ML20236X425
| ML20236X425 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/04/1998 |
| From: | George Thomas NRC (Affiliation Not Assigned) |
| To: | Lakshminaras R NRC (Affiliation Not Assigned) |
| References | |
| TAC-M99711, TAC-M99712, NUDOCS 9808100007 | |
| Download: ML20236X425 (11) | |
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August 4, 1998 MEMORANDUM TO: Raghaven Lakshminaras, Project Manager Project Directorate ll-3 Division of Reactor Projects I and ll FROM: George Thomas, Reactor Engineer / original signed by/
Reactor Systems Branch Division of Systems Safety and Analysis
SUBJECT:
BROWNS FERRY UNITS 2 AND 3 PROPOSED LICENSE AMENDMENT POWER UPRATE REVIEW (TAC NOS. M99711 AND M99712)
Attachment 1 is the Reactor Systems Branch input to the Safety Evaluation (SE) being prepared by your Project Directorate for the subject power uprate license amendment. It is my understanding that you will use this and other technical branch inputs for developing the overall staff safety evaluation for this license amendment. If you have any questions about the attached SE, please call me at 415-1814.
Attachment:
As stated cc: G. Holahan T. Collins F Hebdon CONTACT:
G. Thomas, SRXB/DSSA 415-1814 DISTRIBUTION:
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i ATTACHMENT QEEICE OF NUCLELREFJCTOR REGULATION TECHNICAL SPEGjFICATION CHANGES FOR P._QWE8 UPRATE QP_EB_A_TjQB A TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT DOCKET NO 50-260. 50-296 1 INTRODUCTION Tennessee Valley Authority (WA), the licensee for Browns Ferry Nuc! ear Plant (BFN), Units 2 and 3, submitted a request by letter on October 1,1997 to increase the licensed thermal power >
level from 3293 MWt to 3458 MWt. TVA also responded by letter dated May 20,1998 to provide additionalinformation requested by the staff. This power increase request represents approximately a 5.0% increase in thermal power with at least a 5% increase in rated steam flow. The planned approach to achieve the higher power level consists of: (1) an increase in the core thermal power (with a more uniform (flatter) power distribution to create an increased steam flow, (2) a corresponding increase in feedwater flow, (3) no increase in maximum core flow, (4) a small (less than 3%) increase in reactor operating pressure, and (5) reactor operation primarily along extensions of pre-uprated rod / flow controlI;res. This approach is consistent with the BWR generic power uprate guidelines presenter'in General Electric report NEDC 31897P-1, Generic Guidelines for General Electric Boiling Water Reactor Power Uprate, June 1991 (REF-1). The generic analyses and evaluations in N7JC-31984P, Generic Evaluations of General Electric Boiling Water Reactor Power Upate, July 1991; and ,
Supplements 1 & 2 (Reference 2) are based on a slightly smal,er increase (4.2% vs. 5.0%) than j is requested for BFN 2 and 3. The plant specific analysis for BFNSEP is presented in GE report NEDC-32751P, Power Uprate Safety Anai> sis Report for Browns Ferry Nuclear Plant, Units 2 and 3 (Reference 3). The operating pressure will be i1 creased approximately 30 psi to assure satisfactory pressure control and pressure drop characteristics for the increased steam flow.
2 EVALUATION REACTOR CORE AND FUEL PERFORMANCE 2.1 Fuel Design and Operation i I
All fuel and core design limits will continue to be met by control rod pattern and/or core flow !
adjustments. Current design methods vill not be changed for power uprate. Power uprate will increase the core power density, and will have some effects on operating flexibility, reactivity l characteristics, and energy requirements. These issues are discussed in the following sections. !
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i 2.2 Thermal Limits Assessment Operating limits are established to assure regu'atory and/or safety limits are not exceeded for a range of postulated events as is currently the practice. The operating limit and safety limit MCPR as well as the MAPLHGR and LHGR limits are cyclic dependent and as such will be established or confirmed at each reload as is described in Reference 2.
2.3.1 Power / Flow Operating Map The uprated power / flow operating map includes the operating domain changes for uprated power. The map incluces ne increased core flow (ICF) range and an uprated Maximum Extended Load Line Limit (MELLL). The maximum thermal operating power and maximum core flow correspond to the uprated power and the maximum core flow for ICF. Power has beeri rescaled so that uprated power is equal to 100% rated power. The changes to the power / flow operating map are consistent with the previously approved generic descriptions given in NEDC-31897P-A (Reference 1).
