ML20003B362

Safety Evaluation Supporting Amend 37 to License DPR-68

ML20003B362
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Site:	Browns Ferry
Issue date:	01/12/1981
From:	Office of Nuclear Reactor Regulation
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 37 TO FACILITY OPERATING LICENSE N0. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT NO. 3 DOCKET N0. 50-296 1.0 Introduction By letter dated August 27, 1980 (TVA BFNP TS 148), which was supplemented by letters dated September 23, 1980 and October 14, 1980, and by letters dated Setpember 5, 1980 and October 17, 1980, the Tennessee Valley Authority (the licensee or TVA) requestad changes to the Technical Specifications (Appendix A) appended to Facility Operating License No. DPR-68 for the Browns Ferry Nuclear Plant, Unit No. 3.

The pro-posed amendments and revised Technical Specifications were to:

(1) incorporate the limiting conditions for operation associated with the fourth fuel cycle, (2) reflect new primary containment atmospheric hydrogen monitoring instrumentation being installed during the current refueling outage, and (3) reflect the addition of 480 volt motor generator sets durin the refueling outage to supply reactor motor operated valve (RMOV boards 3D and 3E.

2.0 Discussion Browns Ferry Unit No. 3 (BF-3) shutdown for its third refueling on November 23, 1980.

The initial core loading for BF-3 consisted of 764 of the single water rod 8X8 fuel assemblies, each containing 63 fuel rods. During the first refueling in September 1978, 208 of the fuel assemblies were replaced with 8X8R fuel assemblies containing 62 fuel rods in each.

During the second refueling outage starting in August 1979, an additional 144 of the initial fuel bundles were replaced with P8X8 fuel assemblies, each containing 62 fuel rods.

During the current refueling outage, an additional 124 of the original 8X8 fuel assemblies are being replaced with a like number of new P8X8R fuel assemblies. The prepressurized fuel assemblies (P8X8R) are essentially identical from a core physics standpoint to the two water rod fuel assemblies (8X8R) except that they are prepressurized with about three rather than one atmospheres of helium to minimize fuel clad interaction.

Our evaluation of the P8X8R fuel is discussed in the safety evaluation attached to our letter of April 16, 1979 to General Electric approving the use of this fuel in BWR reload licensing applications.

The larger inventory of helium gas improves the gap conductance between fuel pellets and cladding resulting in reductions The in fuel temperatures, thermal expansion and fission gas release.

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pressurized rods operate at effectively lower linear heat generation l

l rates and are therefore expected to yield perfonnance benefits in terms of fuel reliability. The increased prepressurization also results in improved margin to MAPLHGR limits by reducing stored energy, although TVA is not proposing to take any credit for these beneficial effects in the subject reload application (i.e., they are i

not proposing any changes in the existing MAPLHGR vs. Exposure limits in the existing Technical Specifications).

In pport of this reload

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application for BF-3, TVA submitted by letter f dated August 27, l

1980, a supplemental reload licensing document prepared by the General Electric Company (G.E(j)for TVA and proposed changes to the BF-3 Technical Specifications This initial submittal was supple-

4) dated September y 1980 relating to the GEXL mented by a letter L critical power correlation and a letterd dated October 14, 1980 submitted additional proposed changes to the Technical Specifications (6) to remove references to the power spiking-penalty in the Linear Heat Generation Rate (LHGR) calculations.

j One of the modifications which TVA is accomplishing during this refueling outage and which is discussed herein is a replacement of the primary contair:1ent hydrogen monitoring system. A descriptica of the new

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hydroge. monitoring system and proposed changes to the Technical) Spec-5, 1980 ifications were submitted by TVA's_ letter of_ September

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new hydrogen monitoring system being installed in BF-3 is the same as the new hydrogen monitoring system which was Installed in Browns Ferry Unit No. 2 (BF-2) during the September to November 1980 refuel-4 ing outage.

Use of the new system for BF-2 was approved as part of

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the reload amendment - Amendment No. 58 to Facility Operating License No. DPR-52 dated November 12, 1980. The present and new hydrogen mon-i itoring systems were described in detail in Amendment No. 58 for BF-2 and this descriptive material is incorporated herein by reference.

Another modification which TVA is planning to accomplish during the present refueling outage of BF-3 is to add four 480-volt motor generator (MG) sets to supply reactor motor operated valve boards 3D and 3E.

