RBG-23-317, Annual Operations Rept for 1985
| ML20154P710 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 12/31/1985 |
| From: | Booker J GULF STATES UTILITIES CO. |
| To: | Martin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| RBG-23-317, NUDOCS 8603200401 | |
| Download: ML20154P710 (51) | |
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e fg GULF STATES UTILITIES RIVER BEND STATION ANNUAL OPERATIONS REPORT FOR 1985 8603200401 851231 gDR ADOCK 05000458 PDR N,{}N
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t TABLE OF COIETEIFFS
1.0 INTRODUCTION
2.0 PIJurf PERFORBIABCE SUBBIARY 3.0 OPERATING EVENT SUBSEARY 4 3.1 SCRAMS 3.2 REDUCTIONS IN PONER 4.0 FUEL STATUS -.
5.0 MAINTENANCE 6.0 OCCUPATIONAL RADIATION SUMBRARY REPORT 7.0 SRT HISTORY 8.0 REACTOR COOLANT SYSTERE SPECIFIC ACTIVITY AIEALYSIS i
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.tr 1.0 INTRODUCTIN
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, ,, ANNUAL OPERATING REPORT 1.0 Introduction This River Bend Station Annual Operating Report for 1985 is submitted in accordance with applicable reporting requirements of Technical Specifications 6.9.1.4 and 6.9.1.5, of Appendix A to River Bend Station (RBS) License Number NPF-47. This routine operating report covering operation of the unit during the f previous calendar year also complies with applicable sections of l USNRC Regulatory Guide 1.16 " Reporting of Operating information -
Appendix A Technical Specifications".
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l 2.O PLMtT PERFOR90MCE SIROUutY
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-e 2.0 PLANT PERFOnunur's SUmgARY I. Summary of Operations j River Bend Station received a low power operating license on 1
August 29, 1985, and began fuel loading shortly thereafter.
Initial criticality was achieved on October 31, 1985.
Following initial criticality, the plant received a full power license on November 20 and exceeded 54 thermal po.ver on November 25. The main turbine / generator was rolled on November 26 and synchronized to the grid on December 3. The heat up testing phase of the Start-Up Test Program was completed without major complications. Start-up testing in
) Test Condition 1 continued through the end of 1985. River i Bend experienced three reactor scrans during the month of November; one scram was manually initiated per Technical Specifications and was taken credit for as a training scram, and the other two were from automatic action. Three reactor scrams were also experienced in December; one scram was planned as part of the very successful Loss of Offsite Power startup test, and the other two were from automatic
- actuation. The goal for River Bend was to have reached greater than 200 MWe by year's end. This goal was
. successfully achieved with the plant operating just greater than 200 MWe for a short period.
II. Duration of Activiti,es i
Date Du,rstion l Date low power license received 08-29-85 Fuel loading completed 09-21-85 22 days
- Initial criticality 10-31-85 39 days Full power license received 11-20-85 20 days
- Exceeded 54 power 11-25-85 5 days Synchronized to grid 12-03-85 8 days 200 MWe ( 204) 12-28-85 25 days t
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3.0 OPERAT M q l
l 3.0 OPERATING EVENT SERetARY River Bend experienced six scrams from August 29, 1985, date the low power license was received, to December 31, 1985. The first forced power reduction was utilized to allow for a training scram, (Reactor Scram 85-01). Reactor Scram 85-04 was planned as part of the Loss of Offsite Power startup test. Four scrams were a result of unplanned automatic actuations.
Enclosed are charts of the Power Profile for River Bend Station since initial criticality was achieved on 10-31-85. A discussion behind the cause for each forced outage is also presented.
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A B, C D E N
- 1 '5 lO' 13 20 25 30 Date A. Planned training scram on 11-13-85 at 2340 due to Div. I and II Standby Service Water inoperability.
B. Startup; reactor critical on 11-17-85 at 1438.
C. Unplanned scram on 11-21-85 at 0122 due to low level actuation on loss of feedwater.
D. Startup; reactor critical on 11-21-85 at 0943.
J E. Unplanned reactor scram on 11'28-85 at 1843 on IRM upscale due to a feedwater transient.
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y M to t'5 20 25 :iO Date A. Startup; reactor critical on 12-01-85 at 0321.
B. Loss of of fsite power test conducted with generator synchronized to the grid and reactor at 12% full power. A 9 day planned outage followed this trip.