2.4 Stability The BFN units will implement the Option ill methodology of the advanced digital power range neutron monitoring system to address the stability issue which will incorporate the power / flow map and applicable instrumentation setpoints associated with power uprate operation. Until the implementation of Option Ill, the plant will rely on the revised Interim Corrective Actions for both units until the Option lli is implemented. This is acceptable to the staff. l 2.5 Reactivity Control 2.5.1 Control Rod Drives (CRD) and CRD Hydraulic System The control rod drive (CRD) system controls gross changes in core reactivity by positionirig neutron absorbing control rods within the reactor it is also required to scram the reactor by rapidiy inserting withdrawn rods into the core. The CRD system was evaluated at the uprated steam flow and dome pressure. The increase in dome pressure due to power uprate produces a corresponding increase in the bottom head pressure. Initially, rod insertion will be slower due to the high pressure. As the scram continues, the reactor pressure will eventually become the primary source of pressure to cornplete the scram. Hence, the higher reactor pressure will improve scram performance after the initial degradation. Therefore, an increase in the reactor pressure has little effect on scram time. The licensee has indicated that CRD performance ,
during power uprate will meet current Technical Specification requirements. The licensee will cortinue to monitor by various surveillance requirements the scram time performance as required in the plant Technical Specifications to ensure that the original licensing basis for the scram system is preserved.
For CRD insertion and withdrawal, the required minimum differential pressure between the hydraulic control unit (HCU) and the vessel bottom head is 250 psi. The CRD pumps were i j evaluated against this requirement and were found to have sufficient capacity. The flows j required for CRD cooling and driving are assured by automatic opening of the system control 2
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1 valve, thus compensating for the small increase in pressure. The CRD system will continue to perform all its functions at uprated power, and will function adequateJy during insert and withdraw modes.
3 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 1
3.1 Nuclear System Pressure Relief The nuclear boiler pressure relief system prevents overpressurization of the nuclear system during abnormal operating transients. The plant safety / relief valves (SRVs) with reactor scram provide this protection. The operating steam dome pressure is selected to achieve good control characteristics for the turbine control valves (TCVs) at the higher steam flow condition corresponding to uprated power. The uprate dome pressure increase (30 psi) will require a change in the SRV setpoints. The appropriate increase in the SRV setpoints also ensures that adequate differences between operating pressure and setpoints are maintained (i.e., the l
" simmer margin"), and that the increase in steam dome pressure does not result in an increase in unnecessary SRV actuation.
3.2 Code Overpressure Protection The results of the overpressure protection analysis are containec' :n each cycle-specific reload amendment submittal. The design pressure of the reactor pressu,., vessel (RPV) remains at 1250 psig. The ASME code allowable peak pressure for the reactor vessel is 1375 psig (110%
of the design value), which is the acceptance limit for pressurization events. The limiting pressurization event is an MSIV closure with a failure of the valve position scram. The MSIV !
closure was analyzed by the licensee using the NRC-approved methods (ODYN), with the following assumptions: (1) 102% of the uprated core power and 105% of core flow; (2) the maximum initial reactor dome pressure was assumed to be 1050 psig, which is higher than the nominal uprated pressure; (3) one SRV was assumed out-of-service and (4) the analysis did not take credit for externally actuated mode, via electro-pneumatic mode. The SRV opening pressures were +3% above the nominal setpoint for the available valves. The peak reactor j pressure increases by 42 psig to 1309 psig, but remains below the 1375 psig code limit. This l overpressure analysis is acceptable to the staff.
