piping design per FSAR 10.11.5.1. The increased loading on the pipe and pipe supports is negligible. The difference in pipo stresses (dead weight and seismic) is small and does not exceed USAS B31.1.0-1967 code allowables; however, this fire protection Class II system is not designed for a seismic event. 23

i. 2-87-002-26 Temporary Repair of Fire protection Syatem Piping Pu rpose This temporary alteration was issued to install u temporary patch on a 3-inch high pressure fire protection system pipe to stop a pinhole leak. The location of the patch in unit 2 is at U-line and R11 about 14 feet above the floor. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or , malfunction of a different type than previously evaluated in the FSAR

;                                              was increased or created, and the margin of safety as dtfined in the basis of any TS was not reduced. Installation of temporary patches on the high pressure fire protection system distribution piping does not render the system or portion of the system inoperable and does not invalidate piping design per FSAR 10.11.5.'. The increased loading on the pipe and pipe supports is negligible.

2-87-004-26 Temporary Repair of Fire protection System piping Pu rpose l This temporary alteration was issued to install a temporary patch on a 4-inch high pressure fire protection system distribution pipe to stop a leak. The location of the leak was on the 4-inch line aboue l the CRD rebuild room on tho cast end at elevation 621' of unit 2 reactor building. t Safety Evaluation ! Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the . basis of any TS was not reduced. Installation of temporary patches ! on the high pressure fire protection system distribution piping does not render the system or portion of the system inoperable and does l not invalidate piping design per FSAR 10.11.5.1. The increased loading on the pipe and pipe supports is negligible. l 2-87-005-90 Change CAM Detectors to Meet New Sensitivity Requirements i j Purpose This alteration was issued to make changes to unit 2 CAMS RM 90-249,

                                                -250, and -251. These changes included the replacement of the 24 5

particulate and iodine channel collectors with filter collectors and  ; the replacement of the gaseous channel detectors from GM tubes to scintillation detectors. This alteration will remain installed until .

permanent changes are implemented by Design Change Request 3421.

Safety Evaluation Neither the probability of the occurrence or the consequences of an , accident or malfunction nor the possibility for an accident or malfunction of a different typs than previously evaluated in the FSAR , was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the building effluents will . be monitored by a noble gas monitor (instead of by noble gas, , particulate, and lodine monitors) and significant increase in particulate and lodine activity are always accompanied by an increase in noble gas activity. The alteration does change t.he type of detectors referenced in FEAR Section 7.12.6 but the change is to the type required by the new TS (RETS). 2-87-010-26 Provide Cooling Water and Drainage for Induction Heating Stress Improvements (IHSI) Heat Exchanger Pu rpose This temporary alteration was issued to provide cooling water and drainage for IHSI heat exchangers. This task was accomplished by i attaching a 1.25-inch line (hose) to valve wye located at R T-line on elevation 621' of the unit 2 reactor building and by attaching a 1.25-inch line (hose) to the roof drain HDR located at R U-line on elevation 573' in the unit 2 reactor building. 3 i Safety Evaluation Neither the probability of the occurrence or the consequences of an J accident or malfunction nor the possibility for an accident or

;                                                       malfunction of a different type than previously evaluated in the FSAR j                                                         sas increased or created, and the margin of safety as defined in the basis of any TS was not reduced. The system was supplied raw water by three motor-driven vertical turbine pumpu rated at 2500 GFM each at 300-foot total head. Backup to those pumps is one diesel-engine-driven vertical turbine fire pump rated at 2500 GPM at 340-foot total head. The diesel-engine pump starts only after the three electric-motor-driven vertical pumps fall to supply adequate system pressure. Only or,e pump is necessary to supply the water i                                                        requirement for a fire in one of the safety-relsted areas specified in the TS.                      Two pumps would be required to handle the maximum switchyard fire. There was more than sufficient margin to carry this approximate 20 GPM cooling load and still maintain all fire loads.

The system will be returned to its original configuration when 4 requested by the shift engineer. I 25

2-87-011-303 Installation of Air Conditioning . purpose , E The purpose of this alteration was to provide cooling for unit 2 ~ drywell during reactor building closed cooling water (RBCCW) system outage by installing a temporary air conditioning system. a Safety Evaluation Weither the probability of the occurrence or the consequ2nces of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the - basis of any TS was not reduced. positioning of temporary air conditioning unit and associated equipment does not exceed design

live loading of 400 pounds per square feet. Actual loading is 335 poundo per square feet. The cooling water for the temporary air conditioning unit is supplied by the nonsafoty-related raw cooling water (RCW) system and discharged to the nonsafety-related reactor building roof drain system. Also, the power supply to the unit is from an existing temporary source. Finally, the equipment will be removed prior to fuel load and is not located adjacent to any safety equipment required for this mode of operations. Temporary duct work is routed through primary containment which is not required (per TS 3.7.A) for this mode. All equipment is located in the secondary coatairment.

2-87-012-26 Temporary Repair of Fire Protection System piping Purpose The temporary alteration was issued to install a temporary patch on a 4-inch fire protection system distribution pipe to stop a pinhole leak. The location of the patch is in unit 2 reactor building on elevation 621' and the R-13 T-line. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the a basis of any TS was not reduced. The high pressure fire protection system pipe was not made inoperable by this temporary patch. The patch will prevent further degradation of piping resulting in the l, high pressure fire protection piping integrity being maintained. The ! original design of the piping system will not b= vltered by the patch. l 26

2-87-014-77 provision of Temporary Source of Electrical power to Sump Pump Instrumentation

• Pu rpose This temporary alteration will install a temporary source of 120V AC d

power to the drywell sump level monitoring instrumentation to provide level and alarm indication from LIS-77-1A and LIS-77-18. The normal power supply cables to this system were damagoi in the recent drywell fire and are to be disconnected while repairs are being made. This

alteration will power the drywell drain sump level instrumentation to give operators control room indication and alarm of drywell sump level. Additionally, the high level alarm setpoint on those sumps will be temporarily raised to accommodate more volume before having to pump out the sumps.

Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or ma? function nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. The normal power source to the drywell sump level monitoring instruments will be unkva11able rhile repairs are made to that power source. Providing a temporary source of power to the drywell sump level monitoring instruments will restore these instruments to operations thereby providing the control room operators with alarm indications if there is any leakage within the drywell. Thus, the effects on safety normally provided by this l 5 system are restored. The temporary power supply is the 10A fused, l 120V AC service receptacles on panel 25-52 which has a nonsafety-related power source. Hence, there is no effect on other 1 safety systems. The requirements for operability of these systems is ! not in effect per section 3.6.C.2 of the TS since unit 2 is shutdown 4 and the reactor is defueled. This temporary alteration will b6 removed and the normal power supply to the drywell monitoring l instruments restored and the high level setpoint returned to the original values prior to unit 2 refueling. i 3-87-001-26 Temporary Repair of Fire protection System piping i Pu rpose ) This temporary alteration was issued to install a temporary patch on a 4-inch fire protection system distribution pipe to stop a pinhole leak. The location of this patch is on the fire protection header on 1 the third floor of the unit 3 reactor building at the AFFF unit. i l l l 1 i 27

   - . - - _                -,~ _                                                                                          . -~

Safety Evaluation - l Welther the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident oc malfunction of a different type than p*(viously evaluated in the FSAR was increased or created, and the parcin of safety as defined in the > e basis of any TS was not reduced. Insta11ac t on of temporary patches . on the high pressure fire protection system distribution piping does  ; not rcf. der the system or portion of the system inoperable ar.3 does not invalidate piping design per FSAR 10.11,5.1. The increased loading on the pipe and the pipe supports is nee ;itsible.  ; 3-87-002-82 Use of a com,acreial Grade Relay l

Purpose l This temporary alteration involves replacing DG 3A r.o'/ernce and stop circuit relay (SCR) (type CO-1E series D, class 9050) with an identical, yet commercial model until an acceptable (qualified) model can be procured.

i  ! Safety Evt,'yttign  ; Neither the probability of the occurrence or the consequences of an  ; j accident or malfunction nor the possibility for an accident or malfunction of a differer-( typo than previously evaluatei in the FEAR j l was increased or created, and the margin of sefuty as defined in the ' ) basis of any TS was not reduced. The tir's delay TCH has no safety j

function. The relays only function la to previde annunciation in the l 4

event the diesel does not stop within 30 seconds of a diesel i emergency shutdown initiation. The replacement relay. 41though not

Class 1E or seismically qualified, has the same manufacturer's part number and is identical in form and function to the relav replaced.

Electrical failure of the relay could render the DG inoperable. The

replacement of the relay with the same manufacturer's part numter [

4 provide) reasonable assurance that the reliability of the circuit was  ; not degraded and the DC remained operabic to support engineered safeguted equipment. Identical form and mounting L.' ths relay 1 ensure, .t will not s'fect other components in a selsmic event. A i special condition of this alteration is, however, that all three  : units are defueled and the relay must be replaced with a qualified i device prior to loading fusi in any unit, j i  ! 3-87-003-90 Change CAM Detectors to Meet New Senritivity Requiremento  ; - Purpose This alteration was issued to make changes to unit 3 CAMS RM 90-249, 1 -250, and -251. Thece changes included th> eeplacement of the r k i l 2b l

l particulate and iodine channel collectors with filter collectors and the replacement of the caseous channel detectors. This alteration will remain installed until permanent changes are implement by Design , Change Request 3421. l 1 Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the building effluents will be monitored by a noble gas monitor (inste*4 of by noble gas, - particulate, and iodine monitors) and s1pra icant increases in particulate and iodine activ3ty are always accompanied by increases in noble gas activity. The alteration does change the type of detectors referenced in FSAR section 7.12.6 but the change is to the type required by the new TS (RETS). 3-87-004-26 Temporary Repair of Fire Protection System Piping Purpose The temporary alteration was issued to install temporary patch on an 8-inch fire protection system distribution pipe to stop a pinhole leak. The location of the patch is at P.-16 T-line on elevation 602' next to the stairwell in the unit 3 reactor buildiag. Safety Evalection Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, Ond the margin of safety as defined in the l basis of any TS was nnt reduced. Installation of temporary patches l on the high pressure fite protection system 61stribution piping does L not render the system or portion of the system inoperable and does I not invalidate piping design per FSAR 10.11.5.1. The increased l loading on the piping and supports is negligible. The difference in l pipe stresses (dead weight and seismic) is small and will not exceed ! USAS B31.1.10-1967 code allowables. The fire protection system is a I Class II system and not designed for seismic event. l l 3-87-005-26 Temporary Repair of Fire Protection System Piping Pu rpoga This temporary alteration was issued to install a temporary patch on an 8-inch fire protection system distribution pipe to stop a l 29 l -

pinhole leak. The paten is located at R-16, S-line on elevation 565' in the unit 3 reactor building. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. Installation of temporary patches on the high pressute fire protection system distribution piping does not render the system or portion of the system inoperable and does not invalidata piping design per FSAR 10.11.5.1. The increased loading on the piping and supports is negligible. The differenc'e in pipe stresues (dead weight and seismic) is small and will not exceed USAS B31.1.10-1967 code allowables. The fire protection system is a Class II system and not des 3gned for seismic event. 3-87-007-26 Temporary Repair of Fire protection System piping Pu rpose This temporary alteration was issued to install a temporary patch on an 8-inch fire protection system distribution pipe to repair a pinhole leak. The location of the patch is in unit 3 reactor building on elevation 565' at the R-16 T-line. t Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the high pressure fire I protection system pipe is not further degraded by the patch and it does not render the piping inrperable. The patch will prevent further degradation of piping resulting in the high pressure fire i protection piping integrity fdesign, function, and performance) being

- maintained. All actions taxen have been in agreement with existing TSs.

l l l 30 u

SUMMARY

OF SPECIAL TESTS COMPLETED JANUARY 1, 1981 - DECEMBER 31, 1987 ST86-14 Verification that inoperative Trips are Caused by Circuit-Boards-Out-of-Circuit in the Neutron Monitoting Systems Test Objective This test was performed to determine whether the inoperative trip function of the neutron monitoring system functions as required by TSs when the circuit board is removed. The results of this test will be used to determine whether justification exists for deleting the requirement for removing circuit cards from circuits during the performance of surveillance functional tests. Deletion of this removal step will help prevent equipment degradation. Safety Evaluation Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an ac;ident or malfunction of a different type then previously evaluated in the FSAR was increased or created, and the margin of safety as defit ed in the basis for any TS was not reduced because all testing performed on the neutron monitoring system components was performed in the logic bypassed configuration which is described in the ?SAR and the number of instrument channels allowed to be bypasced by the TS tus not exceeded. ST86-19 Temperature and Dewpoint Traverse of Drywell and Suppression Chamber Test Obiective This special test was performed to obtain temperature and dewpoint area data within the drywell and suppression chamber to determine the best location for containment intergrated Icak rate test (CILRT) inst rumentation. Safety Evaluation Neither the probability of occurrence or the consequences of an accident malfunct!gr yn malfunction of a differentnor typethe possibility than previouslyforevaluated an accident or FSAR in 'he was increased or created, and the margin of safety as defined in the basic for any TS was not reduced because the test was performed while I the unit was in the cold shutdown mode, the drywell head was unbolted , l and vented, and the primary containment was not required to be

maintained, l

l 31

ST86-25 Standby Liquid Control (SLC) Special Sodium pentaborate Analysis Test Test Objective This test was performed to verify that proper dissolving and mixing of sodium pentaborate in the SLC tank.could be achieved by agitating the solution with air. Air sparging was conducted at a constant rate with periodic axial sampling until the difference in concentrat3on between axial samples was acceptable. The sparging time that yields the desired accuracy will be established as the minimum sparging time for all future surveillances. Safety Evaluation Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated ir; the PSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because the SLC system is not specifically required to be operable while each unit is in cold shutdown with fuel removed, but the system was fully functional during testing. The testing did not affect any systems other than the SLC system. The testing did not affect any of the variables or components of the SLC system. ST86-26 Alternate Rod Injection Backup Scram Valve Diagnostics Test Objective The objective of this test included the collection of unit 1 scram valve reaction data on specific hydraulic control units (HCU) during an alternate rod injection (ARI) scram. This data will be used to support the design of the ARI system. Safety Evaluation 4 Neither the probability of occurrence or the consequences of an accident or malfunction nor '.he possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increaced or created, and the margin of safety as defined in the basis for any TS was not reduced because all ine fr91 was removed from the reactor vessel and the test did not affect any system or equipment required to maintain the fuel in a safe configuration. Also, TS 3.1 allows for the disablement of the RpS when fuel has been removed from the vessel. The performance of ST86-26 did not alter or disable any system or equipment necessary to mitigate or prevent any accident or malfunction, which can affect the fuel or spent fuel , pools. 32

L 'f ST87-01 Ventilation System Flow Measurnments - Test Objective The objective of this test was to collect dimensional data for use in establishing the location of ventilation monitoring sample probes and determining the current vent flow rates. Knowing the exact nozzle location, nozzle ID and velocity at the nozzle, sample flow rates can be determined and isokinetic conditions established and existing plant instructions can be properly updated to assure compliance with the radiological effluent technical specification (RETS). Safety Evaluation Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because the probes and continuous air monitors (CAMS) taken out of service served no accident mitigation function, the isolation of secondary containment was not impaired, all work (with the exception of probe measurements on the reactor zone exhaust duct) are outside of this boundary, any exhaust leakage during work on the reactor zone ducts was contained by the refuel floor exhaust system and could be isolated by its isolation features as described in section 5.3 of the FSAR, and because the performance of this test did not place the plant in noncompliance with the TS. ST87-02 EECW System Loss-of-power Test Task Objective The purpose of this test was to measure the amount of drain down leakage that will occur in the EECW system during a loss-of-offsite power. The test configuration simulated the plant configuration that would exist during a loss-of-offsite power by isolating raw cooling water (RCW) and raw service water (RSW) from the EEC'l system and then stopping all of the EECW pumps. The amount of EECW system drain down was measured by a gauge glass installed on the north header at the LS-67-52 location. Safety Evaluation Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FCAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because the EECW system south header was operable during the performance of this test and it was capable 33

of supplying all essential loads and was able to respond to a challenge. Also, the testing did not preclude system function or operability and was retained in a state which ensuced no different accident or malfunction possibilities existed. ST87-03 Offline Liquid Radiation Monitor Isotopic / Transfer Calibration Test Objective This special test was performed to determine 1.n accurate sensitivity and efficiency correction and to obtain transrer source data on the offline liquid radiation monitors required for RETS implementation. Safety Evaluation Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because calibration and operation of the offline liquid radiation monitors will be enhanced and, thereby, an accurate indication of the radioactivity present. ST87-04 Replacement of Surveillance Instruction (SI) 4.2.A-24 Test Objective The purpose of this test was to satisfy TS 4.7.B.3-a, due to the existing plant conditions, by performing FI 4.2.A-24 in lieu ( thereof. The requirement of the TS is to demonstrate the operability of the auto initiation of the SBGT cystem once per operating cycle. l This requirement has been met from a verbatim licensing point of view, but it is felt that the time period since its last performance along with the current operating conditions warrant an operability l demonstration of the logic at this time. 1 Safety Evaluation l Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or I malfunction of a different type than previously evaluated in the FSAR l was increased or created, and the margin of safety as defined in the l basis for any TS was not reduced because this test was used to actually demonstrate the initiation of SBGT system by the pCIS exhaust high radiation or offocale signals. All sis required to l verify operability of the secondary containment were performed prior I to this test within their frequency and their operability was not affected by the test. 1 34

                   .            m-- -    -- -- -

ST87-05 General Electric (GE) Stack Gas Radiation Monitor Isotopic / Transfer Galibration Test Objective The purpose of this test was to perform a primary isotopic and secondary transfer calibration of the GE Model ll7B1681G1 2X2 NaI(Tl) Scintillation Detector and to determine a National Bureau of Standard (NBS) traceable efficiency correction ratio and obtain transfer source data required for the calibration of the main stack gas radiation monitor to support the implementation of RETS requirements. Safety Evaluation , Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because calibration and operation of the stack gas radiation monitor will be enhanced and thereby an accurato indication of radioactivity released to the environment via the offgas vent can be determined. ST87-14 Liquid Radwasto Radiation Monitor Isotopic / Transfer Calibration Test Objectivo The purpose of this test was to perform a primary isotopic and secondary transfer calibration of the GE Model 117B1681G1 2X2 NaI(Tl) Scintillation Detector and to determine an NBS traceable efficiency correction ratio and transfer source data required for the calibration of the liquid radwaste radiation monitors to support implementation of RETS requirements. Safety Evaluation Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a dlfferent type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because calibration and operation of the long range radiation monitor (LRRM) will be enhanced and thereby the radwasto discharge will isolate when the radioactive concentration in the dischargo execeds 2x10-4 pCi/ml. 35