C. Startup; reactor critical on 12-16-85 'at 0500.
D. Unplanned scram on 12-24-85 at 2123 caused by low reactor water level due to a loss of feedwater transient.
E. Startup; reactor critical on 12-25-85 at 1956.
F. Unplanned scram om 12-31-85 at 1011. Lightning striking a 500KV line tripped the turbine on power to load unbalance causing a reactor scram on high reactor pressure.
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3.1 SCRAMS .,
REACTOR SCRAM 85-01 l l
At 1200 on 11/13/85 with the unit in Operational Condition 2 l (Startup) and at approximately 2 percent power, the condition i described below was discovered.
A review of the open items on safety related building punch lists revealed that certain Category I structural elements (lateral supports, etc.) had not been installed in piping tunnels G and H (Reference LER 85-033). The affected safety related piping was part of the Standby Service Water (SSW) system, trains A and B.
The final structural load verification program required the addition of supplemental steel members to the existing structural steel in these areas. Engineering had perviously identified the work to be done on Engineering and Design Change Request numbers P-3172A, P-3059D, P-3064 and C-7188 to bring the tunnels into compliance with the design basis.
Immediate corrective action was taken. Trains A and B of SSW were declared inoperable, a power reduction was commenced per Technical Specification 3.0.3. The reactor was held just critical to allow operation personnel to take credit for a i training scram should a full shutdown be required. At 0000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> on 11/14/85 the structural steel support installation had not been completed and the reactor was manually scrammed with operations taking credit for a training scram. (Reference LER 85-034). By 2100 on 11/14/85 the necessary work was complete and SSW System trains A and B were declared operable.
CORRECTIVE ACTION TO PREVENT RECURRENCE j As discussed in LER 85-033, all open items on safety-related l building punch lists were reviewed in detail to ensure that all !
items which impacted the design basis of the unit were identified l and appropriate action taken. l l There was no single release of radioactivity or single l l radiation exposure specifically associated with the outage which l j accounted for more than 10 percent of the allowable annual l
- values. j i
l OPERATING TIME LOSS l
! 1. In terms of generator-off-line hours.
l Time Loss: 0 There were no generator-off-line hours lost since initial electrical generation occurred after the event.
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- 2. Operating time, loss when reactor was " forced" suberitical, all rods in ..
Time Loss: 86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br /> 58 minutes
- 3. The installation of seismic SSW piping support structures was determined to be critical path for the outage.
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REACTOR SCRAM 85-0.25
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- On 11/21/85 at 0025 with the unit in Operational Condition 2
, (Startup) , Reactor Feedwater Pump (RFP) B tripped for unknown
! causes (Reference LER 85-041). The operator restarted the pump and restored reactor water level to normal. In order to investigate the cause, RFP C was started with its discharge Motor
! Operated Valve (MOV 26C) left shut in the event RFP B tripped again. At 0117 RFP B tripped when its auxiliary oil pump was secured. The operator mistakenly believed that a normal oil i supply was available. An attempt to open the discharge MOV for RFP C failed with the breaker tripping on overload. RFP B and its auxiliary oil pump was restarted, but an attemin to open its discharge valve also failed. At 0123 the reactor scrammed when level decreased to the low level scram setpoint. Reactor Core Isolation Cooling (RCIC) was manually started to restore reactor level. The lowest level indicated on narrow range instrumentation was +2 inches (164 inches above TAF). At 0127 the RFP B discharge valve was opened and RCIC secured.
I J The cause of the initial RFP B trip is believed to have resulted from a main oil pump trip on overload due to low oil temperature. Investigation into the inability to open RFP B and C discharge valves determined that the cause was due to high i
differential pressure across the discharge valves. This would result when the feedwater to condenser recirculation valve l
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(FV104) is open as it was in this case to aid in reactor level control.
i CORRECTIVE ACTION TO PREVENT RECURRENCE In an effort to prevent recurrence, system operating procedure SOP-0009 " Reactor Plant Feedwater" now has a caution inserted concerning RFP discharge valve operation during low power operations which could result in unusually high i differential pressure. The operations staff has also been issued
! a memorandum reminding personnel of the operation of the RFP oil system. The system is being addressed in the current phase of licensed operator requalification training which began 01/03/86.
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, There was no single release of radioactivity or single radiation exposure specifically associated with the outage which accounted for more than 10 percent of the allowable annual values.
