ML20137X956
| ML20137X956 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/31/1985 |
| From: | Gridley R TENNESSEE VALLEY AUTHORITY |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 8603120049 | |
| Download: ML20137X956 (52) | |
Text
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TENNESSEE VALLEY AUTHORITY OFFICE OF NUCLEAR POWER ANNUAL OPERATING REPORT BROWNS FERRY NUCLEAR PLANT i
January 1. 1985 - December 91. 1985 Docket Numbers 50-259, 50-260, and 50-296 License Numbers DPR-33, DPR-52, and DPR-68 Submitted by:
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Sife Director Submitted by:
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a tfnager JE34 y
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TABLE OF C0tGEfES t
fiMLC Plant Modification Summary _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Attachment D Critical Systems, Structures, and Component Tests and Experiments for 1984 _ _ _ _ _ _ _ _ _ _ _ _ _
Attachment C New Procedures and Procedure Changes _ _ _ _ _ _ _ _ _ _ _ _ Attachment A Temporary Alterations to Plant Equipment _ _ _ _ _ _ _ _ _ _ Attachment A Challenges to or Failures to Main Steam Relief and Safety Valves _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Attachment B o
Occupational Exposure Data _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Attachment E e
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ATTACHMEtir A t
NEW PROCEDURES AN
D. PROCEDURE
CHANCES January 1,1985 - December 31, 1985 During 1985, various procedures as described in the Safety Analysis Report were drafted or revised. All changes were reviewed against 10 CFR 50.59, and no procedures were initiated that constituted an unreviewed safety question. The safety evaluations for these procedures are filed at the site and are available on request.
TEMPORARY ALTERATIONS TO PERMANENT PLANT EQUIPMENT January 1, 1985 - December 31, 1985 During 1985, ntnerous temporary alterations were made to on various plant equipment. These alterations were controlled via the use of Temporary Alteration Control Forms. No unreviewed safety questions were involved based on requirements of 10 CFR 50.59 The safety evaluations for these cnanges are filed at the site and are available on request.
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Attachment B
' CHALLENGES TO OR FAILURES OF MAIN STEAM RELIEF VALVES January 1,1985 - December 31, 1985
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Unit 1 1-16-85 at 1440 Due to a loss of normal feedwater low reactor water level resulted and the reactor scrammed. The main steam isolation valves closed and the reactor operator manually opened main steam relief valves 1-22 and 1-23 to control reactor pressure.
Unit 2 e
None Unit '4 None,'
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f Attachment C CRITICAL SYSTEMS. STRUCTURES, AND COMPONENTS (CSSC)
TESTS AND EXPERIMENTS FOR 1985 January 1,1985, - December 31, 1985 ST 84-19 This special test was prepared as an approved method for receipt inspection of Westinghouse Quad + fuel assemblies.
Unreviewed Safety Question Determination
~ Question Is the probability of occurrence or the consequences of an accident or malfbnction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification The probability of occurrence of an accident is not increased. Per Final Safety, Analysis Report (FSAR) section 14.6.4 the design basis accident for fuel handling is dropping an irradiated fuel assembly onto the reactor Work is to be performed on the new fuel inspection stand area, not core.
over the core. The fuel bundle is also not irradiated.
Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification No new accident or malfunction is foreseen.
Question Is the margin of safety as defined in the basis for any technical specification (TS) reduced?
Yes No X
Justification New fuel inspection is not addressed in the TSs.
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b Page 2 1
ST 85-01 The purpose of thin special test is to record the analog signals listed in Table A on a magne. tic tape medium suitable for analysis by Oak Ridge National Laboratory at several different test plateaus under varying conditions of power, flow, dual and single recirculation pump operation,
. and load line in order to define an operating region for stable single loop operation, and to gather baseline data for operation in that region.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification Recirculation pumps will be started and stopped in accordance with Operating Instruction (OI)-68 (except that step III.F.2 is not applicable) and these transients are bounded by FSAR section 14.5.5 and 14.5.6.
The equalizer valve positions will not be altered per FSAR section 4.3.3.
The stability analysis of section 7.17 predicts no divergence or limit cycle oscillations for single ' loop operation. However, one of the purpo:ss of this test is to measure the stability of the system at various operating conditions.
Precautions in the procedure denote what actions to take to l
return the unit to stable operation if instabilities are observed. The procedure provides for monitoring various parameters to quickly detect if the system becomes unstable. This is not expected to happen. Most of the signals to be recorded on magnetic tape are already available at the i
startup test panel with the exception of sufficient average poder range monitor (APRM), local power range monitor (LPRM) jet pumps No.
10 and No.
11 DP, and jet pump loops flow signals. With regard to jet pumps signals, the worst thing that could happen is loss of indication for the particular instruments. This would have no effect on the ability to safely shutdown and cool the reactor. The necessary APRM and LPRM signals will be obtained from the meters for 2 LPRM and 5 APRM cabinets in panel 9-14, utilizing the test connections in the meters. These connections for the APRMs and LPRMs will only be utilized during the actual recording of data. Otherwise, they will not be connected to the meters.
Again, the worst thing that can happen is the shorting of the meters, causing loss of indication. This would not affect the reactor protection system (RPS) or the ability to safely shutdown and cool the reactor. The short duration of the test, particularly at the more severe test conditions, plus the precautions in the procedure are sufficient to ensure no damage to the internal equipment of the reagtor will result. Therefore, no increase in probability or consequences of an accident or equipment malfunction is foreseen.
Question Is the possibility for an accident or malfunction of a different type than any evaluated previousl/ in the Safety Analysis Report created?
Yes No X
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Justification The effects of natural circulation, single loop, and dual loop recirculation have previously been evaluated in the FSAR, as shown in the references.
No additional possibilities are foreseen.
Question Is the margin of safety as defined in the basis for any TS reduced?
Yes No X
Justification Operatior. in single loop mode is permitted by TS 3.6.F.1 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The test will be completed within that time span. While TS 3.6.F.2 could be interpreted to prohibit opening the discharge valve of the inactive pump, FSAR section 4.3.4 clarifies this TS. Based on that clarification, this test does not affect the margin of safety inherent in this specification.
Surveillance Instruction (SI) 4.6.E-1 will be used to verify the requirements of TS 3.6.E just prior to performing the test. SI 4.6.A.6 and 7 will be satisfactorily completed prior to restarting a pump as required by OI-68. Therefore, the margin of safety as defined by these TSs is not reduced.
ST 85 Reactor pressure vessel (RPV) level and pressure and drywell temperature data will be recorded from plant permanently installed instrumentation during normal cooldown, subsequent repressurization from 70 psig, and then to cold shutdown. This will be done in an attempt to recreate the events associated with the water level discrepancy observed on unit 3 startuos on November 20. 1984. and February 19. 1985. and thereby. determine the cause of the level discrepancies.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification Cooidown will be performed per normal plant procedures. Cooldown and heatup rates will-be limited to 90 F/hr. Data recording will be from permanently installed instrumentation. Control rods will be fully inserted prior to repressurization which will be limited to 90 psig.
In addition to the emergency core cooling system required at these conditions, control rod drive and condensate will provide makeup.
Based on this, neither the probability nor consequences of equipment malfunctions or accidents is increased.
Page 4 Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification The FSAR in sections 14.5 and 14.6 evaluates abnormal operation transients and design basis accidents which bound this special test.
Question Is the margin of safety as defined in the basis for any TS reduced?
Yes No X
Justification IC the level discrepancy is recreated, control rods will already be fully inserted. Heatup and cooldown rates will be observed; and with RPV pressure less than 122 psig, high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) are not required to be in standby readiness, and with RPV pressure 105 psig, 1 loop of RHR may be removed from LPCI standby for 24-hours to prepare for SD cooling (TS 3.5.B.2).
The unit will be placed in cold shutdown shortly afterwards. The RHRS may be in shutdown cooling below 105 psig.
ST 85-03 The purpose of this test is to investigate the cause of the reactor water level discrepancies observed between LI-3-60 and LI-3-53/LI-3-206 ouring startups of unit 3 on November 20, 1984, and February 13, 1985. On February 13,1985, LI-3-53 and LI-3-206 were observed reading as much as 20 inches higher than LI-3-60. The problem appears to be due to a decrease in the reference leg of LI-3-53 and LI-3-206.
Leakage of the reference leg at the panels or instrument drain lines may be proved or discounted by observing a level discrepancy trend or known reference leg leak rate by isolating the instruments at the panel and monitoring the reference leg leak rate for change. This test will detenmine that.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
e Justification With the unit in cold shutdown and all rods in scram, anticipated transient without a scram ( ATWAS), automatic depressurization system ( ADS), HPCI, RCIC, feedwater control, and control room indications are not required.
With primary containment not requ 4, autor, tic primary containment
Page 5 isolation system (PCIS) is not required.
Special requirement No. 2 preserves the level of reliability of,the functions which are affected and still required.
Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification Instrument line breaks outside primary containment are analyzed.
Question Is the margin of safety as defined in the basis for any TS reduced?
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No X
Yes
}$sti rication With the reactor in cold shutdown, energy release, cladding temperature, and radioactivity releases are bounded by loss of coolant accident (LOCA) and break outside containment analyses.
STA 84-04 Ultrasonic testing of A and B scram Barton/GEMAC instrument reference legs will verify legs full and determine approximate location and size of air bubbles, if any, trapped in reference legs prior to startup. During startup RPV pressure, level, temperature, and drywell temperature data will be taken to characterize the level discrepancy if it reappears or to document that it does not reappear.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
i Justification Startup and heatup will be performed per normal plant procedures. Data recording will be from permanently installed instrumentation. Ultrasonic testing will provide additional verification that these instruments are operable.,No increase in consequences or probability is created by these activities.
Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
b Page 6 Justification This special test consists of nondestructive testing, normal plant startup, and data recording from permanent instrumentation readings. These activities do not create any possibilities of malfunctions which are not already evaluated in the FSAR.
Question Is the margin of safety as defined in the basis for any TS reduced?
Yes No X
Justification Data recording does not affect the operability of the instruments. Startup is per normal procedure. If a level anomaly appears, rod withdrawal will be suspended and the appropriate TS complied with.
ST 85-05 The purpose of this test is to establish the calibration drift tendencies of a Static-0-Ring, model 103As-B203-NX-JJTTX6 differential pressure switch.
A Static-0-Ring model 103AS-B203-NX-JJTTX6 switch will be substituted in place of 3-LIS-27-51 and calibrated to 22" WC on the rise.
The calibration will be checked every month for twelve months and recalibrated as necessary to establish the calibration drift tendencies of the Static-0-Ring switch.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification Level switch 3-LIS-27-51 is not addressed by the Safety Analysis Report and the functions performed by this switch do not affect condenser operability.
i Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification Level switch 3-LIS-27-51 is not addressed by the Safety Analysis Report, and the functions performed by this switch do not affect condenser operability.
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Page 7 Question Is the margin of safety as defined in tlue basis for any TS reduced?
Yes No X
Justification Level switch 3-LIS-27-51 is not addressed by the TS.
ST 85-08 The purpose of this test is to monitor and record the electrical current characteristics of the unit 3 HPCI suppression pool outboard suction valve (3-FCV-73-27) motor with 27 psig minimum differential pressure applied across the valve.
During this test, 3-FCV-73-27 will be cycled and the motor current monitored during two different wiring configurations. One wiring configuration will be the method shown on TVA as-constructed drawings 45N714-2RB, 45N3711-3RA, and 45N3711-5RA. 'In the other wiring configuration to be used, the armature leads A1 and A2 will be reversed and the field winding leads S1 and S2 will be reversed at the 250-volt reactor motor operated valve (MOV) board. This condition duplicates the "as-found" wiring configuration.
Results of this test are needed by the office of engineering to evaluate the operability of 3-FCV-73-27.
Unreviewed Safety Queation Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification This test includes temporarily wiring the motor for 3-FCV-73-27 in two different wiring configurations and monitoring the operating characterization of each. Since HPCI will be inoperable during the time this test is performed, no increase if foreseen.
Question Is the possibility for an accident or malfbnction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification The worst gase failure would be the possibility of introduction of air into the water supply line for HPCI. Section 6.4.1 of the FSAR addresses this problem; however, stating that constant pressure on the water supply line prevents air pockets from forming. Since no air injection into the HPCI system is foreseen, no new possibilities are created.
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Question Is the margin of safety as defined in the basis for ar.y TS reduced?
Yes No X
Justification This test will be perfonned when HPCI is not required for service.
The bases for TS 3.5.E states that HPCI is considered inoperable when the reactor is being started from a cold condition.
Since this teat will be performed with the unit in cold shutdown and HPCI considered inoperable, the margin of safety as defined in the basis of the TS is not reduced.
ST 89-11 The HPCI degraded voltage test will consist of evaluating HPCI syston reliability while operating under degraded voltage conditions simulati.ng low station battery voltage. HPCI system operation while ur. der degraded.-
voltage conditions will be accomplished by varying the d.c. supply voltage to the HPCI inverter and the HPCI turbine governor control system.
The degraded voltage test will consist of-two basic parts.
Part A will perform.a static calibration of the HPCI turbine governor control systas and also evaluate HPCI turbine governor response under degraded voltage conditions.
Part B of the degraded voltage test will evaluate actual HPCI turbine operating response while varying the d.c. voltage level using auxiliary boiler steam as the motive force.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident of malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No _ L_
Justification The reactor will be in the cold shutdown mode during performance of this special test, thus HPCI operation will not be required. The appropriate HPCI sis will be performed after the special test is completed and before reactor startup from cold shutdown to ensure HPCI op?rability, Question Is the posgibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
b Page 9 Justification Since the HPCI systen will be returned to normal and its operability verified after the special test, no other type of failure possibility will be created.
Question Is the margin of safety as defined in the basis for any TS reduced?
Yes No X
Justification The special test will' be performed with the reactor in the cold shutdown mode and HPCI operability will be proven before reactor startup, thus the TS basis is not changed.
ST 8G14 This test is to determine the mixing time required for representative sampling of the floor drain sample tanks.
Copper nitrate will be added to the tanks and the contents will be recirculated until the copper concentration stabilizes.
Unreviewed Safety Question Determination Question
'Is the probability of occurrence or the consequences of an accident or malfbnction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification Performance of this test will not significantly increase the probability of a loss of tank contents. In addition, none of the design safety features listed in the FSAR will be affected.
Question Is the possibility for an accident or malfbnction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification Systens operation during the special test is not significantly different from normale operation and will not result in any increase in the radioactivity in the waste liquids.
Question Is the margin of safety as defined in the basis for any TS reduced?
Yes _
No X
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I Page 10 2gstification it The'IS addresses only the radioactive releases from the plant which are not
~ impacted by the proposed special test.
ST 85-15 The purpose of these tests is to determine the mininum recirculation time necessary to obtain representative samples from the waste sample tanks.
A minimum of one test will be needed with a maximum of three tests anticipated.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfbnction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justificatiou Performance of this test will not significantly increase the probability of a loss of tank contents. In addition, none of the design ~ safety features listect in the FSAR will be affected.
Question i
Is the possibility for an accident or malfunction of a different type than sa any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification Systems operation during the special test is not significantly different from normal operation and will not result in any increase in the-radioactivity in the waste liquids.
Oudstian
+. Is the margin of safety as defined in the basis for any TS reduced?
Yes No X
a Justification The TS addresses only the radioactive releases from the plant which are not impacted by the proposed special test.
ST 85-18 1'7 I
This test shall be used to detect a tube leak in the R4CU non-regenerative heat exchangers. This test will look for increases in shell pressure after isolation.
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Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification This test will not affect the capability of any system to perform its intended safety function. The probability or consequence of an accident will not be increased.
Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification No types of accidents can occur other than those described in the Safety Analysis Report.
Question Is the margin of safety as defined. in the basis for any TS reduced?
Yes No X
Justification Performance of this test does not alter any TS requirements, thus the margin of safety is not affected.
ST 85-20 To evaluate HPCI operation under degraded voltage conditions and record various operating parameters of the HPCI turbine governor controls per instructions in special test 8520.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Anakysis Report increased?
Yes No X
lustification The reactor will be in the cold shutdown mode during performance of this special test, thus HPCI operation will not be required. The appropriate HPCI sis will be performed after the special test is completed and before reactor startup from col' 'hutdown to "nsure HPCI operability.
C-Page 12 Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification Since the HPCI system will be returned to normal and its operability verified after the special test, no other type of failure possibilities will be created.
Question Is the margin'of safety as defined in the basis for any TS reduced?
Yes No X
Justification The special test will be performed with the reactor in the cold shutdown mode; and HPCI operability will be proven before reactor startup, thus the TS basis is not changed.
ST 85-21 The pu'rpose of this test is to obtain information about the actuation forces of valves 3-FCV-74-58 and 3-FCV-74-59.
These forces will be used to help answer questions concerning the legitimacy and accuracy of the results of leak rate tests performed on these valves where pressure is applied under the globe (reverse direction of actual flow) because these valves are presently installed backwards on unit 3.
Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification Since this test will be conducted in compliance with TS 3.5.B.9 which states that one loop of RHR may be removed from service, provided the other loop and associated diesel generators are operable, the probability or consequence of an accident or malfunction is not increased.
Question Is the posgibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
b Page 13 Justification One loop of RHR will be available throughout the test. Operation with one loop of RHR out of service is covered in TS 3.5.B.9.
Therefore, no possibility for an accident or malfunction of a different type than evaluated in the FSAR will be created.
' Question Is the margin'of safety as defined in the basis for any TS. reduced?
Yes No X
Justification No TSs will be violated during this test; therefore, the margin of safety is not reduced.