ST87-15 Unit 1 Off-Gas posttreatment Radiation Monitor. Isotopic / Transfer Calibration Task Objective The purpose of this test was to perform a primary isotopic and secondary transfer calibration of the GE Model 117B1681G1 2X2 NaI(Tl) Scintillation Detector and to determine an NBS traceable efficiency ratio and obtain transfer source data for the Unit 1 off-gas posttreatment radiation monitor to support the implementation of RETS requirements. Safety Evaluation Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because calibration and operation of the off-gas posttreatment radiation monitor will be enhanced and thereby provide an accurate indication of radioactivity and enough information to easily control the release rate. ST87-16 Unit 2 Off-Gas Posttreatment Radiation Monitor Isotopic / Transfer Calibration Test Objective The purpose of this test was to perform a primary isotopic and secondary transfer calibration of the GE Model ll7B1681G1 2X2 NaI (Tl) Scintillation Detector and to determine an NBS traceable efficiency ratio and obtain transfer source data for the unit 2 off-gas posttreatment radiation monitor to support the implementation of RETS requirements. Safety Evaluntion Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because calibration and operation of the off-gas posttreatment radiation monitor will be enhanced and thereby provide an accurate indication of the radioactivity and enough information to casily control the release rate. 36

ST87-17 Unit 3 Off-Gas Posttreatment Radiation Monitor. Isotopic / Transfer Calibration Test Objective The purpose of this test was to perform a primary isotopic and secondary transfer calibration of the GE Model 117B1681G12X2 NaI(T1) Scintillation Detector and to detarmine an NBS traceable efficiency ratio and obtain transfer source data for the unit 3 off-gas posttreatment radiation monitor to support the implementation of RETS requirements. Safety Evaluation , Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because calibration and operation of the off-gas posttreatment radiation monitor will be enhanced and thereby provide an accurate indication of the radioactivity and enough information to easily control the release rate. ST87-18 Radwaste Disposal System Representative Sample Determination Test Objective The purpose of this test was to establish a minimum recirculation (mixing) time necessary to ensure representative sampling of radwaste sample tanks. This information is required to establish proper sampling techniques prior to discharging contents or, as applicable, to the condernate storage tt.nks. Safety Evaluation Neither the probability of occurrence or the consequences of an accident or taalfunction nor the possibility for an accident or l malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the i basis for any TS was not reduced because none of the design safety l features listed in the FSAR were affected, system operation during l the special test was not significantly different from normal i operation, and the test was performed so there was no way for the occurrence of an unmonitored release to the environment. The TS addresses only the radioactive release from the plant which is not l impacted by the special test. The quantity of chemical traces l (sodium nitrite) used was small enough to ensure the nitrite additions did not exceed the reportable quantity limits in 40 CFR, ! Part 117. l l l l 37 t

ST87-21 Spreading Room Backdraft Damper Test . Test Objective This special test was performed to verify operability of the two backdraf t relief dam; rs (2-31-2020 and 3-31-2021) located in the cable spreading rooms. The test was designed to verify relief setpoints by pressurizing the upstream plenums with the dampers in place an.d to ensure CO2 would be contained. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. Fire suppression is the safety function of the backdraft dampers. Specifically, they serve to contain CO2 within the room of discharge, maintain adequate CO2 concentrations, and prevent migration of the gas into the turbine building. A continuous firewatch was maintained during the performance of this test and 125 pound (minimum) portable fire extinguishers were in place. Doors 470, 477, 534, and 537, which were blocked open for access of an air hose, were covered by firewatches while the doors were open. Security was maintained for doors 534 and 537 by posting a public safety officer at each while the doors were blocked open. Caution statements were included to suspend testing at 1.25-inch WG to prevent overpressurization if the dampers were incorrectly set. ST87-23 Diesel Generator Excitation Test Test Objective This test was performed to obtain data on DG excitation system for use in completing the dynamic analysis of the DG system. Data obtained included data from the maximum field voltage the exciter can produce with the generator unloaded and from the exciter and generator during the start of a large motor (RHR pump) onto an isolated DG feed bus. Safety Evaluatien Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type that previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because this test removed DG 3C and i loop 2 of the unit 3 RHR system from service. Provisions and limitations for removal from service or loss from service of this l I l l 38

equipment are contained and analyzed in the FSAR. ~Also, it is assumed in the FSAR that the DG is tested to verify performance and the FSAR contains provisions for possible failure. The design of the electrical system and the TSs allow for systems to be removed for testing. ST87-27 Yard Loop and Cooling Tower Loop Hydraulic performance Evaluation Test Ob.iection The purpose of this test was to validate a major procedure revision to SI 4.11.A.1.h. prior to issuing it as an approved SI. Safety Evaluation Neither the probability of occurrence or the cotisequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in tha FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because the valves and equipment operations required by this test did not affect the probability of pressure boundary failure in any safety-related area, or of flooding of safety-related equipment due to any other system malfunction since the test was designed to ensure the automatic functions of the high pressure fire protection system were not impaired or altered. Valve alignments during the test simulated single pipe failure on the yard loop. Such single failures have been analyzed in the FSAR. The procedure required that the system be returned to normal alignment in the event of a fire. The fire protection system remained operable throughout the test. Flow and pressure conditions stayed within the design capability of the fire pump and the water distribution maina. ST87-29 Optimum Hop and Rag Disposal Technique Determination l ' Test Objective The purpose of this special test was to determine the optimum disposal technique for contaminated mops and rags using the drum i compactor. l Safety Evaluation Neither the probability of occurrence or the consequences of an accident or malfunction nor the possibility for an accid.nt or [ malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis for any TS was not reduced because the operations of the drum compactor (radwaste) is independent of the safety equipment evaluated in the FSAR. The test did not involve equipment important to safety 39

nop did it involve the breach of any component or system that may allow an uncontrolled or unmonitored release of radioactivity to the environment. ST87-30 DG Excitation (3C, B. D) Test Objective The purpose of this test was to obtain additional data on the DG excitation system to resolve discrepancies between the results of ST87-23 and previous runs of SI 4.9.A.1.B. The test results were required for completing the dynamic analysis of the DG system. Safety Evaluation During this test DGs 3C, B, and D, loop 2 of the unit 3 RHR system, and loop 1 and 2 of the unit 2 RHR system were removed from service. Provisions and limitations for removal from service or loss of this equipment are contained and analyzed in the FSAR, so this testing did not increase the consequence or probability of an analyzed event. Also, the margin of safety was not reduced because the design of the electrical system and the TSs allowed for this system removal for testing. Since it is assumed in the FSAR that testing of the DGs is performed to verify performance and contains provisions for possible failure, the possibility for an accident or malfunction of a different type than previously evaluated was not created. Since the fire protection system remained operable throughout the test and the test instructions provided for a return to normal alignment in the event of a fire alarm, the margin of safety was not reduced. ST87-32 Valve Leak Rate and Maximum Pressure Differential Test Test Objective The objective of this test was to prove that eight EECW System sectionalizing valves and 10 RHRSW heat exchangers inlet valves met acceptance criteria in that the leak rate across the valves internal pressure boundary (flapper) at design pressure was minimal, and that' the flapper could withstand the maximum differential pressure that it l would be exposed to during a system hydrostatic test. l l l Safety Evaluation l Neither the probability of occurrence or the consequences of an ! accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the l 1 40

basis for any TS was not reduced because this testing provided a means to verify the valves met test acceptance criteria prior to installation, plant configuration as described in the FSAR was not affected, the valves were not subjected to conditions that exceeded TS limits, and if the test results had been found unacceptable for any valve tested, that component would not have been installed in the plant without further evaluation. ST87-35 DG D Emergency Load Test Test Objective This special test was performed to verify the capability of the DG D to accept its unit 2 emergency loads in incremental steps by supplying power to its assigned loads fed from 4KV SD BD 'D', DSL AUX BD 'B', and 480V SD BD 2B under loss of offsite power / loss of coolant accident conditions and then testing the 480V load shedding feature with selected 480V loads of units 1 and 2. This was the third test in a series of four. Each step increment represents approximately 250KW starting from a bare of 2100KW. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because there were no provisions of the FSAR which were violated by the performance of this test. This test was a derivative of SI 4.9. A.1.b-4 which is performed routinely for load acceptance testing of DGs. It has a valid safety evaluation. Also, no modification to the plant was involved in order to perform this test. 41 4

l i NEW PROCEDURES ISSUED j JANUARY 1, 1987 - DECEMBER 31, 1987 l FP-0-39-PM-001 Preventive Maintenance for Horizontal CO2 Activated l I Fire Damper in DG Building Reason The instruction is used to perform preventative maintenance on fire dampers that are CO2 activated. This instruction allows for the inspection, lubrication, and manual cycling of these dampers to insure operability in the event of CO2 discharge. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this procedure will provide periodic inspection and maintenance on the dampers to ensure that they are capable of performing their design functions and comply with FSAR 10.11.6. This procedure does not alter the design or operability of the dar.pers and complies with TS. MMG-034 Wilden Model 4100 Dewatering Pump Disassembly, Inspection, and Reassembly Re_ason This instruction describes the assembly and disassembly of the Wilden Model M4100 dewatering pump. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the involved equipment la not safety-related nor is it connected to any safety-related equipment previously evaluated in the FSAR. The activity involved is not considered in the basis for any TS. The activity is not addressed in any TS nor does it require the use of any equipment that may violate TS. 42 4

SDSP 12.3 Plant component Identification - Reason Implements procedure to gain consistency and accuracy in functional description of plant equipment. Procedure outlines actions and responsibilities for reviewing and upgrading component identification tags and name plates on or associated with plant operational equipment. Safety Evaluation Neither the probability of the oc urrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because procedures will provide more accurate functional descriptions on equipment labels and generating instructions, therefore, probability of accident or malfunction should decrease. Safety margins should improve by enhancement of operatot ability to correctly understand instructions and implement control actions. SDSP 16.5 Control of Material After Issue from Power Stores Reasqn This procedure establishes the responsibilities for preparing instructions for control of material af ter issue from Power Stores and sets forth the minimum requirements which must be addressed in these instructions. Implemented in response to condition adverse to quality report (CAQR) BFQ870131, this CAQR identified the need for an SDSP to adequately control critical structures, systems and componento af ter issue and prior to installation or end use. Safety Evaluation Feither the probability of the occurrence or the consequences of an accident os malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this program does not alter the method of or requirements for procurement or installation of permanent plant material or equipment. This is a program enhancement which provides additional assurance and is not a change to the TS. 43

SDSP 26.1 Special Nuclear Material Control and Accountability Reason The procedure defines those programmatic requirements necessary to ensure the receipt, inspection, handling and shipment of special nuclear material (SNM) are performed according to properly approved written instructions. This instruction ensures that the location and status of all SNM are known and can be verified at all times, and that applicable regulations are adhered to regarding inventories, inspections, and reporting requirements. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or . malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because fuel assembly insertion error during refueling and fuel har.dling accident are the only possible accidents which result from implementing this procedure and these accidents are discussed in the FSAR. There are no changes to systems or procedures which affect system performance which might reduce margin of safety. All implementing procedures all in accordance with TS. SDSP 27.2 Responsibilities of the Independent Safety Engineering Group Reason New procedure implements PKp 604.05. Independent Safety Engineering Group (ISEG) Evaluation. ISEG performs independent reviews, surveillances, and audits of nuclear safety-related activities, programs, and events. ISEG was originally developed for NTOL plants after the Three Mile Island incident. Safety Evaluation Neither the probability of the occurrence or the consequences of an l accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the j basis of any TS was not reduced because SDSP 27.2 is an i administrative procedure and doec not affect plant equipment. The procedure is not related to accidents or malfunctions, does not effect sny margins of safety, and is not addressed in the TSs. 44 f 1 i

RWI-001 Administration of the Radioactive Material and Radwaste Packaging and Transportation Program, Unit 0 Reason This inst'uction, r used in conjunction with the radioactive material shipment manual, establisnes Browns Ferry Nuclear Plant (BFN) policy to ensure that shipments of radioactive material and radwaste from BFN meet all appropriate regulations. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than pecviously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because RWI-001 is the administration of the radioactive material and radwaste packaging and transportation program. This procedure is strictly administrative. During the implementation of this procedure, no safety-related equipment will be operated or affected in anyway. No physical activities vill take place in the plant which may increase the possibility or create a malfunction of a different type previously evaluated in the FSAR. RWI-101 Identifying Types of Solid Radwaste, Unit 0 Reason This instruction provides guidance to plant personnel in identifying the various types of solid radioactive waste material commonly generated at BFN. Safety Evaluation Neither the probability of the occurrence or the consequenses of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the' basis of any TS was not reduced because the procedure merely identifies different types of radwaste to help each employee be aware of their individual involvement to help reduce the amount of unneccessary radwaste generated at BFN. During the implementation of this procedure, there will be no activities performed which may jeopardize or impair any safety-related equipment as e$aluated in the FSAR. No activities will take place which may create an accident that has not been previously evaluated in the FSAR. 45

RWI-102 Use of Radwaste package Control Tags, Unit 0 . Reason The purpose of this procedure is to outline the use of tags which establish traceability of all solid waste material received for processing by radwaste. This procedure is applicable to all persons involved with radwaste packages. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or , malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the procedure merely outlines personnel responsibility to help maintain control of solid radioactive waste going to radwaste. No activities will be performed which will increase the probability or consequences of an accident. During the implementation of this procedure, no activities will take place which may create ..n accident that has not been previously evaluated in the FSAR. RWI-103 Removal and Routing of Radioactive Waste from Radiologically Controlled Originating Areas, Unit 0 Reason This instruction provides guidance in the removal and routing of radioactive waste material from the radiologically controlled originating area to the proper areas for handling and processing. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this procedure merely provides a guideline for all plant employees to follow while routing radioactive waste materials from the originating area to the proper area for handling and processing. There will be no activities performed which may increase the probability or occurrence of an accident. This procedure does not require the use of any equipment or material which may increase the possibility or consequences of an accident that has not been previously evaluated in the FSAR. 46

4 RWI-104 Trash Frisk and Segregation Facility Operation, Unit 0 Reason This instruction provides guidance for the day-to-day operation of the trash frisk and segregation facility. This facility permits the inspection and sorting for reuse, decontamination, and clean release of materials that would otherwise be disposed of as radwaste. T1.e primary purpose of this facility is to prevent the inclusion of clean materials as radwaote, thereby reducing radwaste volume. Safety , Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the operation of safety-related equipment will not take place or be impaired by the implementation of this procedure. The equipment used during this operation can in no way affect any safety-related equipment. There is no possibility for a radiological release into the environment or the occurrence of an accident which has not been previously evaluated in the BFN FSAR. RWI-105 Packaging of Acceptable Noncompactible Materials in Drums, Unit 0 Reason This instruction provides for the packaging of acceptable noncompactible material into drums. Safety Evaluation ! Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or l malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because there will be no safety-related equipment operated or impaired nor will the probability or consequences of an accident be increased. No activities will take place that may cause an accident which has not l previously been evaluated in the FSAR. l l i l 47

RWI-106 packaging of Acceptable Material in Noncompactible Boxes Reason This instruction provides the method for the packaging of acceptable noncompacted materials into boxes. Safety Evaluation Neither the probability of tbs occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because there will be no safety-related equipment operated or impaired nor will the probability or consequences of an accident be increased. No activities will take place that may cause an accident which has not previously been evaluated in the FSAR. RWI-109 Loading Radioactive Waste for Shipment, Unit 0 Reason The purpose of thic instruction is to provide guidance in the proper techniques for loadit.g and inspecting radioactive waste packages prior to shipment from BFN. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not redu:ed because implementation of the procedure will not require the use of any equipment important to safety. Also, the probability or consequence of an accident previously evaluated in the FSAR will not be increased. There will be no work performed during this instruction which may create the possibility of an accident or malfunction which has not been previously evaluated in FSAR. RWI-110 Shipment of Radioactive Material Unit 0 s Reason This instruction outlines the requirements for shipping radioactive material from BFN to other TVA locations, or to outside organizations. This instruction is applicable to shipments involving radioactive material such as samples, sources, nonieradiated incore detectors, laundry, contaminated plant bacdware, etc. l l 48

Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfur.; tion of a dif ferent type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the implementation of this procedure will not require the use of any safety-related equipment nor will it require any work to be performed in the plant. Therefore the probability of occurrence or the consequences of an accident cannot be created. RWI-111 Storage of Radioactive Waste and Materials, Unit 0 Reason This inatruction outlines the requirements for both short-term and long-term storage of low icvel radioactive waste and materials onsite. It provides guidance on the preparation, transport, storage, and removal from storage of materials. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a diffetent type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the items in this procedure do not create any conditions which may increase the potential of an accident that are different than normal operating conditions. Strict adherence to this pcocedure will ensure that the possibility for an accident or malfunction of a different type than any evaluated previously in the BFN FSAR will not be created. RWI-112 Container Marking, Unit 0 Reason This instruction provides information needed to properly mark radwaste containers intended for disposal. This instruction is applicable to any radwasto drum, box, or resin liner. It includes marking information for both the Barnwell, South Carolina, and the Richland, Washington, disposal sites. Safety Evaluation 1 I ! Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or l 49

malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this instruction is strictly an administrative procedure. There will be no safety-related equipment operated or affected. Also, no activities will be performed which may increase the possibility or consequences of an accident which has not been previously evaluated in the current BFN FSAR. This procedure does not require the use of any equipment which may create an accider t or malfunction of a different type. RWI-ll3 Mechanical Filter Solidification Unit 0 Reason - This instruction outlines the method for solidification of mechanical , filter elements which contain some free liquid. This instruction is applicable only to filters which are waste class and have a specific activity of less that lu Ci/cc of isotopes with greater than 5 year half-lives. Safety Evaluatien Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because implementation of the procedure will not require the use of any equipment important to safety. Also, the probability or consequence of an accident previously evaluated in the FSAR will not be increased. There will be no fork performed during this instruction which may create the possibility of an accident or malfunction which has not been previously evaluation in the FSAR. RWI-115 processing Unacceptable Material Into Drums and Boxes. Unit 0 Reason This instruction provides the method for the processing of unacceptable material into noncompacted drums or boxes. Safety Evaluation l Neither the probability of the occurrence or the consequences of an I accident or malfunction nor the possibility for an accident or l malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the 50 L

basis of any TS was r.ot reduced because implementation of the procedure will not require the use of any equipment important to safety. Also, the probability or consequences of an accident previously evaluated in FSAR will not be increased. There will be no work performed during this instruction which may create the possibility of an accident or malfunction which has not been previously evaluated in the FSAR. RWI-120 Quality Assurance Program for Resin Liner Dewatering, Unit 0 Reason This instruction provides the means to ensure that resin liners ,which have been filled and dewatered meet the disposal site and regulatory criteria for free standing liquid. This instruction is applicable to TVA steel liners, General Electric (GE), cask steel liners, and Chem-Nuclear Systems, Inc., high integrity containers 14-195, 8-120, and 14-170. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this procedure is strictly administrative. It provides the means to ensure that criteria are met. No safety-related equipment is involved. This procedure in no way can reduce the margin of safety. RWI-121 Leak Test for GE Model 589 Shipping Cask, Unit 0 Reason The purpose of this instruction is to perform a test to verify that the GE model 589 cask is leak tight. This instruction applies only to the model 589 cask. The test shall be performed annually or each time the lead 0-ring seal on the cask is replaced. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR l was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because although the shipping cask involved in carrying out this procedure are considered to be critical structures, systems, and components while onsite, they are not connected to any safety-related equipment. Extreme caution is exercised during this test to prevent the possibility of an accident. 51

