ML20235M533
| ML20235M533 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/31/1986 |
| From: | Robert Lewis, Walben J TENNESSEE VALLEY AUTHORITY |
| To: | |
| References | |
| NUDOCS 8707170227 | |
| Download: ML20235M533 (39) | |
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TENNESSEE VALLEY Alml0RITY OFFICE OF NUCLEAR POWER 4
ANNUAL OPERATING REPORT BROWNS FERRY NUCLEAR PLANT l
January 1, 1986 - December 31, 1986 .
t Docket Numbers 50-259, 50-260, and 50-296 License Numbers DPR-33, DPR-52, and DPR-68 Submitted by:
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sSite Director 4 e
Submitted by: ,
$ hPlant Manager w s4 sko/A7 (
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8707170227 061231 PDR ADOCK 05000259 (k, i
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' TABLE OF CONTENTS 10 CFR 50.59 Requirements i New Procedures and Procedure Changes . . . . . . . . . . . . . . Attachment A Temporary Alterations to Plant Equipment . . . . . . . . . . . . Attachment A )
1 Critical Systems, Structures, and Component Tests and Experiments for 1986 . . . . . . . . . . . . . . . Attachment B ,
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Plant Modification Summary . . . . . . . . . . . . . . . . . . Attachment C Technical Specification 6.7.1.b Requirements Challenges to or Failures to Main Steam Relief and Safety Valves. . . . . . . . . . . . . . . . . . . Attachment D Occupational Exposure Data . . . . . . . . . . . . . . . . . . . Attachment E .
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- 1 ATTACHMENT A l NEN PROC'EDURES AND PROCEDURE CHANGES
-January 1, 1986 - December 31, 1986-Procedure: This procedure addresses a breach of Wheeler Dam added "
to the Emergency Plans Manual.
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Description:
The new procedure lists actions to be taken in the event
. a breach of Wheeler Dam occurs.
Listed actions are based on applicable portio'ns'of the Safety Analysis:
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Safety Analysis Report (SAR) and place the plant in a safe condition. A breach of Wheeler Dam is addressed in the SAR and the procedure will implement actions consistent with the SAR.
Procedure: Upgrading of Emergency Operating Instructions (E0I) one-and two. -
Description:
E01-1 (Reactor Control) has undergone a general revision to include directions.for mitigating anticipated transients without scram (ATHS) events, to incorporate.
procedural improvements, to implement control room design review recommendations and to change the -
procedure format.
Safety Analysis: Safety Evaluation on next page.
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ATTACHMENT A I
New Procedur,es and Procedure Changes (continued) l l
l Safety Evaluation and Justification l The requirements of 10 CFR 50.59 state that a change cannot be made_without NRC approval if an Unreviewed Safety Question (USQ) is involved or a technical specification change _is required. The NRC has reviewed and approved (see '
references 5 and 6) the generic Emergency Procedure Guidelines (EPGs) (revision 3) for the upgrade of E0Is. The E0Is have been updated to a' draft revision. A detailed comparison of these E0Is to the EPGs was performed in order to identify the deviations of the E0Is from revision 3 of the EPGs and to verify-that the ,
deviations do not constitute a USQ. Discussion of the' deviations follow. The $
deviation 1s underlined. f In EPG step RC/L, if any control rod is not inserted to or beyond position 02, the automatic depressurization system (ADS) logic is bypassed. The reason for this is -
to prevent automatic depressurization during an ATWS event. Flooding with the l reactor not shut down would cause a severe reactor power excursion. This action.is- ]
determined not to involve a USQ for two reasons. Revision 3 of the-EPGs requires l ADS to be bypassed for other events. NRC reviewed and approved this action because j specific directions are given in the E0Is when a manual ADS is required. This ]
action for ATHS follows the same logic. The second reason is this action is taken- 4 only in the event of an ATWS, an event which lies outside the design bases of BFN.
Therefore, this action does not conflict with the Final Safety Analysis Report (FSAR) or technical specification.
In EPG step RC/L, and elsewhere, directions are given to bypass the reactor core q isolation cooling (RCIC) low pressure isolation setpoint. This setpoirt is~not !
part of the primary containment isolation system and is in place for equipment l protection, i.e., to prevent the RCIC turbine from stalling. This action is called )
for when no low pressure injection systems are available, a condition outside j design bases (multiple failures). No assumptions in the FSAR are based on this- 1 logic. Therefore, a USQ is not involved. -l In EPG step C4, Reactor Flooding, directions are given to bypass the high level turbine trip for RCIC. The technical specification bases do not provide a basis for this trip. The FSAR~section 4.7 does not provide a design basis'for the RCIC high level turbine trip. Section 7.4 states the basis for the high level trip for high-pressure coolant injection (HPCI),is that HPCI has performed satisfactorily and to protect the turbine from moisture carryover. This action is' required only j if reactor water level cannot be determined and if conditions for flooding cannot be established using the preferred low-pressure motor injection systems. This i condition is outside the design basis conditions; and since RCIC is not assumed to !
operate during a design basis event, this action does not constitute an USQ.
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i ATTACHMENT A a New Procedur,es and Procedure Changes (continued)
Safety Evaluation and Justification (continued)
In EPG step RE/L-2, directions are given to bypass the HPCI high suppression pool water level suction transfer logic if suppression pool temperature is greater than 140*F. This action is specified because HPCI could be lost due to high bearing oil temperature or loss of net positive suction head (NPSH). The technical specification bases state that NPSH may not be available if suppression pool temperature reaches 140*F without containment overpressure. Containment "
overpressure is never assumed per the General Design Criteria and, therefore, i operation of HPCI with suction from the suppression pool with a temperature of or' ;
above 140*F is outside the design bases and does not constitute a.USQ.
The alternate shutdown cooling steps are not implemented._ This was perceived to be a nonemergency action and there are no assumptions in the FSAR dependent on alternate shutdown cooling.
The steps requiring flooding because of elevated containment pressure are omitted. -!
Further analysis by the Boiling Water Reactors Owners Group (BWROG) Emergency Procedure Committee (EPC) drew the conclusion that no additional containment '
depressurization benefits could be obtained by flooding'subcooled water out a -!
, break. There are no conditions or events in the FSAR which depend on this action. !
l Caution 14, which states not to depressurize below the HPCI isolation setpoint (100 psig) unless motor driven pumps are available, is omitted. The HPCI turbine by itself uses enough steam to depressurize the reactor below 100 psig and directions are given to bypass the low pressure isolation on RCIC. -
EPG steps calling for an alternate means to inject boron are omitted. Based on the low probability of an ATWS event coupled with the low probability of standby liquid control failure, it was decided not to implement this step.now. There are no requirements in the FSAR or technical bases for this action.