OPERATING TIME LOSS
- 1. In terms of generator-off-line hours.
Time Loss: 0 There were no generator-off-line hourg lost since initial electrical generation occurred after tNo event.
- 2. Operating time, loss when reactor was " forced" suberitical, all rods in.-
Time Loss: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 21 minutes
- 3. There was no major safety related maintenance performed during this outage.
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j REACTOR SCRAM 85-03 f '
on 11/28/85 at 1845 with the unit in Operational Condition 2 l (Startup), the reactor scrammed on Intermediate Range Monitor (IRM) upscale trip (Reference LER 85-047) . During a turbine roll the number 3 bearing was showing signs of high vibration forcing the turbine to be manually tripped at 1821 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.928905e-4 months <br /> on 11/28/85. At this time the number 1 turbine bypass valve was 30 percent open.
In anticipation of breaking condenser vacuum reactor power was ,
reduced. At 1833 condenser vacuum was broken to accelerate '
braking of the turbine. At 1840 the turbine bypass valves were tripped closed on low vacuum and reactor pressure began to increase slowly. As Main Steam Line drains were opened to reduce reactor presnure the level swell caused the feedwater level
- control valves to shut. Additionally, the increased voids caused power to decrease and the IRMs were down ranged to keep flux levels onscale.
As the pressure reductions slowed and level decreased the !
feedwater level control valve began to open. With reactor pressure nearly stable and with cold feedwater entering the vessel, reactor power began to increase. The reactor operator, having previously down ranged the IRMs, failed to range up the IRMs in time to prevent a reactor trip.
) t CORRECTIVE ACTION TO PREVENT RECURRENCE Since similar transients (i.e. turbine trip, loss of condenser vacuum, level and pressure transients). have occurred without a reactor scram, this is determined to be an isolated event. All operating personnel have been notified of this event
! and are aware of the conditions which cause reactor power transients. No further corrective action was taken.
There was no single release of radioactivity or single radiation exposure specifically associated with the outage which
, accounted for more than 10 percent of the allowable annual i
values.
j OPERATING TIME LOSS
! 1. In terms of generator-off-line hours.
i l Time Loss 0 There were no generator-off-line hours since initial electrical generation occurred after the event.
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- 2. Operating time, loss when reactor was " forced" subcritical, all rods in.-
Time Loss: '54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> 38 minutes
- 3. Startup was delayed because of blown fuses found on the backup scram valve circuit. Reverse DC polarity at the valve was determined to be root cause of the blown fuse problem.
Resolution of this item proved to be critical path for the outage.
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REACTOR SCRAlt 85-04 On 12/06/85 at 2234, as a result of required startup testing, .
a preplanned reactor scram occurred due to a turbine trip. The i i turbine trip at 12 percent of full power with the generator synchronized to the grid was part of the Loss of Offsite Power
- startup test. Both the Loss of Offsite Power test and scram recovery proceded without incident.
CORRECTIVE ACTION TO PREVENT RECURRENCE None required; pre-planned scram to support Start-Up Test Program.
OPERATING TIME LOSS
- 1. In terms of generator-of f-line hours.
Time Loss: 16 days
- 2. Operating time loss when reactor was " forced" suberitical, i
all rods in.
! Time Loss: 9 days
- 3. This outage was utilized to complete Surveillance Test Procedures required prior to entering Mode 1. Surveillance Test Procedures proved to be critical path for the outage.
Other Outage Items Worked:
- Replace Reactor Core Isolation Cooling trip throttle valve because of sticking.
- Replace five control rod drive position indication probles.
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REACTOR SCRAM 85-05 l
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On 12/24/85 at 2123 with the unit at 1 percent rated power in l Operational Condition 2 (Startup) , the reactor scrammed on vessel low level 3 (+9.7 inches) 172 inches above top of active fuel (Reference LER 85-060). The low level resulted from the trip of 4
the running feedwater pump C. The B feedwater pump was immediately started in an effort to prevent a scram but its discharge valve failed to open. Approximately 5 minutes later with level at +16 inches (178 inches above top of active fuel) i l
the remaining feedwater pump A was started and began injecting i
water just as the low level 3 scram initiated. Reactor vessel level was restored immediately with the minimum level reached )
being +9.0 inches (171 inches above top of active fuel) . !