ST 85222 Experimentally measure the standby liquid control (SLC) system suction pressure by installing an inline pressure transducer and performing a flow test.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
h tification Special ie t (ST) 85-22 requires the operation of the SLC system in a manner consistent with approved operating procedures.
Question Is the possibility for an accident or malfun'ction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification ST 85-22 provides for flow testing of the SLC system in the same manner as SI 4.4. A.1 with the addition of a pressure transducer installed in the SLC system suction piping.
The addition of this pressure transducer has been examined and found not to degrade the systems integrity or reliability.
Question LIs the margin of safety as defined in the basis for any TS reduced?
Yes No X
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'Page 14 Justification ST 85-22 does not degrade the SLC system to a greater degree than the SLC systen surveillance.
ST 85-23 The HPCI degraded voltage test will consist of evaluating HPCI system reliability while operating under degraded voltage conditions simulating low station battery voltage. HPCI system operation while under degraded voltage conditions will be accomplished by varying the de supply voltage to the HPCI inverter and the HPCI turbine governor control systen.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated on the Safety Analysis Report increased?
Yes No X
Justification The reactor will be in the cold shutdown mode during performance of this special test, thus HPCI operation will not be required. The appropriate HPCI sis will be performed after this special test is completed and before reactor startup from cold shutdown to ensure HPCI operability.
Question Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification Since the HPCI system will be returned to normal and its operability verified after the special test, no other type of failure possibility will be created.
Question Is the margin of safety as defined in the basis for any TS reduced?
Yes No X
Justification
'The speciak test will be performed with the reactor in the cold shutdown mode, and HPCI operability will be proven before reactor startup, thus the TSs basis is not changed.
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Page 15 ST 85-24, -25, and -26 were performed on non-CSSC.
ST 85-27 The HPCI degraded voltage test will consist of evaluating HPCI system reliability while operating under degraded voltage conditions simulating low station battery voltage using a 440 ohm dropping resistor in place of the normal 500 ohm dropping resistor. HPCI systen operation while under degraded voltage conditions will~be accomplished by varying the de supply voltage to the HPCI inverter and the HPCI turbine governor control system.
Unreviewed Safety Question Determination Question Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated en the Safety Analysis Report increased?
Yes No X
Justification The reactor will be in the cold shutdown mode during performance of this special test, thus HPCI operation will not be required. The appropriate HPCI sis will be performed after this special test is completed and before reactor startup to ensure HPCI system operability.
Question Is the possibility for an accident or malfbnction of a different type than any evaluated previously in the Safety Analysis Report created?
Yes No X
Justification Since the HPCI system will be returned to normal and its operability verified after this special test, no other type of failure possibility will be created.
Question Is the margin of safety as defined in the basis for any TS reduced?
Yes No X
Justification This special test will be performed with the reactor in the cold shutdown mode; and HPCI operability will be proven before reactor startup, thus the plant TS basis is not changed.
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I ATTACHMENT D PLANT MODIFICATIONS
SUMMARY
JANUARY 1, 1985 - DECEMBER 31, 1985 Modifications Safety Evaluation ECN P3000 - Primary Replaced pressure transmitters PT-64-50. -51, and -67 Containment System -
on unit 1; PT-64-51 and -67 on units 2 and 3.
The ECN Units 1,2, and 3 was completed for all three units.
The new Foxboro transmitters are better than the original transmitters; however, they are not environmentally qualified. Their use is an interim solution only until permanent qualified instruments may be procured. ECN P3058 will replace these transmitters. The Foxboro transmitters do perform the same function as the original GEMAC and did not adversely affect any safety-related function.
ECN P3'002 - Sampling ECN P3002 replaced the cristing solenoid valve and Water Quality FSV-43-13 with new valve. ASCO model No. X206-380-3F.
System - Unit 3 This portion of the modification had been previously implemented. During unit 3 cycle 5 outage, verification of the solenoid valve mounting configuration for seismic qualification was implemented. The ECN was completed for unit 3.
The new equipment is environmentally qualified for its respective environment. The function and qualification of the equipment remained the same. A seismic analysis was performed for the replacement solenoid valve to ensure the original design requirements were met.
ECN P3087 - High Replaced level switches LS-73-57A and -57B.
The ECN Pressure Coolant was completed.for unit 3.
Injection System -
. Unit 3 The replacement switches meet or exceed all the qualifications as the old equipment with the addition of environmental qualifications. A seismic analysis was performed to ensure the original design requirements were met. The function of the equipment remained the same.
ECN P3130 - Control Air Solenoid valves FSV-32-62 and -63 were previously System - Unit 3 replaced with environmentally qualified ones per ECN P3130. During unit 3, cycle 5 outage, verification of the solenoid valve mounting configuration for seismic qualification was performed.
The ECN was completed for unit 3.
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2 PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation The function and qualifications of the equipment remained the same. A reismic analysis was performed for the replacement valves to ensure original design requirements were met.
ECN P3134 - Containment Modified the mounting brackets for 3-FM-84-198 and Atmospheric Dilution 3-FM-84-208.
The ECN was completed for unit 3.
System - Unit 3 The modification assures the seismic mounting of the FMs. Therefore, the probability of occurrence or the malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased.
ECN P3136 - Various The ECN was issued and implemented for documentation Systems - Units 1, 2, only to "as-construct" drawing series 47W225, sheets and 3 00 through 76, to show the "as-constructed" design conditions for areas subjected to harsh and mild environments for units 1, 2, and 3.
The ECN was totally completed.
Analyses were performed per IE' Bulletin 79-01B.
The documentation provided radiation and temperature environments resulting from the various types of accidents. No physical work was performed at the plant. The margin of safety was not reduced.
ECN P3139 - Core Spray Replaced the core spray pump room cooler fan motors.
System - Unit 3 The ECN was completed for unit 3.
The existing fan motors were not environmentally qualified. The replacement motors are environmentally and seismically qualified. The new motors are of equal or better quality than the existing motors. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR was not increarad.
ECN P5018 - Residual
, Removed packing leakoff valves and plugged leakoff Heat Removal System -
connection on FCV-74-67 and FCV-74-53.
The ECN was Unit 1 totally completed as it only covered unit 1.
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b 3
PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation The valves allowed plant personnel to dotormine if the valve packing was leaking excessively. The valves were not required for safety-related functions of the injection valves or the RHR system. Removal of the valves did not adversely affect any mode of the RHR system operation.
ECN P5042 - Residual The ECN was issued and implemented for documentation Heat Removal System -
only for as-constructing drawings and to clear the TACF Unit 1 that installed lifting lugs to the ceiling beam in the southwest corner room for the RHR pump motor IC.
The ECN was completed for unit 1.
The added lifting lugs do not interfere with any safety-related functions. The modification did not degrade the structural or seismic integrity of any existing structural members.
The margin of safety was not reduced.
ECN P5051 - RHR Service Installed a flow rostricting orifice between the Water - Common RHRSW/EECW pump discharge line and the associated air release / vacuum rollof valve for each pump. The ECN was completed.
The air release and vacuum roller capabilities of the air release valve were maintained.
Installation of the orifice will reduce leakage from the RHRSW/EECW pump dischargo line in the event of a failure of the air release valve.
Installation of the orifice did not adversely affect pump oporability.
ECN P5052 - 4160V Bus The ECN was issued and documented to reflect the
- Tie Board - Common "as-constructed" configuration of supply cables between the 4kV bus tie board and the 4kV cooling tower switchgear. The ECN was completed.
Since this was a documentation change only and the plant work had alcoady boon analyzed and performed por ECN P0297, no now type of failure was created by this change. Based on this, the possibility for an accident
'or malfunction of a different type than any evaluated previously in SAR is not created.
ECN P5058 - Asfociated Installed new cable tray supports in the unit 3 reactor Electrical Equipment building, R18, U, elovation 565.
The ECN was completed Syrtem - Unit 3 for unit 3.
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4 PLANT MODIFICATIONS
SUMMARY
Nodifications Safety Evaluation During an inspection, the existing hangers were found not to be seismically qualified. The hangers were modified in such a way that the final configuration is seismically qualified. The modification did not adversely affect any safety-related system.
ECN-P5070 - Reactor The ECN was issued and documented to "as-construct" Core Isolation Cooling drawings.
Leak-off valve 71-575 was not installed on System - Unit 1 FCV-71-8 on unit 1 as the drawings showed. The ECN was completed.
Having the leak-off line capped without having a valve in the line did not adversely affect the operation or function of the valve FCV-71-8.
The margin of safety as defined in the basis for any Technical Specification was not reduced.
ECN P5096 - Standby Nodified the diesel control circuit so the units 1 and Diesel Generators 2 diesels can be paralleled with the unit 3 diesels in System - Units 1 and 2 the presence of a preaccident or common accident lignal. The ECN was completed.
The Design basis for the plant to accomplish tho' safe shutdown is to have capability to parallel units 1 and 2 diesels with the unit 3 diesels after 10 minutes..