By letter dated May ll,1979, we issued Amendments Nos. 51, 45 and 23 to Facility Licenses Nos. DPR-33, DPR-52 ana DPR-68 for the Browns Ferry Nuclear Plant, Units Nos. 1, 2 and 3.

The Amendments added a condition r

to the license for each facility authorizing TVA to perform certain modifications (as described in TVA's submittals and the Safety Evalua-tion related to these Amendments) to change the power supply for certain LPCI valves for Units Nos.1, 2 and 3 and to eliminate the loop selection logic for Unit No. 3.

Our letter of May 11, 1979 noted that TVA had i

committed to complete the modifications for BF-3 by the end of the second refueling outage and to submit proposed Technical Specification i

changes with the reload amendment request for each unit.

For 8F-3, the modifications consisted of the following:

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l l Elimination of the Low Pressure Coolant Injection (LPCI) system's a.

recirculation loop selection logic, revision of the logic and closure of the Residual Heat Removal (RHR) cross-tie valve and a recirculation equalizer valve; and b.

Changing the power supply te the recirculation pump discharge valves, LPCI injection valves, RHR pump minimum flow bypass j

j valves, and RHR test isolation valves. The change also modifies independent valve a.c. power supplies, and modified d.c. power supplies to 4kV shutdown board control power to provide adequate independence such that a station battery failure does not jeopardize core cooling capabilities.

During the second refueling outage of BF-3 (August 24 to December 8, 1979) all of the electrical changes related to the LPCI modification were completed except for the addition of the MG sets.

Due to a strike at the manufacturer's facility, the MG sets were not delivered in time to be installed during the last refueling outage.

There are two 480-V ac Reactor Motor-0perated Valve (RMOV) Boards that contain motor-generator (M-G) sets in their feeder lines. These 480-V ac RMOV boards have an automatic transfer from their normal to alter-natepowersource(480-Vacshutdownboards). The M-G sets act as electrical isolators to prevent a fault from propagating between elec-trical divisions due to an automatic transfer.

The 480-V ac RM0V boards involved provide motive power to those valves necessary for automatic operation of RHR injection (Recirculation pump discharge valves, LPCI injection valves, RHR pump minimum flow bypass valves and RHR test isolation valves) and will interface with the division-alized 480 V shutdown boards through the M-G sets.

Each RM0V board will have two sets, and although only one M-G set will normally supply l

power to the RMOV board, both M-G sets will run at all times to assure readiness of the alternate M-G set to accept load if required.

By letter (0) dated October 17, 1980, TVA submitted proposed changes to the Technical Specifications to reflect the addition of the 480 Volt MH sets to specify surveillance and operability requirements for this equipment.

3.0 Reload This refueling (Reload 3) is the second for BF-3 to incorporate GE's P8X8R fuel design on a batch basis.

The description of the nuclear and mechanical design of the Reload 3 P8X8R fuel and the exposed unpressurized 8X8 and 8X8R fuels, used in the initial and first reload cores, is contained in GE's generic licensing topical report for BWR reloadsl9). Reference 9 also contains a complete set of references to topical reports which describe GE's analytical methods for the nuclear, thermal-hydraulic, transient and accident calculations per-formed for this reload together with infonnation on the applicability

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of these methods to cores containing a mixture of different fuel designs.

Portions of the plant-specific data, such as operating conditions and design parameters, which are used in transient and accident calculations, have also been included in the topical report.

The use and safety implications of prepressurized fuel are presented in Reference 9 and have been found acceptable per Reference 10 (Enclosed in Appendix C of Reference 9).

Values for plant-specific data such as steady state operating pressure, core flow, safety and safety-relief valve setpoints, rated thermal power, rated steam flow, and other design parameters are provided in Reference 9.

Additional plant and cycle dependent information is provided in the reload application (Reference 2) which closely follows the outline of Appendix A of Reference 9.

Reference 10 includes a description of the staff's review, approval, and conditions of approval for the plant-specific data.

The above-mentioned plant-specific data have been used in the transient and accident analysis provided with the reload application in compliance with Reference 10.

Our safety evaluation of the GE generic reload licensing topical report has also concluded that the nuclear, and mechanical design of the 8X8R and P8X8R fuels, and GE's analytical methods for nuclear and thermal-hydraulic calculations as applied to mixed cores containing 8X8, 8X8R and P8X8R fuels, are acceptable.