'4 The initial trip of' the C feedwater pump resulted from gear increaser bearing failure due to overheating. Approximately 30 minutes before the scram the C feedwater pump had been started in order to make repairs on the feedwater pump A minimum flow line.
Upon investigation it was discovered that there was a lack of I sufficient cooling to the lube oil system. Step 4.2.8 of Station '
Operating Procedures SOP-0009 " Reactor Feedwater System"' requires that the Closed Cooling System (CCS) flow to the lube oil coolers be adjusted to maintain oil temperature from the coolers between {
80 and 120 degrees F. There is no similar step to be performed after the start of the feedwater pump. Manually operated gate ;
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valves are used to control CCS flow through the oil coolers.; The subject valve was found throttled nearly closed due to the CCS water temperature being colder than normal (at approximately 50
- degrees F). Although this valve position maintained proper
! temperature prior to the feedwater pump start, it was not adequate to maintain lube oil temperature with the pump running.
i j CORRECTIVE ACTION TO PREVENT RECURRENCE 1
A better design and one which would have prevented the above
- failure is an automatic lube oil temperature control.
Modification requests 86-0030, 86-0031 and 86-0032 have been i initiated to install automatic temperature control circuits for i these gate valves. In the interim, procedure revisions to SOP-0009 are in place to ensure careful monitoring of lube oil temperature and adjustment of CC8 flow to maintain lube oil at j
the proper temperature after the start of the feedwater pump.
The inability to open the a feedwater pump discharge valve against the high differential pressure was attributed to
- incorrect torque switch settings. Thgtorque switch settings
! were reset and the valve ratested satishetorily. A similar
- feedwater pump B discharge valve failure to open occurred on l 11/21/85 and was reported in.LER 85-041. At that time it was not 4
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i known that the t.o'rque switch settings were incorrect and the failure was attributed to the high differential pressure.
There was no single release of radioactivity or single radiation exposure specifically associated with the outage which accounted for more than 10 percent of the allowable annual i
values.
- OPERATING TIME LOSS
- 1. Generator-off-line hours - 0
- 2. Time loss when reactor was " forced" suberitical, all rods in.
Time Loss: 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> 23 minutes
- 3. Torque switch settinge on the B feedwater pump discharge j valve were determined to be incorrect and were reset prior to
- reactor startup. Resetting of these torque switches proved
] to be the critical path for this outage.
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REACTOR SCRAM 85-06 At 1015 on 12/31/85 with the unit in Operational Condition 1 )
(Power Operation), with reactor power at 20 percent and the :
turbine generator synchronized to the grid, the four turbine I 1 control valves and four turbine intercept valves were given a l
fast closure signal. The increased pressure caused the turbine
! bypass valves to open. Fifty-one seconds later the four turbine l stop valves tripped which caused a turbine generator trip.
- Approximately seven seconds later, the reactor scrammed on high pressure (Reference LER 85-063).
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) Investigation of the incident revealed that the actuation of the turbine control / intercept valve fast closure was due to a false indication of a turbine generator power to load imbalance.
This sensed power to load imbalance was caused by a combination 4
of two separate occurrences. A pressure transducer sensing impulse pressure to the low pressure turbine stage had failed prior to the scram and was documented on a Maintenance Work Request (MNR) to be reworked. At the time of this event the pressure transducer was failed high. Also, a transient was j
introduced on the Gulf States Utilities (GSU) grid by a lightning strike on a 500 KV transmission line. The power to load imbalance relay requires a power differential between the steam input to the low pressure turbine and the electrical output of the generator of 40 percent differential and a sufficient rate of j
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change of current before it will trip. The failed pressure transducer provided the 40 percent differential and the lightning strike provided a sufficient rate of change in current.
Once the turbine control / intercept valves were given the fast closure signal they tripped and immediately tried to reset. With all eight valves trying to reset simultaneously there was a sufficient loss of hydraulic pressure in the Emergency Trip System to cause the turbine bypass valve to open and the four turbine stop valves to shut. This caused a turbine generator trip and reactor scram on high pressure. ,
CORRECTIVE ACTION TO PREVENT RECURRENCE In an effort to prevent recurrence the pressure transducer has been replaced via Maintenance Work Request 6637. A retrofit
' of the reset logic circuitry for the turbine intercept valve fast closure is being investigated. This retrofit would sequence the i reopening of the turbine intercept valves one at a time to
! minimize the hydraulic pressure reduction. Modification Request l 86-0129 has been initiated to request thig design change.