The failure anlysis for this mode has been analyzed and has been found acceptable.
The modification did not increase the probability of occurrence or the consequences of an accident or the malfunction of equipment important to safety which was previously evaluated in the FSAR.
- ECN P5099 - High Noved TS-73-2R 90 degrees from its present location Pressure Coolant to allow room for pipe hanger. The ECN was completed.
Injection System "
Unit 3 The TS maintained all its existing requirements.
Relocating the switch did not change its function of detecting an HPCI steam line break. The TS maintained all of its existing requirements.
ECN P5111 - Emergency hevised drawings to show the "as-constructed" Equipment Cooling Water configuration of the EECW system. The change was for System - Unit 3 valves 3-67-733 and -734 to the unit 3 diesel generator e
C 5
PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation building shutdown board room coolers which changed their status from normally closed to normally open.
The ECN was comploted.
The modification was performed to eliminate the significant amount of operator action required to align the EECW supply to the chillers when RCW was unavailable. Since the EECW now operates continuously and EECW water flows through the chiller supply lines, the overall reliability of the plant is improved.
ECN P5114 - Standby The ECN was issued and documented to as-construct Diesel Generator System drawings.
The drawings woro revised to show correct
- Unit 3 de-energized stato on relay contacts 5-6 and 7-8 on relays TRA-3. TRB-3, and TRC-3, and corrected the referenced contact loctions on relays TRB-3 and TRC-3.
The ECN was completed.
The relay contacts, in the de-energized state, are arranged to isolate the main control room circuitry when backup control is necessary. This ECN corrected the drawings so this function is shown. The ECN did not change the function or operation of the diesel generators or any other safety-related equipment.
ECN P5123 - Primary The ECN was issued and implemented for documentation Containment System -
only to verify MR-64-110 and PR-64-116 had been removed Units 1, 2, and 3 from all three units and to revise drawings to dolote the instruments from the drawings. The ECN was completed.
The instruments were not required for normal operation of the plant. They did not provido any sofoty-related input into any logic. Therefore, the possibility for an accident or malfunction of a difforont type than previously evaluated was not created.
ECN P5124 - High Replaced existing leak detector cables damaged during Pressure Coolant the unit 3, cycle 5 outato. Pulled and terminated now Injection System -
cables 3M151, 3M152, 3M153, and 3M154 for conduit Unit 3 3M168. The type of cable was changed from WDD to kDV-1.
The ECN was completed.
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6 PLANT MODIFICATIONS
SUMMARY
l Nodifications Safety Evaluation The new cable is superior to the damaged cable it replaced and exceeds or meets all requirements.
It is class 1E and environmentally qualified for its service. No new accidents or malfunctions are foreseen. The margins of safety in the Technical Specifications were not affected.
ECN PS140 - Standby The ECN was issued for documentation only to Diesel Generators -
as-construct drawings.
The changes were Units 1 and 2 as-constructing wiring in mode control logic on elementaries and adding missing wiring reference on DG "C" logic panel connection drawing. The ECN was completed.
Both changes were minor and did not affect the operation of the logic circuits. The modifications were covered by previous ECNs and USQDs; only the drawing revisions were omitted. The changes did not affect the operation of the logic circuits.
ECN PS141 6 Reactor The ECN was issued and implemented for documentation to Core Isolation Cooling allow the use of alternate material in the turbine System - Unit 3 exhaust check valve, 3-71-580.
For part #15, a washer, the ECN allowed type 410. ASTM A-276 stainless steel and for part #19, a lock bracket, the ECN allowed type 304, ASTN A-240 stainless steel. The ECN was completed.
A seismic evaluation showed the components are seismically acceptable and the vendor confirmed the availability and performance of the valve was not affected by these changes.
preferred transformer TUP 3 to a RPS dedicated Unit 3 regulating transformer.
The transformer receives its power from the 480V RMOV Board 3B.
The breaker feeding the new transformer is class 1E current actuated isolation device. The ECN was completed.
The maintenance feed still performs the same function.
'The new transformer meets the BFNP Class II criteria.
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r 7
PLANT NODIFICATIONS
SUMMARY
Modifications Safety Evaluation The ECN was totally completed and is a temporary change until ECN P0712 can be implemented.
ECN PS143 kV The ECN was issued and implemented to as-construct Shutdown Board -
drawings only.
Changed CASA-6 relay contacts 3-4 to Unit 3 1-2 and swapped the notes next to contacts 1-2 and 3-4.
The ECN was completed.
There was no physical work performed.
This was a documentation change only.
The probability for an accident or malfunction of equipment important to safety previously evaluated in the SAR was not increased.
ECN P5144 - Main Steam Revised drawings to reflect that valve number 1-522 is System *- Unit i not physically installed. The ECN was completed.
No physical work was involved. Only the drawing was revised to show actual configuration of the plant. The valve was to be a packing leakoff isolation valve for valve 1-523.
It would have served as an operational convenience and performed no safety-related function.
ECN PS147 - Reactor The ECN was issued and implemented for documentation Recirculation System -
only to as-construct drawings. The drawings were Unit 2 revised to show jumper installed on 4-kV reactor recirculation pump board 2, compt. 3, terminal ZE-1 to ZE-2 and compt. 5, terminal ZD-11 to ZD-12.
The ECN was completed.
The jumper had been previously installed to allow low oil pressure, temperature, and flow limit annunciators to operate prior to the breakers being closed.
No safety-related function was adversely affected by the modification.
The modification only provided operational information for the operators.
ECN P5149 - Reactor Installed one additional conduit support, unit 3 Building Conduit drywell, elevation 584'.
The ECN was completod.
l I
Supports - Unit 3 The Civil Engineering Branch performed and documented
'an evaluation showing the additional support did not degrade the seismic qualifications of the drywell structural steel, any other supports or equipment, nor
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I 8
PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation the conduit for which it was designed. No change in the function, operation, or qualifications of any safety-related equipment or structures occurred due to this modification.
ECN P5150 - Reactor The ECN was issued and implemented for documentation Water Cleanup System -
only. Deleted stem leakoff valves69-582, 69-554, and Unit 1 69-555 from drawings as they did not exist in the plant. The ECN was completod.
The leskoff valves were not required for any safety-related function of the flow control valves or the RWCU system. Removal of the valves did not adversely affect the RWCU system operation. The margin of safety was not reduced.
ECN P5151 - Off-Gas The ECN was issued and implemented for documentation System - Unit 1 only. Revised drawing to show wire number 14A6(55) connected to torque switch number 17 at terminal 55.
This depicts "as-constructed" status of valve FCV-66-2.
The ECF was completed.
The change was a documentation' change only to correct drawings. No physical work was performed.
Valve FCV-66-2 still performs its intended function. No safety-related equipment or function was adversely affected by the change.
ECN P5155 - Standby The ECN was lasued and implomonted for documentation Liquid Control System -
only to revise drawings to show the actual plant Units 1, 2, and 3 configuration. The drawings were revised to show valve 63-538 " locked closed" instead of " closed" only. The ECN was completed.
The modification did not change the valvo's function of isolating the check valve test lino from the SLC system's main process flowpath. The change did not alter any of the valve's present design requiraments.
The probability of an accident or a malfunction of equipment Important to safoty was not increasod.
ECN PS160 - RHR Service the ECN was issued and implemented for documentation Water System - Units only. Revised drawings to show valves23-514 and 1, 2, and 3 23-534 lockod closed. The ECN was completed.
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C 9
PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation The valves are normally closed and are not required to be opened for proper operation of the RHR or RHRSW systems. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated was not increased.
ECN P5172 - Residual The ECN was issued and implemented for documentation Heat Removal System -
of use of a shim (spacer) on the shaft side bushing Unit 3 between the hinge arm and bushing on the 24-inch RHR testable check valves. All the work was performed per MMI-51. The work was performed for valves FCV-74-68 and FCV-74-54.
The ECN was completed.
The modification was performed due to FCV-74-68 falling its LLRT due to the valve disc being slightly off centor. Insertion of the spacer centered the disc and should eliminate further problems. The valves function exactly as before and their design was not degraded.
No new accidents or malfunctions are foreseen.
ECN PS176 - Standby The ECN was issued and implemented for documentation Liquid Control System -
only to as-construct drawings only. The revision will Units 1, 2, and 3 show that valve 63-537 is locked closed instead of closed only. The ECN was completed.
The valve being locked closed ensures that the SLC tank does not inadvertently drain. The modification did not change the function of the valve nor alter any of the valves present design requirements. The modification did not degrade the operating condition of the plant.
- ECN PS182 - Emergency The ECN was issued and implemented for documentation Equipment Cooling Water only to "as-construet" drawings. The drawing revisions System - Common changed the unit prefix number from 3 to O for valves 67-279A through 67-294A, 299A, and 300A. The ECN was completod.