RWI-122 Sampling Procedures Waste Classification, Unit.0 Reason This procedure describes routine sampling methods to be used to obtain reasonably representative samples of plant solid waste for spettroscopy measurements. The methods were developed to ensure that shipments of radioactive materials from the plant site are properly characterized as to their radionuclide composition and activity levels. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined.in the basis of any TS was not reduced because the implementation of this procedure will not increase the or.currence or consequences of an accident or impair the operation of any safety-related equipment. No equipment will be operated noc will activities be performed which may create an accident. RWI-123 Use of Casks, Unit 0 Reason This instruction provides basic information relating to the use of NRC licensed packages for disposal shipments. Safety Evaluation Neither the protability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this procedure is strictly administrative. It merely outlines the use of casks. No safety-related equipment is involved. Therefore, this procedure in no way can compromise the margin of safety. RWI-124 Utilization of Scaling Factors, Unit 0 Reason To describe methods of utilizing scaling factors to calculate the concentrations of hard to detect isotopes as required by 10CFR61 for determining radioactive waste classifications. I 52

4 Safety Evaluation . Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this procedure is strictly an outline of the method to calculate isotopic concentrations. The implementation of this procedure does not require any work inside the vital' area. Therefore, the possibility of an accident does not exist. > RWI-151 Verification of Radioactive Waste package Contents (Noncompacted) , Reason New procedure to provide for the inspection of radwaste packages containing noncompacted material in order to ensure compliance with all applicable regulations and requirements. j Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the unpacking inspections and repacking of radwaste packages will not in any way impact nuclear safety. This new instruction does not in any way change the facility i or any plant operating characteristic. RWI-152 Verification of Radioactive Waste package Contents (Compacted) i Reason { New procedure to provide for the inspection of radwaste packages l containing compacted material in order to ensure compliance with all applicable regulations and requirements. l Safety Evaluation Neither the probability of the occurrence or the consequences of an ! accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR l was increased or created, and the margin of safety as defined in the l basis of any TS was not reduced because the unpacklag inspections and repacking of radwaste packages will not in anyway impact nuclear safety. This new instruction does not in anyway change the faellity or any plant operating characteristic. 53 _.._.,.,.__._,~___--.__._m - _ . , _ . _ _ _ _ _ _ _ _ _ . _ _ . , . , . . . _ . - . _ _ - . _ _ . - - , -

a PLANT INSTRUCTION REVISIONS JANUARY 1, 1987 - DECEMBER 31, 1987 The following procedures were revised to comply with the BFN Nuclear Performance Plan by ensuring they were properly scoped, formatted, based on the best technical sources, corrected for technical inaccuracies, and written in clear, easy to understand, sequential steps using well defined terms, including applicable commitments, and separated iaformation, cautions, and warnings from actions. Also, unit 2 procedures were reviewed to ensure they detailed the proper steps for the safe operation and shutdown of the plant. Procedures covering a general operation or surveillance were broken down into individual instructions for the applicable unit. 0-AOI-32-1 Loss of 1-0I-71 Reactor Core 2-AOI-57-5A Loss of Control Air, Unit 0 Isolation Cooling System, I&C Bus A Unit 1 0-0I-65 Standby Cas 1-0I-92 Source Range 2-AOI-57-SB Loss of Treatment OI, Unit 0 Monitors OI I&C Bus B 1-AOI-2-1 Reactor Coolant 1-0I-92A Intermediate 2-AOI-68-1 Recirculation High Conductivity, Unit 1 Range Monitors OI Pump Trip, Unit 2 1-AOI-57-5A Loss of I&C 1-0I-92B Average Power 2-AOI-68-2 Jet Pump Bus A Range Monitoring OI Failure Unit 2 1-AOI-57-5B Loss of I&C 1-0I-92C Rod Block 2-AOI-68-3 Recirculation Bus B Monitor OI Loop A or B Speed Control Failure, Unit 2 1-A0I-70-1 Loss of Reactor 1-0I-94 Traversing 2-A0I-70-1 Loss of Building Closed Cooling Incore Probo System, Reactor Building Closed Water, Unit 1 Unit 1 Cooling Water, Unit 2 1-AOI-85-4 Loss of RFIS 1-0I-99 Reactor Protection 2-AOI-85-4 Loss of RPIS, Unit 1 System OI, Unit 1 Unit 2 1-AOI-92-1 RBM Failure 1-SI-4.6.A.6&7 Reactor 2-AOI-92-1 RBM Failure Recirculation Pump Start Limitations 1-AOI-99-1 Lone of Power 2-AOI-2-1 Reactor Coolant 2-AOI-99-1 Loss of Power to one RPS Bus, Unit 1 High Conductivity, Unit 2 to one RPS Bus, Unit 2 2-01-37 Gland Seal Water 3-AOI-92-1 RBM Failure AOI-90-1 Area Radiation System, Unit 2 High, Unit 2 2-0I-70 Reactor Building 3-AOI-99-1 Loss of Power OI-24 Raw Cooling Water Closed Cooling Water, to one RPS Bus, Unit 3 System, Unit 2 Unit 2 54

2-0I-71 Reactor Core 3-0I-71 Reactor Ccre . OI-27 Condenser Isolation Cooling Isolation Cooling Circulating Water System, Unit 2 System, Unit 3 System, Unit 2 2-OI-92 Source Range 3-0I-92 Source' Range OI-34 Vacuum Priming Monitors OI Monitors OI System, Unit 0 2-OI-92A Intermediate 3-OI-92A Intermediate OI-53 Demineralizer Range Monitors OI Range Monitors OI Backwash Air System, Unit 0 2-0I-92B Average Power 3-0I-928 Average Power OI-75 Core Spray System, Range Monitoring OI Range Monitoring OI Unit 2 2-OI-92C Rod Block Monitor 3-OI-92C Rod Block Monitor OI-82 Standby Diesel OI OI Generator 2-0I-94 Traversing Incore 3-0I-94 Traversing Incore SI 4.5.B.14-01 Probe System, Unit 2 Probe System Unit 3 Recirculation Pump Discharge Valve Cycling Unit 2 2-0I-99 Reactor Protection 3-01-99 Reactor Protection SI 4.6.A.5-01 RPV System OI Unit 2 System OI, Unit 3 Temperature Monitoring with Head Tensioned (Cold Shutdown), Unit 2 2-SI-4.6.A.6&7 Reactor 3-SI-4.6.A.6&7 Reactor SI.4.7.B.1.g Standby Cas Recirculation Pump Start Recirculation Pump Start Trectment Filter Limitations Limitations Pressure Drop Test, Unit 0 3-AOI-2-1 Reactor Coolant AOI-57-2 Station Blackout SI.4.7.B.2.d Standby Cas High Conductivity, Unit 3 Unit 0 Treatment System Train Operation, Unit 0 3-A0I-57-5A Loss of I&C AOI-57-6 Loss of SI.4.7.B.3.b Standby Cas Bus A Nonpreferred, Unit 0 Treatment System Filter Cooling Bypass Valve Operability, Unit 0 3-AOI-57-5B Loss of I&C AOI-85-3 CRD System l Bus B Failure, Unit I l l 3-A01-70-1 Loss of Reactor AOI-85-3 CRD System j Building Closed Cooling Failure, Unit 3 ! Water, Unit 3 ! 3-AOI-85-4 Loss of RPIS, A01-85-3 CRD System Unit 3 Failuro. Unit 2 The following procedures were revised in response to the revision of S -1. SI-1 was revised to ensure that TSs were correctly translated into survelliance l 55 l L

acheduling requirements. Thesa scheduling requirements were implemented .8.n the following sis by this revisio . SI 3.1 Inservice Pump Testing SI-3.1.3 RHRSW Pump Performance Required by ASME XI, Unic 6 SI-3.1.1 Cors Spesy Pump Performance SI 3.1. 7 S m' Pump Performance,. Units 1, 2, and 3 SI 3.1.11 EECW pump B1seline Data SI-3.2 Inservice Valve Testing Acquisition and Evaluation, Unit 0 Required by ASME Section XI SI-3.1.12 HPCI System Pump SI 4.2.B-45B RHR System Baseline Data Evaluation Logic-Tic.e Dolay Calibration, Unit 2 only SI-3.1.13 RCIC System Pump SI.4.7.A.3.b Suppre:sion Baseline Data Evaluation Chamber - Reactor Brilding Yacuum Breakars The following procedures were revised in response to an NRC unresolved item in report 84-17-05. TVA evaluated unsupervised carbon dioxide system r,ctuation I circuits to check if additional surveillance was required in order to increase reliability of the automatic carbon dioxide fire suppression syatems. This revision changed performance frequency for part of the SI which deals with CO2 initiation in the DG rooms. Added note to NA sections on semiannual performance due to ir erensed performance frequency. Also, changed local panel number. SI 4.11.C A&5 Fire protection System Testing of Heat and Smoke Detectors, Unit 1 SI 4.11C1&5 Fire Protection System Testing of Heat and Smoke Detectors, Unit 3 TS Amendments 138 (U1), 134 (U2), and 109 (U3) revised sectior. 6.8.1. This revision made provision for any applicable procedure, detailed in Appendix A of regulatory guide 1.33, revision 3 (February 1978), othcr then administrative procedures to be reviewed by process of technical review rather than by way of the Plant Operations Review Committee (PORC). Revision of the following procedures coincides with the amendments to TSs, providing for technical review. SDSP 12.2 Development of System Test Specifications l 56 _ . . - - - _, _ _ _ . . - . - --p,- +

a -- l SDSP 2.1 Site Procedures . and Instructions SDSP 2.11 Implementation and Change of Site Procedures and Instructions. The following procedures were revised in response to NRC report 87-29. This report provides instructions to all personnel involved about the responsibilities pertaining to the control and accountability of special nuclear material. Revisions to the following procedures outline those responsibilities, gMI-35 Traversing Incore Probe Detactor Replacement, Units 1, 2, and 3 SMI 192.2 LPRM Maintenance Instruction, Units 1, 2, and 3 SMI 192.4 IRM System Maintenance Instruction, Unit 1 SMI 192.5 SRM Maintenance Instruction, Unit 1 The followins procedures were revised to correct discrepancies found during review of 31 4.5.C.1. These discrepanc.les include verification of calibration for instruments and tecting frequency inconsistent instruments with TS. These discrepancies are detailed in cos.Jitions adverse to quality reports (CAQRs). SI-4.5.C.1 (3) R:(RSW Pump and Header Operability and Flow Test, Unit 2 SI-4.5.C.1 (3) RHRSW Pump and Header Operability and Flow Test, Unit 0 The following pre edures were revised in response to NEC inspection report 86-38. Report 86-38 details the mispiccement of five fission counters at BFN. The revisions establish better control and accountability of special nuclear 3 material. c SMI 192.2 LPRM Maintenance Instructions Unitt 1, 2, and 3 SMI 19?. 4 IRM Maintenance Instructions Units 1, 2, and 3 51

2 SMI,192.5 SRM Maintenance . j Instructions, Units 1, 2,yand 3 The following procedure was revised to incorporate valve changes made per ECN P0392. Changed frequency from once per operating cycle to once per 18 months, added step to notify unit operator before commencing the instruction, changed to notify shift engineer rather than the assistant shift engineer, added two annunciators (steps 4.6 and 7.5.23), changed range of torque wrench (step 5.21), clarified voltage range (step 5.3.4). SI 4.1.A-8 CAL RPS High Water Level in Scram Discharge Tank The following procedure was revised to incorporate valve changes made per ECN P0392. Added step to notify unit operator before commencing the instruction, changed to notify shift engineer rather than assistant shift engineer, added two annunciators (steps 4.6 and 7.5.16), clarified voltage range (step 5.2.2). SI.4.1.A-8FT RPS High Gater Level in Scram Oischarge Tank Functional Test The following Precedures were revised ln response to NRC IE notice 86-048. This notice outlines generic problems found by NRC daaling with boron concentration and SLC tank levels. A special test (8625) was run as a result of this notice to establish and verify proper mixing of the SLC tank with applied air sparging of the solution. Revision complies with the recommendation from this special test. Revised to reflect the beginning and end air sparge time. CI 463.1 Sampling Sodium Pent 4 borate Solution from the SLC Storage Tank. Unit 0 SI.4.4-08 SLC System, Unit 0 The following procedures were revised in response to NRC inspection report 87-27. This report identified a failure to track and control radioactive byproduct material. Revision made to comply with a commitment to track and control radioactive byproduct material made as a result of this report. SDSP 23.2 Radioactive Source Control, Unit 0 58 4 _ _ ,, _. _ _ . - _ _ _ . - ~ . - _ _ _ - , ., , - , _

SI 4.8.E Miscellaneous . Radioactive Material Sources, Unit 0 The following procedures were revised as a result of commitment in sicensee Event Report 50-259/85050. This is a commitment to review and revise SI-1 to ensure TS requirements are correctly implemented. This revision implements the upgraded version of SI-1 for unit 3. 1-SI-1 Surveillance Program, Unit 1 2-SI-1 Surveillance Program, Unit 2 3-SI-1 Surveillance Program, Unit 3 The following procedures were revised in response to NRC generic letter 82-12. This letter outlined the NRC's policy on factors causing fatigue of operating personnel. Revision made to comply with NRC policy. BF 12.24 Conduct of Operations. Units 1, 2, and 3 SI 4.2.B-32 Instrumentation that Initiates or Controls the CSCS RCIC Steam Line Space High Temperature, Ualts 1, 2, and 3 The following procedure was revised in response to a July 13, 1987 letter from TVA to NRC. The commitment was to revlse SDSP 12.1 and BF 1.10 to denote the i NSSS vendor, GE, as a member of the joint test group. This revision deleted the ! statement that the joint test group subchairmaa shall have the authority to sign i for PORC chairman on prescribed forms. Also, added modifications with an ( associated design change notice or engineering change notice issued (SDSP 8.4) to the review list of the werkplan sub;ommittee. l f BF 1.10 Plant Operations l Review Committee l Unit 0 j The following procedure was revised in response to NRC inspection report 84-26 14. This report questions the QA controls applied during maintenance of open systems to verify foreign maticial is exempted. BF 3.10 revised to require evaluation and when necessary the application of methods to exclude foreign articles from critical plant piping systems. 59 1 , \

                  - ~                    .-              -       .                           .
    )     .

Also, revised to incorporate a temporary change and incorporate revisions due to the periodic 2-year review. BF 3.10 Cleanliness of

                , Fluid Systems, Units 1, 2, and 3 The following procedure was revised in response to NUREG 1000. This NUREG concerns the generic implicationa of ATWS events at the Salem Nuclear power plant. This revision made to comply with items found within NUREG 1000 applicable to BFN.

EMI-7J Overhaul and Test procedure for GE Medium Voltage Circuit  ; Breakers - Types AM-4.76-350/250 and AM-4.76-250 Magna-Blast The following procedure was revised as a result of the completion of a 2-year review as required by SDSp (DR 860622). General revision made as result of this 2-year review. GOI 100.10 Operation with Torus Drained Unit 3 The following procedure was revised to implement the details of ISI-7. ISI-7 1 (paga 363 o SI 4.6,0) outlines the relief from inspection requirements of NRC for inaccessible welds in piping penetrations and under rigid pipe restraints. The flued head to process pipe welds are inaccessible for any type of l examination. The pipe welds located under rigid pipe restraints are inaccessible for volumetric examination. ISI-7 identifies the specific locations at BFN whet. these conditions occur. MAI-23 was revised to include a note outlining the conditions which fall under the scope of ISI-7. l  ! l NAI-23 Support and i Installation of piping Systems in Category I Structures. 1

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The following revision was made to ensure the verification of bolt material crade for QC, verificatioa of bolt tightness, inspection of surface beneath baseplate, and clarification of requiremerns for distance between holes in use and abandoned holes. All revisions made in accordance with the applicable general construction specification G32 revision 12. - r 60 I

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MAI4 Bolt Anchors Set in Hardened Concrete Structures, Units 1, 2, and 3 The following procedure was revised to ensuro segregation of defective equipment from nondefective equipment per Nuclear Managers Review Group (NMRG) report R-86-02-NPS finding L4. This finding notes the possibility of mistakenly using defective equipment in the place of nondefective equipment. This could occur in the case of defective equipment losing its tag and being stored with nondefective equipment. Segregation reduces this risk and is accomplished per revision to MMI-102. MMI-102 Rigging Equipment and Portable Holst Program, Unit 0 The followira procedure was revised to correct sequences and instructions for receiving permission from the shift engineer and unit operator before starting this instruction. Also, revised as result of periodic 2-year review. MMI-17-02 Main Steam Isolation Valve Assemb1;- (Inspection, Maintenance and Repair), Units 1, 2, and 3 The following procedure was revised in response to NRC inspection report d6-32. This report indicated deficiencies in the technique of lifting the refueling platform for inspection and repair. This revision outlined the correct , technique and requirements for lifting the refueling platform, i MMI-34 Refueling Platform Inspection and Repair, Unit 0 The following procedure was revised to delete the requirements for ECN packages l bei

• retained in the shift engineer's office. Also, revised due to j

int ; oration of SDSP 8.2 into SDSP 8.4 (Preparation and Processing of Workplan and inspection Records). PMI-8.2 Plant Design Change Review, Unit 0 The following procedure was revised in response to NRC IFI 87-12-02. This revision ensuras qualification of test engineers, test procedure used is of i latest revision, and establishing controls on test performance. Revision also l incorporated the new conditions adverse to quality (CAQ) program. SDSP 12.1 Restart Test Program 61 l .