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i There was no single release of radioactivity or single radiation exposure specifically associated with the outage which accounted for more than 10 percent of the ' allowable annual values.
OPERATING TIME LOSS
- 1. In terms of generator-off-line hours Time Loss: 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 19 minutes
- 2. The failed pressure transducer which was replaced prior to l reactor startup and was the critical path item for restart.
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l ah 3.2 REDUCTIONS IN PONER There were no forced reductions in power other than those reported in 3.1 of this report.
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v 4.0 FUEL STATUS 1
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- 4.0 FUEL STATUS i River Bend began receiving its initial core load of fuel from General Electric on February 1, 1985. Final receipt, channeling, inspection, and initial storage was completed March 8, 1985. The loading of the 624 bundle initial core began on August 31, 1985
- and was completed September 21, 1985. Initial criticality was achieved on October 31, 1985 and River Bend continued to perform start-up testing through the end of 1985 in accordance with j Chapter 14 of the FSAR.
i The first power generated by River Bend occurred December 3, 1 1985. The station has generated 129,928 MWH thermal which equals
- 1.90 effective full power days through the end of 1985. The remaining energy for Cycle I is estimated t- be 307.8 effective full power days.
I River Bend Cycle I contains 5 fuel bundle types described as follows:
4 76 tuel bundles of 0.7 weight percent of U-235 108 fuel bundles of 0.9 weight percent o* U-235 l 120 fuel bundles of 1.6 weight percent of U-235 280 fuel bundles of 2.5 weight percer.t of U-235 40 fuel bundles of 2.8 weight percent of U-235 TheRiverBenddesignencompassesGeneralk'lectric'sControlCell Core and Barrier Fuel concept. However, the station did not operate at levels high enough in 1985 to benefit from the l expected capacity factor improvements. Both water chemistry and offgas monitoring clearly show that no failure. of fuel has occurred during the 1985 operations.
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5.0 MADrfatANCE This section covers the paried from initial criticality (10-31-85) through the and of 1985.
During the first two months of power ascension, problems have arisen as systems are operated as an integrated electrical generating facility. These system interface problems are not as critical as those that result in outages but effect future availability as we increase in power and achieve commercial operation. To address these problems task forces are assembled as the systems problems are identified. These task forces are charged with investigating and determining modifications to -be implemented to improve the efficiency of system operations.
This section describes the problems that have occurred and work being performed to correct them.
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REACTOR WATER CLEANUP ISOLATIONS Problems have been experienced with inadvertent isolations of the Reactor Water Cleanup system (RWCU) . These problems have been categorized into the areas of hardware deficiencies, personnel errors, and procedural deficiencies. Specific hardware problems include:
- 1. Water flashing downstream of a flow element causing erratic flow oscillations.
- 2. Trip setpoints on the differential flow instrument that does not taken into account density changes.
A task force has been assembled to address the RUCU problems, and needed design changes are being implemented. To date the following modifications have been performed:
- 1. Some orifices in the RWCU system were replaced to eliminate flashing downstream of the flow element.
- 2. An isolation bypass switch was installed for the purpose of preventing inadvertent isolations of RWCU during performance of related surveillances.
As a result of these modifications RWCU isolations have been reduced.
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COIfTAIlOG3fT AIRLOCK DOORS Numerous pr'oblems have been encountered with the operation of the containment airlock doors. River Bend has two airlocks consisting of interlocked doors. Each door uses two inflatable seals. Problems encountered result from hardware deficiencies and improper usage by untrained personnel and include:
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- 1. Seal ruptures
- 2. Seal retaining piece bolt failure
- 3. Leaking air valves and tubing A task force has been assembled to review the airlock door design and provide a long-term solution. In the short term, the following has helped.to reduce door failures:
- The seal design was changed by the manufacturer to prevent fatigue failure of retaining bolts through repeated inflate / deflate cycles.
- All leaking valves have been fixed and a heavier gauge tubing installed.
- Containment entries are being more closely controlled to minimize use.
- Personnel training on operation of the doors has been conducted.
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6.0 OCCUPATIg gMTIN sUsecutY REPORT l
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6.0 OCCUPATIOttAL RADIATICII
SUMMARY
REPORT Enclosed in this section are two reports on Occupational Radiation Exposure for River Bend Station personnel for the year 1985 as required by 10CFR20.407, Reg. Guide 1.16, and Technical !