The modification did not alter any physical or safety-related aspects of the plant. The change did
'not alter the design or functional requirements of any of the affected valves.
The chango in no way degrades the operating conditions of the plant, f
10 PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation ECN PS211 - Reactor The ECN was issued and implemented for as-constructing Recirculation System -
drawings (valve marker tabulations) to depict current Units 1, 2, and 3 piping configuration for the reactor recirculation system for valve tag numbers68-227 through 68-295.
The ECN was completed.
The modification revised existing documentation to include these valves.
No physical work was involved; ao safoty-related function was affected. The possibility for an accident or malfunction of a different type than any evaluated previously in the SAR was not created.
ECN P0708 -Reactor Cracked piping was ropaired as necessary by weld Recirc'., RHR, Coro overlays to allow continued operation of unit 1.
ECN Spray, and RWCU P0708 was issued for documentation of OE evaluations Systems - Unit 1 for the weld overlay repairs for various class I wolds.
The Office of Engineering performed stress and seismic analyses and documented them, stating "the weld overlay repairs performed on the unit 1 piping did not adversely affect the structural integrity of the affected systems."
The evaluations performod are for one cycle of operation only (outage to outago). If operation for more than one cycle is desired with these overlays, the USQD will be revised. The overlay work was performed previously. The ECN paperwork was completed in 1985.
ECN P0719 - RHR The 16" RHR 3C heat ozchanger RHRSW supply isolation Service Water System -
valve was replaced with a 16" Series 1400 Wafer Pratt Unit 3 Butterfly Valve.
The valve was tagged HCV-3-23-37.
A seismic analysis was performed on the now valvo and piping configuration to assure the solamic integrity of the RHRSW system. The now valve meets or ozcoeds all the requirements previously established for its use.
The now valvo performs the samo function as the replaced valve. No safety-related system or equipment function was adversely affected.
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11 PLANT MODIFICATIONS
SUMMARY
Nodifications Safety Evaluation ECN P0730 - Residual Removed RHR head spray piping, hangers, and associated Heat Removal System -
work from the bulkhead to the pipe flange on'the RPV Unit 3 head spray. The ECN was not completed. The electrical portion of the modification was not laplemented.
The removal of the head spray line is part of the IGSCC studies being done on Browns Ferry Nuclear Plant. The final configuration of the penetration for the head spray line must be quallfled as a primary containment pressure boundary.
ECN P3761 - Vital Area Fabricated and installed barriers in the Reactor Barriors - Units 1, 2 Building, elevation 565'.0", refuel floor, and the and 3 intake pumping station for all HVAC duct and piping penetrations in vital areas. The ECN was only partially implemented.
All barriers necessary to implement the new physical security plan have been installed. The Office of Engineering is responsible for the analysis, evaluation Lor post-modification tests of the.HVAC systems affected by this change to show that the system design requirements are still met with the modification implemented.
Provided the analysis, evaluations, or tests show the design requirements are still satisfied, the change will not create any new failure possibilltles because the original plant design conditions will be maintained.
ECN P0602 - Residual Fabricated and installed orifice plates on the unit 2 Heat Removal System -
RHR pump test lines and performed post-modification Unit 2 tests. The ECN was completed.
The orifice installation affects the RHR torus cooling mode flow rate only. No functional change was involved. The PMT demonstrated that operating two RHR pumps in the same loop in the torus cooling modo resulted in a maximum flow of at least 13,000 GPMs and that vibrational levels at varying flow rates were acceptablo. An analysis was performed by GE and a revision to the USQD was written for verification that the ortflee plate installaticn did not adversely affect plant nuclear safety.
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12 PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation ECN P0612 - Control Installed 1/2-inch stainless stool fler hose from the Air System - Unit 3 control air permanont line to the pilot star,e assembly of the MSRVs. The ECN was completed.
The flex hose was selsmically quallfied to function during a seicmic event. The modification did not affect the seismic qualifications of other equipment connected to the hose.
The margin of safety was not reduced because system function remained the same.
ECN P0614 - Main Steam Rerouted 1-inch HPCI vent line in the unit 3 steam System - Unit 3 tunnel. Reroute of the line was required due to interferenc9 with equipmont transfer during MSIV work.
The ECN was not completed. Work still remains to be implemented on the MSIVs.
The roroute of the vent line did not adversely affect the venting capability for the llPCI pump discharge piping. An analysis of the rerouted vont line by OE shows the solsmic qualifications of the vent line were not adversely affected.
Based on this evaluation, the modification did not adversely affect plant safety.
ECN P0616 - Reactor Replaced the reactor rectre. pump motors' oil cooler Building Closed Cooling cooling water inlet / outlet brazed fittings with high Water System - Unit 3 pressure threaded brass fittings to facilitate motor maintenance.
The ECN was completed.
The function of system was not altered by the replacement fittings. The now threaded fittings are as good or better than the brazed fittings.
The probability of an accident or malfunction of equipment Important to safety was not increased.
ECN P0621 - McIn Steam Profabricated and installed locking tabs and limit System - Unit 3 switch platos for the main steam isolation valvos in unit 3.
Only a portion of the work covered by the ECN was completed.
The modification did not affect the operation of the HSIVs.
Seismic analysis performed by OE veriflod seismic qualifications of the MSIVs or the main steam lines were not adversely affected.
Thorofore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety was not increased.
13 PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation ECN P0396 - Standby Replaced the stiff 300 mcm cables with more florlble Diesel Generator System 1/0 cables from the 125V (diesel generator batteries)
Units 1 and 2 to the distribution cabinet.
The ECN was completed.
An analysis was made to assure the new cables are capable of handling the loads necessary to start and control the diesel generators. The new cables are seismically quallfled; electrical separations were maintained. The operability of the diesel generators was not lessened by replacement of the cables.
ECN P0073 - Reactor Installed thrust roller bearing, self-lubricating Building Crane -
bushing, grosso seals, and and closure to the 5-ton Common auxiliary hoist hook block of the reactor building 125-ton crano. The ECN was completed.
The modifications did not impair the design capacity of the crane.
The modifications eliminated the twisting of the double cable and cable groove jumping by reducing the friction between the hoist-hook and the hoist-hook block. The margin of safety was not reduced.
ECN P0334 - Fuel Storage Replaced the oziating air compressor on the refuel and Handling System -
platform with a larger capacity compressor (220/480V, Unit 1 6011Z, 3 phase, 3/4 HP). The ECN was completed.
The compressor location and oporation affects no safety-related system and prior analysis showed the new compressor did not add any excessive load to the power supply. The margin of safety remained unaffected.
ECN P0520 - liigh Replaced automatic dolugo valve to llPCI fire protection
'Prossure Fire Protection system (FCV 26-37).
The valvo is UL approved for fire System - Unit 1 service uso and was replaced to correct a deviation cited in NRC Inspection reports. The ECN was completed.
The new valve moots or orcoeds design requirements such as design pressure, temperaturo, valve class, etc.
The modification did not degrado any safety-related system. No other accident or mm1 function of a different type than previously evaluated in the FSAR is forosoon or created, f
F 14 PLANT MODIFICATIONS
SUMMARY
Modirleations Safety Evaluation ECN P0369 - High Installed a MOV bypass valve around FCV-73-3.
The ECN Pressure Coolant was completed.
Injection System -
Unit 3 The piping, valvo, and the associated equipment were seismically qualified.
The valve is a primary containment isolation valve and has manual controls that are independent from FCV-73-3 but has automatic logic that is identical to FCV-73-3.
The change reduced the probability of occurrence of HPCI steamline break outside containment because tho transients associated with the steamline pressurization and warmup were greatly reduced.
ECN L1382 - Reactor Revised RBCCW restraint drawings.
Drawings were made Building Closed Cooling adding required details from as-constructed Water System - Unit 1 measuromonts.
Somo rostraints were modified to assure the restraints were as-constructed por design.
The added review and modification of the restraints'per the ECN help assure that the RBCCW does not fall during system operation or a solsmic event. The modification did not creato the possiblity for a different type of accident or malfunction.
ECN P0501 - Reactor Redesigned foedwater pipo support Mark No. RFWS-48 to Foodwater System -
reduce stressos and prevent damage to the 18-inch Unit I feedwater pump discharge pipe. The ECN was completod.
The now support distributes the thormal stress loadings over a larger surface area of the pipe. This decreases the possibility of crack initiation. The now support has no effect on the operation of any safety-related system or component. No Technical Specification was affected by the modification; therefore, the margin of safoty was not reduced.
ECN P0584 - Fuel Pool Revised fuel pool cooling logic so one pump may be run Cooling System - Unit 2 without continuous alarm. The ECN was completed.
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15 PLANT MODIFICATIONS
SUMMARY
Nodifications Safety Evaluation The modification did not alter the operation of the fuel pool cooling system.
The new logic still provides annunciation if the running pump or pumps discharge pressure falls below the setpoint. The nonclass II circuitry is isolated from the Class IE' portion.