In EPG step C1 and subsequent steps, operation of systems is to be accomplished regardless of NPSH. This is done only when adequate core cooling or the containment is threatened. This action is implicit in the EPGs but is spelled out in the E0Is. A condition requiring this is outside the BFN design bases.
In EPG step C2-1.3, the step does not specify using the alternate reactor pressure vessel depressurization systems in the order which will minimize radioactive release to the environment. This cannot be known in advance and because it is more ,
important to get the reactor depressurized quickly (some systems may require lengthy time to line up) to prevent failing fuel or failing containment. Because this difference is incremented due to human factor concerns, it is not in a disagreement with the intent of the EPGs. In addition, this action is only required for events which are outside the design bases.
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ATTACHMENT A ,
1 New Procedur,es and Procedure Changes (continued)
Safety Evaluation and Justification (continued)
In EPG step C5 (Level / Power Control) and C4 (Reactor Flooding), steps call for J bypassing the low-pressure coolant injection (LPCI) timer to allow throttling of the LPCI system to prevent rapid ficoding for a core which is not shut down.
Although not explicitly called for in the EPGs, it is implicit in the instruction ,
to " commence and slowly increase injection . . ." Therefore, it is concluded this action has been reviewed and approved. "
In EPG step C5 (Level / Power Control), actions are predicated on the flow stagnation water level rather than the reactor flow stagnation power. After approval of Revision 3 of the EPGs, Electric Power Research Institute, Oak Ridge National j Laboratory (see reference 14), and others determined the reactor power could be as I high as 17 percent with the water level at top of active fuel. In attempting to i control reactor power above eight percent but as low as practicable, it was feared I the core might be partially uncovered. Therefore, at the recommendation of the BWROG EPG, the actions were changed to be predicated on level. This accomplishes ;
the intent of the EPG steps to ensure adequate core cooling but to minimize reactor power as much as possible during an ATHS.
In EPG step C5-1, directions are given to bypass the main steam isolation valve 2 (MSIV) interlocks (water level) if any MSIV is open. This is for the purpose of maintaining the condenser as a heat sink when the level is purposely lowered below ) '
the isolation setpoint. Additional required conditions are at least one MSIV already being open to indicate the MSIVs have not closed on low reactor pressure in RUN mode, high steam flow, high radiation, or high tunnei temperature due to a leak. If the MSIVs closed on any of these, they are not reopened. This agrees with the philosophy in the override preceding EPG step RC/P-2 and thus has received NRC approval.
In EPG step.C7, an additional requirement of reactor water level at two-thirds core 1 height has been added before injection is terminated from sources external to l primary containment. General Electric Company studies have shown that for BHR-4s !
core spray by itself may not provide adequate core cooling for a completely uncovered core in a steam environment. If level is maintained at two-thirds core height, adequate core cooling is assured and the event falls within the design I bases. If level cannot be maintained .at or above two-thirds core height, the event is outside design bases and the containment must be flooded to establish adequate core cooling. This does not conflict with any assumptions in the FSAR.
In E0I-2, step PC/P-5 states to vent regardless of the offsite radioactivity release rate. While this is not explicitly stated in the EPG step, review of ,
reference 3 reveals this is part of the step's basis. Therefore, this action has l received NRC approval. Also note that the action is only called for in an event outside design bases (i.e., primary containment pressure exceeds 55 psig).
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ATTACHMENT A New Procedures and Procedure Changes (continued)
Safety Evaluation and Justification (continued)
In_ E0I-2, the containment is vented regardless of offsite radioactivity release rate before drywell sprays are used regardless of whether adequate core cooling if assured. In the EPGs the steps are reversed. The reason for this is that in the EPGs it was assumed the primary containment pressure limit was the pressure at which the containment would rupture (or twice design pressure) if not vented. As an open item from reference 5, the criteria for defining containment venting pressure needed to be determined. It was determined that the main steam relief valve and the containment vents would be inoperable at the primary containment rupture pressure. The drywell coray step is executed only if venting is unsuccessful.
Therefore, this step arrangement agrees with the intent of the EPG steps to protect containment. Again neither of these actions is taken as long as design basis conditions are not exceeded. This concludes the analysis of deviations from the EPGs.
None of the deviations constitute a USQ because (1) the actions are the correct -
things to do if events exceed design basis event assumptions and they are not required unless conditions do exceed design basis events, or (2) the actions have received previous NRC approval (references 5 and 6).
I The plant-specific action levels have been calculated and independently verified (references 9-11). Control room walkthroughs, desk top reviews, and s'imulator exercises have been completed to verify that the E0Is (1) perform the intent of the I technical guidelines, (2) are applicable to the plant in terms of equipment available and calculation of graphs and action levels, and (3) can be performed by the operator. Based on NRC approval of the EPGs and on the analysis of the deviations from the EPGs involving no USQ, it is concluded these procedures may be implemented.
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ATTACHMENT A New Procedures and Procedure Changes
- References
- l. Emergency Procedure Guidelines Revision 3 BWR l-6, December 22, 1982.
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- 2. Emergency Procedure Guidelines Appendix A, Revision 31, March 1984.
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- 3. Emergency Procedure Guidelines. Appendix B. Revision 3A, March 1984.
- 4. Emergency Procedure Guidelinas Appendix C, Calculational Procedures, 'l Revision 3, June 1, 1983 (and subsequent updates).
- 5. Memorandum from Darrel G. Eisenhut, NRC, to BWR Licensees, dated February 8, 1983, subject: SAFETY EVALUATION OF " EMERGENCY PROCEDURE GUIDELINES, REVISION 2", NED0-24934, June 1982. ;
- 6. Memorandum from D. M. Crutchfield, NRC, to Tom Dente, Chairman, BHROG, dated November. 23, 1983, subject: SAFETY EVALUATION OF " EMERGENCY PROCEDURE ;
GUIDELINES, REVISION 3." l
- 7. Letter from L. M. Mills, TVA, to H. R. Denton, NRC, dated June 22, 1984, .;
which transmitted tne Emergency Operating Procedures Generation Package. ?
- 8. NUREG 0899, " Guideline for the Preparation of E0Ps."
- 9. Memorandum from G. T. Jones to J. A. Coffey dated November 29, 1984, !
subject: BFN-E0I Calculations. -
- 10. Memorandum from T. F. Ziegler to G. T. Jones dated January 25, 1985, lj subject: BFN - E0I Calculations.
- 11. Memorandum from G. T. Jones to J. A. Coffey dated November 29, 1984, subject: BFN - E0I Calculations .