The following procedure was revised to change substitution of filler material. Substitution of filler material for welds was found unpermissible in ASME code work where toughness requirements are invoked. Revision prohibits the substitution of E705-6 for E7035-3 filler metal in applications requiring impact testing. (Corrective action for SCR BFN NEB 8616.) SDSP 13.6 Fleid Instruction for Welding Techniques and Repair for G-29M Application The following procedure was revised to provide adequate technical instruction for preparation and processing of FCRs. SDSP 8.9 Field Change Request Unit 1, 2, and 3 The following procedure was revised to incorporate temporary changes and to correct a procedural deficiency. This instruction failed to outline requirements for qualification of the QC inspector (CAQR BFT870669). This revision outlines the qualification requirements for the QC inspector. SDSP 9.6 Mechanical and Instrument and Controls System Walkdown Revision to the fo11owin5 Procedure incorporated signoff verifying that all prerequisites have been met. Revised to require notification of applicable unit operator of testing being performed. Also, revised test apparatus sketch and procedure to reflect change in apparatus. (Revision in accordance with condition identified by CAQR BFQ 870059.) t SI 3.2.9 Testing of Section XI Relief valves Units 1, 2, and 3 The following procedure was revised to correct discrepancies found during review of SI A.5.B.2. The stated discrepancies include inconsistency in requirements for completion of data sheets, nonclarification of qualification requirements for inspectors, and various other procedural discrepancies and inconsistencies. These discrepancies are corrected by this revision. (DR 87-0217) SI 4.5.B Residual Heat Removal System Units 1, 2, and 3 62

The following procedure was revised to correct discrepancies found during review of SI 4.5.C.1. These discrepancies include inconsistency with controlling document, improper logging of SI data, and improper terting techniques (CAQR BFQ870044). SI-4.5.C.1 (2) EECW Pump Operation Surveillance Instruction, Unit 3 The following procedure was revised in response to NRC inspection item 87-02-07. This report concerns the monitoring of the reactor building ventilation. This revision incorporates the following comments made in this report.

1) The SI lacks a note for signoff by Shift Engineer, Assistant Shift Engineer, or Unit Operator to verify equipment lineups.

2) Step 6.4 requires that reactor building indoor air temperature be recorded; however, no temperature instrument was referenced for rete! eval of this data.

3) Step 6.9 states, "Verify indicating lights on panel 9-25 for FC 064-36, the drywell/ torus bypass dampor." This indication is not provided on panel 9-25.

SI 4.7.C-05 Secondary Containment Units 1, 2, and 3 The following procedure was revised to update procedure for tracking charcoal samples from CREV unit. procedure updated to better measure flow rates of CREV units. Clarified steps for start and stop times for CREV units. Also, revised to require notification of unit operator before commencing instruction (LER 259/82032). i SI-4.7.E CREV The following SI was revised to add attachments for alpha activity calculations and steps to tie the SI to the CI. Also, revised to delete a temporary change. l l SI 4.8.B.2-2 Aleborne Effluent - particulate Filter Analysis (monthly gross alpha) The following procedure was re ised to include inspection of the power boost current transformer connections '4R 259/8 7005) . This LER documents the events pertaining '.o the burring out of this connection on March 3, 1987. Revised to include spe:ifics for inspection of cc aact finger and inspection of bolted 63 O, - - - - - , _.r--r - - _ . . , , . - ,-

connection as a result of the consequences of a phase-to-phase short between contacts in the DG control cabinet for the 3ED DG that occurred on April 20, 1987. Also, revision added requirement for check of actuator coupling. SI 4.9.A.1.6 Diesel Generator Annual Inspection Units 1, 2, and 3 The following procedure was revised to clarify RHR seal cooler operability requirements. TI 33-06 EECW Flow Verification Units 1, 2 and 3 The following procedure was revised in response to NRG commitment 86-29. TI-77 was found to contain the following problems.

1) The scope of the document was too broad.

2) Procedure not specific in its directions to the user.

3) Procedure contains instructions inappropriate for a technical instruction.

4) The procedure format requires the user to repeatedly refer to a number of sections within the proceduro in order to acquire all necessary information on a particular task. ,

Corrective action was taken to revise TI-77 and resulted in the deletion of this instruction. Radwaste tasks previously accomplished through TI-77 are now accomplished through a series of RWIs. Each RWI deals with a specific i radwaste instruction. TI-77 Radwaste Packaging and Shipping. l l 64 i

RESTART TEST PROGRAM TEST INSTRUCTIONS ISSUED JAKUARY 1, 1987 - DECEMBER 31, 1987 2-BFN-RTP-002 Main Condensate System Objective The objective of this instruction is to perform testing to verify the proper clean and precoat sequence of the unit 2 condensate demineralizers. Safety Evaluation This test instruction does not involve a change in the f acility design or plant operating characteristics from that described in the FSAR and which could impact nuclear safety. The system being tested, main condensate system, is nonsafety-related and non-CSSC, and the test does not require any operations that deviate from normal. 2-BFN-RTP-023 Residual Heat Removal Service Water system Obiective The purpose of this test instruction is to provide an intergrated systems test based on TS SI of the RHRSW system which are listed below in addition to the backup control test which is describe

• in 2-BFN-RTP-BUC. During the test, valves and pumps operability and syst an logic including the reservoir level monitoring will be verified.

SI-4.2.B.67 RHR Service Water Initiation Logic SI-4.5.C.1(3) RHRSW Pump and Header Operability Flow Test SI-4.9.A.3.a Common Accident Signal Logic System SI-4.2.B-39A Core Spray System Logic SI-4.5.C.1(1) RHRSW and EECW System Valve Operability Test SI-4.5.B.1.c Residual Heat Pomoval System SI-4.2.H.1 Reservoir Level Monitoring Functional Test Safety Evaluation The functions demonstrated by performing this test, residuel

• sat removal service water system, are descelbed in the FSAR and the TSs. fherefore, this test is bounded by the evaluation in the FSAR and plant TSs. Also, this test instruction does not involve a change in the f acility design or plant operating characteristics from that dencribed in the FSAR and which could impact nuclear safety.

65

2-BFN-RTp-024 R0 Raw Cooling Water System - Objective The purpose of this restart test procedure is to specify the functional testing required to demonstrate that the raw cooling water system will meet the design baseline evaluation requirements and minimum operational requirements. This procedure will include verification of proper valve opening and closure upon loss of air and , transferal of system usage. Also, RCW pumps and booctar pumps will be tested. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created because the accident conditions, functions, and tests demonstrated by this test of the RCW system are within the descriptica of the RCW system in the BFN FSAR. The RCW system does not affect TSs. Therefore, the margin of safety for any TS is not reduced by this test. 2-BFN-RTp-025 Raw Service Water System Objective The objective of this restart test program instruction is to provide a description for performing integrated systems testing of the raw service water hip.h pressure fire protection and radwaste as required by the Design Baseline and Verification program (DBVp). The test involves the determination of proper operability of the systems, valves, pumps, and logic including certain pCIS isolation signal responses and local leak rato measurements. Safety Evaluation Neither the probability of the occurrence or the consequences of an l I accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created because the tests performed on these systems are system functions described in the FSAR toe the systems affected. These systems fur.ctions described in this test instruction are within the bounds of the TSs applicable for the system tested. Therefore, this test will not reduce the margin of safety for any TS.

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66 l l t

2-BFN-RTP-027 Condenser Circulating Water . Objective -- The objective of the condenser circulating water (CCW) restart test program test instruction is to verify operability of the crosstle valves, condenser waterbox outlet butteefly valves, and the condenser waterbox vacuum priming system. It also provides test instructions I for verifying instrumentation loops for the continuous warm water 4 channel level and forebay / warm water channel differential level indications in the control room bay. Lastly, the test provides for verifying the vacuum breaking capability of the condenser circulating water system. Safety Evaluation Neither the probability of the occurrenct; or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created because no test conditions are specified or actions will be conducted that conflict with the functions of the CCW or vacuum priming system as described in the FSAR. The TSs are not J affected b/ the test. No margin of safety for any TS is reduced by this test. This test will be performed to meet the safe shutdown i requirements as identified by the DBVP test requirements. 4 i 2-BFN-RTP-030 DG Building and Reactor Building Ventilation System objective ! The purpose of this restart test procedure is to verify proper l operation of the DG exhaust fans with their associated damper motor operators and battery vent hood exhaust fans, determine the air l temperatures surrounding an idling and fully loadins diesel for ) analyzing heat loads, and determining the proper operation of the stairwell and battery rooms exhaust fans. These tests are to verify that the system can meet the functional requirements for mitigation and shutdown from events resulting from modes identified in the scope j of the DBVP. I Safety Evaluation i This test instruction does not involve a change in the faellity design or plant operating characteristics from that described in the FSAR and which could impact nuclear safety. The system being tested l is the DG building and reactor building ventilation system. This test identifies normal system operation. Hence, neither the

probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a l

l 67 i

B different type than previously evaluated in the FSAR.was increased or created, and the margin of nafety as defined in the basta of any TS was not reduced. 2-BFN-RTP-032 Control Air /Drywell Control Air Objective The purpose of this test instruction is to describe the tests necessary to demonstrate and document that the control air system (CAS) and drywell control air system shutdown features can perform their required safety function for the safe shutdown of unit 2. Specific objectives of the tests address testing the CAS primary containment isolation valves for proper closure upon receipt of a PCIS isolation signal, loss of control air, or loss of power; verifying adequacy of the control air supply to the MSRV accumulators and their storage capacity for supporting the automatic depressurization system (ADS) function and the MSIV accumulators for one closing cycle of these valves; and to check the seals and door interlocks for the equipment access door. Leak rate tests are also inherently addressed in this testing. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because test conditions are specified or actions conducted that do not cotiflict with the automatic or manual operation of the system as described in the FSAR and are not different from normal testing or previously conducted tests. Also, testing of the ADS accumulators is performed by MMI-42 (which has been previously evaluated for safety considerations) which requires the unit to be in cold shutdown. This test is bounded by accidents and malfunctions previously analyzed in the FSAR and will be conducted within the bound of the TSs. 2-BFN-RTP-057-1 125V Direct Ourrent (DC) System Test Objective The overall objective of this instruction is to insure that testing of the 125V DC system can provide its design function in support of reactor safe shutdown. This will be accomplished by demonstrating that the 125V DC batteries are capable of supplying their asalsned loads for safe shutdown by performing a battery capacity test and by demonstrating the 125V DC battery chargers' output voltage rippio is within design limitations by monitoring the voltage during battery recharge. 68

Safety Evaluation . Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created because this test utilizes BI-4.9.A.2.c alone to verify proper operation of the batteries. No other test steps are required. Hence, strict adherence to the TS is assured. The FSAR assumes that testing of the batteries will be performed to verify performance, thus this condition is no different than those analyzed for the FSAR. The design of the electrical system and TS allow for systems to be out of service for testing, thus the margin of safety is not reduced. 2-BFN-RTP-057-3 250V DC Distribution System-Unit Batteries Objective This restart test instruction is designed to determine and document that the 250V DC dintribution system-unit batteries perform as required to support the safe shutdown of the reactor. This test includes discharge capacity testing of the units 1, 2, and 3 batteries. The most recently performed battery discharge test data for the units 1 and 2 batteries will be reviewed for acceptability to satisfy the discharge test requirements for those batteries. A ripple test will be performed on 250V DC battery chargers 1, 2A, 2B, and 3. Chargers 1, 2A, and 2B will be tested to verify their functional requirements upon receipt of a 480V load shed signal. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this test utilizes approved ' sis alone to verify proper operation of the 250V DC unit batteries as allowed by the FSAR, section 8.6.2.5 and 8.6.5. The conduct of the test causes sections of 250V DC unit battery system to be inoperable from the TS aspect, but allowances for these conditions are considered by the FSAR and are within the bounds of the TSs. 2-BFN-RTP-057-4 480V Distribution System Objec tive The purpose of this test is to demonstrate the 480V AC dist.ribution system is adequate to support the safe shutdown of BFN as described in the Safe Shutdown Analysis. This will be accomplished by verifying automatic and proper transfer of 480V, reactor motor 59

ope *ator valve and control bay vent boards normal sources to their a1tornate supply of power on loss of voltage to the normal feed. The tesh is also designed to verify automatic load shedding and timo-delayed reenergization of decignated 480V loads under accident conditions (LOP /LOCA) and automatic shedding of designated loads of 480V shutdown boards 2A and 2B under sustained loss of potential con 61tions. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or mt1 function nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in tho basis of any TS was not reduced because this test was designed to confirm design requirements as described in the FSAR for the 480V AC power supply system. No test conditions are specified or actions conducted that conflict with the automatic or manual operations of the system as defined by the FSAR. The conduct of the test causes sections of 480V AC power supply system to be inoperable from the TS aspect, but allowancos for these conditions are considered by the FSAR and are within the bounds of the TS. 2-BFN-RTP-057-5 4.16 KV Distribution System Objective This instruction is designed to demonstrate that the AKV power distribution system can meet the functional requirements for safe shutdown. Specifically, verified during this testing will be that the shutdown buses, unit boards, and the 4KV distribution and common boards

     - will transfer from normal to alternate upon loss of power to the normal source, and that with either the normal or alternate breaker closed the respective open breaker will not close, and that the 4KV distribution system wi'41 provido auto start signals to the RHRSW pumps upon receipt of various designated signals;
     - can procesc reactor recirculation pump and ATWS signals and generate trip signals to the pump and MG set;
     - can effec $. control and indication for DGs from both control rooms and the shutdown boards and successfully parallel the DGs between units and with the grid under normal and emergency conditions;
     - will provide input start signals to the DGs with loss of voltage, with bus voltage degraded, and with common accident signal; 70

                                                            - will provide input to shutdown boards to shed loads and to sequence on preselected loads for shutdown boards after receipt of sustained undervoltage, degraded voltage signal, and LDCA signal when the DGs are providing power;
                                                           - and that the AKV distribution system controls for each shutdown bus and board can be transferred from control room to the respective bus or board and operated and then transferred back to the control room.

Safety Evaluation Neither the probability of tha occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because the test is conducted to verify the performance of the 4KV power distribution system and is consistent with the design as described in the FSAR. No test conditions are specified or actions conducted that conflict with automatic or manual operations of the system as defined by the FSAR. The conduct of the test cauwes sections of 4KV distribution system to be inoperable from the TS aspect, but allowances for these conditions are conwidered by the FSAR and are within the bounds of thu TSs. 2-BFN-RTP-057-7 250V DC Shutdown Batteries Ob_iective This restart test instruction will be performed to document the 4 battery discharge capacity for shutdown board batteries A, B, C, D, ! and 3EB. The shutdown board battery char $ers SB-A, -B, -C, -D, -3EB, and -Spare (portable) will be tested to verify the ripple voltage does not exceed design specification. This test instruction, will be used to determine that the 250V DC shutdown board batteries SB-A, -B,

                                                            -C, -D, and -3EB are capable of supplying the 250V DC loads for safe shutdown and that the shutdown board battery chargers SB-A, -B, -C, j                                                             -D, -3EB and -Spare outputs do not contain excessive ripple.

1 Safety Evaluation ) Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this test utilizes j surveitlance procedures alone to verify proper operation of the i 250V Dc shutdown batteries as allowed by the FSAR, section 8.6.2.5 1 and 8.6.5. During the conduct of the test, sections of 250V DC shutdown battery system will be inoperable from the TS aspect, but allowances for these conditions are considered by the FSAR and are within the bounds of the TS. i 71

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2-RTp-063 SLC System Objective This restart test instruction will be performed in an ef fort to ensure that the SLC system is capable of meeting established operability and safety requirements in support of unit 2 operation. The test instruction includes demonstrating the neutron absorbing effectiveness of the sodium pentaborate solution contained in the SLC storage tanks, verify the operability of the SLC pump 2A and 2B suction header heat trace control loops, demonstrate that the SLC pump 2A and 2B discharge relief valves will lift within the acceptable range to ensure that system over pressurization will not occur, and by functionally testing SLC pump 2A and 2B interlock and controls simultaneous with the operation of each of the pumps in a boron solution recirculation mode. The functional ability of the SLC will be tested by injecting demineralized water into the reactor vessel. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. The SLC is required to be available at all times when it is possible to make the reactor critical as stated in section 3.8 of the BFN FSAR. The SLC is not required to be operable at times when there le no fuel in the reactor vessel. This test will be performed under these conditions. Hence, compliance with the TS and FSAR are assured. The ability to shutdown the reactor without control rods, using the SLC system, is analyzed as a special event in the FSAR. Assumptions and analyses made will not be violated. 2-BFN-RTp-065 R1 Secondary Containment Objective The purpose of this test instruction is to test the SBGT system and secondary containment system in a manner that will verify their design function of limiting the discharge of radioactive material to the environs. The test will ensure that all three SBGT trains start automatically upon receipt of a PC 15 (Group C) signal and that all combinations of two out of three SBGT trains maintain the specified negative pressure while taking suction from zones requiring secondary containment. Operability of the SBGT dampers and components, heaters, filters, f ans, and secondary containment dampers will be de t e rmined. The secondary containment will be tested to determine in-leskage flow and pressure drup across the filters of ebch SBGT train. 72

Safety Evaluation . Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. This test instruction does not involve a change in the facility design or plant operating characteristics from that described in the FSAR and which could impact nuclear safety. The system being tested, secondary containment system, conforms to applicable TSs. f 2-BFN-RTP-067 EECW System Objectivis The objective of this restart test instruction is to perform the testing identified in system test specification which will provide verification that the EECW performs as designed to meet the safe shutdown requirements. Specifically, this test is required to perform or verify that the EECW system will supply essential equipment with the required flow rates and p'. assures 67d that the nonessential cooler supply valve closing setpoints are adequate; the units 1 and 2 control bay emergency chiller unit can be supplied when the control bay chillars vee valved out; EECW water inventory condition in the north ';.eader during a loss of offsite (AC) power event is adequate; operability of the EECW pump discharge strainers and backwash valves, motor a erated pump discharge valves, automatic nonessential equipment supply valves, and chillers cooling water discharge valves; and that either of the two EECW headers are capable of providing water to any unit's fuel pool. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. This test instruction toes not involve a change in the facility design or plant cperating characteristics or procedure from that described in the FSAR and which could impact nuclear safety and is hence bounded by the analysis provided in the FSAR and TS. 2-BFN-RTP-070 Reactor Duilding Closed Cooling Water (RBCCW) System Obj ec tivs 't Performance of the test described in this instruction will result in , functional verification of the proper operation of the RBCCW system and its components. Operations of the RBCCW components from inside 73

e and outside the control room will be tested. Also, operability of spare pump suction and discharge valve interlocks, automatic start of pump B, 3 seconds after failure of pump A to start, automatic closure of the sectionalizing MOV due to low pressure, and the drywell atmosphere cooling system will be determined. Other related tests will evaluate the RBCCW logic and primary containment and prov3de additional supportive information for the overall evaluation of the RBCCW system. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. The test is conducted to verify that the performance of the system is consistent with the design as described in the FSAR, section 10.6. No test conditions or actions are specified that conflict with the automatic or manual operation of the system as described in the F3AR and which provide the basis for the accident analysis in the FSAR. The test is conducted within all TS requirements, although none specifically address the RBCCW system. Thus the margin of safety provided by TS is not reduced. 2-BFN-RTp-071 RCIC System Ob_iective The purpose of this restart test instruction is to describe those tests which are necessary to demonstrate snd document that the unit 2 RCIC system can perform its required safety function as well as perform its design function. This includes the fact that proper operation of each RCIC system motor-operated valve, solenoid valve, air-operated valve, and air-operated check valve must be verified for automatic and manual operation from the control room and from outside the control room as well as proper operation of the RCIC system instrument line excess flow check valves; RCIC system initiation logic with RCIC in the standby readiness configuration and in the flow test configuration utilizing the design input initiatica sigetis; RCIC system isolation logic utilizing the design input isolation signals; RCIC system under flow conditions in the test configuration and that it is capable of delivering rated flow and pressure; and RCIC turbino overspeed trip mechanisms. Other aspects of this restart test requirement include a RCIC system reactor coolant pressure boundary components hydrostatic prer9ure tect RCIC primar," containment penetration isolation valves leak rate testing, and that the ECCS analog trip unito power supplies currently designated as RCIC components supply the correct division I and II components. l 74

Safety Evaluation

• Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced because this tes'c is conducted to verify that the performance of thre system is consistent with the design as describsd in the FSAR, section 4.7 and 4.8.5.2. No test conditions or actions are specified that conflict with the automatic or manual operation of the system as described in the FSAR. This testing is bounded by TS section 3.5.F and tables 3.2.B and 4.2.B.