Therefore, the margin of safety was not reduced.
ECN P0585 - Standby Performed post-modification functional testing of speed Diesel Generator sensors and speed switches installed on Diesel System - Units 1 and 2 Generators A, 8. C, and D.
The ECN was completed.
The ECN replaced the existing speed sensing panels located in the excitation cabinet with solid-state speed sensing panels. The existing relays were unreliable and beginning to show component fatigue.
The solid-state speed sensing replacement panels are more reliable. The new panels are slessically qualified, Class IE.
The function of the panels was not changed. The margin of safety was not reduced.
ECN P0590
- Main Control Added additional fluorescent fixtures to the AC, Room Emergency diesel-backed, standby system for emergency lighting.
Lighting - Unit 1 Relocated selected DC incandescent light fixtures to better serve deficient areas. The ECN was completed.
The modification increased the emergency light levels on safety-related panels.
The affected systems remained seismically qualified. The diesel generator capacity will not be exceeded and the control bay HVAC was not affected since the modified lighting system heat load was bounded by the normal lighting system heat load. No margins of safety have been reduced.
ECN P0591 - Neutron Replaced the existing (R3) 150K OHM, 10 percent, 1-watt Monitoring System -
resistor, located in the lon chamber power supply of Unit 2 the neutron monitoring system with a 150K OHM, 5 percent 2-watt resistor. The ECN was completed.
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F 16 PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation The function of the ICPS was not altered.
The possibility of a reactor scram during normal operation was reduced by replacing the resistor. Technical Specifications do not address the Neutron Monitoring-ION system; therefore, it does not reduce the margin of safety.
ECN P0569 - Reactor Replaced air-operated valves FCV-3-98 and -99 with Feedwater System -
air-operated double disc gate valves with pressure seal Unit 1 bonnets. Also, installed associated hangers. The ECW was completed for unit 1.
A seismic analysis was performed by the Office of Engineering. The new valves meet or exceed all the design requirements that the old ones met. The new valves perform the same function but are less likely to experlence leakage.
The margin of safety was not reduced.
ECN P0185 - Standby Rewired the diesel generator protective relaying Diesel Generator circuitry to enable diesel generator loss of fleid trip System - Units 1, 2, to be bypassed during accident' conditions. Performed and 3 post-modification test at the end of work on each diesel. The ECN was completed.
The modification was performed due to NRC-0IE Circular 77-16.
The existing D.G. circuitry permitted a D.G.
trip upon loss of field voltage even when an accident signal was present. The modification assures the appropriate D.G. protection trip circuits are provided with automatic bypass features that prevent them from negating automatic starting or tripping of D.G.'s during fast start or emergency operations.
ECNs P0688 and P0690 -
Modified valves FCV-76-19 and FCV-76-18 to allow the Containment Inecting flange side that cannot be isolated from primary Systems - Units I containment to be tostable. The modification provided and 3 two gaskets (por valve) with a pressure tap between the gaskets to allow the flanges to be tested. The ECN was completed.
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(~~I 17 FLANT NODIFICATIONS SUNNARY Nodifications safety Evaluation Operability of the valves was not affected by the change. The seismic qualification was maintained and leak testing was provided. Therefore, the margin of safety was not reduced.
ECN P0495 - Containment Installed gaskets and drilled test connections in the Atmosphere Dilution inboard flange to allow inboard flange to be testable System - Unit 3 for FSV-84-8A, -88, -8C, and -8D.
The ECN was completed.
The modification was implemented so the valves could pass the leak rate test in meeting NRC requirements on testing primary containment teolation valves. No new possibilities of any new accidents occurring from the modification are created as the function of the valves was not changed. The margin of safety was not reduced.
ECN P0676 - Nain Steam Nodified the Teledyne-Republic Company pulsation dampers System - Unit 3 in instrument sensing lines of main steam pressure switches (PS-1-72, -76, -82, and -86).
The modification included the addition of a shin and a a
buffer material between the pulsation damper cap and the pulsation dampers pin (s). The ECN was completed.
The modification reduced the frequency of spurious NSIV isolation signals and MSIV closure. Only the snubber Internals were modified so pressure boundary remained intact. The snubbers perform the same function as before belns modified. The switches will function properly for their intended use.
ECN P0666 - Residual Installed a one-Inch bypass line around valves74-674
' Neat Removal System -
and -675.
The bypass line included the addition of Units 1 and 3 two check valves (74-829 and -830) and one isolation valve (74-028). The ECN was completed.
The modification did not affect any of the safety-related functions of the RNR system. The addition of the check valves provide double isolation between the RNR system and the condensate system.
A selsmic analysis was performed to demonstrate the Installation of the bypass line and the associated valves did not adversely affect system operation or integrity.
I 18 PLANT MODIFICATIONS
SUMMARY
Modirleations Safety Evaluation ECN P0653 - Main Steam Modified the 10-inch MSRV vacuum breakers and installed System - Units 1 and 3 the modified valves. The changes were installing new modified hinge arm, hingo shaft, hinge bearings and associated equipment; adding a spacer between the 10-inch mounting flange and the valve body to move the valvo pallet (disc) out of the flow stream when the valve is in the open position; and removing valve position indicator. The ECN was completod.
The modified vacuum breaker valvo will perform the same function as the old GPE vacuum breaker valvo. It has the same code classification and solemic qualificat' n.
Testing of the modified valves demonstrated their ability of withstanding worso case service conditions. The probability of the vacuum breakers malfunctioning was reduced. The modification assures the MSRVs are capable of proper operation; thorofore, the margin of safety was not reduced.
ECN P0664 - Standby Fabricated and installed the diesel generator Diesel Generator ventilation hangers SDGV-1 through -6.
System - Units 1 and 3 The ductwork in the Diesel Generator Building covered by the ECN had previously been installed.
The ECN was completed.
The modification was performed to divert the air exhaust from the diesel generators away from the electrical cabinets which are located near the diosol generators. The new ductwork met all design conditions applicable to the diesel generator (seismic, tornado, etc.) so it did not compromise the safety function of the diesel generator. A post-modification test was perto.med to verify the intake cooling air flow requirements to the diesel generators are still being met.
Provlous ovaluations in tho Safety Analysis Report woro not affected; no safoty margins were reduced.
ECN P0662 - Fuel Placed the cavity shleid blocks over the top of the llandling and Storago equipment pool to roduce radiation exposure from the System - Unit 3
' vessel Internals during the unit 3 outage.
Returned the shield blocks back to their normal storago positions. The ECN was completed, f
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i 19 PLANT NODIFICATIONS
SUMMARY
Nodifications Safety Evaluation s
The plant procedure (NMI) was revised to cover this modification. During refueling outages the modification will be implemented per plant procedure to reduce radition esposure from the vessel internals.
The modification was implemented by following the guidelines set forth by NRC in NUREG-0612. The margin of safety was not reduced.
ECN P0631 - Radiation Relocated the RHRSW and RCW in-line detectors, wells Monitoring System -
and associated cables, and preamp 11fers to the south end Unit 3 of the service water discharge tunnels. The ECN was completed.
Relocation of the detectors will allow them to perform their intended function without the possibility of spurious alarms due to a high background radiation source. The detectors were seismically mounted and meet the requirements of IE Bulletin 70-018. The modification allows lower levels of radiation to be detected earlier and more realistic alarm setpoints.
The function of the detectors remained the same.
ECN P0245 - Main Steam Replaced Fischer-Porter transmitters FT-1-13. -25. -36 System - Unit 2 and -50 with Rosemount type. The ECW was completed.
The modification was needed to improve transmitter dampening and reduce noise in the feedwater controls.
The replacement transmitters are of an equal quality and qualifications.
It is espected that improved system control and reliability will be realized.
ECN P0598 - Control Prefabricated and installed intruder barriers in the
- Bay Structure System -
ductwork in the control bay.
The ECN was not completed.
t Common Barrlers have been innta11ed in all other areas to I
meet NRC requirements.
The barriers do not restrict the air flow significantly. No electrical systems were adversely i
affected.
The barriers were analyzed to assure the ductwork and supports were seismically quellfled.
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F 20 PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation ECN P0461 - Control Installed fire dampers in the control bay ventilation Bay HVAC System -
ducts which penetrato the perimeter walls of the nos. 2 Common and 3 battery room complexes, elevation 593. The ECN was completed.
The two new dampers installed are seismic Category I, same as the existing dampers.
Installation of the dampers did not negatively affect HVAC flow patterns in the battery room complexes. Also, the dampers will help TVA meet its commitments for fire protection.
ECN PO422 - Reactor Modified RPS 120V AC power systems to provide redundant Protection System -
class 1E protection between the non-class 1E power Unit I supplies and the class IE battery board RPS distribution cabinets. The ECN was completed.
The modification helps assure the RPS operates properly during an event requiring an automatic scram. All equipment installed met all class IE and environmental qualifications.
Therefore, the margin of safety was not reduced by the modification.