Specification section 6.9.1.5a of Appendix A to River Bend Station Operating License Number NPF-47.
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Occupational Radiation Summary Report For Calendar Year 1985 i
Listed below is the report which sumarizies the number of personnel who '
received deep dose equivalent in the various annual dose ranges. This '
report is required by the requirements of 10CFR20. "
Annual Whole Body Exposure Rangeft) Number of Individuals (rems)
- In Each Range No measurable exposure 1,518 Measurable exposure less than 0.1 142 0.1 rem but less than 0.25 ram 15
, 0.25 rem but less than 0.5 rem 1 0.5 rem or greater None Total number of individuals reported 1,676 This total,1,676, is the number of individuals for whom personnel monitoring (was provided during calendar year 1985 as per 10CFR20.407, paragraph a)(2).
Dwight M.'Ross
&%.WfY Radiological Health Supervisor (1) Individual values exactly equal to the values separating exposure ranges are reported in the higher range.
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7.0 SAFETY RRLTMF VALVE HISTORY
i 7.0 SAFETY RELIEF VALVE HISTORY Enclosed is documentation of all challenges to safety / relief valves and a summary of their performance as required by Technical Specification Section 6.9.1.5.b of Appendix A to River Bend Station License Number NPF-47.
During 1985, challenges to the Safety Relief Valves (SRVs) were limited to that required as part of the Startup Testing Program at River Bend Station. All SRV cperations were preplanned and there were no failures of SRVs to perform as designed. Included in this section is a listing of the SRVs and the time and date of each actuation. In addition the Reporting Operating Information on Main Steam Safety / Relief Valves for Institute of Nuclear Power Operations has been provided.
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l SRV ACTUATION EVENTS
- SRV DATE TIME IB21*F041A 12-06-85 1058 1D21*F041B 11-19-85 1558 12-06-85 1105 1B21*F041C 11-19-85 1552 12-06-85 1204 1B21*F041D 11-19-85 1542 12-06-85 1110 1B21*F041F 11-19-85 1555 12-06-85 1154 1B21*F041G 12-06-85 1147 1B21*F041L 12-06-85 1159 1B21*F047A 11-19-85 1536 12-06-85 1213 1B21*F047B 12-06-85 1053 IB21*F047C 11-19-85 1549 12-06-85 1040 IB21*F047D 12-06-85 1047 1B21*F047F 12-06-85 0803 IB21*F051B 12-06-85 1033 1B21*F051C 12-06-85 1230 1B21*F051D 12-06-85 0452 1B21*F051G 11-19-85 1546 12-06-85 0813
- All SRV actuations took place as part of the Startup Test Program.
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l REPORTING OPEItATING INFORMATION ON MAIN STEAM LINE SAFETY / RELIEF VALVES FOR INSTITUTE OF NUCLEAR POIMR OPEllATIONS INITIAL RECOItDS
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8.0 REACTOR COOLANT. SYSTEM SPECIFIC ACTIVITY ANALYSIS During 1985 analysis of specific activity of primary coolant indicated that the Limiting Condition of Operation of specification 3.4.5 " Specific Activity" of River Bend Technical Specification was never exceeded.
As delineated by Technical Specification section 6.9.1.5.c of l Appendix A to River Bend License Number NPF-47, no further information is required or enclosed.
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GULF STATES UTILITIES COMPANY RNER BEND STATION POST OFFICE BOX 220 ST FR ANCISVILLE LOUISaANA 70775 AAE A CODE 504 635 6094 346 8651 March 3, 1986 RBG- 23,317 File Nos. G9.5, G9.25.1.5 Mr. Robert D. Martin, Regional Administrator U.S. Nuclear Regulatory Commistion Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011
Dear Mr. Martin:
Unit 1 RiverBendStation-h Docket No. 50d4 v Enclosed is the River Bend Station Annual Operating Report for 1985. This report is submitted in accordance with Technical Specifications 6.9.1.4 and 6.9.1.5 of Appendix A to River Bend Station (RBS) License Number NPF-47.
Sincerely, J.E. Booker Manager-Engineering Nuclear Fuels & Licensing River Bend Nuclear Group M
JEB/DAS/ebm cc: Director of Inspection & Enforcement j U. S. Nuclear Regulatory Commission l Washington, D. C. 20555 l w wx J