3.4 Reactor Recirculation System Power uprate will be accomplished by operating along extensions of rod lines on the power / flow map with no increase in maximum core flow. The cycle-specific core reload analyses will be performed with the most conservative core flow. The evaluation by the licensea of the reactor recirculation system performance at uprated po;ver determined that the core bow can be maintained with less than 1.3 % incmase in pump speed. The BFN units are licensed for increased core flow (ICF) operation. TVA dces not typically utilize ICF as part of the plant i operational strategy and therefore has not compiled any substantial history involving operation at higher pump speeds. Therefore, TVA's experience with higher pump speed and/or vibration problems is limited. Operational limitations involving isigher recirculation flow and/or vibration
! will be documented and resolved. Vibration monitoring is provided on Units 2 and 3 for the f Recirculation pump motor, pump shaft, and pump case. The licensee estimates that the 3
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l required pump head and pump flow at the uprate condition willincrease the power demand of the recircubtion motors and the pump NPSH but these increases are within the capability of the equipment. The cavitation protection interlock will remain the same in absolute thermal power, since it is based on the feedwater flow rate. These interlocks are based on subcooling in the external recirculation loop and thus are a function of absolute thermal power. With power uprate, slightly more subcooling occurs in the external recirculation loop due to the higher RPV dome pressure. It would therefore be possible to lower the cavitation interlock setpoint slightly, but this change would be small and is not necessary. The licensee concluded that uprated power operation is within the capability of the recirculation system. The licensee will be required to perform power uprate start up testing on the recirculation system to demonstrate flow control over the entire pump speed range to enable a complete l calibration of the flow controlinstrumentation. ,
3.7 Main Steam Isolation Valves (MSIVs)
The main steam isolation valves (MSIVs) have been evaluated by the licensee.
The MSIV operating conditions under power uprate remain within the MSIV design conditions. The BFN units evaluation results are consistent with the bases and conclusions of the generic evaluation. Performance will be monitored by surveillance requirements in the Technical Specification to ensure original I licensing basis for the MSIV's are preserved.(EMEB concurrence is required for this section).
3.8 Reactor Core Isolation Cooling System (RCIC)
The reactor core isolation cooling system (RCIC) provides core cooling when the reactor l pressure vessel (RPV) is isolated from the main condenser, and the RPV pressure is greater j than the maximum allowable for initiation of a low pressure core coolir,g system. The RM j system has been evaluated by the licensee, and is consistent with the bases ana concluws of )
the generic evaluation. The system was found to have the capability to deliver its design rated J fiow at the increased reactor pressure resulting from the increase in the SRV setpoint pressure and the allowable SRV setpoint tolerance of +3 %. The increase in reacto, pressure resulting from these changes, increases the maximum required pump operating head from 2800 feet to 2930 feet. In order for the RCIC system to have the capability to deliver its design rate flow at the higher pump discharge head requirements associsted with power uprate, the maximum specified pump and turbine speed has been increased from 4500 to 4600 rpm. Also, the surveillance test range is increased from 1010 psig and 920 psig to 1040 psig and 950 psig, consistent with the 30 psi increase to the nominal reactor operating pressure.
In response to a stnff request, the licensee has indicated by latter dated May 20,1998 that the recommendations of GE SIL No. 377 are not needed on the RCIC system on each BFN unit.
This recommended modification is intended to achieve the turbine speed control / system
- reliability desired by Sll 377, and is consistent with the requirements in the staff SE of the generic topical report. The purpose of the modification is to mitigate the concern that i
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a clightly higher steam pressure and flow rate at the RCIC turbine inlet will challenge the system trip functions such as turbine overspend, high steam flow isolation, low pump suction pressure and high turbine exhaust pressure. The SIL identifies modifications pr!marily intended i
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for the larger GS-2 model turbine. Although the same modification would dampen the start up transient observed in the smaller GS-1 turbine used in the BFN units. Operating experience with the GS-1 indicates that it is not as susceptible to a overspeed conditions during a quick start. The increase in the maximum RCIC system operating pressure resulting from power uprate is not expected to result in transient speed that require a modification to that described in GE SIL 377. This is acceptable to the staff.
3.9 Residual Heat Removal System (RHR)
The residual heat removal system (RHR) is designed to restore and maintain the coolant inventory in the reactor vessel and to provide primary system decay heat renioval following reactor shutdown for both normal and post-accident conditions. The RHR system is designed to operate in the low pressure coolant injection (LPCI) mode, shutdown cooling mode, suppression pool cooling mode, and containment spray cooling mode. The effects of power uprate on these operating modes are discussed in the following paragraphs.