- 12. Emergency Procedure Guidelines, Draft revision 4AB, October 31, 1985. .
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- 13. Letter from R.'Gridley, TVA, to D. R. Muller, NRC, dated March 26, 1986.
- 14. Letter report from R. M. Herrington, ORNL, to Dr.. T. J. Walker, NRC,' dated November 13, 1985, " Evaluation of Operator Action Strategies for Mitigation of j MSIV Closure Initiated ATHS."
- 16. BFN TS, Unit 1, 2, and 3. l 6 l
e Attachment A
. TEMPORARY ALTERATIONS Alteration: Change the setpoint of two pressure switches (PS) from 450 psig to 400 psig in the Core Spray Cooling System.
Description:
The setpoints were lowered to allow the alarms to go off sooner. ,
4 Safety Analysis: The setpoint of PS-75-24 and PS-75-52 will be lowered 50 psig which will allow the alarms PA-75-24 and PA-75-52 to be received at an earlier system pressure. The lower setpoint will allow the operator more time to react before the relief valve setting is reached. No other equipment will be operated or adjusted outside of normal design parameters.
Alteration: Install an alternate route of yard system fire protection to the offgas building.
Description:
Normal supply has excessive leakage.
Safety Analysis: The offgas post-treatment isolation function will remain intact. Because a roving fire watch is assigned there will be no fire protection lost.
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Attachment B CRITICAL SYSTEMS, STRUCTyRES, AND COMPONENTS (CSSC) TESTS AND EXPERIMENTS ST85-06 '
Special. Test: This test is to provide documented data on local' leak rate-testing globe valves with pressure applied aver and under the disc.
Description:
The test consists of bench testing various size globe valves ;
using air and water as test media. The results of the test I will be used to ensure present test methods-are sufficient.
Safety Analysis: 'This test will not be performed on equipment while it is installed.in the plant. Therefore, there are not-any affected systems and safety is not infringed upon.
ST85-21 Special Test: This special test is to obtain information about the actuation forcesHof two Flow Control Valves (FCV) in the -
Residual Heat Removal System (RHR).
Description:
The data obtained from this test will be used to determine the validity and accuracy of local. leak rate tests of these valves.
Safety Analysis: The testing of FCV-74-58 and 59 will be conducted in compliance with TS 3.5.B.9. One loop of RHR will be available throughout the test.
S705-20 Special Test. This test is to evaluate High-Pressure Coolant Injection (HPCI) System operation under degraded voltage conditions and record various operating parameters of the HPCI turbine governor controls.
Description:
The reactor will be in the cold shutdown mode. This'special test will prove HPCI operability before reactor startup'in the degraded voltage condition.
Safety Analysis: The test will be performed during cold' shutdown mode and the HPCI system will be returned to normal with operability verified after the test.
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i Attachment B j ST85-27 Special Test: This tes,t evaluates operation of the HPCI system with a 440-ohm dropping resistor in place of the 500-ohm resistor on the HPCI system. ]
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Description:
Previous special tests have demonstrated that the HPC1 system will not operate properly during degraded voltage conditions (STEAR 8520). If the 440-ohm resistor passes the test, it will be recommended as a permanent fix. 'T Safety Analysis: The test will be performed during cold shutdown mode and the HPCI system will be returned to normal with +
operability verified after the test.
ST86-08 i
Special Test: This test is to perform a proposed revision to the monthly Surveillance Instruction (SI) 4.9.A.1.a which demonstrates operability of a diesel generator monthly.
Description:
The test is to verify the SI's adequacy in determining monthly operability requirements of a diesel generator. .
Safety Analysis: In the present condition of the plant, making one emergency generator inoperable to perform the test is allowed. This test does not place the plant in any condition outside the analysis provided for in the FSAR.
ST86-09 Special Test: This test is to verify the adequacy of a proposed operability test revision. ,
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Description:
The test is to assure that the emergency diesel i generator will perform as designed after required maintenance is conducted. This includes starting an 3 available diesel,, running and checking all auxiliary ]
systems and verifying technical specifications can be satisfied.
. Safety Analysis: In the present condition of the plant, making one emergency generator inoperable.to perform the test is ,
allowed. This test does not place the plant in any l condition outside the analysis provided for in the FSAR.
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Attachment B ST86-15 Special Test: This test is to add copper nitrate solution to liquid radwaste tanks, recirculate the tanks, and sample them periodically.
Description:
This test applies to the floor drain sample tanks.
Safety Analysis: System operation during the special test will not result in any increase in the radioactivity in the waste liquids. In addition, none of the design safety features listed in the FSAR will be affected.
ST85-22 Special Test: This test will measure the standby liquid control (SLC) ,
system pump suction pressure by installing a pressure transducer and performing a pump recirculation flow test.
Description:
The test 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.
Safety Analysis: The addition of the pressure transducer has been ,
examined and found not to degrade the system's integrity or reliability. The test requires the operation of the SLC system in a mant.er consistent with approved operating procedures.
ST86-06 >
Special Test: The test is to check the logic downstale input condition for the reactor ,and the refueling zone ventilation radiation monitors. This test is performed by removing low trip relays from service, and verifying the associated radiation monitors will not cause misoperation of the Primary Containment Isolation System.
Description:
The test is to verify that the ventilation radiation monitors do not cause malfunction of the primary containment isolation system or secondary containment.
Safety Analysis: Any logic rendered inoperable by this test is tripped in the safe direction thus the ability of these monitors to initiate a Group VI and secondary containment isolation is not impaired and the test remains in the bounds of the FSAR.
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Attachment B ST86-10 ,
l Special Test: This test determines the time required to depressurize '
the scram air header using only the backup scram valves.
Description:
The test will disable the reactor protection system I (RPS) which is allowed by technical specification 3.1 when fuel has been removed from the vessel. 1 Safety Analysis: The performance of this test will not degrade the safety of the plant because all fuel is removed from the e reactor vessel as a prerequisite and will not affect any system or equipment required to maintain fuel in the spent fuel pools in a safe configuration.
I ST86-16 Special Test: This test is to recover parts of HCV-23-31 which may have entered the heat exchanger. ,
Description:
The test will flush the RHR heat exchanger 2A at rated .:
flow for five minutes.
Safety Analysis: Since the reactor fuel is removed from the vessel the only heat load that could be placed on the RHR system is {
the fuel pool cooling system. In the present plant j configuration, RHR is not required to meet this -
t possibility. 1 ST86-20 .
Special Test: The test chemically decontaminated the isolated reactor water cleanup (RHCU) pump for unit 1.