Because of our review of a large number of generic considerations related to use of 8X8R and P8X8R fuels in mixed loadings, and on the basis of the evaluations which have been presented in Reference 9, only a limited number of additional areas of review have been included in this safety evaluation report.

For evaluations of areas not specifically addressed in this safety evalution report, the reader is referred to Reference 9.

3.1.1 Nuclear Characteristics For Cycle 4,124 fresh pressurized type P80RB265L fuel bundles will be loaded into the core.

The remainder of the fuel bundles in the core will be a combination 8X8, 8X8R and P8X8R fuel bundles exposed during the previous three cycles.

The fresh fuel will be loaded and the previously peripheral fuel will be shuffled inward so as to constitute an octant-symmetric core pattern, which is acceptable.

Based e, the data provided in Sections 4 and 5 of Reference 2, both the control rod system and the standby liquid control system will have an acceptable shutdown capability during Cycle 4.

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3.1.2 Thermal-Hydraulics i

3.1.2.1 Fuel Cladding Integrity Safety Limit MCPR As stated in Reference 9, for BWR cores which reload with GE's P3X8R fuel, the allowable minimum critical power ratio (MCPR) resulting frum either core-wide or localized abnormal operational transients is equal to 1.07.

When meeting this MCPR safety limit during a transient, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The 1.07 safety limit minimum critical power ratio (SLMCPR) to be used for Cycle 4 is unchanged from the SLMCPR previously approved for Cycles 2 and 3.

The basis for this safety limit is addressed in Re erence 9, while our generic approvals are given in Reference 10.

3.1.2.2 Operating Limit MCPR i

Various transient events can reduce the MCPR from its normal operating level.

To assure that the fuel cladding integrity safety limit MCPR will not be violated during any abnormal operational transient, the most limiting transients have been reanalyzed for this reload by the licensee, in order to determine which event results in the largest reduction in the minimum critical power ratio. These events have j

been analyzed for both the exposed 8X8, 8X8R, and P8X8R fuel and the fresh P8X8R fuel. Addition of the largest reductions in critical power ratio to the safety limit MCPR establishes the operating limits for each fuel type.

The transient events analyzed were load rejection without bypass, feedwater controller failure, loss of 100*F feedwater heating and control rod withdrawal error.

3.1. 2. 2.1 Abnormal Operational Transient Analysis Methods The generic methods used for these calculations, including cycle-independent initial conditions and transient input parameters, are described in Reference 9.

Our acceptance of the cycle-independent values appears in Reference 10. Additionally, our evaluation of the transient analysis methods, together with a description and summary of the outstanding issues associated with these methods, ap1 ears in Reference 10.

Supplementary cycle-dependent initial conditions and transient input parameters used in the transient analyses appear in the tables in Sections 6 and 7 of Reference 2.

Our evaluationU0) has also addressed the methods used to develop these supplementary input values.

I 3.1. 2. 2. 2 Transient Analysis Results The transients evaluated were the limiting pressure and power increase transients, generator load rejection without bypass and the feedwater controller failure (loss of 100 F feedwater heating), and the control rod withdrawal error.

Initial conditions and transient input parameters as specified in Sections 6 and 7 of Reference 2 were assumed.

As noted above, the calculated system responses and reductions in CPR during each of the operational transients have been provided in Sections 9 and 10 of the GE Supplemental Reload Licensing Submittal (Reference 2). On this topic, it is acceptable if fuel specific operating limits are established for prepressurized fuel (Appendix C, Reference 9). On this basis, the transient analysis results are acceptable for use in r

the evaluation of the operating limit MCPR.

Based on this, the proposed Technical Specification modifications to operating limit MCPR are acceptable.

The following table gives the limiting CPR reduction as calculated by GE, the event for which limiting CPR reduction occurs, and the required operating limit MCPR for each fuel type:

1 Fuel Type Most Severe CPR Redution Operating Limit MCPR i

8x8 0.17 Load Rejection w/o Bypass 1.24 8X8R 0.18 Load Rejection w/o Bypass 1.25 P8X8R 0.18 Load Rejection w/o Bypass 1.25 Thus, when the reactor is operated in accordance with the above operat-ing limit MCPRs the 1.07 SLMCPR will not be violated in the event of the most severe abnormal operational transient. This is acceptable to the staff per the finding of the previous section. On this basis, operating limit MCPR Technical Specifications have been established.