3.9.1 Shutdown Cooling Mode The operational objective for normal shutdown is to reduce the bulk reactor temperature to 125 F in approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, using two RHR loops. At the uprated power level the decay heat is increased proportionally, thus slightly increasing the time required to reach the shutdown temperature to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This increased time is judged to have an insignificant impact on p! ant safety. Regulatory Guide 1.139, " Guidance for Residual Heat Removal". requires demonstration of cold shutdown capability (200 degrees F reactor fluid temperature) within ;
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. For power uprate, the licensee did not perform a plant specific BFN evaluation for I shutdown cooling based on the criteria of Regulatory Guide 1.139. However, as noted above
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the licensee stated that the reactor can be cooled to less than 125 degrees F in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, which '
meets the 36-hour crierion.
3.9.2 Suppression Pool Cooling and Containment Spray Modes The Suppression Pool Cooling (SPC) and Containment Spray Cooling (CSC) modes are designed to provide sufficient cooling to maintain the containment and suppression pool i temperatures and pressures within design limits during normal operation and after a blowdown {
in the event of a design bam LJCA. This objective is met with power uprate, since the peak l suppression pool temperature analysis by the licensee (described in Section 4.1.1 of the l licensee submittal) confirms that the pool temperature has not increased and will stay below its design limit at uprated conditions. There is no increase in the containment spray temperature.. ;
This has a negligible effect on the calculated values of drywell pressure, drywell temperature, i and suppression chamber pressure, since these parameters reach peak values prior to actuation of the containment spray. The licensee stated that the capability of the CSC mode is therefore acceptable for puwer uprate. The effect of power uprate to the above mentioned l cooling modes of the RHR system are acceptable to the staff.
l 3.10 Reactor Water Cleanup (RWCU) System I
The RWCU system pressure and temperature will increase slightly as a result of power uprate.
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The licensee has evaluated the impact of these increases and has concluded that uprate will {
not adversely affect system integrity. The cleanup effectiveness may be diminished slightly as {
a result of the increased faedwater flow to the icactor; however, the current limits for reactor
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water chemistry w!!! remain uachanged for power uprate. These effects on the RWCU system are acceptable to the staff (Chemical Engineering Branch concurrence is required).
4 ENGINEERED SAFETY FEATURES 4.1 Containment System Performance (Add input from SCSB).
4.2 Emergency Co,c Cooling Systems (ECCS)
The effect of power uprate and the increase in RPV dome pressure on each ECCS system is addressed talow. The effect of suppression pool temperature on the NPSH of the RHR pumps during the SPC and CSC moces is also discussed in Sectk3n 4.2 of the licensee submittal. The results show that there is adequate NPSH margin for the RHR and core spray (CS) pumps.
Since the peak temperature is the same as pre-uprate conditions, the NPSH available at peak (
temperature conditions is not adversely affected and uprate will net affect compliance to the ECCS pump NPSH requiremuits.
4.2.1 High Pressure Core injection System (HPCI)
The HPCI system has been evaluated by the licensee, and is consistent with the bases and conclusions of the generic evaluation. The licensee has indicated that they have implemented the guidance contained in GE SIL 480 on the HPCI system for on each unit. The licensee stated that the HPCI is capable of delivering its design flow at the uprate conditions. The increase in reactor pressure resulting from the power uprate increases the maximum required pump set operating head from 2800 feet to 2930 feet. In order for the HPCI system to have the ability to deliver its design rate flow at the higher pump set discharge head requirements associated with power uprate, the maximum pump and turbice rated speed has been increased from 4000 to 4100 rpm. The licensee will be required to perform startup testing on HPCI during the initial startup alter being licensed at uprated power. The surveillance test range is increased from 1010 psig and 920 psig to 1040 psig and 950 psig, consistent with the 30 psi increase to the nominal reactor operating pressure. The licensee stated in their May 20,1998 letter that the reliability of the HPCI system will be monitored in accordance with the criteria developed to comply with the Maintenance Rule that was implemented in accordance with 10 CFR50.65. This is acceptable.