Description:
This test evaluates commercially available chemical 1 decontamination. solvents to best determine a solvent for i plant use. )
Safety Analysis: The RWCU pumps are not critical systems, structures or component items, and they will be adequately isolated i from the remairnier of the system during i decontamination. Pumps are not an integral part of the plant reactor safety system.
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1 Attachment B ST86-05 Special Test: The test is to. verify the operability of the radwaste box compactor upon completion of maintenance.
Description:
The compactor is located in the turbine breezeway at '
T 14 1/2 A line, elevation 577.
Safety Analysis: T.he compactor is not safety related nor is it located in an area where it could interfere with plant safety '
systems. .
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ATTACHMENT C PLANT MODIFICATIONS
SUMMARY
l JANUARY,1, 1986 - DECEMBER 31, 1986 Modifications Description and Safety Evaluation Summary Engineering Change Replaced Foxboro Model 611 DM transmitters, .l Notice (ECN) P0244 PDT-68-65 and PDT-68-82,.with Rosemount Model '*
Reactor Water 1151 HP7822PB. Performed postmodification Recirculation test. The ECN was completed for unit 2. The System - Unit 2 modification has not been implemented on units l' and 3. _ j The Rosemount transmitters will serve the same functions as the Foxboro model and their use will not affect the reactor water recirculation system I performance. The Rosemount models have been proven to be more reliable, therefore, their use will enhance the function served by the transmitters. No technical specification safety margins were affected.
ECN P5044 Redesigned the radiation konitor shielding box Reactor Building for RE-90-131 and support. The ECN was' totally .
Closed Cooling completed as it only covered unit 3.
Water (RBCCW)
System - Unit 3 The function and performance of the radiation monitor and the RBCCW system were not degraded by the -!
modification. A seismic' analysis was. performed to 1 ensure the shielding box and support met seismic _ l Category I requirements. The' margin of safety was not reduced.
ECN P5133 The modification was a documentation change only RWCU . to correct "as-constructed" drawings for unit 3.
System - Unit 3 The change involved adding a note to the drawing (47W 810-1) that FE-69-101 and test connections are ,
for unit 2 only and the equipment is not on unit 3.
The ECN was totally completed as the ECN only covered unit 3. ,
1 Adding the no'te to the drawing or not having FE-69-101 in the RWCU line did not adversely affect any safety-related equipment or function.
This change did not affect any present technical specification. Thus, the margin of safety as defined in the basis for any technical specification was not reduced.
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( .I ATTACHMENT C Modifications _ Descr'iption and Safety Evaluation Summary ECN P5181 Fabricated and installed permanent supports for !
-Drywell Penetrations instrument lines at inboard penetrations and in Unit 3 unit 3 drywell as required. The ECN was totally .
completed as it only covered unit 3.
The permanent supports will maintain physical i separation between the stainless steel instrument i lines and the carbon steel sleeve. The. supports' i ensure that the slope of the instrument lines is )
adequate through the penetration sleeves. The seismic qualification and operability of.the .
instrument lines were not adversely affected by the addition of the supports. Based on the safety .
evaluation, the margin of safety was not reduced- . -j
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ECN PS220 Provided seismic support for junction box (JB) 5786. 1 Standby Gas The ECN was totally completed as.it only covered Treatment (SGTS) unit 1. >
System - Unit 1 I The junction box mounting now meets the seismic Class I. -- l requirements. No safety-related function was adversely affected. The function of SGTS'was maintained as intended. Implementation of the. ,
modification assured that the affected SGTS cabling I will nc:t fail due to a seismic event. Based on this, .
the margin of the safety:was not reduced. -
ECN PS247 Added a diagonal channel brace to panel 25-5A. The Reactor Protection System (RPS) - Unit 2 channel brace was added to the panel in a location i adjacent to the diagonal brace on panel 25-58. The i ECN was totally completed as it only covered unit 2. .)
The addition of the brace provided a separate diagonal channel brace for both panels 25-5A and 25-58. The modifications. did not adversely. affect the operation or qualification of the instruments located on th.e panel. The addition of the brace provided the required seismic qualifications-for the
. panel. Therefore, the margin of safety was not reduced.
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ATTACHMENT C Modifications Descr"iption and Safety Evaluation Summary j I
ECN P5249 Replaced all'the 1/2-inch thick rail clamps and the, 1 Reactor Building associated bolts on the reactor building overhead Overhead Crane - ,rane with one-inch thick rail clamps and the Units 1, 2, and 3 appropriate bolts. The ECN was completed. ,
The modification restored the seismic qualification of the crane and did not adversely affect any other equipment or components. The modification.did not I adversely impact operability of the crane.
Therefore, the margin of safety was not reduced.
ECN PS237 Modified the fuel-pool cooling pumps discharge flange Fuel Pool Cooling connections to eliminate the possible overst.'ess that System - Units 2 and 3 resulted from-bolting the pump discharge flange to ;
the discharge p.ipe flange. :The modification removed '
the raised face' from the flange on the pump discharge piping so that the pump flange is bolted to the pipe 4 flange in a flat face to flat face configuration.
The ECN was completed for units 2 and 3. The i modification has not been performed on unit 1. i The modification met or exceeded all the design t requirements for the spent fuel cooling system. The !
change did not affect the ability of the system to fulfill its design basis. The change'only' altered the flange configuration. No new failure possibilities are foreseen. The margin of safety was-not reduced.
ECN P5258 Redesigned the bonnet drain line vibration support Residual Heat Removal for valve HCV-3-74-55 to withstand the RHR System ,
(RHR) System - Unit 3 service conditions. The ECN-only covered unit 3, but !
was not completed as painting of the new steel has not been implemented.
The modification improved the support's ability to perform its f. unction of preventing failures due to excessive system induced vibrations. Division of Nuclear Engineering (DNE) seismic' evaluation ensured that the affected system and components remained seismically qualified. The margin of safety was not '
recced.
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Modifications Descr'iption and Safety Evaluation Summary i ECN P5304 Installed new seismically qualified holddown studs-Diesel Generator for the diesel generator battery racks for units 1,
-Battery Racks - 2, and 3. )
I Units 1, 2, and 3 The ECN was completed. The old holddown studs were of material not suitable for welding. The 11ew j holddown studs are of a compatible material with l embedded plates which allowed them to be welded
- j together. No safety-related function or equipment I was adversely affected. The modification allows the ]
racks to be seismically qualified. The margin of l safety was not reduced. No technical specifications were affected by the modification.
ECN P5354 A workplan was written as authorization for the use Torus /Drywell of Valspar 78-00 as a coating material for use in Coatings - Unit 2 repairing / replacing existing torus vent header i coating. The vent header coating is being done in accordance with construction specification N/A-919 for the air space of the torus. The ECN was totally .
completed as it only covered unit 2.