3.1.2.3 Fuel Cladding Integrity Safety Limit LHGR The control rod withdrawal error and fuel loading error events were reanalyzed by the licensee to also determine the maximum transient linear heat generation rates (LHGRs). The results for BF-3 Cycle 4 are giyag in Appendix B of the Supplemental Reload Licensing Sub-mittalt21 The calculated Fuel Loading Error LHGR is 16.9 kW/ft for a rotated bundle and 18.1 kW/ft for a misplaced bundle. These results 3

indicate that the fuel type-dependent and exposure-dependent safety limit LHGRs, shown in Table 2-3 of Reference 9, will not be violated should these events occur. Thus, fuel failure due to excessive cladding strain will be precluded. We find these results, which a<iequately account for the effects o

'uel densification power spiking, to be acceptable.

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- 3.1.3 Accident Analysec 3.1.3.1 ECCS Appendix K Analysis In our safety evaluation of Reference' 9, we concluded that "the continued application of the present GE ECCS-LOCA (" Appendix K") models to the 8X8 retrofit reload fuel is generally acceptable and in our Reference 10 evaluation we extended that conclusion to prepressurized fuel. On these bases, the MAPLHGR limits,which remain unchanged from the previous cycle, are acceptable.

1 3.1.3.2 Control Rod Drop Accident For Cycle 4, the key plant-specific and cycle-specific nuclear character-l istics for the worst case control rod drop accident (CRDA) occurring i

during both cold and hot startup conditions are conservatively bounded by the values used in bounding CRDA analysis given in Reference 9.

The Reload Licensing Submittal (gre presented in Section 16 of the Supp results of G.E.'s analysis 4

The bounding analysis, which includes i

the adverse effects of fuel densification power spiking, shows that the peak enthalpy will not exceed the 280 cal /gm design limit. Therefore, for Cycle 4 of BF-3, the peak fuel enthalpy associated with a CRDA from the hot and cold startup condition will also be within the 280 cal /gm design limit.

l Thus, we conclude that the peak enthalpy associated with a control rod drop accident occurring from any in-sequence control rod movement will I

be below the 280 cal /gm design limit.

3.1.3.3 Fuel Loading Error The GE method for analysis of misoriented and misloaded-bundles has been reviewed and approved by the staff and is part of the Reference 2 methodology.

Potential fuel loading errors involving misoriented bundles and bundles loaded into incorrect positions have been analyzed by this methodology and the r9sylts are reported in Section 15 of the supplemental reload submittalt21. The analyses determined that a rotated

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P8X8R fresh fuel assembly was the most limiting loading error event; the ACPR for this event was 0.17.

This is the same, or almost the same, as the ACPR for the limiting transient which determines the safety limit MCPR. As shown in Section 3.1.2.2.2, above, the ACPR for the limiting transient is also 0.17 for 8X8 fuel and 0.18 for 8X8R and P8X8R fuel assemblies.

During the recent refueling (September - November 1980) of Browns Ferry-Unit 2, it was discovered that two 7X7 fuel assemblies had gone thru cycle 3 misoriented 90 and that there is presently a 7X7 fuel element (core location 11-06) in Browns Ferry Unit 1 that is misoriented 90".

During recent refuelings, there has also been a significant number of-misoriented-fuel assemblies detected at the final core verification stage. By letter dated November 6, 1980, TVA committed to make changes

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in their fuel handling procedures.

These new procedures are being t

followed during the fuel shuffling operations for BF-3.

For BF-an independent QA inspector on the refueling bridge is being used to a

i check fuel element orientation and location.

i 3.1.3.4 Overpressure Analysis f

For Cycle 4, the licensee has reanalyzed the limiting pressurization i

event to demonstrate that the ASME Boiler and Pressure Vessel Code l

requirements are met for BF-3.

The methods used for this analysis, when modified to account for one failed safety valve, have also been previously approved by the staff. The acceptance criteria for this event is that the calculated peak transient pressure not exceed 100%

The reanalysjs which is presented of design pressure, i.e.,1375 psig.

inSection12ofthesupplementalreloadsubmittal(2),showsthatthe peak pressure at the bottom of the reactor vessel does not exceed 1299 psig for worst case end-of-cycle conditions, even when assuming the r

effects of one failed safety valve. This is a decrease of 1 psig from the previous fuel cycle and is the reason for the changes on pages 30 and 225 of the proposed Technical Specifications. We conclude that j

there is sufficient margin between the peak calculated vessel pressure and the design limit pressure to allow for the failure of at least one l

valve. Thuefore, the limiting overpressure event as analyzed by.the licensee is considered acceptable on the bases outlined in Reference 9.