4.2.2 Low Pressure Core injection System (LPCI mode of RHR)
The hardware for the low pressure portions of the RHR are not affected by power uprate. The upper limit of the low pressure ECCS injection setpoints will not be changed for power uprate, therefore the low pressure portions of these systems will not experience any higher pressures. ;
The licensing and ' design flow rates of the low pressure ECCS will not be increased. In addition, j l the RHR system shutdown cooling mode flow rates and operating pressures will not be increased. Therefore, since the system does not experience different operating conditions due <
to power uprate, there is no impact due to power uprate. The licensee stated that the BFN l;
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i units are bounded by the gerraric analyses presented in Section 4.1 of Reference 2. This is acceptable to the staff. l l
4.2.3 Core Spray System (CS) i The hardware for the low pressure core spray are not affected by power uprate. The upper limit I of the low pressure ECCS injection setpoints will not be changed for power uprate, therefore the low pressure portions of these systems will not experience any higher pressures. The licensing and design flow rates of the lov/ pressure ECCS will not be increased. Therefore, since thase systems do not experience different operating corsditions due to power uprate, there is no I impact due to power uprate. Also, the impact of power uprate on the long terin response to a LOCA will contiaue to be bounded by the short term response. The licensee stated that the BFN units are bounded by the generic analyses prestnted in Section 4.1 of Reference 2. This is acceptable to the staff.
4.2.4 Autumatic Depressurization Systems (ADS)
The ADS uses safety / relief valves to re'fuce reactor pressure following a small break LOCA with HPCI failure. This function allows low pressure coolant injection (LPCI) and core spray (CS) to flow to the vessel. The ADS initiation logic and ADS valve control are adequate for uprate. Plant design requires a minimum flow capacity for the SRVs, and that ADS iMtiate after a time delay on either low waier level plus high drywell pressure, or on low water level alone.
The ability to perfctm either of these functions is not affected by power uprate.
4.3 ECCS Performance Evaluation The emergency core coc!ing systems (ECCS) are designed to provide protection against hypothetical Joss-o!-coolant accidents (LOCAs) caused by ruptures in the primary systems piping. The ECCS performance under all LOCA conditions and their analysis models must satisfy the requirements of 10 CFR 50.46 and 10 CFR Appendix K. The fuel, used in BFN Uriits 1 and 2, was analyzed by the licensee with the NRC-approved methods (SAFER /GESTR). The results of the base ECCS.LOCA analysis using NRC-approved methods is presented in NEDC-32484P (Reference 4), the plant specific ECCS-LOCA results for the BFN units. I The licensee used the staff approved SAFER /GESTR (S/G) methodology to assess the ECCS l capability for meeting the 10 CFR 50.46 criterka. The S!G-LOCA analysis for the BFN Units 1 and 2 was performed by the licensee with the appropriate reload fuel in accordance with NRC requirements to demonstrate conformance with the ECCS acceptance criteria of 10 CFR 50.46 and Appendix K (Reference 4). The base S/G-LOCA and the analyses were performed at a nominal power level 3458 MWt (105% of the current rated power of 3293 MWt) and an Appendix K power level of 3527 MWt (102% of 3458 MWt)in anticipation of future power I
uprate. Therefore, this analysis bounds the requested power uprate of 3458 MWt. The analyses demonstrate that the limiting licensing basis PCT occurs for the recirculation line break DBA as documented in the base report (Reference 4) remain applicable to the planned power uprate analysis to 3458 MWt. The analyses performed in Ref. 4 are in accordance with j the NRC requirements and demonstrated conformance with the ECCS acceptance criteria of 10 I i
CFR 50.46 and hence it is acceptable. The licensee must submit the results of the plant 1
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specific LOCA analysis at each reload.
Single loop operation (SLO) is not currently licensed for the BFN units, and is not being requested under this power uprate license amendment request.(PM to verify) The licensee will submit any MCPR or MAPLHGR multipliers with the core operating limits repod (COLR) as is the usual practice.
6.5 STANDBY LIQUID CONTROL SYSTEM The Standby Liquid Controi System (SLCS) is designed to assure reactor shutdown, from full power operation to cold subcritical by mixing boron with the primary reactor coolant, in the event when no control rods can be inserted. The ability of the SLCS boron solution to achieve and maintain sate shutdown is not a direct function of core thermal power, and therefore, is not affected by power uprate. SLCS shutdown capability is re-evaluated for each reload core. The SLCS pumps are positive displacement pumps, where small pressure changes in the SRV setpoint would have no effect on the rated injection flow to the reactor. The system was found to have the capability to deliver its design rated flow resulting from the increased SRV setpoint pressure and the allowable setpoint tolerance of +3 % The pump surveillance test pressure is being changed from 1275 psig to 1325 psig, to account for the increase in system injection pressure at power uprate conditions.