The Mobil 78 coating passes the test criteria for a design buis accident and has the best record reported in the industry for coatings in Mark I type The Valspar coating is equivalent to containments. I the Mobil coating. The coatir:g of the torus /drywell '
is not defined in the basis for any technical specification. Considering this and the fact that ,
there are no failures created by this change that !
could affect the operation of other systems or l functions, the margin of safety as defined in the basis fcr any technical specification is not reduced.
l ECN P3028 Replaced existing pressure switches PS-84-21 and Containment Atmosphere PS-84-22 with environmentally qualified Dilution System - (static-0-ring) pressure switches. The ECN was Unit 2 completed for unit 2. Work had been implemented previously on unit 3, but is remaining to be done on unit 1.
A seismic analysis was performed to ensure the original design requirements were met. The function of the equipment remained the same, therefore, the ;
margin of safety was not reduced. !
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ATTACHMENT C Modifications Descr"iption and Safety Evaluation Summary i
ECN P3061 Replaced level switches LS-73-56A and LS-73-56B. The j High Pressure Coolant ECN was completed for unit 3. Work is being Injection (HPCI) implemented on unit 2 and remains to be implemented System - Unit 3 for unit 1. ~
The new equipment meets the same requirements and i performs the same functions as the original ]
equipment. The new equipment is environmentally ~ '
qualified for their respective environment. The technical specification addresses the availability of i the HPCI system. Since the function of the system was not changed, the margin of safety was not reduced.
ECN P3106 Replaced flow transmitter FT-84-19 and -20, added Containment Atmosphere zener diodes, and replaced resistors. The ECN was Dilution System - completed for unit 3. The modification is being Unit 3 implemented on unit 2 and work remains to be done on unit 1.
The new equipment 1s environmentally qualified for .
, their respective environment. The function of the equipment remained the same. Therefore, the safety objectives defined in the technical specifications were not reduced.
ECN P3138 Replaced the RHR pump's cooler fan motors. The ECN ,
Residual Heat Removal was completed on unit 3. The work has been completed System - Units 2 and 3 for pumps 2A and 2C and is in progress on pumps 2B and 20. None of the modification has been implemented on unit 1.
The replacement motors are environmentally and l seismically qualified. The modification was performed so that the minimum number of RHR pumps was always available, therefore, the margin of safety was not re6uced.
ECN P3139 Replaced the core spray pump room cooler fan Core Spray System motors. The installation work was completed on Unit 2 unit 2, but the postmodification test remains to be completed. The ECN was completed previously on unit 3. Work is remaining to be implemented on unit 1.
The replacement motors are environmentally and seismically qualified. The modification was performed so that the. minimum number of core spray pumps were always available, therefore, the margin of safety was not reduced.
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ATTACHMENT C Modifications Description and Safety Evaluation Summary ECN P3134 Replaced the mounting brackets for 2-FM-84-198 Containment Atmospheric and 208. The ECN was completed for unit 2. The Dilution (CAD) System - modification has been implemented on unit 3, but is j Unit 2 remaining to be done on unit 1. ,
i Implementation of this ECN assured seismic qualification of.FM-84-198 and.FM-84-208. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to, safety previously evaluated in the SAR is not
, increased.
ECN P0707 Replaced the 24V DC power supplies PX-99-A1, Reactor Protection PX-99-A1A, PX-99-A2, PX-99-A2A, PX-99-B1, PX-99-81A, System - Unit 2 PX-99-82, and PX-99-82A.in the RFS panels 9-83, 9-84, 9-85 and 9-86 with Nutherm model 55320 power supply.
The ECN was completed for unit 2. The work is remaining to be implemented on units 1 and 3.
The new power supply provides a system improvement -
that decreases maintenance and service problems. The modification did not create the possibility for an accident or malfunction of a different type than'any evaluated previously in the SAR. The TS values for the RPS were unaffected by the modification. -
ECN P0709 Installed a throttling.(globe) valve ~in the emergency Standby Diesel equipment cooling water (EECW)-supply to the diesel Generator System - generator engine cooler; locked the existing Unit 1, and 2 downstream balancing valves open. The ECN has been ;
.- completed for units 1, 2,.and 3 except for the i prefabrication of a spare steel head for the diesel generator engine coolers.
A postmodification test was performed to demonstrate the required flow rates for the engine coolers are ,
provided with,the new valves installed. A selsmic analysis was performed to demonstrate that the addition of the valves did not adversely affect the seismic qualification of the EECW supply to the standby diesel generator engine cooler. The margin of safety was not reduced.
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ATTACHMENT C Modifications Description and Safety Evaluation Summary ECN P0711 The ECN is for documentation only for as-constructing Automatic Fire drawings to depict a substitute for Walter Kidde Protection / Smoke models FT-200, CPD-1212 and CPD-1201 smoke detectors Detection - with Gamewell "Sensmoke" model 29969. The old
- Units 1, 2, and 3 detectors will be replaced on an as-needed basis.
The ECN was totally completed.
The existing detectors are no longer manufactured.
The new detectors are of as good or better quality than the original. They will utilize the same power supplies. The new detectors meet all technical specification requirements, thus, the margin of safety was not reduced.
ECN P0720 Prefabricated the new jet pump instrumentation l Reactor Recirculation replacement piping from the jet pump instrumentation System - Unit 2 seal to the reinstalled existing piping. A small portion of the work was completed. Work is in progress on unit 2. ECN P0720 covers unit 2 only. .
The design requirements for the replacement piping meet or exceed the design requirements for the existing piping. The change of piping material will not adversely affect operation of the recirculation system. The margin of safety as defined in the bases for any technical specification is not reduced. :
ECN P0741 The ECN is for documentation only for the Containment Inerting authorization of replacing sampling and sensing System - Units 1, containment valves FSV-76-49 through FSV-76-68 with 2, & 3 Valcor, model V526-6002 valves on an as-needed basis. The ECll was completed on all 3 units.
This new valve has a soft-seat feature which reduces the required maintenance and offers a greater standardizati.on of parts. The replacement valves are vendor approved and will perform the same safety functions as the original valves. No safety-related function or equipment will be adversely affected by the modification. Technical specification 3.7/4.7 i define the requirements for primary containment '
isolation valves. However, there are no restrictions which would be affected by this modification. Based on this, the margin of safety was not reduced. I 19
l ATTACHMENT C l
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[ Modifications Descr'iption and Safety Evaluation Summary ECN'P0753 ECN P0753 is being implemented on unit 2 to provide HPCI/ ADS Cable adequate HPCI and ADS cable separations. Only a Separation - Unit 2 portion has been completed at this-time. Cable separation tags have been installed, wire lifts and terminations are being implemented and postmodification testing is remaining. All work is I remaining on units 1 and 3.