2-BFN-RTp-075 Core Spray System Objective The objective of this test is to demonstrate that the core spray system (75) performs as designed to maintain the reactor in a safe mode. This includes thorough intergrated tests of the core spray system valves and pumps operability and system logic. Safety Evaluation Neither the probability of the occurrence or the consequencoc of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. This test instruction does not involve a change in the facility design or plant operating characteristics from that described in the FSAR and which could impact nuclear safety. This procedure change doas not add anything new that has not been previously evaluated. The testing described in the procedure is addressed in TS. 2-BFN-RTp-079 Fuel Handling and Storsge Objective The purpose of this test instruction is to functionally test the fuel handling and storage system. The test will verify proper electrical logic whose design intent is to provide safe fuel handling and prevent inadvertent criticality during refueling operations. The I test will ensure that the refueling platform main grapple "loaded," frame mounted hoist "loaded " and monorail mounted hoist "loaded" setpoints are verified. 4 75

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1 Safety Evaluation j

 ;                                                Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. This test is consistent with the design as described in the FSAR as corrected by the baseline program test requirements document for system 079. No test conditions are specified or actions conducted that conflict with the automatic or i                                                 manual operations of the system as defined by the FSAR as mo.fified by the baseline program test requirements document for system 179. The operation of the fuel handling and storage system as condu ted by J                                                  this test are within the bounds of the FSAR (as modified by the baseline program test requirements document for system 079) and TS.

2-BFN-RTP-082 R1 Standby DGs i Ob3ective I The objective of this test is to insure that the standby DGs, the funi oil system, and the diwsel starting air system during the restart test program perform in accordance with design requirements. l Specifically, the test is designed to demonstrate that the DGs will start automatically and manually and thereby provide a 4KV standby source of onsite power; will automatically provide the power when the normal and alternate power sources are unavailable and will override 1, the parallel operation mode selector switch and automatically configure the voltage regulator and governor for single unit

operation; can supply its assigned emergency loads; can reject its full load and its largest assigned emergency load without exceeding l

! designated speed limits as well as operate at full load for at least i 74 hours, and can be resynchronized and paralleled to a DG fed i emergency bus to offsite power, and transfer the load to offsite l power without interrupting the supply to the loads; that the DUs of units 1 and 2 can be operated in parallel with the corresponding DGs of unit 3, and that total loading will be shared equally while l operating in parallei; each DG breaker can be operated from the 4KV ! shutdown board when the transfer switch is positioned in "Emergency" i and all other feeder breakers are open and that the DG breaker control from the main control room is disabled in "Emergency" position only; each bank of the diesel starting air system (DSAS) can r start its associated diesel and has sufficient capacity at minimum normal operating pressure to support the normal and alternate j automatic DG starting sequences; a common accident signal will l initiate proper sequencing of the emergency loads with normal voltage available and all protective functions except the differential l overcurrent and overspeed are bypassed when this signal is present; l the fuel oil systen automatically provides fuel to the diesels, can . l transfer fuel oil between 7-day storage tanks, will operate properly i l

76

on loss of plant control air with service air avallable to transfer fuel oil and be controlled and monitored by attendant level detecting equipment, and that voltage is provided to operate the fuel transfer pumps for continued operation of the DGs; a signal indicating DG cooling requirement is provided to initiate RHR service water pumps;  ! start circuits are operational and can be switched for activating multiple consecutive start attempts; to obtain DC voltage and frequency stability data when both fuel transfer pumps (FTM1 and FTM2) are started simultaneously; and fuel oil consumption rate data. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. This test was conducted to verify the performance of the standby DGs and associated support systems. It is consistent with the design as described in the FSAR. Na test conditions are specified or actions conducted that conflict with the automatic or manual operations of the systems as defined by the FSAR. The conduct of the test causes individual DGs to be inoperable < from the TS aspect, but allowances for these conditions are considered by the FSAR and are within the bounds of the TS. 2-BFN-RTp 085 CRD System Objective The purpose of this instruction is to provide tests needed to verify that the CRD system can meet the safe shutdown requirements and readiness to support startup from the unit ? cycle 5 outage. Reactor manual control nystem (RMCS) and rod position information system (RP1S) testing will also be performed to ensure the reliable operation of components required to support the CRD system. Specifically, the test will verify that the scram discharge system will function to allow reactor scram by providing sufficient volume for CRD over piston area and ge11 leakage water, snd verify the successful completion of modifications performed to the scram ~ discharge system by various ECNs; verify the satisfactory performance 4 of the RFIS by ensuring that displays and logic will operate properly during system operation; verify that the scram pilot air header switches will provide reactor scram upon receipt of low indicated header pressure, the SDIV level instrumentation will provide reactor scram while sufficient volume exists to accommodate control rod scram exhaust water, satisfy the inspection requirements for the CRD housing support installation and alignment, and provide scram insertion times for all 185 control rods during plant startup; . l 1 11

verify instructions necessary to perform rod block functional testing of the RMCS and diagnostic testing for each CRD 60 determine satisfactory performance; verify control response and stability for CRDH system flow control valves replaced by ECM P0596. Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. This test instruction does not involve a change in the facility design or plant operating characteristics from that described in the FSAR and which could . l impact nuclear safety. 2-BFN-RTP-090 Radiation Monitoring System (RMS) Obj ective The functional test requirements provided by this system test instruction outlines tr.e testing which is necessary to determine whether the RMS can provide for safe shutdown from anticipated , tran61ents and accidents. Specifically, the testing will be used to  ; verify thnt the main steam line radiation monitor provides a trip  ; signal to the RPS on high radiation in excess of setpoint and on loss r of voltage to the monitors, verify that the reactor building . ventilation monitors provide an isolation signal to the PCIS on receipt of a high radiation signal and separately on a loss of voltage signal from the monitors, and verify that the refueling zone ventilation monitors provide a trip signal to the PCIS on receipt of a high radiation signal and separately on a loss of voltage signal from the monitore Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accideal or malfunction of a different type than previously evaluated in the FSAR war increased or created, and the margin of safety as defined in the basic of any TS was not reduced. The test is conducted to verify that the performance of the system is consistent with the design as described in the FSAR, section 7.12. No test conditions or actions are specified that conflict with the automatic or manual operation of the system as described in the FSAR. Also, the test is conducted within all TS requirements for the RMS. 18

2-BFN-RTp-244 Backup Control Communications System . Objective The objective of this test is to verify that tne backup conteol communications system can roceive and transmit both signal (howler) and voice as required from primary and alternate sound powered stations. This includes verifying each fixed station sound powered telephone operates properly for transmission and reception, is connected to the system only when push to talk button is depressed, selector switch and howler operates, and can transmit and receive voice transmission using portable sound powered chest sets tied to the system. , Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility fer an accident or malfunction of a different type than pceviously ovaluated in the FSAR was increased or created, and the margin of safety as defined in the basis of any TS was not reduced. The communication system has no direct interface with any of the systems described in Chapter 14 or Appendix G of the FSAR accident analysis. Alsc, there is nc accident or malfunction considered in Chapter 14 or Appendix G of the FSAR that concerns this system. The communications system is not addressed in the TS so a test of this system will have no affect on the margin of safety. 79

CHANGENOTICESFORRESTARTTESTPROCEDUREhISSUED JANUARY 1, 1987 - DECEMBER 31, 1987 2-BFN-RTP-0.14 R1 Raw Cooling Water (RCW) System Reason The purpose of this restart test procedure is to specify the functional testing required to demonstrate that RCW system will meet the design baseline evaluation requirements and minicum operational requirements. Unito 1 and 2 control bay air chiller temperature control valves are operable and fail open on loss of air. RCW system provides pressure boundary integrity to EECW by verifying that normally open RCW check valves shut when flow in transferred from RCW to EECW and EECW header pressure is maintained. EECW alignment will ensure the autostart on a simulated RCW header low pressure. RCW pumps ID and 3D backup control from outside the control room is operable. RCW pumps auto start on low RCW discharge pressure and auto stop when RCW discharge pressure is sufficient. RCW booster pump auto start and auto stop functions are operable. Long layup time and corrosion problems have not degraded the RCW heat exchanger and cooler flows snd that the component flow is sufficient to support unit startup. RCW pumps and RCW booster pumps supply adequate discharge head.

Safety Evaluation Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or

malfunction of a different type than previously evaluated in the FSAR i

was increased or created because accident conditions, functions, and tests demonstrated by this test are in the description Jf the RCW system in the EFN FSAR. The RCW system does not affect TSs. i Therefore, the margin of safety for any TS is not reduced by this test. 2-BFN-RTP-025 CN-1 Raw Service Water (RSW) System Reason Because of installed instrument locations and the lack of installed i flow elements, volumetric flow obtained is suspect. Also, given that

                                       ' shutoff head' measurements meet acceptance criteria and that a                                                                                                                                                             i; single RSW pump can raise level in the RSW head tank, flow curve

! comparisons are unnecessary. Hence, this test instruction was revised to delete the requirement for the RSW pump flow to be within 50 spm of the design curve. i 80 1

Safety Evaluation . This change is bounded by the approved unecviewed safety question determination (USQD) and does not constitute a change in procedural detail in the FSAR or TS or any other safety analysis that supports a licensing document. Although this intent change deleted flow rate acceptance criteria for the RSW pumps, flow rate for the RSW pumps is not addressed in the FSAR (section 3.11) or the TS. This test does not confirm a safe shutdown requirement or mode for the RSW system.

  .s 2-BFN-RTp-057-3 CM-1 250V DC Distribution System-Unit Batteries Reason This change notice was issued to implement a tighter ripple criteria specification for testing the 250V DC battery chargers, f

Safety Evaluation This change is bounded by the appro 'ed USQD and does not constitute a change in procedural detail in the FSAR or TSs. The change does not alter overall test objectives or any of the conditions established during testing. It does, however, tighten the acceptance criteria from "two percent or loss" to "one percent or less for ripple." 2-BFN-RTp-057-4 CN-4 480V AC Distribution System Reason The reason for this intent change was to increase the neope of the test to include 480V shutdown board for units 1 and 3 in addition to those for unit 2. All 480V shutdown boards have been determined to be required for unit 2 safe shutdown. Therefore, instead of having to test only unit 2 shutdown boards, units 1 and 3 shutdown boards must be tested and verified operable. Safety Evaluation This change is bounded by the approved USQD and does not constitute a change in procedural detail in FSAR or TSs because this intent change does not change test objectives, acceptance criteria, or any of the conditions established during testing, 81

2-BFN-RTP-057-4 CN-5 480V Distribution System Reason This intent change notice was issued to add instruction clarification and verification and documentation of the 480V control bay vent board A and the 480V load shedding capabilities of the 480V distribution system. This revision also included changes of the acceptance criteria to agree with test requirements documentation. Safety Evalaution This test instruction revision does not involve a change in the , facility design or plant operating characteristics from that described in the FSAR and which could impact nuclear safety. It is bounded by the approved USQD for the origiaal instruction. The changes to the system being tested, 480V distribution system, involves use/ additions to the PORC approved procedure and does not exceed the bounds of the design bases. 2-BFN-RTP-057-4 CN-6 480V Distribution System Reason The basic test instruction (2-BFN-RTP-57-4) is performed concurrently with SI 4.9.A.1.b. The SI initiates the load shedding and this test instruction is used to document the events. Hence, this test instruction was revised because of SI 4.9. A.1.b revisions to accommodate load shed documentation. Safety Evaluation This test instruction does not involve a change in the facility design or plant operating characteristics from that described in the FSAR and which could impact nuclear safety. The change being made to the system tested, 480V load shed logic which is initiated by SI 4.9.A.1.5.4, will not exceed the bounds of the design bases, t 2-BFN-RTP-057-5 CN-17 4.16 KV Distribution System l l Reason l This change notice was issued to correct specific alignment and l restoration steps regarding the configuration of the bus tio board ! and allow the test to be performed in its abnormal configuration. l This instruction change included operation requirements for board transfer change, clarification of procedure to conform with TVA nomenclature change, and correction of chart speed. l 82

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Safety Evaluation . 1 Neither the probability of the occurrence or the consequences of an accident or malfunction nor the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR was increased or created because the configuration of the plant electrical distribution system is in an analyzed condition. In worst case, a loss of all offsite power or en electrical fault on either shutdown bus or the bus tio board, the shutdown boards would be isolated and start the DG just as they would in normal configuration. TS 3.9.B.5 allows all unit 1 and unit 2 shutdown boards to be fed from one shutdown bus for a limited amount of time. Therefore, the margin of safety will not be reduced.

<                             2-BFN-RTP-057-5 CN-21 4.16 KV Distribution System Reason This change notice was issued to revise the extent of unit 3 diesel load acceptance testing as previously described in this test ins t ruc tion. Specifically, the test will be limited to residual heat removal service water (RHRSW) pump sequencing and shedding. This limited testing in conjunction with unit 3 condition will not allow performance of SI 4.9.1.A.1.b emergency load acceptance test. Unit 2                 ;

restart testing required will be contained within 2-BFN-RTP-082 and will be appended to this procedure. Safety Evaluation i This change is bounded by the approved USQD and does not constitute a change in procedural detail in the FSAR or TS. This chango does not alter test objectives, acceptance criteria, or any of the conditions established during testing as analyzed in the existing USQD. 2-BFN-RTP-057-5 CN-22 4.16 KV Distribution System Reason 4 The purpose of this change was to delete test instruction steps which were not required for the test performance and documentation of test results. These steps dealt with the use of a SI review cover sheet , to document the performance of supporting tests. , 83 j

l l Safety Evaluation - l l This change is bounded by the approved USQD and does not constitute a change in procedural detail in the FSAR or TS. This intent change does not change test objectives, acceptance criteria, or any of the conditions established during testing, l

,     2-BFN-RTP-057-5 CN-23 4.16 KV Distribution System Reason Steps were deleted because testing as originally proposed in supporting tests (DPSO-SMI 1-A.4 through 1-3ED.4) did not cover, load shed properly. Current DPSO-SMI test instructions now cover this requirement. Thus, this instruction change deleted steps which became redundant and served no useful purpose. Also, typographical errors were corrected by this change notice.

Safety Evaluatiott This change is boJnded by the approved USQD and does not constitute a change in procedural detail in the FSAR or TS because this intent change does not change test objectives, acceptance criteria, or any of the conditions established during testing. 2-BFN-RTP-057-7 CN-1 250V DC Shutdown Batteries Reason This change notification war issued to implement a tighter change ripple criteria specification for testing the 250V DC battery chargers. Safety Evaluation This change is bounded by the approved USQD and does not constitute a change in procedural detail in the FSAR or TS. The change does not alter overall test objectives or any of the conditions established i during testing. It does, however, tighten the acceptance criteria from "two percent or 1 css" to "one percent or less for rippic." 2-BFN-RTP-070 CN-1 Reactor Building Closed Cooling Water (RBCCW) i Reason 1 ! The test objective to verify primary containment integrity by means l of a 'ocal leak rate test was deleted because it is no longer a test i l i 84

for requirement for the RBCCW system. primary containment integrity will be verified by testing the containment penetration system. Safety Evalaution This test instruction does not inialve a change in the f acility [ design or plant operating characteristics from that described in the L FSAR and which could impact nuclear safety. The section deleted is no longer a requirement of this test. 2-BFN-RTP-082 CM-4 Standby DGs Reason Fuel oil transfer pump steps were deleted and replaced because the initial steps in the procedure did not result in properly priming the transfer pumps. The revised steps utilized an air hose priming method. This also required replacement of existing data sheets 7.1 thru 7.8 to correspond to these revised steps. Minor procedural clarification was also made to properly describe the stop switch as a pair of stop buttons. Safety Evaluation This change is bounded by the approved USQD and does not constitute a change in procedural detail in the FSAR or TS. This intent change does not change test objectives, acceptance criteria, or any ot the conditions established during testing. 2-BFN-RTp-082 CN-10 Standby DG Reason This revision was made to address the calibration of overcurrent relay in accordance with preestablished DPSO procedures. Safety Evaluation This test instruction does not involve a change in the facility design or plant operating characteristics from that described in the FSAR and which could impact nucicar safety. Relay setting changes allow performance of the emergency power system-diesci lubo oli subsystem tests previously evaluated in the USQD for 2-BFN-RTP-082. i I l 85 i 3

2-BFN-RTP-082 CN-17 Standby DG + Reason This revision was made to ensure proper load testing of DG 3A by including the 480V shutdown board 3A from alternate to normal feed transfer in accordance with 0I-57. i Safety Evaluation This test instruction does not involve a change in the facility design or plant operating characteristics from that described in the FSAR and which could impact nuclear safety. The change does not interfere with operating characteristics of the plant or require data to support DBE test requirements. This change is bound by the USQD for system test 2-BFN-RTP-082 R0. i 2-BFN-RTP-090 CN-05 Radiation Monitoring System Reason This revision was issued to ensure that the test requirement to verify that a trip signal is sent to the CREV system on receipt of a simulated high radiation signal was properly cross-referenced to another instruction. Hence, the requirement to satisfy this step is included in another procedure. Safety Evaluation This change is bounded by the approved USQD and does not constitute a  : , change in procedural detail in the FSAR or TS because this intent  ; 4 change does not change test objectives, acceptance criteria, or any of the conditions established during testing.  ; { 1 J

2-BFN-RTP-090 CN-3 Radiation Monitoring System Reason i

Test instruction step 5.1.5 was inadvertently lef t out during s preparation and has been added in this revision. This step adds the use of a SI to assist in testing the reactor auto scram channels A-D initiation signals to the RPS logic from main steam radiation i monitors upon receipt of high radiation in excess of setpoint signal. i 1 l 4 86

        ~ --                                                , - -        e n              - , - , , ,,-,---------r ,- --n- ,    ---. ..,c   , ,--+---,,e  ,-- --,.w -
                                                                                                                                                                      --w

i Safety Evalaution . This change is bounded by the approved USQD and does not constitute a change in procedural detali in the FSAR or TS. This change does not change test objectives, acceptance criteria, or any of the conditions established during testing. 2-BFN-RTp-090 CN-06 Radiation Monitoring System Reason Test objectives were added to verify that all main steam line and all reactor building ventilation exhaust line radiation monitor channels respond when exposed to gamma flux sources. This verification was added because the test objectives, test descriptions, and acceptance i criteria were not clear as to the requirement to check i detector / monitor response. Changes were made to clarify the procedure. Safety Evaluation This change is bounded by the approved USQD and does not constitute a change in procedural detail in the FSAR or TS. This change does not change test objectives, acceptance criteria, or any of the conditions established during testing. i 2-BFN-RTP-090 CN-07 Radiation Monitoring System Reason This change incorporates the revised acceptance criteria and lists i the steps to test the requirements. Verification that a PCIS group 6 i isolation signal overrides all manual-electric operations of valves with the exception of emergency open steps were added because of the acceptance criteria changed during the revision of the test, but the changed criteria was not incorporated into the test Latil this change notice. Step 6.24, verification that the isolation valves of the drywell radioactive particulate, iodine, and gascous monitors close on receipt of group 6 isolation signal from PCIS was deleted because this step does not apply to this test and requirement to perform this step was transferred to 2-BFN-STS-p064A. The use of SI 4.2.A.10 was deleted as the instruction for verifying that reactor building ventilation monitors provide a (reset) signal to the PCIS when the monitor loses voltage. 1 l 1 i f 81

p Safety Evaluation . l This test instruction does not involve a change in the facility design or plant operating characteristics from that described in the 4 FSAR and which could impact nuclear safety. The test instruction < changes are still within the scope of the criginal USQD. l 3 I s  ; a , l i 1 j i 4 1 i i 88 , i . u I

  - , _ _ _     . _ _ _ . _ , , _ _ . _ _ , _ ~ _ . . . _ _ . _ , _ . - , _ . , . , _ , .._._,..___ . _ _.- _ _-, _ ...__,__                                                                       _,, , . _ - - _ , , _ .         _.