ECN L1952 - High Repaired HPCI restraints R-24 and R-35 on HPCI pump Pressure Coolant discharge piping. The ECN was completed.
Injection System -
Unit 3 The modifications assured the physical configuration of the restraints is adequate to withstand the design seismic loadings. The modification did not adversely affect the operation of the HPCI system.
ECN PO462 - Primary Fabricated strongback fixture and installed on the Containment - Unit 1 inside door of the personnel airlock, Reactor Building.
unit 1, elevation 565'.
The ECN was completed.
The strongback fixture was placed in such a way to prevent structural damage to the inner airlock door.
The modification did not degrade the airlock's seismic qualifications.
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21 PLANT MODIFICATIONS
SUMMARY
1 3 Modifications Safety Evaluation ECN P0441 - Standby Provided physical barriers in the Diesel Generator Diesel Generator Building HVAC ductwork to preve.it access into a vital System - Units 1, 2, area. The ECN was completed.
and 3 The protective barriers do not restrict the air flow significantly. No electrical systems were adversely affected. The barriors were analyzed to assure the HlAC ductwork and supports were seismically qualified.
ECN P0345 - Control Bay The ECN was issued and implemented for decumentation HVAC System - Unit I to as-construct drawings for work perfc..aed on TACF 1-79-64.
The change was to document installation of a check valve in the vent and oil recovery system for the control bay air-conditionir.g chillers. The ECW was completed.
The normal operation of the control bay chillers was not altered by the addition of the check valve.
Adequate cooling will still be provided when necessary.
The possibility for an accident or malfunction of a different type was not created.
ECN P0203 - Reactor The ECN was issued and implemented for documentation Core Isolation Cooling only to as-construct drdelngs and remove TACF. The work System - Unit 2 which was previously performed removed cable 2ES1215 from service and installed wiring in panel 25-31.
Th t, ECN was completed.
Implementing the modification did not affect the starting capability or any operation of the RCIC system.
Reliability was not reduced as the system is a single train system.
l ECN P0384 - Primary Changed the stroke time of cantainment purge valves I
Containment System -
2-FCV-64-17
-18, -19.
-29,,-30. -32, and -33 to 2-1/2 Unit 2 seconds and 2-FCV-76-24 to 5 seconds or less by replacing solenoid valves 2-FSV-64-17. -18. -19. -29
-30. -32, -33 and 2-FSV-76-24.
Replaced air supply tubinf, to 1/2-inch and rotated valves 2-FCV-64-29
-30,
,4nd 76-24.
The ECN was completod, f
l
[
l
F 22 PLANT MODIFICATIONS
SUMMARY
Nodifications Safety Evaluation The purge line isolation valves still perform the same function in the same manner and the more rapid closure timo did not affect this. The reduced stroke time brings the valvo closure time into conformance with NRC Branch Technical Position CSB 6-4.
ECN P0242 - Main Steam Removed packing blood-off valves from the MSIVs and System - Units 1 and 2 capped the bleed-off lines.
Lantern rings were removed during valve maintenance as required. The ECN was completed.
The MSIV stem packing is a portion of the primary coolant loop boundary.
Its function is to decrease the likelihood of a steam leak along the stem. The modification did not alter the function nor affect the MSIV performance. The modification will assure that the leakage rates are not likely to be exceeded, thus the margin of safety was not decreased.
ECN PO450 - Jet Pump Replaced the existing jet pump holddown beam assemblies Assemblies - Unit I with now improved assemblies. The improvement is the type of metal being used for the bolts. The type metal was changed from 304 stainless steel to 316L stainless stool. The ECN was completed. (The work was done previously but documentation was completed in 1985.)
The function of the assembly was not changed and the improved assembly was soismically qualified.
The modification reduced the probability of cracking and jet pump failuro.
ECN L1666 - Radiation Installed process liquid radiation monitor sample Monitoring System -
nortles in RBCCW line on olevation 593' in the reactor
- Unit 1 building and in RCW offluent and RHR heat exchanger effluent lines at south wall of the reactor building on elevation 565' in unit 1.
The ECN was completed.
Removing the on-lino monitors and replacing them with off-line monitors increased the sensitivity of the radiation detection in theso linos and help assure that
,the releases from the plant are within the limits.
f
I 23 PLANT MODIFICATIONS
SUMMARY
Nodifications Safety Evaluation ECN L1848 - Rod Sequence Modified rod sequence control system auxiliary buffer Control System - Unit 2 board panels to provide better access for maintenance.
The ECN was completed.
General Electric reviewed the modification and stated, "the presence or abcence of the access holes do not affect any safety-related function, nor does it adversely affect the panel in which it is located." No other system is affected; therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
ECNs PO440 and PO471 -
Provided inteuder barriers in control bay HVAC ductwork Control Bay 'HVAC System -
in the reactor building. The ECNs were completed.
Common The barriers are seismically qualified and do not negatively affect HVAC air flow. The change will enhance securl'ty.
The modification was performed to comply with 10 CFR 73.2(f) and Physical Security Plan which was identified during an NRC security audit.
ECN P0623 - Various Replaced existing GE multipoint recorders 2-TRS-66-77, Systems - Units 2 and 3 3-TR-73-54, 3-TR-56-4, 3-TR-80-1, 3-TRS-74-80, 3-TR-68-58, 3-TR-68-84, 3-TR-68-97, 3-TR-68-98, and 3-TRS-66-77 with L&N Speedomax Series 250 multipoint recorders.
The ECN was not completed. Other recorders on various systems are to be replaced.
The replacement recorders are as good or better quality than the old GE recorders which are no longer being manufactured. Based on this, the margin of safety was not reduced.
DCR Core Component 24 -
Replaced discharged fuel assemblies with P8 x 8R Fuel Assemblies -
assemblies.
The fuel loading was performed per RTI-3B.
Unit 3 The DCR core component is a continuing item until all fuel assemblies are used; therefore, it was not completed for unit 3.
The fuel assemblies were evaluate'd for neutronic thermal-hydraulic, thermal-mechanical, chemical, and material property characteristics to ensure that the f
margine to safety design limits are maintained.
I I
24 PLANT MODIFICATIONS
SUMMARY
Modifications
. Safety Evaluation ECN P0272 - Neutron A workplan was written for documentation to Monitoring System -
"as-contruct" drawings and clear JIWR and TACF. The Unit 2 work which had been previously performed' rerouted wiring for IRN E (cable number 2NM51-IA) through spare connector 27B instead of 31B (penetration BB) and rerouted IRN F (cable number 2NM84-IB) through spare connector 28B instead of 31B (penetration BD). The ECN was completed.
The rerouting of the cable affected no safety-analysis. The change did not alter the system function nor was the penetration changed.
ECN P0126 - Reactor Installed ECCS ATU inverters with associated cables Protection and and conduits.
Performed functional test on the Emergency Core Cooling inverters. A very small portion of the work covered by Systems - Unit 3 the ECN was completed.
The ECN authorizes the installation of the analog transmitter / trip unit' system for Engineered Safeguard sensor trip inputs. The system replaces pressure, level, and temperature switches with analog transmitter / trip unit combinations.
The modification did not adversely affect any safety-related systems or functions. There were no changes in design basis, protective function, redundagcy, trip point setting or logic. Nothing changed with the implementation of the modification except that the likelihood of failure will be reduced.
ECN P0027 - Fuel The ECN replaces the original fuel storage racks and Handling and Storage control rod storage racks in the spent fuel pool with System - Units 1, 2, new, higher-capacity fuel storage racks and new
'and 3 temporary control rod storago provisions. The ECN was completed on unit 1 but was only partially implemented on units 2 and 3.
Seismic qualification of the racks and their arrangements were evaluated and approved. NRC approval was obtained prior to installation of the new racks.
f
25 PLANT NODIFICATIONS SUNNARY Nodifications Safety Evaluation ECN P0361'- Primary This ECN covers the modifications for the long-term Containment System -
torus integrity program. The major portion of the Units 2 and 3 fieldwork covered by the ECN had been implemented in previous refueling outages., Various modifications to-hangers and supports were implemented during 1985 and will be continued being modified during the unit 2 cycle 5 outage and the unit 3 cyle 58 outage of 1986.
The modifications upgrade and strengthen the torus.
The changes will result in a new, intermediate configuration which is better than the original configuration thus leaving the torus in a safe condition. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety was not increased.
ECN P0370 - Various Installed support for masonry wall 105 on control bay, Nasonry Walls -
elevation 617, between columns R17 and R18. The ECN Unit 3 was not completed.
Analysis was perf'ormed to assure the repaired wall did not adversely affect the existing seismically qualified structure. Therefore, no new accidents are foreseen.
ECNs L1896 and L1916 -
Revised power sources for FSV-32-28A and FSV-32-298 and Control Air System -
made final electrical tie-in for FCV-32-91.
Power Unit 3 source for FSV-32-28A was changed from the plant nonpreferred AC system B to an I&C AC system bus.