This modification is necessary to meet the design H basis for the HPCI and ADS systems. This change will not alter the function or operation'of the HPCI va.lves or ADS valves, but rather will ensure that a single fire could not. render both HPCI and ADS inoperable. Based on this, the' margin of safety will not be reduced.
ECN P0761 Installed intruder barriers in the control bay,.
Vital Area Barriers - elevations IC and 3C. The ECN was only partially Unit 1, 2, and 3 implemented. Other barriers are.being installed at various locations for all heating,~ ventilation,~and .;
i air-conditioning (HVAC) duct and piping penetrations in vital area boundaries.
The modification is necessary to implement the new physical security plan. Provided the analysis, !
evaluations, or tests show the design requirements. I are still satisfied, the change will not create any riew failure possibilities because'the original plant' design conditions will be maintained.
ECN P0831 .- Installed nameplates on the unit preferred motor Unit Preferred generator (MG) set 2 cabinet, in the reactor building, System - Unit 2 elevation 593, battery board room 2. Designated the emergency fused disconnects as normally open. The ECN was totally completed as it only covered unit 2.
The mcsdificat. ion will help prevent a fire-related accident from causing an unacceptable malfunction of the unit preferred MG set 2. This ensures compliance with NRC guidelines and continued plant safety. No technical specifications are affected.'
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ATTACHMENT C
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l Modifications Descr'iption and Safety Evaluation Summary
'ECNs P0613,and P0614 Documented changes to MMI-17 to incorporate portions Main Steam System - of ECNs P0613 and P0614 for the rework of poppets and Unit 1 bonnets on the MSIVs for the installation of the 2" stems. The ECNs were not completed.
The modifications are being performed to the MSIVs to reduce their leak rates which are measured during refueling outage. leak rate testing. The operation of the MSIVs was.not changed. The change did not alter the MSIVs function or closing' time. Therefore, the irargin of safety was not reduced. j 1
ECN P0631 Removed / relocated RE-90-132, -133, -134 and
' Radiation Monitoring RM-90-132, -133, and -134 to the south end of the System - Unit 2 service water discharge tunnels. Performed a 4 postmodification test, PMT-137. The ECN was I completed on unit 2. P0631 was previously imple.mented on unit 3 but work is remaining on unit 1.
Relocating the detectors allows them to perform their :
, intended function without the possibility of spurious i alarm due to a high background radiation source. The detectors were seismically mounted and. meet the-requirements of IE Bulletin 79-018. No safety-related equipment'or function was adversely affected. Based on this, the margin of safety was-not reduced.
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ECN P0646 Installed isolation valve on each air supply line to Standby Diesel air starting motors on' diesel generators A, B, C, and Generator System - D for units 1 and 2. The ECN was completed for Units 1 and 2 units 1 and 2. The ECN has not been implemented on unit 3.
The valves are locked-open manual valves and do not degrade the seismic qualifications of the system. The s function of the system was not changed; the margin of J safety was not reduced.
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ATTACHMENT C
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I Modifications Description and Safety Evaluation Summary j ECN P0673 Routed new conduit and cable to replace' existing i RHR System - Unit i embedded conduit 1R1674 which'was' cut while. core i drilling. The ECN was totally completed as it only ']
covered unit 1. 1 The new conduit was run to bya. : the damaged section. The cables located in conduit IR1674 are ;
not needed for proper operation of the RHR systera. l The margin of safety was not reduced. <
ECN P0685 The ECN covered modifications"to the five-ton hook-Refuel floor tie to ensure the proper operation of the limit l Overheadbow switch. The modifications were performed-by l Crane - Common Temporary Alteration Control Form (TACF) 0-83-CD and 4 Maintenance Request No. A166361. A workplan was written for documentation to remove.the T/CF and to as-construct drawings. The ECN was totally completed.
The modifications ensure the hoisting rope remains in the proper position and.that the limit switch trips-at the correct time. This improved the' safety and. -
reliability of the overhead crane five-ton hook and' its associated bow tie during normal operation. ..The function of the five-ton auxiliary hoist was;not changed. The margin'of safety was not reduced.
ECN P0698 This modification replaced an open conductor (green Process Radiation wire)'with a spare conductor (red-black wire) to Monitoring System - 2-RM-90-280A.
Common The ECN was totally completed.
Conductor AA20 (green wire) was open causing 1 RM-90-280A to be inoperative. The replacement '
conductor is in the same cable (2RM267) and. meets all requirements of the green wire. Spare conductors were included in cables for this type of situation. )
No technical specification basis was affected.
ECN P0570 Replaced viscous-type vibration damper with gear-type Emergency Diesel vibration damper on the unit 1 and 2 diesel generator Generator - Units 1 & 2 engines. The ECN was-completed for units ~1 and 2. 1 The work is being implemented on unit 3. 1 The. exchange of diesel generator vibrational dampers I (a gear-type for a viscous-type) did not reduce the ability of the diesel generators to perform their function or increase the probability of their failure. The margin of safety was not reduced.
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q ATTACHMENT C
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Modifications. Description and Safety Evaluation Summary ECN P0417 Performed postmodification testing of the fire pump Low Level Radwaste which was installed for the LLRW facility. Since (LLRW) Fire the LLRN facility is not now in use at Browns Ferry,.
Protection System - the fire pump tested will be used for warehouse and '
Common the training complex protection. The ECN was completed.
The LLRN fire protection system is completely
- independent of the main Browns Ferry fire protection- _
system. The alternate water supply added by this ECN does not interface with any plant systems and is located at a distance where plant structures will not be affected.
ECNs P0489 and P0590 Performed postmodification testing on the standby Standby Lighting Main lighting which was. previously installed in the unit 1 ]
Control Room - Unit 1 control room. Workplan 8657 contained insufficient data to perform the PMT as outlined by PMT-106. The 3 remaining testing will be implemented under a new j workplan. 1 The modification implemented by the ECNs increased the emergency light levels on safety-related panels.
The affected fixtures remained seismically qualified i (seismic mounting). The added load did not exceed j the limits specified by the Electrical Engineering _
Branch. The control bay heating, ventilation, and air-conditioning (HVAC) was~not affected since the modified lighting system heat load was bounded by the.
normal lighting system heat load. No margins of
. safety were reduced. ;
ECN P0399 Pulled cables for ILC Buses A and B. A small portion Instrument and of the work covered by the ECN was completed for :
Controls (I&C) - Unit 2 unit 2. The ECN provides for the addition of regulating transformers and the modification is still in progress on unit 2. Only a small portion of the work has been implemented on units 1 and 3.