1987 REEEASE

SUMMARY

I Eiseous Releases  ! Eiquid Releases i Fissions & I lodines l Particulates l Tritium i Gross 1 Tritium i Dissolved i Gross i MON 1H l Activation i l l l Radiodctivity l l Noble Gases l Alpha i I (C1) I (C1) l (C1) 1 (C1) 1 (C1) 1 (CT) I (C1) l (C1) l I I I I I I I I I January I <1.21E 02 1 <1.88E-04 1 <8.99E-04 1 1.86E-01 1 2.82E-02 1 3.92E-01 l <1.02E-03 1 <2.45E-04 I I I I I I I I I I february I <1.21E 02 1 <3.57E-04 1 1.30E-04 1 2.37E-01 1 1.46E-02 l 2.46E-01 l <6.20E-04 1 1.91E-04 I I I I I I I I I I March I 3.10E-05 1 <2.91E-04 1 3.69E-04 1 3.01E-02 1 2.26E-02 1 2.17E-01 1 <9.90E-04 i <1.75E-03 I I I I I I I I I I April  ! 8.16E-03 1 <5.16E-04 l 1.22E-04 1 2.76E-02 1 2.72E-02 1 1.72E-01 1 <6.19E-04 i <1.42E-04 l 1 1 I I I I I I I May I <9.44E 01 1 <9.63E-05 1 3.99E-05 1 5.69E-02 1 2.50E-02 1 1.62E-01 1 <1.08E-03 1 <1.05E-04 I 1 I I l l l l l 1 June I <1.41E 01 l <6.10E-04 1 2.28E-04 1 1.28E-01 1 3.82E-02 1 1.74E-01 1 <1.84E-03 l <1.13E-04 I l l l l 1 I I I I July I <1.08E 02 1 <1.32E-01 1 <7.40E-02 1 1.52E-01 1 2.67E-02 l 1.70E-01 1 <1.80E-03 1 <1.80E-04 1 i i i i l I i i I August I <1.04E 02 1 <6.69E-04 1 3.63E-04 1 9.2SE-02 l 2.64E-02 l 1.61E-01 1 <1.19E-03 1 <1.71E-04 I l I I i i i i I l September I <1.46E 02 1 <2.18E-04 1 <6.87E-04 1 9.88E-02 1 2.12E-02 1 8.89E-02 1 <8.26E-04 1 2.79E-04 l l l 1 1 I I I I I October I <9.38E 01 1 <8.46E 01 1 3.01E-04 l 8.69E-02 1 5.02E-02 1 1.06E-01 1 <1.89E-03 1 4.39E-04 I I I I l l 1 1 I I November i <1.77E 02 1 <3.78E-05 1 4.30E-05 l 6.37E-02 l 2.94E-02 1 8.74E-02 1 <1.15E-03 1 1.10E-03 1 I I I I I I I I - I December i 3.14E-01 1 <3.50E-05 l 1.88E-04 l 2.00E-02 1 2.47E-02 1 4.98E-02 1 <9.04E-04 1 6.52E-04 l 1 I _ i I i I I I i Variation in the data for gaseous releases have been correlated with the numbers of operating fans. There were no excursion of interest nor releases which exceeded TS limits. - 89

TRANSMISSION LINE CORRIDOR HERBICIDE USAGE JANUARY 1, 1987 - DECEMBER 31, 1987 The herbicide Spike (dry flowable [DF)) was used on a section of the BFN-West Point 500-kV transmission line in calendar year 1987. Spike

    >      DF is a preemergence herbicide manufactured by Elanco Products Company under environmental protection agency (EPA) registration No. 1471-147 (specimen label'and product safety data cheet attached).

The BFN-West Point 500-kV transmission line right-of-way is 200 feet wide; however, vegetation is controlled on a 150-foot wide section leaving a 25-foot wide buffer rone on each outer edge of the

          .right-of-way. The right-of-way is maintained by a combination of mechanical clearing and herbicide application.

The herbicide Spike (DF) was applied on approximately 10 miles of this right-of-way by the banding method to eliminate danger to trees and to maintain the ISO-foot right-of-way. The banding method consists of applying a single band, approximately 2 inches wide, along each outer edge of the cleared section of right-of-way. A total of 122 pounds of Spike (DF) was applied by the banding method of 24.24 acres at the rate of 5 pounds (4.25 pounds active ingredients) per acre with water as the carrier. The combination of mechanical clearing and herbicido treatment has proven effective and cost efficient in controlling undesirable vegetation along transmission line rights-of-way. l j 90

t: ELANCO . Herbicide 8pika Dry Flowable ( A preemergence and postemergence herbicide E:g=== for control of brush and .....e.,,,...,m,, w....,.............. weeds in such areas as: - - . ..

                                     ~_.

A_a ssa~ rynts <if-w a t u~er as a a~ -o,e

                                                                           ,..e-.....

Lancscao.ng is parred e

                                                                                                             -s_, s u m r,..s Ra.ecad yucs Tum tarms                         At me ease of hviaay                  Dity.c ara s
                                     %ay roarecs and banast                guatora.ts, syqcs:s, and          Fret, reams itanoac rqnt.s+t way                  rrarm er                         Fence rows Fbas sacuders enero re            At the base of tanswsson ws98EJuG1 ts Cesdod                lO*efsandPOes TTe oenrae are cbrat.on of control rray vary with tre a.e of cre% cal arpi ed. Sod teature, and l                                     h EFv74Gert.

seraEunaron; ht5-(t,1drhethy.etnyl).1,3 Othtad.1201-2 y@NN4methyturea . .

                                                                                                                                , 85 0*.

M svyeoent: . , , t 5 @N Cersaris 3 4 pounds a:two ingrec ent cor 4 counds $PlKE e-ine reg stores tracerr.c.rit for Canco Prauts teoweason WARN!P+G: KEEP CL.1T OF RE ACH CF CHiLOREN See caca panel for addtenal cawicn statsswanti Net Weight 4 Pounds EPARg te. 1471147 k Elanco Products Company . A Division of Eli Lilly and Company . Indianapolls,IN 46285, U.S. A. 12

AVO!D NONTARGET DRIFT OR PRODUCT MOVEMENT. PRECAUTIONARY STATEMENTS DO NOT APPLY WHEN WINDS ARE GUSTY OR UNDER  ! ANY OTHER CONDITION WHICH WILL ALLOW DRIFT OR PRODUCT MOVEMENT. DO NOT APPLY TO AREAS WARNING (, Human: Harmfulif swallow'ed. Avoid contact with skin, eyes or WHERE SOlt MOVEMENT BY WATER EROSION AND.OR NATURAL OR MECHANICAL MEANS IS LtKELY. DO NOT , clottung. Avo>d inhaung dust from product in cze of contact, APPLY TO AREAS WHERE WIND IS LIKELY TO CAUSE ffusn wun water. SOIL MOVEMENT UNLESS A SOIL SEALANT IS USED. DRIFT OR ANY FORM OF PRODUCT MOVEMENT FROM Environmental: Do not contaminate any body of water. TREATED AREAS MAY CAUSE DAMAGE TO ANY VEGE-ponos or streams as death or intury may occur to vegetation TATION TO WHICH TREATMENT IS NOT INTENDED. imgated by such. Do not ccataminate water by c!aaning of , ecu:pment or d:sposal of wastes. Oltchbank Usage - Do not apply SPIKE Dry Flowable to any pomon of the ditchbank that will come into cirect contact with Storage and Disposal: The hero: cidal properties of SPIKE water as movement of SPIKE Dry Flowable in trus water to Dry flowabte require caution in hanoling, storage and non target plant species may result in the injury or death of transportation of this product Do not contaminate food or feed those plants. Do not apply on ditches used to transport irnga-by s,orage or cssoosal Open dumping is prehested. Do not tion or potable water. Keep from contact with other pesticides reuse emp y container. Dispose in an incinerator or land fdl and seeds. accroved for pestcade containers or bury in a noncropfand area away ficm desuable~ plants, trees and water supply. Thoroughly clean a!! traces of SPIKE Dry Flowable from apphcation ecuipment atter use. DO NOT EMPTY RESIDUES The manufac:urer makes no warranties, express or implied, CLEANED FROM APPLICATION EQUIPMENT ON AREAS concemsng tNs product or its use which extend beycnd the WHERE THEY MAY COME IN CONTACT WITH THE desenp: ion on the label. All statements made conceming tms ROOTS OF DESIRABLE VEGETATION OR THE WATER product apply onty when used as directed. SPIKE Ory Flow- SOURCE FOR SUCH VEGETATION. aele must be appned according to Elanco's wntten instruc-bons, indLcng, but not hmited to, recomtnend ed rates. Failure Woody Plant Control- Grazing is allowed in areas receiv-to do so may result in poor weed control or plant injury. Elanco ing band or individual plant treatments with 4.70 pounds per acre or less of SPIKE Ory Flowable. In areas recetvi.ng band or expressly crsclaims any warranty, either express or implied, indmdual ptant treatments with 4.70 pounds per acre or less of for the use of SPlKE Ory Flowable, alone or in combination with other products, when that use is not in sinet compliance SP(KE Ory Flowable, grass may be cut for hay one year after application. wt:h Elanco's wnr.en recommendations. SPIKE Dry Flowable may injure or suppress certain herba-ceous vegetation in the treated area. Therefore, do not apply w nere suen injury cannoi te tolerated. Do not appry brosecast PRECAUTIONS armam d SME % Rwade where brage or mae (- SPIKE DRY FLOWABLE IS INTENDED FOR nance of a grass com is desm M m mesmemace t NONCROPLAND VEGETATION CONTROL IT ' # #' ** * *PP

• E*"**

I l IS AN EXTREMELY ACTIVL HERBtCIDE 8 9" "*

• IL WHICH WILL KILL TREES, SHRUBS AND OTHER FORMS OF DESIRABLE VEGETATION HAVING ROOTS EXTENDING INTO THE TREATED AREA. DIRECTIONS FOR USE: Read All FEEDER ROOTS OF MANY SPECIES OF DESIRABLE DirectlOnS Careful ly BeiOre Applying.

VEGETATION EXTEND MANY FEET BEYOND THE DRIP , LINE OF THE BRANCHES, AND A VERY SMALL AMOUNT Total Vegetation Control: SPIKE Dry Flowab!e is a pre-OF SPIKE Cry F'iowab'e IN CONTACT WITH ONE FEEDER emergence and postemergence heroicide for total control of ROOT OF ATREE, SHRUB OR OTHER DES!RABLE VEGE- vegetation in such noncropland areas as: airport runways, TATION MAY CAUSE SERIOUS INJURY OR DEATH TO utihty substations and rights-of way, and concrete pavements THE ENTIRE PLANT, where no future landscaping is planned, at the base of high-way guardraits, sign posts and markers, at the base of trans. AN ARBCRICULTURIST (TREE EXPERT) SHOULD BE mission towers and poles, around industnal building 5, lumber-CCNSULTED TO HELP YOU TO DETERMINE IF THE AREA yards, railroad yards, ditchbanks, firebreaks, and fence rows, CF PROPOSED APPLICATION IS FREE OF ALL ROOTS OF DESIRABLE VEGETATION. THE EFFECT OF SPIKE Ory For total vegetation control in areas not treated the previous Fbnade ON DESIRABLE VEGETATION MAY BE tRRE- season with SPIKE Ory Flowable or other residual hercicides, VERSIBLE AND ITS PRESENCE LN THE SOIL MAY PRE- apply SPIKE Dry Flowable pnor to or just after emergence of VENT GRCWTH OF OTHER DESIRABl E VEGETATION plants as follows: FOR SCME YEARS AFTER APPLICATION. 4 5 pounds per acre, SPIKE Dry Flowable wdl control the READ THE EffTIRE LABEL BEFORE USING SPlKE Ory hi Fbwable TO DETERMINE IF THIS PRODeCT IS SUITABLE FOR THE DESIRED PURPOSE. f,'t rf heath t e gra s Do not use SPIKE Dry Flowable on areas such as wa%s, Aster, wh:te heath Burc!over Bar:ey, htt!e Buttercup, sma!!!!ower drueways, streets, lawns, patios, tennis courts, swimming Bedstraiv Camphorweed pools, cemetenes, or other landscaped areas, or unest Carrot, wi!d asonart or concrete pavement where future tanoscaping is Biuegrass, annual Bluegrass, Kentucky Catsear, spotted 4( j planned. Do not apply on field crops. Do nct apply on any area into which the roots of field crops or other desirable vegetation Bouncingbet Cheat may extend. ROOTS OF TREES, SHRUBS. AND OTHER Bromegrass, downy Chickweed Bromegrass, hpgut Clover, red DESIRABLE VEGETATION MAY EXTEND FAR BEYCND Bromegrass, smooth CocMetsur THE DRIP UNE OF THE PLANT'S BRANCHES. 13

Creeper, Virge.,a Nightshade, say:rleaf P pperweed Swe:telover Crowfootgrass Oat,wdd Pigwied Thistb, Canada Dock, curty Parucum, Texas Ragweed, common Woodsonet, yeilow Dogfennel Pepperweed, Virginia Smartweed, Pennsylvania buckhom At 3 pounds per acre, SPIKE Dry Flowable will control the - (; Fs , rattast an

                                                                                   * *9 '

Fido!eneck, coast Puncturrone Ragweed, giang Goldenrod Spurge Filatee Filaree. redstem Raspberry, red in areas of rainfall greater than 25 inches per year, the 3 Fleabane, annual Ryegrass, Italian pounds per acre maintenance rate should be used for all weed Foatail Sedge, annual species listed above.

• Gaularcia, rosenng Shepnerdspurse Geraruum, Carotma Sida, pncMy For the mamtenance of total vegetation controlin noncrop.

Go6 dented Sowitwstle, annual land areas west of the Rocky Mountams which were treated Grape Spikeweed the previous season with SPIKE Ory Flowab!e or other resid-Gumweed Spurge ual herocdes, apply SPlKE Dry Flowable poor to or just atter Hemocx, poison Spurge, spotted emergence of plants as fonows:(Some of the species listed Her. ort Starintstte, yedow may show erratic control depending on the time between HoneysueMe, Japanese Strawberry application and weed germination.) pYe At 1.5 pounds per acre SPIKE Dry Flowable wdl control the i ptan Kocha Timothy ICUC*i"9 Lamoscuarters Trumpetereeper Bassia, fivehook Pigweed Lupme Velvetgrass Cheat Plantain Mecac, blac$( Vetch Cudweed Ryegrass, annual Mommggiory Witchgrass Fortad Saltbush Mulletn, common Lettuce, prickly Shepherdspurse Oat, wild Witchgrass At 7.0 pounds per acts, SPIKE Dry Flowable wdl control the Oxtongue, bristly I U#*"9# At 2.0 pounds per acre SPIKE Dry Flowable will control the Awngrass, triple ivy, poison gonowing: Barley, fortad Reed, common Buttercup Mustard Bromegrass, Japanese Sandbur, field Ragweed, westem Smartweed, swamp Canarygrass, reed Canarygrass, reed Knapweed, Russian Starthistle, yellow Carperweed Sowthistle, perennial Telegraphplant Knotweed Chicory Spurge, prost ate g0# (1 Cmquefoil, common Sumac, staghom Sweetetover, white At 3.0 pounds per acre. SPIKE Dry Flowable will control the followog: At 9.5 pounds per acre, SP!KE Ory Flowable will control the Barley Sida, a!ka!i fouowmg: Gumweed Smartweed, swamp Bamyarograss Oxtongue, bristry Puncturevine Pu a , common in areas of rainfali greater than 25 incnes per year, the 3 ass pounds per acre maintenance rate should te used le all weed Jornsongrass seec!ing Ragweed, common species listed above. Lovegrass Sattbush APPLICATION DIRECTIONS At 18.75 pounds per acre, SPIKE D y Flowable will control the Appiy SPIKE Dry Flowable in 15 to 150 ganons of water per fcDowmg " acre before or dunng the penod of active growth of plants to be Bermudagrass Lookinggfass, Venus controlled, trutial controlis enhanced by ramfalt. Cwweed Vaseygrass

• Dalhsgrass in areas of low annual rainfall (less than 15 inches per year) wa le s e app rt e tm of year For the maintenance of retal ye9etacon controt in noncrop- when the predominant portion of that rainfall occurs, A mmi-land areas east of the Rocky Mountams wtuch were treated - mum of 1 to 1 % 1nches of rainf atils required to actrvate SPIKE the previous season wth SPIKE Dry Flowabte or other resi- Dry Flowable and place it in the pnmary weed seed germina-cual hert= cades, apply SPIKE Dry Flowable pnor to or just att er tion zone, emergence of plants as foHows:(Some of the species hsted may snow enatac control depending on the time between To maximize performance under a vanery of conditions, application and weed germination ) SPIKE Dry Flowable is recommended for use in tank mix combmation with any of the following products; amitrole, At 2.0 pounds per acre, SPlKE Ory Flowable wdl control the Atratol8 80 W. Banvel' 720, Banvel* W.S., Karmex
• followmg: .