FSV-32-298 power was changed from the I&C bus to the plant nonpreferred bus. The ECNs were completed.
The modification increased the reliability for keeping FSV-32-28A and -298 open during normal plant operation. The implementation of the ECNs will prevent spurious scram due to failure of one of the buses.
This prevention will add to the' safety of the plant.
ECN P0079 - Reactor Disconnected cable 3R952 from breaker 311 on panel 9-9
~
Feedwater System -
in the unit 3 control room and tagged the breaker as a Units 2 and 3 spare.
Disconnected cable 2R952 from breaker 311 on
)anel 9-9 in the unit 2 control room and tagged the breaker as a spare.
The modification changed the feedwater system logic which was being fed from I&C Bus e
B to I&C Bus A in order to prevent a unit scram. The ECN was completed.
p-26 PLANT MODIFICATIONS
SUMMARY
Modifications Safety Evaluation The modification increased the reliability of the system by assuring the three logic channels are powered from separate sources. This reduced the probability of an unnecessary transient.
ECN P0533 - Torus This ECN provided the design, procurement, and Temperature Monitoring installation of an improved temperature monitoring System - Units 1 and 3 system for the torus. All major field work was implemented. This system is for control room indication only.
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. The modification provided assurance that the torus temperature is within the prescribed limits set forth in the Technical Specifications; therefore, the margin of safety was not reduced.
ECN P0275 - Off-site Replaced existing degraded voltage relays with new Power - Units 1 and 2 relays for the 4-kV shutdown boards A, B, C, and D.
Changed out the 100K resistors'with 200K OHM resistors. Only a small portion of the work covered by the ECN was imp 1mented. The major portion of the work had been previously implemented.
The system modification as a whole also reduced the probability of losing off-site power due to more sources and no transferring of loads.during startup and shutdown.
a f
NUMBER OF FER$02NEL AND C AN.REN BY CCRK AND J!J FUICTION 1
PLA2T1 BROW %S FERRY huCLE AR PLAMT 1985 10158 TIESDAYe.FEBRUAPY 18, 19R6 ATLKH12E E CTIBER EF PERSONNEL 4 > 10 0 M.R ED )
TOT 4L MA%. REM
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- j GRCUP STATION UTILITY CONTRACT 10TAL STATION UTILITY CCNTRACT TCTAL EMFLOYEES EMPLOYEES A NC OTUERS PERSONS E MP LO YE ES E MP L OY E E S AND CTHEES P* REMS j
MAINTENANCE FERSONNEL 738 48 30 816 35.749 3.931 2.3C6 41.986 3PERATING PE 950NhE L 77 2
0 79 13.*36 0.280 0.0C0 16.216 HE ALTH PHYSICS PERSONNEL 96 4
111 211 26 681 0.859 42.411 69.951
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SUPIRVIS?RY FERSONhEL 15 0
1 16
?.131 0.t00 0.015 2.146 ENGINEERING PERSONNEL 133 12 44 189 17.519 1.792 6.1 03 25.414 I
.......g....
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93 1059 66 186 1311 94.016 6.862 5f.635 15t.713
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UTILITY CohTPACT 10TAL STAT!ON UTILITY CChTRtCT TCTAL
'l E MPL O Y E E S E MPLO Y E E S AND OTHERS AERSONS EMPLOYEES tMFLOYEES Akt CTHERS P. REPS
.I 9tINTENANCE FERSON4EL P54 44 94 992 24*.317
- 4. F. 4 4 2P.151 2P2.352 OPERATING PERSONNEL 75 2
0 77 5.654 C.007 0.CC0
- 5. eel MEA LTH PHYSI CS PER SONNEL 92 3
109 204 11.692 0.522 18.P?4 32 10e SUPE 9 V I S )k Y FERS01MEL 13 1
2
. 16 1.217 0.000 0.24T 1.464 E%GINEERING FERSONNEL 137 12 43 192 25.903 0.865 2 604 25.372
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M3 1171 62 248 1981 293.783 6.238 S t.9 36 350.957
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............................................... MO SPECIAL MAINTENANCE...................'...=............................
d GROUP ST AT ION UTILITY CONTRACT EDTAL STATION UTILITY CCf.T R A CT TCTAL EMPt0 VEES EMFLOYEES AND OTHERS RERSONS EMPLOYEES EPPLOYEES AhC CTFEPS P. REPS MAINTENANCE FER$0NhEL 664 73 126 863 17*>.13 6 41.738
?F.323 315.187 S
OPERATING PERSONNEL 31 0
e 31 1.147 0.CCD D.C 3t 1 147 HE ALTH P4YSICS PERSONhEL 56 0
100 156 4.440 0.000 2a. 6 05 34.245 SU PE; VIS S R Y FEPSONNEL 11 1
2 14 2 559 0 314 e.143 3.C13 ENG14EERING PERSOMNEL 96 5
22 123 15.327 0.273 4.732 21.332
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1187 199.609 42.325 135.000 374.*34
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r.aEMS CAINT[4A1CE PERSONNEL 129 0
-6 135 4.613 0.000 C.5te 5 173 J7E4ATING PEF 50t.NEL 13 O
0*-
13 2.089 C.0C0 C.003 2.029 l'
HE ALTH P1YSICS PERSONhEL 62 1
42 105 1 959 0.000 t'. P l ?
2.672 SUPERVJSORY PERSCNNEL 0
0
- l 1
3.000 0.000 C.CCC C.000 ENSIMEER ING PERSONNEL 3
C
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PLA;T1 BRIWMS F ER;.Y NUCLE A~l PL A67 1995 19358 TUESDAY,'FrPPUsPY 18, 1986'
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-EMPLOYEES AND OTHERS P ER SONS EPPLOYEES EMP L OYEE S ANC OTHEPs Pa#Er!
R4INTENANCE FERSONNEL 88 0
0 88 4.282.
0.CCC 0.CCC 4.292 4
OPERATING PEASONNEL 22 1
e 23 9.891 C.004 E. C.*?
C.f95 HC ALTH P'tTS I CS PER S0*.t.EL 15 0
21 36
!.C47 0.000 C.T35 C.782 SU P ER V ISO R Y 6Eq$0NNEL 1
0 0
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0 30 D.464 C.000 C. S C!
0.464-b 10 9
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21 178
- .684 C.C04
- 3. T M 6 423
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0 223 25.717 0.291 C.((0 26.0[6 hE ALTH PHYSICS PERSONREL 321 9
383 712 44.719 1 381
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6 48 5.907 0.314 0.4E2 6.623 b
(NGIN?E9It.G PERS0hhEL 399 E9 113 541 61 219 2.930 13.999 77.14e
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TOTAL NUMBERS OF INDIVIDUALS I
GROUP
,S T A TICN UTILITE CONT R A CT TOTAL b
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0 78 HE ALTH PHTSICS PERS0khEL
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11 0 216
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1 TENNECSEE VALLEY AUTHORITY
)
CH ATTANOOGA. TENNESSEE 37401 SN 157B Lookout Place J !' ' 5 A10 - l 4 p r2.
February 28, 1986 hlY}$) ?)
l U.S. Nuclear Regulatory Commission Region II ATTN:
Dr. J. Nelson Grace, Regional Admittistrator 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30303 p
Dear Dr. Grace:
- $NNUAL' OPERATING REPORT - JANNANT -1 1964s 30. nace 31 1984 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 - eOCKEI EOS, 50-259, 50-260, 50-296 p
OPEHATING LICENSES DPR-33, DPH-52, AND DPH-68 Purcuant to Browns Ferry Nuclear Planit Technical Specification 6.7.1.b.
enclosed is the annual operating report for browns Ferry units 1, 2, and 3 for the period January 1 - December 31, lugs.
Very truly youru, TENNESSEE VALLEY AUTHoh1TY H.
ridley Manager of icencitig Enclosures (2) cc:
See page 2 1
gi oc s@f-t Cf - JY An Equal Opportunity Employer
(, y ni
n.
e v.
a Dr.'J. Nelson Grace February 28, 1986 cc (Enclosures):
Mr. James Taylor, Director (2)
'gOffice of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Mr. Harold R. Denton, Director (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
205S5 Director of Nuclear Reactor Regulation (2)
Attention:
Mr. D. R. Muller, Project Director BWR Project Directorate No. 2 Division of Bolling Water Heactor Licensing U.S. Nuclear Hegulatory Commission Washington, D.C.
2c355 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Director, Nuclear Engineering and Operations Department Electric Power Research Institute P.O. Box 10412 Palo Alto, California 94303 Director (2)
Office of Management and Program Analysis U.S. Nuclear Hegulatory Conaission Washington, DC 20555 Director (1)
Alabama Water Improvement Commission Public Health Services Building Montgomery, Alabama 36130 Chief, compliance Branch (1)
Enforcement Division Environmental Protection Agency Region IV 345 Courtland Street,ll.E.
Atlanta, Coorgia 30308
_m