The safety objectives and bases for technical specification 3.9/4.9 will not be violated. The changes will actually make the power supply more reliable.
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ATTACHMENT C Modifications Description and Safety Evaluation Summary ECN P0555 Applied protective coating and the heat curing for Pressure Suppression the unit 2 torus. The ECN was completed for unit 2. '
Chamber (Torus) - The ECN was previously completed on units 1 and 3.
Unit 2 The coating applied is of the organic-based thermosetting resin type as specified in the basis for 3.7/4.7. The margin of safety as defined in the basis for this section or any*other section is not reduced.
ECN P0533 The ECN provides for the design, procurement, and Temperature Monitoring installation of an improved temperature monitoring System - Unit 2 system for the torus. A small portion of the work has been completed during the unit 2, cycle 5 outage. The modification is still in progress. The ECN was previously completed on unit 1 except another postmodification test must be performed. The major portion of the ECN has been completed on unit 3. .
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, q
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ECN P0158 Documentation change only, lowered the setpoints for Reactor Feedwater/ PS-3-74B, PS-68-95, and PS-68-96 from 500 psig to Reactor Water Recir- 450 psig. The ECN was completed for all 3 units.
culation Systems -
Unit 1, 2, and 3 The change was required to correct a discrepancy between the drawings and the technical specification. The technical specifications were changed when the LPCI modification was performed. No l physical work was involved.
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ATTACHMENT C Modifications Description and Safety Evaluation Summary ECN P0112 Installed internal wiring in panel 9-14 necessary to Neutron Monitoring - ' install thermal power monitoring. Performed Unit 1 postmodification test and calibration of the average power range monitor including the thermal power monitor. Only a portion of the work covered by the ECN was completed. Hiring of trips to the first-out recorder still remains to be done. The same work is remaining on all 3 units. It is in progress on T unit 2.
All equipment added met Class 1E qualification.
- Seismic qualification was not affected. The modification only changes the average power range monitor (APRM) flow reference scram system and not the fixed setpoint portion of the APRMs. The probability of any previously analyzed event is not increased.
ECN P0157 Capped and removed the inlet and outlet emergency Core Spray Cooling equipment cooling water (EECW) pipes to the core System - Unit 3 spray coolers. Valves 543, 547, 586, and 590 were also removed and replaced with caps. Work was previously implemented which removed the coolers for .
bearing lube oil in the core spray pump motors. The ECN was completed for unit 3. The ECN has been implemented on units 1 and 2, but the documentation is in the closeout cycle.
Removal of the coolers and cooler inlet and outlet- '
lines did not change previous evaluations or introduce a condition not previously evaluated. The EECW flow to the coolers had been valved off previously per modification performed by ECN L1994; i therefore, the coolers and lines were no longer l required. No technical specification basis was l affected by this modification.
ECN P0371 Installed emergency feed to 480V reactor MOV boards I 480V Reactor Motor and performed postmodification test. All field work Operated Valve (MOV) covered by th.e ECN has been completed on all three Board- Unit 3 units.
The modifications permit the use of manually l controlled temporary maintenance feeders from the "C" {
Boards to the "D" and "E" Boards for use during cold shutdown conditions only. Technical specification section 3.9.C discusses equipment availability when the unit is in cold shutdown. The basis to this section was reviewed and this modification supported the basis. Thus the margin of safety was not reduced.
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ATTACHMENT C Modifications Descr'iption and Safety Evaluation Summary ECN-P0381 Replaced GEMAC transmitters-with Rosemount. The Various Systems - transmitters are being replaced on an as-needed Unit 2 basis. Transmitters LT-63-1 and PT-35-17 were replaced on unit 2. The ECN has not been completed "
on any unit, but transmitters have been replaced on all three units.
The Rosemount transmitters are compatible with '
existing equipment, perform the same. functions as the GEMAC transmitters, and meet the requirements specified for the functions performed. No~ technical-specification safety margin was adversely affected. 3 ECN P0354 Installed embedded conduits and ground cables for the Radiation Monitoring off-gas stack monitor building located northeast of i System - Common. the off-gas stack. A small portion of the j modification covered by the ECN was completed. The j work is in progress now during the unit 2, cycle 5 outage.
This ECN is' installing a permanent radiation monitoring system at the plant stack to meet NRC extended range effluent monitoring requirements'of NUREG 0737. The monitoring system being. installed ;
only provides an indication on the level of radiation !
released out of the stack during and after a '
potential accident-.- This system does not. provide any safety-related function, nor does itl interface with any safety-related equipment.
ECN P0353 < A workplan was written for documentation only to Control Rod Drive as-construct drawings which were omitted when the (CRD) Hydraulic physical work was implemented. .The ECN provided a System - Units 1, crosstie between the vent lines on the east and west 2, and 3 scram discharge headers and installed an isolation valve in the crosstie line. The ECN has been totally completed for,all three units.
The modification adds redundancy to an existing function. No new function is performed. The piping, valve, and fittings have the same seismic '
qualifications as those portions of the existing vent system. No technical specification is affected since this system performs no safety-related function and-is only used for diagnostic purposes. Therefore, the margin of safety is unaffected.
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l ATTACHMENT C Modifications Description and Safety Evaluation Summary i
ECN P0392 This ECN covers the "long-term" modifications to the j CRD System - Unit 2 CR0 System and to the clean radioactive waste.(CRW) J drain system. The modification ~was' partially 1 implemented during unit 2 outage, cycle 4 and is in- 'l progress during unit 2, cycle 5 outage. The CR0 ,"
scram discharge volume header water level monitor and the associated conduits and cables removal has been )
completed and panel 9-91 and associated instruments J (eight-level switches) have been' installed. Support i modifications and postmodification testing is
remaining to be completed on units 1 and 3.
Following implementation of the change, the systems )
will perform all of the same functions that they 1 presently perform. No possibility for accidents or j malfunctions different from those previously {
evaluated is foreseen. The margin of safety will.not -)
be reduced. j ECN L1970 Replaced the four-inch carbon steel piping and check EECW System - valves with stainless steel. The piping changed was .
Unit 1 Diesel Generator from the.first check valve to the upstream flange on :
Building suction throttling valve. A'small portion of:the work was completed for unit 1. The ECN has been partially implemented on units 1, 2, and 3.
The material replacement did not change the operation or affect the ability of the EECW system to respond I to an accident situation. The margin of safety was not reduced. The integrity of the EECW piping system was not reduced by the new material.
l ECN L1647 Installed flow element in the discharge line of the l RWCU - Unit 2 nonregenerative heat exchanger on the RWCU side. l The ECN has been completed on units 1 and 2. They ,
were completed before unit 2, cycle 5 outage, but the i documentation was just completed. The ONP has requested the,ECN be cancelled for unit 3.