3.1.4 Thermal Hydraulic Stability l

A thermal-hydraulic stability analysis was perfonned for this reload using the methods described in Reference 9.

The results, which are presented in Section 13 of the Supplemental Reload Licensing Submittal (2) i show that the fuel dependent channel hydrodynamic stability decay ratios i

and reactor core stability decay ratio at the least stable operating state (corresponding to the intersection of the natural circulation power curve and the 105% rod line) are 0.29 (8X8R/P8X8R), 0.36 (8X8) i and 0.85 respectively. These predicted decay ratios are all well i

below the 1.0 Ultimate Performance Limit decay ratio proposed by GE.

Prior to Cycle 3 operation, the staff as 'an interim measure, added a requirement to the BF-3 Technical Specifications which restricted planned plant operation in the natural circulation mode.

Continuation of this restriction will also provide a significant increase in the reactor core stability operating margins during Cycle 4.

On the basis of the foregoing, the staff considers the thermal-hydraulic stability of BF-3 during Cycle 4 to be acceptable.

3.1.5 Startup Test Program The licensee has not changed his startup test program from that approved for the previous cycle. This program therefore remains acceptable.

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. 3.2 Addition of MG Sets As noted in the Discussion above, the modifications to the LPCI systems at Browns Ferry Units 1, 2 and 3 were approved by Amendment Nos. 51, 45 and 23 issued May 11, 1979.

The modification for BF-3 was completed during the previous refueling outage except for the addition of the MG sets in the feeder lines to the RMOV Boards.

The reload amendment for the previous fuel cycle (cycle three) - Amendment No. 28 to Facility License No. DPR-58 issued November 30, 1979 - included changes to the Technical Specifications resultin,q_from this modificat_ ion as.well as.

our evaluation of TVA's reanalysis of the Loss of Coolant Accident (LOCA) for BF-3 with the LPCI modifications in place.

The only changes to the Technical Specifications related to the modifications in this amendment is the addition of operability and surveillance requirements for the new MG sets and the bases therefore. We have reviewed the proposed additions to the Technical Specifications and find them acceptable.

3.3 Hydrogen Monitoring System The proposed changes to the BF-3 Technical Specifications involve the number of gas analyzer systems in the drywell (changed from 2 to 1 system) and th time interval between performing a calibration test on the gas analyzer systems (from once a month to quarterly).

Hydrogen concentrations in the containments of the Browns Ferry reactors are currently measured by hydrogen electrode sensors installed in the drywell and in the suppression pool torus of each reactor.

TVA proposes to replace these sensors with thermal conductivity gas analyarrs located outside the primary containment for easy access.

Gas sample lines will lead from the upper part of each drywell and torus through existing penetrations to a sampling cabinet outside the primary containment. The sample will pass through approximately 100 feet of 1/2-inch stainless steel pipe, a water trap and chiller to remove entrained moisture, a bellows pump and through either of two independent thermal conductivity l

sensors, exhausting back into the drywell.

l The hydrogen monitoring system will be cperating continuously during j

reactor operation and the sample will reach the sensor in less than 2 minutes at the pumping speed of the bellows pump.

The sensor will begin I

to respond in 3 seconds and will reach two-thirds of its steady reading l

in 21 seconds. The sensitivity of reading is 0.4 volume percent hydrogen (2% of 20% full scale).

We have fnund the proposed hydrogen monitoring system to be acceptable after evaluating it against the acceptance criteria and requirements listed in the following:

1

. (1) Standard Review Plan 6.2.4, " Containment Isolation System."

(2) Standard Review Plan 6.2.5, " Combustible Gas Control in Containment."

(3) NUREG-0578, " Lessons Learned Task Force Status Report and Short-Term Recommendations," Sections 2.1.5, 2.1.8.a and Appendix A (2.1.8a),

" Improved Post-Accident Sampling Capability."

(4) NUREG-0737, " Clarification of TMI Action Plan Requirements."