80W, MSM A, paraquat, Princeps 80W. Roundup *, SURFLA?4' 75W. 2,4 0 or Oust'. Where appheations are Bluegrass, annual Fleabane, annual made to existmg vege.ation, the contact or buming prc,perties Bluegrass. Kentucky Horse *eed of paraquat or MSMA provide the rapid top kill while SPIKE (, Carrot, wdd MuHein Dry Flo*aue gives the residual Long term control desired. Cruckweed, common Panicum, fatl Apphcations to areas wh,ch are mtested with certain d.tticult to Croton Parsnip, wild control perentual weeds (such as Johnsongrass, termuda. 14

SPIKG Ory Flowabis applied at tha rate of 1.25 pounds per grass, quackgrass, horsetail, bindweed, dandebon or acts wat control the following spect:s. nutsedge) will bemfit from tank. mixing SPIKE Ory Flowable

         ' with Roundup, amitrole. Banvel, Banvel 720. or 2,4 0. The                                                                                                    Haplopappus tenussectus          (Burroweed) addition of SURFLAN. Oust, Karmex, Atratol or Pnncep wdl                                                                                                   Larrea indentata                 (Creosotebush) improve SPIKE's pertormance on certain annual broaoleaf                                                                                                     Afimosa bruncitera               (Wait a. minute bush) and grass weeds such as foxtail, kochia, Russian thistfe, or

(. seedting johnsongrass. SPIKE Ory Flowable applied at the rate of 2.50 pounds per

• acre will control the following species.

Read the SPIKE Ory Flowable laoel and labels of products to Ai/anthus a/Dssima (Tree of. heaven) be tank mixed carefully before usmg. Note all wamings, Aloysia lycloides (Whitebrush) caut ons, precautions, and limitations of warranty on aillabels. (Big sagebrush) Anemisia Indentata Apply with any sprayer that wdl apply the spray uniformly. Carya glabra (Pignut hickory)

• Check the sprayer before and dunng use to insure proper Celtis occicentalis (Westem hackberry) calibration and uniform application. Oatura ciscolor (Desert thomapple)

Lycium berlandien (Behager wodbeny) To mix, fdl spray tank ha!! full of water. Start agitation and Aforus rubra (Red mu: berry) continue dunng the entire mixing process. Aod required P;nus monticola (Westem whtte pine) amount of SPIKE Ory Flowable and allow to rnix when tank Pinus spp. (Pine) , mixing. If additional product is a wertab'e powder, add to tank Prunus emarginata (Bitter cheny) and allow to mix thoroughly. If additional product is a liquid, Rhus glabra (Smooth sumac) add slowly while filhng remainder of tank with water. Con. Robinia pseudoacacia (Black locust) tinueus agitatic,n in the spray tank is required to keep the Rosa multiflora (Multiflora rose) matenals m suspensica throughout application. Salvia leucophylla (Black sage) Agitate by mechanical or bypass (hydraulic) means in the Symphoncarpos orbiculatus (Buckbrush) spray tank. If bypass or return agitation is used, it should ulmus amencana (Amencan elm) terminate at the bottom of the tank to mmimize foaming. Vaccinium spp. (B!ueberry) Gamsada sm Wh@ For treating small areas, a tank type hand sprayer or sonnkhng can may be used. Before app 6 cation determme the amount of SPIKE Ory Flowable applied at the rate of 3.50 pounds per water and chemical necessary to cover uniformly the area to acre will control the following species. ' be tre'ited. Shake or stir frequently, Abies balsamea (Balsam fit) Acacia /arnesiana (Huisache) Acer saccharum (Sugar mante) Alnus rugosa (Speckied alder) WOODY PLANT CONTROL Betula popuhtolia (Gray birch) SPIKE Ory Flowabf e is an eff ective herbicide for the control of Carya texana (Black hickory) ( f,' brush and vines. SPIKE Ory Flowable can be apphed either as Certis pallida (Granieno) a broadcast spray, banded apolication of as an indtvidual plant Condaha obtusifolia (Lotebush condatia) treatment depending upon the size, density and location of #er vomstone (Yaupon) brush to be controlled. Lanz lancina (Tamarack) Pecea glauca (White spruce) SPIKE is to be applied to the soil (Not the Foliagei) where it is Populus balsamilera (Balsam poplar) atsorted by the roots of plants. Effects are slow to appear and Populus ce/toides (Eastem cottonwood) wdl not become apparent untd sutticient moisture has camed Overcus douglasn (Blue Oak) SPtKE Ory Flowaufe into the root zone. The time required to Overcus manlandica , (Blackjad oak) acnieve controlis dependent on soit type, amount of ramfall (Post oak) Overcus stellata and rootmg depth of target species. Some species may go SaSxspp. (Watow) through several defoliations and refoliaticns over a pened of (Desert yaupon) Schaetteria cuneifolia approximately two to three years pner to cying. Spiraea tomentosa (Hardhack) SPIKE Ory Flowable can be applied anyt me except when the Ulmus alata (Wmged e!m) ground is frozen or the soil is saturated with moisture. For optimum results, apphcations should be made just pnor to the SPIKE Ory Flowable applied at the rate of 4.70 pounds per acre wdl control the following species. resumption of acuve seasonat growth in the spnng and'or

• penods of rainf atl. For apphcations made m the late summer or Acacia berlandieri (Guajillo) earty f Clin areas of average annual rainfall of greater tnan 25 Acacia greggd (Catetaw acacia) inches, higher rates should be used and inconsistent control Acacia rigidula (B!ackbrush acacia) may res .Ct on densety infested brush areas and hard to control Acacia tortuosa (Twisted acacia) species. Acernegundo (Borelder)

Adenostoma lasciculatum (Chamise) SPIKE Ory Flowable may be used on cut brush but for Alnus rubra (Red alder) cotimum results time should te attowed for the brush to re. CJMsis radicans (Trumpeteroper) sprout to a height of approximately 5 feet pnor to apphcation. Carya avata (Shagbarx hickory) SPIKE requires an actively growing plant to be effective. The Cerrocarpus betuloides (Birchleaf targer the resprouts the more bPIKE that will be taken up by mountainmahogany) the plant and the more ettective and consistent the control wat Colubnna tesensis (Texas colubnnal g,, Condalia obovata (Bluewood condatia) Comus drummondd (Roughleaf dogwood) t(, For the control cf woody plants and vmes. the foftowing rates of SPIKE Ory Flowable are recommended. These rates can Crataegus spp. (Hawthom) Eysennarctia tenana (Texas kidneywood) vary depending upon sod type rainf all, time of appbcation and fagus grance/o/ia (Amencan beech) sire' density cl the woocy plants, 15

~ ~ - ~ ~ -- ---- ---- --- Jatropha droica (Leatherstim) For treating small areas, a t:n'it typs hand sprayer may bo used. Before apphcation, determme the amount of water and Leucochyrium frutescens (Centzo (Texas sdverteafl) Uquedambat styraci#ua (Sweetgum) chemical necessary1o cover uniformly the area to be treated. Parthenocissus quinque /otia (Virgtnia creeper) Shake or stir frequently. Populus granc4 dentata (Bigtooth aspen) I Do not apply taroadcast applications of SPIKE Ory Flowable

                                                           $o"'lm**o*r "c'nokeenerryi            *n' lo'*9' o' ***nanc' o'

• 9'85$ cov" i$ d*5'*d-

                         $*v"s* 'r"ofn7'Ea "*

Pseudotsuga menziesi (Dountas ist) Puerarra /cData (Kudzu) BANDED APPLICATION Quercus dumosa (Cahlomia scrub Cak) SPlKE Ory Flowable is recommended for the control of woody Overcus patustris (P n oak) plant species in noncropland areas (such as utstify, radroad, . Overcus ivora (Red oak) and pipeline nghts of way, ditchbanks and fence rows) by Overcus wgm<ana (bve oak) appbcahon of a senes of parallel bands to the sod surface. Rhus typrana (Staghom sumac) Individual bands should be spaced at intervals from 4 to 10 Rubus ailegnernen?is (AWegneny blackberry) feet and at the currently labeled rate range of 2.5 to 7.0 pounds

                    . Salvia ba!/ofaeffora                  (Shrubby blue salvia)                per acre dependmg on the woody species to be controlled.

Actual heroicide bands should be kept as narrow as possiele SPIKE Dry Ficuase acaled at the rate of 6.00 pounds per dunng application to acNeve minimal iniury or control of ccre wat contret tne fotlow ng species: herbaceous vegetation. Apply SPIKE Dry Flowable to the sod Acer macrophy3cm (Big! eat maple) surface in 5 to 75 gallons of water per acre in a senes of Acer p/aranordes (Norway maple) parallel bands with spacmg between bands ranging from 4 to (Sdver maple) 10 feet. In areas such as brush infested fence rows on utility Acer racenannum Bacchans spp. (Groundsel tree) rights of way, a single band may be applied. Control is (Flowenng dogwood) dependent upon root systems interceptog bands. Therefore Comus #onda Frarmas pennsylvanica (Green ash) larger stems should be treated individually when using single Gaulthena shallon (Salal) bands. Juniperus wgmtana (Eastem redcedar) Band spacing should be selected based on the size of the woody plants in the area to be treated and the amount of injury Un e or ptera ( ree) *'***"*'*'h*'*********9*b**'"'****'*!*'*0' Metaleuca quequenervia (Mefaleuca) Pmus canksiana (Jack pme) Where control of young or seedting woody plants is desired, Pit,us echmara (Shortteaf pine) bands should be spaced closer together. TNs wdl acNeve Pmus resmosa (Red sne) maximum exposure to their limitec root systems. Where larger Pinus wgsniana (%rgmia pine) more mature woody species are to be controlled, bancs Platanus cecicentahs (Amencan sycamore) should be spaced at the wider end of the recommended g Prunus serotma (B!ack Cherry) spacing range. Overcus a!ca (WNte oak) in addmon to allowing adequate exposure of the mere exten. Rubus lacmratus (Evergreen blackberry) sive root systems of these larger woody species for control, Rubus occidentalis (Blaca raspberry) use of the wider spacings wdl furtner reduce injury or controt of Sch.nus tereem:Motius (Bra:ihan peppertree) herbaceous vegetation m the area of treatment. SPIKE Dry Fbwame app 5ed at tne rate of 7.0 pounds per acre within the treated band nearty all vegetation, woody and wdl controd the fcucrwsng spec;es. hereaceous, wdl be killed. Some herbaceous vegetation close (%ne mapte) to the treated band with roots extendmg into it may be severely Acer c:rcinafum Arctostagnysos paruta (Greenteaf manzamta) injured or knied. However, since herbaceous species tend to (Wedgeleaf ceanotrts s have testricted root systems, most species outside the treated Ceanornus cuneatus band will not be affected. Banded applications in areas of Ceanomus leucocemus (Whitetnom chaparray Crataegus crussalli (Cockspur hawthorn) steep terrain should be applied across existmg slepes in order Elaeagnus angesefona (Russianotve) to prevent sod erosion.

                                                                          }                         Apply with equipment designed to deliver the spray unformly

[3)yh y in the bands. To maintain the integnty of the individual herbi.

 .                          Rhustauana                        (l aurel sumac)                       cide bands, straight stream nozzies fitted with internal stabiliz.

Sapum secr/erum (Tatlowtree) ing vanes or their equivalent are recommended. Operating Smdas rotuncircoa (Common greenoner) pressures should also be kept as low as wdl provide un form Ulmus parvs/cha (CNnese etm) dehvery of the spray soluton. Pressures in the range of 10 to Ulmus ruora (St.ppey elm) 40 psi should be accQuate. Pressures in excess of 40 psi wdl tend to cause ;he individual bands to break up. BROADCAST APPLICATION Wnen appucanons are made in an area where noz:tes are A099 SPIKE D4 Flowab!e in 15 to 150 9artons of water 0er the indavidual spray stre3ms may occur. If Conditions do not . acre with any property Cahbrated herbtCide sprayer Check the permit delsvery of intact spray streams to the soil surf ace this sprayer tefore and dunng use to insure proper cakbrabon and method of apphcanon should not be used. uniform apphcabon. Add the recommended amount of SPIKE Ory Flowaele to ctesn water m the spray tank cunng the fithng Fill the spray tank hatf futt of water. Start ag:taton and con. operat:on. Material must te kest m suspensten at au bmes by tanue dunng entire mixmg and spraying operabon Add the g constant ag tahon. Agitate by mechanical or bypass (hy. required amount of SPIKE Ory Flowable and allow it to mir araunci means in the s pray tana. it bypa ss or return ag tation is thoroughty whde completag the spray tank lilting. ;f hand held used, d should termmate at the bcttom of the tans to rrunimite or back pack type sprayers are used, shake vigorously after loamcg. f,ltmg ano penodically during appbcabon to mamtain product t r,

i a suspension. A master shut.ott switen for the cntito spraying system cnd nozzle check vciv:s are recomm:nded on brush species End conditions encountered. For indmdu:1 commercial spray equipment. ,a@ a O s m M W p a M Matenal must be kept in suspenston at ad times by continuous u d m cam at N W d mM

• ( agitabon. Agitate by mechanx:al or bypass thyorauhc) means in the soray tank. It bypass or tetum agitabon is used. it should terminate at the cottom et the tank to minimiz e foaming. Chect The Spot Gun is prepared for individual plant treatment by the sprayer treoventry tefore and cunng use to insure p oper mixing 2 pounds of GPtKE Dry Flowable in sufficient water to cahbraton and uniform aopwaten. octain 1 gallon of spray soluten. Set the Spot Gun to dehver 8 milliktors of this solution for every 1 to 2 inches of stem di.

ameter at tne case of the unwanted woody plants. For applica. INDIVIDUAL (SPOT) APPLICATlDN tion on steep slopes or other sensitive areas the Spot Gun can be equipped with a sod proce to allow injection of the SPIKE SP!KE Dry F1cwaole may be acched in high or low volumes of water to seettveiy centrolinomcual woocy plants. Recom. Dry FlowaDfe solution beneath the sod surf ace. Placement at a menced ratas wi.I vary cepending upon ute conditens with soil depth of 2 to 4 incnes will eliminate any suriace movement and reduce injury to herpaceous vegetation, the Pugher rates rieeoed for GtGCurt to control species, large plants, heavier sons. tas aopicauons and cut brusn. Consult At the presented rates. a 4 Dound bag of SPIKE Ory Flowab!e your 'ocal E'anco SPIKE estnbutor to cetermine the best rates will treat approximately 950 stems 1 to 2 inches in diameter. for your tocaten. Because of its non volati!e nature and low potential for enft this SPIKE application technicue can be used for treating un. Fct Figh volume apolcanora, mrx 1 pound of SPIFE Dry wantea woody ptants growing on non cropland areas aciacent FlowaD!e in enougn water to make 10 ga!!cns of solution, to sensitive crops. (See facel precautions.) A white spot Acpty to ounces of matenal to the sol per every 2 to 4 inches of stem diameter, should be visible at the base of each treated stem which

                                                                                           *         *     #0          # *E For low volume applicabons, mut 1 pound of SPlKE Dry Flow.

C AUTlD N: 00 NOT USE SPIKE Dry Flowable IN THIS MAN. ante in enough water to maxe 1 gallon of soluten. Apply 1 ounce of matenal to the sol per every 2 to 4 incnes of stem NER IN ANY AREA WHERE DESIRABLE SPECIES ARE IN THE VICINITY OF THE PLANTS TO BE ELIMINATED. A diameter. SMAlt. AMOUNT OF SPIKE Ory Flowable IN CONTACT When tresong large stems, apply the multiple treatments in Y ITH THE ROOTS OF DESIRABLE TREES OR OTHER even spac:ng around the stem. WOODY SPECIES MAY CAUSE SEVERE INJURY OR DEATH. THE ROOTS OF SUCH PLANTS MAY EXTEND For apptying S PIKE Ory Fbwable in banded or individual p! ant FAR BEYOND THElR DRIP LINES. treatment, hvo pieces of equement are suggested; the Solo Model 425 bacx pack sprayer for both cancing and indmdual SPIKE Dry Flowable will iniure or control other herbaceous plant treatment and the Spot Gun for individual plant vegetation in the treated area. Therefore, do not apply where j . treatment. such injury cannot be tolerated. See the list of herbaceous vegetaten controlled by SPIKE Dry Flowat's under the Total The Solo sprayer is prepared for spraying by adding the Vegetation Control section of this booklet. pre-slumed contents of a 4 pound bag of SPIKE Dry Flowa0!e and wa:er to tha tank. Fill to capacity with add.tional water and

                                                                                   ,s m e.-i   , e r w e,. w .c o m snake vigorousty. Equip the Solo sprayer with a 0003 SS straight stream ror:te and the Solo pressure regulator with         g     gyEg g y e... ._ , ,,, c% c.

the green (10 ps) pressure remrung sonn9 To band SPIKE Dry Fbwa.W at 5 pounds per acre, wa:k at 3 mph (264 feet per w

                                                                                        =.-won. e L om c. hemma a con' trarmte) wnh the Soeo on contsruousty and space the cands 5          g f *,*'* 7 ' d %

feet apart. Adi ust the rate ard walking speed accorcing to the ow-vamwu neww t t osos o. newn a cm.3 Ik 4 , 17 Es slo 106

e .

                                             ~

Spike @ Dry Flowable ID 5943; FN 3084 7 SPIK ES DRY FLOWABLE is a preemergence and postemer. 11. STABILITY AND STOR AGE F gerte herbicide for control of brush and weeds in noncrop The herbicidal properties of SPIKE DRY FLOWABLE

               "" 'S' require caution in handling, storage and transportation f this product. Store in original container only. Do

1. PHYSICAL AND CHEMICAL PROPERTIES not contaminate water, food, or feed by storage or dis-

   !                      A. Active Irgro6ent Ganetic Name                                                              posal.