The technical specifications require that the RWCU system must be able to maintain the reactor coolant chemistry within certain limitations for continued 1 plant operation and during shutdown. This l modification had no effect on system isolation valve j or isolation function performance. Therefore, the i modification did not jeopardize the margin of safety I that is defined in the technical specifications.
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ATTACHMENT C Modifications Descr.iption and Safety Evaluation Summary ECN P0192 Relocated the RHRSW vent valves by welding the vents Residual Heat directly on the RHR heat exchanger heads. The ECN Removal Service was completed for unit 2. The ECN was implemented Water (RHRSW) - Unit 2 previously on unit 3. Work is remaining for IC heat exchanger on unit 1.
The modification did not affect the vent function.
RHRSH performance was not affected. The modification eliminated the cutting of the vent lines when the heat exchanger heads are removed and rewelding when the heads are replaced. No technical specification was affected, therefore, the margin of safety was not reduced. ,
ECN P0324 Provided conduits, grounds, junction boxes, tees and Radiation Monitoring cables for instrumentation to the containment System - Unit 1 monitoring system in unit 1. The ECN was partially implemented on unit 1, is in progress on unit 2, and the modification has not been implemented on unit 3.
The high range containment radiation monitors to be -
added by the ECN will be used to provide an indication of the radiation levels in the containment atmosphere after an accident. This information will i help the operator understand postaccident t.onditions inside containment. The monitors will not initiate any automatic function required to mitigate an -
accident. Therefore, addition of high range I containment radiation monitors will not adversely !
affect plant safety.
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ECN Ll742 The ECN covered the installation of accelerometers, Local Power Range wiring, conduit, and hardware necessary to establish Monitor Surveillance a surveillance system for monitoring incore i System - Unit 1 instrument tube-channel impacting. The ECN covered units 1 and 2. The modification was implemented l previously on units 1 and 2, but documentation was i only recently. closed out on unit 1. )
This modification was an NRC requirement. NRC approval was received before implementation..
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ATTACHMENT C Modifications Description and Safety Evaluation Summary ECN P0361 This'ECN covers the modifications for the long-term Primary Containment torus integrity program. The major portion of the System - Unit 2 fieldwork covered by the ECN hd been implemented in previous refueling outages. Various modifications _to hangers and supports were implemented during 1986 and will continue being modified during the unit 2,-cycle "
5 outage. - Units 1 and 3 Torus modifications were completed before the last startup of each unit, however, recent inspections by NRC resulted in identification of small discrepancies which will'be corrected.
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 1 or the consequences of an accident or malfunction of equipment important to safety were not increased.
ECN L1800 Installed Honeywell actuators on the reactor zone i
Reactor Building supply fan inlet dampers. Hookup and checkout Ventilation System - of the actuators were never implemented. Revision.1 Volt 1 of the USQD was issued by DNE to cover the partial -1 implementation of the ECN for the work performed on unit 1. The modification has not been implemented on units 2 and 3.
The partial implementation of ECN L1800 will not affect secondary containment integrity or SGTS operation; no safety concerns exist. The margin of safety was not reduced.
ECN P0126 ECN P0126 authorizes the installation of the proposed Various Systems - analog transmitter / trip unit system for engineered Unit 2 safeguard sensor trip inputs. The system replaces pressure, level, and temperature switches with analog transmitter / trip unit combinations, which provide continual mon.itoring of critical parameters in addition to performing basic logic trip. operations.
A small portion of the work has been completed on unit 2. The work'is in progress during-unit 2, cycle 5 outage. The ECN has been partially implemented on units 1 and 3.
The instrument functions will not be changed by the modification. The modification package was reviewed and approved by NRC before implementation. The likelihood of failure will be reduced by' implementing the ECN.
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ATTACHMENT C Modifications Description and Safety Evaluation Summary )
i ECN P0697 Fabricated and installed intruder barrier in the j Control Bay Air communication room supply ductwork which penetrates l Conditioning System - into the control bay 2C spreading room. A small l Common portion of the work covered by the ECN was )
completed. The work is in progress during unit 2, " l cycle 5 outage. l l
There are no technical specification requirements l associated with the changes being made to the control {
bay HVAC system by this ECN. The margin of safety 2 will not be reduced.
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Attachment D CHALLENGES TO OR FAILURES OF MAIN STEAM RELIEF VALVE January 1, 1986 - December 31, 1986 Unit 1 I None l
Unit 2 ,
None
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i Unit 3 J None 1 All three units were in cold shutdown during the entire reporting period.
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Attachment E OCCUPATIONAL EXPOSURE DATA 1"
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TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374o1 SN 157B Lookout Place JUL 101987 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk l
Washington, D.C. 20555 Gentlemen:
In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 ANNUAL OPERATING REPORT FOR JANUARY 1, 1986 - DECEMBER 31, 1986 BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 1, 2, AND 3 - DOCKET NOS. 50-259, 50-260, 50-296 -
OPERATING LICENSES DPR-33, DPR-52, AND DPR-68 Pursuant to Browns Ferry Nuclear Plant Technical Specification 6.7.1.b and 10 CFR 50.59 and 10 CFR 20.407, enclosed is the annual operating report for Browns Ferry units 1, 2, and 3 for the period January 1 - December 31, 1986.
The delay of this submittal was discussed with G. E. Gears of your staff who concurred with our request during the week of April 20, 1987, at TVA's BFN site. _.
Please refer any questions to M. J. May, Manager of Site Licensing, Browns Ferry Nuclear Plant, at (205) 729-3566.
Very truly yours, TENNESSEE VALLEY AUTHORITY
./M
. Gridley, Director Nuclear Cafety and Licensing Enclosure ec: See page 1 -
An Equal Opportunity Employer l l
c i
~ 2-i U.S. Nuclear Regulatory Commission JU[ j () g337 .
s
- cc (Enclosure)
Mr. G. G. Zech, Assistant Director Regional Inspections
}E :l Division of TVA Projects Office of Special Projects U.S. Nuclear Regulatory Commission Region II 101 Marietta St., NW, Suite 2900 Atlanta, Georgia 30323 k: ,
Browns Ferry Resident Inspector Browns Ferry Nuclear Plant
: P.O. Box 311 Athens, Alabama 35611 Mr. J. A. Zwolinski, Assistant Director for Projects Division of TVA Projects '
Office of Special Projects
} U.S. Nuclear Regulatory Commission +
4350 East West Highway EWW 322 Bethesda, Maryland 20814 e
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