9 REACTOR SAFETY PERFORMANCE FEATURES 9.1 Reactor Transients Reload licensing analyses evaluate the limiting plant transients. Disturbances of the plant caused by a malfunction, a single failure of equipment, or personnel error are investigated according to the type of initiating event. The licensee will use its NRC-approved licensing l analysis methodology to calculate the effects of the limiting reactor transients as identified in the I generic guidelines. The limiting events for the BFN units were identified as those analyzed in Reference 1. The relatively small changes in rated power and maximum allowed core flow are not expected to effect the selection of limiting events. The events explicitly evaluated for the power uprate analysis are:
Loss of Feeuwater Heating (LOFWH)
Feedwater Controller Failure (FWCF)
Generator Load Rejection without Bypass (GLRWOB) I Turbine Trip without Bypass (TTWOB)
Rod Withdrawa! Error (RWE) i Slow Recirculation Flow increase Inadvertent HPCI Actuation ;
The limiting events which establish the minimum critical power ratio (MCPR) operating limits are j for the power uprate conditions are GLRWOB, TTWOB and FWCF, The analyses for the !
limiting transients were performed at 100 % power. These events are analyzed with staff approved methods ODYN code and GEMINI methodology which include allowance for core 8
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phase. The review, update of the EOls for any changes, and training are scheduled to be completed prior to the power uprate start-up for Browns Ferry unit 3. This is acceptable.
Maine Yankee Lessons Learned (PM SHOULD VERIFY WITH OTHER BRANCHES)
The BFN power uprate amendments were reviewed with consideration given to the recommendations from the Report of the Maine Yankee Lessons Learned Task Group, dated December 5,1996. This report is documented in SECY-97-02, " Response to OlG Event Inquiry 96-04S Regarding Maine Yankee," dated February 18,1997. The Task Group concluded that a power uprate procedure should be developed in light of the Maine Yankee findings. Although a Maine Yankee lessons learned power uprate procedure has not been developed, the recommendations of the report were considered in the review of the BFN power uprate.
The staff requested that the licensee identify all codes / methodologies used to obtain safety limits and operating limits and how they verified these limits were correct for the appropriate uprate core. The licensee was also requested to identify and discuss any limitations associated with these codes / methodologies that may have been imposd by the staff. In the TVA letter dated May 20,1998 TVA responded to the staff request. TVA identified all the codes / methodologies used for the power uprate analyses and confirmed that all the models/ methodologies are used appropriately for the power uprate evaluation.
In letter dated (PM to add), TVA confirmed that they audited GE to assure that the codes are used by GE correctly for power uprate conditions and the limitations and restrictions were <
followed by GE appropriately.
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The main findings centered around the use and applicability of the code methodologies used to support the uprated power. TVA has made an effort to verify that the codes are appropriate and applicable to the plant given the uprated conditions. TVA indicated that the LOCA and transients analyses conform with the generic analyses approved by the staff for power uprate.
This is acceptable.
FSAR CHANGES TO BE COMPLETED BY THE PM CONCLUSION We conclude that the analyses and evaluaticas supporting the power uprate follow the generic guidelines approved by the staff and hence is acceptable.
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REFERENCES J
- 1. GE Nuclear Energy, " Generic Guidelines F:r General Electric Boiling Water Reactor Power Uprate", Licensing Topical Report NEDO-31897, Class I (non-proprietary), February 1992; and NEDC-31897P-A, Class 111 (Proprietary), May 1992.
- 2. GE Nuclear Energy, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31984P, Class 111 (Proprietary), July 1991; NEDO-31984, Class I (Non-proprietary), March 1992; and Supplements 1 and 2.
- 3. GE Nuclear Energy," Power Uprate Safety Analysis for the Browns Ferry Nuclear Plant Units 2 and 3, LicensingTopical Report NEDC-32751P, Class lll (Proprietary),
September 1997.
- 4. GE Nuclear Energy, Browns Ferry Nuclear Plants Units 1,2 and 3 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis (Revision 1), NEDC-32484P, February 1996.
- 5. SECY-91-401," Generic Boiling Water Reactor power uprate program," December 12,1991.
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