Tetuthiuron B. Chemical Name ll. UNUSUAL FIRE AND EXPLOSION HAZARDS [ N-{5-(1,1 dimethylethyl).1,3,4 thiadiatol 2 yl) . None known N N*-dimethylurea e i C. Product Components IV. SPILL INFORMATION J Tebuthiuron 85.0% in case of leak or spill, contain material and dispose as Inert ingredients 15 0% waste. Do not contaminate any body of water, Sweep up material. Place it and damaged unusable containers K D. DOT Classificateon in a landfill approved for pesticides in accordance with Nonregulated applicable regulations. F

 ?                                                                                                                       Large spills due to traffic accidents, etc, should be re.

E * *

• r ga tr spersible granule with a mild E

odor ucts Company for assistance. Prevent spilled material 7 from flowing on's adjacent land or into streams, ponds E F. Auto Ignition Temperature E Decomposes at 320*F (160*C) I V. PROTECTIVE EQUIPMENT REQUIR*,MENTS

 ~

• During manuf acture, wear gogf es to"p'rotect eyes, wear s

6mpermeable gloves and protective equipment to avoid direct contact with skin. Use NIOSH (1) approved dust f H. Explosive Limit Does not yield dust rHpaator,

 $_                      l. Solubility                                                                      VI. FIRE FIGHTING INFORMATION Dispem m watn Considered nonfiammable. May emit toxic fumes when

• J- Threshou Limet Value ' # *P ' #"' "'"***"'""'"

Not estat6shed from fire site to enter nearby streams, ponds or lakes. l Keep containers cooled with water spray, b K. pH facpeous 50/50)

66 V

L-hof e: In sees af an wgerut involvene ensman nrngeenen es seatect ts4 the et Lldy sad Compeay telepheme speteter,(31T) 2612033 fee refer 64 le she phyenclam em t all, M - O ?

         . . . . . .                             , .                                        .- .      ...               1 1 ...        H.                        . ~.            .. .

v v . .

                                                                                                        .                                 .o v

Vll. TOXICOLOGY Children 1 to 5 years 15 mi A. Acute Exposure (SPIKE DRY FLOWABLE) Adults and older children $5 30

1. Eyes-Moderate ocular irritation occurred (2 tablespoons,1 or,)

when SPIKE DRY FLOWABLE was placed in the eyes of rabbits. All treated eyes cleared Then give at least 10 or of water to children and within three days posttreatment, 24 or to adults. Do not induce vomiting or give

• anything by mouth to an unconscious person.

2. Skin-SPIKE ORY FLOWABLE caused no systemic toxicity or dermal irritation when X, applied to the skin of rabbits at a dose of LABEL STATEMENTS 2000 mg per kg of body weight. A. Container Disposal

3. Inhalation-Rats were exposed "nose only" Triple tinse (or equivalent). Do not reuse. Then for four hours to a solid particulate aerosol puncture and dispose of in a sanitary landfill, away containing 4 84 mg of post mill blend SPIKE from desirable plants, trees and water supply, or DRY FLOWABLE per liter of air, based on a by incineration, or, if allowed , state and local total gravimetric concentration basis. All rats authorities, by burning. If burned, stay out of survived the exposure and 14 day postexpo. smoke.

sure observation period. Signs of toxicity in. cluded ataxia, body weight loss, dry nasal 8. Precautionary Statements exudate, reduced activity, labored respiration, This product will kill trees, shrubs and other forms and poor grooming that disappeared within of desirable vegetation having roots extending into five days af ter exposure. SPIKE DRY FLOW. the treated area. Read precautions on label care. ABLE was tested as a post mill blend for in, fully before using. halation toxicity because only a low percent. age of the bulk SPlKE ORY FLOWABLE 1. Human-Kee,a out of reach of children. May forms a respitable aerosol. be fatal if swallowed. Causes moderate eye irritation. Harmful if absorbed through the

4. Ingestion-The median lethal dose for rats skin. A-oid breathing dust and contact with given a single oral dose of SPIKE DRY FLOW. Skin or eyes. Use eye protection and protec.

ABLE was estimated at 4S8 mg per kg of tive clothing, such as coveralls, a long sleeved body weight. Signs of toxicity included leg shirt, and impermeable gloves when handling weakness, lethirgy, tremars, ataxia, coma. this product. Wash thoroughly with soap and reduced ac tivi t y, and salivation. Survivors water af ter handling and before eating, drink. appeared normal by 72 hours. ing, or using tobacco. Wash exposed clothing

5. Sensitiration-There was no indication of con.

tact sensitiration when guinea pigs were ex. 2. Environmental-Do not contaminate any posed topically to SPlKE DRY F LOWA8LE. body of water, ponds or streams as death or injury may occur to vegetation irrig3ted by Vill. HUMAN HEALTH we Nt contaminate water by cleaning of equipment or disposal of wastes. Laboratory animal studies that have been conducted with tebuthiuron indicate that the use of tebuthiuren EPA Registration Number: 1471 147 does not present a hazard when recommended handling procedures are followed.' Cr.emical Abstract Registry Number: 24014 18 1 IX. FIRST Af D (Statement of Practical Treatment) XI. REFERENCES A. Eyes-Flush eyes with plenty of water and call a (1) 1985 NIOSH Certified Equipment Guide physician if irritation persists. NOTE: This information applies only to SPIKE ORY

8. Skin-Wash exposed areas with plenty of soap and F LOWABLE which is sold in the U.S.

water. Wash all contaminated clothes before reuse. Call a physician if irritation persists. 'For user handling procedures, refer to product label; for manufacturing handling procedures refer to NACA Guide, C. Inhalation-Remove individual to fresh air. If breathing dif feculty occurs, provide cardiopulmon. * * ' #C' turing and Formulation of Pesticides. ary resuscit.ition assistance and get medical atten. tion. D. Ingestion-May be f atal if swallowed. Call a physi- . cian or Poison Control Center. Drink one or two glasses of wate~ and induce vomiting by touching back of throat with finger, or, if availat.le, by ad-ministering one to two tablespoons of syrup of ipt c:.c; SPIKES (tebuthiuron, Elanco) Issued 11/85 v

b REACTOR VESSEL FATIGUE USAGE EVALUATION l I The cumulative usage factors for the reactor vessel are as follows: Usame Factor i Location Unit 1 Unit 2 Unit 3 f Shell at water 1Lne 0.00620 0.00492 0.00431 , Feedwater nozzle .0.29782 0.21329 0.16139 Closure studs 0.24204 0.17629 0.14360 i h I [ I I i ll h i I e i t i 91 y- ---.-,v__ _ -..- - - --

CHALLENGES TO OR FAILURES OF MAIN STRAM RELIEF VALVE ( l JANUARY 1, 1987 - DECEMBER 31, 1987 ' Unit 1  ! t

                                     +

None

                                                                 !Ullt I t

None Unit 3 ' None j t All three units were in cold shutdown during the entire report.ing period. , i I I I t h i i 6 I l 92 h

                                                                                                                                       . - . . ._-___-_L

t4t;F E R OF *EP507.!*EL A ?." fr: 'tEr ?v ) L os a at:3 Joa r Lt.C TI C*; rL AfJT: FearJL FErst tg LEAR PL&f 1 1CE7 10:42 rot 4 CAY. FEPRUAtt 1 19;f f.U".E ER OF PE RS CN*.!L DI CC *-REM TO TAL PAF-4EP

        . - . - - - . - - - - - . - - . - . - - - - - - - - - - - - . - - - - - -                                -- =t=FEA TOL ce? SUPVIILLANCE ---                                        - - - - - - - - - - - - - - - - - - - - - - - - . - - - - - - . . ~ . -

6ROUP !TATION UTILITY C et.TP ACT 1CTAL ST AT ICt. UTILITY CO NTR ACT TOTAL [PiPLOYEE! EMPLOYEES A fd

• O T HER5 F EE 5072! EPPL O YEE! EPPLcVEES AtJO OTHEF5 P-F E r5 M AIN T E N &r:CE PER50NNEL 775 3* 2P E27 55.229 1.516 C.513 57.258 OPE R & f it;G P E* LOW N EL 65 1 F 74 7.766 C.CE7 1.108 F . ?61 NEALTM PHYSICS PERSONNEL 132 2 41 17E 45.Ef* C.203 10.229
• f . 4 C1 SUPER V I SO R Y P ER50 t;N EL 13 2 15 1.2?3 C.1F4 I 0.053 1. E20 EnCI N L E RI t.; P ER50 7.1sEL 132 172 1 3 06 10.C21 0.017 20.EE6 3!.704 70 1117 40 241 1393 120 27f 2.C07 40.6E9 152.954

        --.        .--.-- .- - - - -              = = - . . - - - . - - - . - - - -
                                                                                                          .-.--. FO: R ou T I AE P A I A* T E A A s.C E                      ----          - - - - - - - - - - - - - - - - - . . - - - - - - - - - - - - - - - - - - - -

GRCUP STATION UTILIT Y CotTR ACT TOTAL ST AT ICN UTILITY C0 tJTR A CT T3TAL [MPLOYEE5 EMPLOYEES AND C TPERS P EP50 *J5 ["P' 0YEES

                                                                                                                                                                                         -                    CEPLOYEES              AND OTFERS               *-9EPS PAINTEtJ A%CE PERSDP:NEL                                      895                          34                       4F                                9 77               410.C51                      5.387                 5.908               421.346 CPER A TING PER!0NN EL                                            63                          1                       7                                  71                 1 .07?                    C.084                 0.C56                    1.219 MEALTH PMY5ICS PERSONNEL                                      129                             1                     4C                                3 70                16 365                      0.006                 2.984                  1E.255 5UP ER V ILD RY P ER5 0 tA EL                                     11                          1                       3                                  15                0.P91                      C.C00                 C.024                    C . 915 EfJO I P:EERIhG PER50 fen [L                                   122                             5                    155                                2 82                17.444                      0.60 9               16.*42                  34.995 70                                                         1220                             42                     252                              1515                  445.830                      6.026               25.914                477.830
                                                                                                 =-=
                                                                                                                --- M r= I N-5[ R VI C E I t:5FECTION                          - - - - - -

6ROUP STATION UTILITY CONTR ACT TOTAL STAT!cN UTILITY CO ETR A CT TOTAL EPFLOYECS E P:PLOYEES AtJ3 O THERS P EE 50 N 5 [MPL O YEES E MPL OY E ES AND OTHERS P.-AEMS FAINTEN ANCE PCP50P:MEL 138 27 3 16R 12.258 5.977 C.499 12.734 OPER AT I AG PER50tJNEL 3 0 1 4 0. 003 0.000 0.025 C.028

          ' NE AL TM PMT52C5 PE R50NMEL                                        59                           1                     22                                  82                0.386                       0.005               0.174                   0 565 SUPE R V I SO R Y PERS O NNEL                                          5                       0                       0                                    5              C.4e7                       0.000 EN;INEERING PER507."NEL                                                                                                                                                                                                    0.000                    C.487 13                          3                     20                                   3G               0.426                       0.039               6.803                   7 268 l             PD                                                            218                         31                         46                               295                13.560                        6.C21               7.501                 27.022 l

1

        =                       - - - - - -     ---.--

l

                                                                                                                  =.-

w0=5PEC I AL P A I N T E R A NCE ------ --- - - - - - - - - - - - + - - - - . - - - - - - - - - - - -- l GROUP STATION . UTILITY CONTR ACT YOTAL ST AT ION UTILITY CONTRACT TOTAL [MPLOYECS EMPLOYEES AtJD OTHERS PER50N5 [MPL O YE ES EPPLOYEES AND OTHERS *t-R E ES MAINTENANCE PERSONNEL 637 21 115 773 CPER A t ! N; P E R SON N EL

                                                                                                                .                                                                     177.342                       6.927               89.462                273.731 35                           0                       1                                 36
             *TALTH PHYSICS PERSONNEL 0 .797                     0.000                 0 0C0                    C.797 106                              1                     37                               144-                 24 .36S                    0.001                11.114                  35.483 .

SupCR V IS"R Y P ER50 Nf.[L 8 0 2 9 1 .239 S.C00 0.030 1. 269 ECGINEER4NG PERSONNEL 92 3 75 1 70 15.450 0.128 16.745 32.223 PO 878 25 229 , 1132 21?.196 7.556 117.351 344.103 i

!__,._, .--            - - - - - - . . . . ,          . . . , . , _ - . - - - - - - - - -                                      -       - - - - - ~ ~ - - - - - - - ~ " ~ ~ ~                           ~ ' " " ~ ' ~ ~ ~ ~ ~ ~                    ~         ~

w NurBER OF *ER509.NEL AtO PAN-pef

• BY 4 ' 09 M AND J0P FLWCTION Ptahf1 8 E 0 W40 FE R R Y % UCL E A R FLani 1987 16 :4 2 MO RO AY , FEBRUARY 1, 1? Et kJMEE4 CF PERSONE.TL (>1CD M-REri TOTAL mat-RCM GROUP OTATION UTILITY CONT *AET TC1AL STATICA - UTILITY CCNTRACT 107 5L

[rpLcYEES EMPLCTEES AND OTFERS P E ttSONS EFPLO Y CES IMPLOYCES AteD CTFERS P-REMS as1NTEn ANCE PE Rt0NNEL 2589 116  !?! 28?6 663.69% 1?.807 ?6.709 720.211 i crER ATING PC1:02 EL 208 2 16 223 11.219 C.171 1 194 12.*24

      *eEALTH Pwv 1ES PER!ONNEL      526                   5             If 3              (*4             88.872             0.215            25.076          114.1(3 S urta Y1 sca Y PER00NuCL        41                   3                5                49              4.224           C.184             0.107              4.*15 EsGINEEatt1G FERSONNEL         3E6                 1 *,            454               MS3             44.395             1.2?9            69.629          11*.382 i                                                        ===                                                                  ======           =======        = = = = s = =s j                                    ====                                 ...            ....             . - . . . .

2750 13? 231 4720 812.405 21.675 192.775 1021 255 l l l l t W l O h e

• 9JUM PE 8t f5 F F E R SO Ah EL A ?.3
• AM-RC P PY L 04 A Ar:0 JC P, FUNCTIO *8 FL ANT
• 620LOS FCE.PY ALCL E Aa PLANI 1967 16*42 MONCAY, FEBRUART 1 1988 kUFBER OF PERSONt!L 0100 F-R T F) TO T AL MA N-R EM

 - - - - - - - - - _ - - ~ ~ - - - - - - - -
                                               ------=
                                                                    -=--- = --------- M 0 = W A S T E F R O C E S S I Pt G - ----- ---- ------ ---- ------ -- ----- --                  - - - - - - - - - - - - -     --

GROUP ~5 TA T IOP3 UTILITY C 0 7.T E A CT TOTML ST AT ICM UTILITY CONTR ACT TOTAL [*PLGYEEE EFFLCYEES Af40 OTHERS

• ER SON S [FOL O YC ES EFPLOYEES AND 01kERS P-P[PS PAINTLMANCE PERSONNEL S3 0 1 I CS 2.844 0.C00 OPER A T ING P CR EONN EL C.327 3.171 12 0 1 13 1. 004 0.000

       + E AL T o. PHYSICS PCR50kkEL                           72                                                                                                                 0.005                 1.CCS 0                     11                 P3            1.2*1                C.000 Sue [Rv50 R Y P E RSO NtCL                                                                                                                                                  0.050                 1.341 4                0                        0                 4            0.214               C.CCO                0.050                 0.214 Et:1NLE RING P[RSO NNEL                                  10                0                      15                 2*s           C.481               C.CCC                C.207                 C. Eta ED                                                     191                  C                     24              2 25            5. 794               C.000                C.625                 5.423
             ---------------------------------------------MC=K[                                                 FUEL - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ~

Gn0UP IT AT Ion UTILITY CONTR A CT 70TAL ST ATI ON UTILITY C04 TRACT TOTAL E FPL OY C E S EFPLCYCES AND of t[R P E R S 0*J S [ FPLO Y[E S [P PL O YCE S AMD OTHERS P-R[FS

• AINT[N ANCE PE PSO NNEL 51 0 0 51 5.971 C.CCC OPER AT ING PE R$0ta[L 3D C.000  ! . 571 0 0 30 C. 5 70 0.000 C.CCD C.573 NEALTH PHYSICS PER $0 ANCL 28 0 12 40 0.633 C.CCO Ct.GINEER ING PE RSONNEL 17 0.385 1.C18 1 16 34 0 573 0.CC5 C.326 C. T C4

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J f1LFf EP OF PER OfWrt A P.1) Pau-*!* MY L 0RP A'JO JC h F L% CT I CtJ Plati: B5,0 W ! FCFPY fUCLEAR PLAF1 1987 16:42 t*0t40AV. FEERUARY 1, 1983 *-i TCTAL t:t% E ER S O F IN C IY100 A LS M OOF ST AT10% UTILIT Y CONT F ACT TOTAL PAIMTCtJ ANCE PER50nt.[L  ?*6 44 12 1 1111 Or[RKTInG P[R50kt;EL 65 1 9 75 HCALTH vHYSJC5 PERSONNEL 134 0 , 27 1 61 SUPERVISOPY PER$CnNCL 14 0 3 17 ChGIt.ECRItG PERSONCEL 115 2 15 4 271

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TENNESSEE VALLEY AUTHORITY CH ATTANOOGA. TENNESSEE 37401 5N 1578 Lookout Place APR 051988 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Hashington, D.C. 20555 Gentlemen: In the Hatter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260

                                                         )                      50-296 BR0HNS FERRY NUCLEAR PLANT (BFN) UNITS 1, 2, AND 3 - ANNUAL OPFFATING REPORT FOR JANUARY 1, 1987 - DECEMBER 31, 1987 Enclosed is the Annual Operating Report for BFN for the period of January 1 to December 31, 1987. This report is being submitted to satisfy the requirements of 10 CFR 50.59, BFN Technical Specifications Appendix A, sections 6.9.1.2 and 6.9.2, and BFN Technical Specifications Appendix B, section 3.2.2. It contains a summary of plant inodifications, special tests, procedures issued and revised, occupational exposure data, reactor vessel, fatigue usage, and herbicide usage.

If you have any questions, please telephone H. J. Hay at (205) 729-3566. Very truly yours, TENNESSEE VA EY AUTHORITY l' R. Gridley, Di ector Nuclear Licen ing and Regulatory Affairs Enclosure cc: See page 2 7 An Equal Opportunity Employ er

2-U.S. Nuclear Regulatory Comission APR 051988 cc (Enclosure): Mr. K. P. Barr, Acting Assistant Director for Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NH, Suite 2900 Atlanta, Georgia 30323 Mr. G. G. Zech, Assistant Director for Projects TVA Projects Division c U.S. Nuclear Regulatory Comission One White Flint, North 11555 Rockvill6 Pike Rockville, Maryland 20852 Director, Nuclear Engineering and Operations Department Electric Power Research Institute P. O. Box 10412 Palo Alto, California 94303 INPO Records Center Institute of Nuclear Power Operations Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30389 Browns ferry Resident Inspector Browns Ferry Nuclear Plant Route 12. Box 637 Athens, Alabama 35611 J}}