We find the proposed thennal conductivity method has adequate sensitivity and is at least as reliable as the currently used hydrogen electrode method.

j The thermal conductivity method, including the sampling system, will be in continuous operation before and after the initial phase of the accident with readings and controls in the control room. The system is designed to be operable under accident conditions and at negative pressures down to 2 psi below ambient atmospheric pressure and will measure H2 concen-tration in the range of 0.1% V/a to 20% V/o. The response time of about 2 minutes is sufficiently rapid to provide timely warning of hazardous hydrogen concentrations in containment after an accident. The require-ment in NUREG-0578, Appendix A (Section 2.1.8a) is for sampling and analysis within an hour after an accident.

Exposure to the operator during sampling meets the "As Low As Reasonably Achievable" levels because there is no need for an operator to be near the analyzer during its operation.

If maintenance after an accident is required, radioactive gases can be purged out of the analyzer remotely from the control room.

The location of the sensors outside primary containment also makes them more accessible for maintenance and inspection during normal reactor operation.

No additional penetrations of u m primary containment will be required since the new sampling lines will pass through unused existing spare penetrations. The power circuits to operate the hydrogen monitoring system will meet the safety requirements of engineered safety features.

The redundancy requirement will be met by providing two independent thermal conductivity sensors to which gas samples from either the dry-well or the torus atmosphere may be directed. The licensee has verbally i

indicated that each of the sampling lines contain two automatic isola-tion valves that will automatically isolate upon receipt of a containment i

isolation signal.

The operator will then manually open these lines l

within 30 minutes of initiation of safety injection to sample the l

containment atmosphere.

Based on our evaluation, we conclude that the proposed changes in the hydrogen monitoring systems are acceptable and meet the requirements of General Design Criterion 41 (Containment Atmosphere Cleanup) of t

Appendix A to 10 CFR Part 50.

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. The system being replaced had two gas analyzer systems in the drywell and one gas analyzer system in the wetwell, and was calibrated monthly using standard gas samples. The new system will have one gas analyzer system for the drywell and one analyzer for the wetwell with monthly channel functional tests and quarterly channel calibration tests using standard gas samples.

Redundancy is provided by the drywell purge system which limits the hydrogen concentration difference between the drywell and wetwell to 0.2V/o. Therefore, if either gas analyzer fails the operator can still be able to measure, to an acceptable degree of accuracy, the hydrogen concentration in both the wetwell and drywell using only one gas analyzer.

The interval between calibration of the gas analyzer system was lengthened from one month to 3 months because gas analyzer systems that use the thennal crJ.ouctivity method are inherently more stable and less susceptibla to drift.

4.0 Environmentti Considerations We have detern.ined that the amendment does not authorize a change in effluent types or "otal amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination we have further concluded that the amendment involves an action aich is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

5.0 Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety nargin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.

Dated:

References i

l 1.

Tennessee Valley Authority letter (L. M. Mills) to USNRC (H. R. Denton) l dated August 27, 1980 (TVA BFNP TS 148).

2.

" Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Power l

Station Unit 3, Reload No.

3," Y1003J01 A03 dated August 1980.

l 3.

" Proposed Technical Specification Changes, Browns Ferry Nuclear Plant Unit 3" submitted as enclosure 1 to TVA letter (L. M. Mills) to USNRC (H. R. Denton) dated August 27, 1980.

l I

4.

Tennessee Valley Authority letter (L. M. Mills) to USNRC (T. A.

]

Ippolito) dated September 23, 1980.

5.

Tennessee Valley Authority letter (L. M. Mills) to USNRC (H. R. Denton) dated October 14, 1980 (TVA BFNP TS 148).

6.

" Proposed Changes to Technical Specifications, Browns Ferry Nuclear l

Plant Unit 3" submitted as enclosure 1 to TVA letter (L. M. Mills) to USNRC (H. R. Denton) dated October 14, 1980.

Tennessee Valley Authority letter (J. L. Cross) to USNRC (H. R. Denton) l 7.

dated September 5, 1980 (TVA BFNP TS 148).

i 8.

Tennessee Valley Authority letter (L. M. Mills) to USNRC (H. R. Denton) dated October 17, 1980 (TVA BFNP TS 148).

9.

" General Electric Boiling Water Reactor Generic Reload Application,"

NEDE-240ll-P-A, August 1979.

10.

Letter, T. A. Ippolito (USNRC) to R. Gridley (GE), April 16, 1979 and enclosed SER.

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