ML20135C981
| ML20135C981 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/14/1997 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML20135C833 | List: |
| References | |
| NUDOCS 9703040373 | |
| Download: ML20135C981 (88) | |
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ATTACHMENT 6
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j FARLEY NUCLEAR PLANT UNITS 1 AND 2
.j POWER UPRATE PROJECT BOP LICENSING REPORT 1
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9703040373 970214 PDR ADOCK 05000340 PDR w
p BOP UPRATE LICENSING REPORT FNP - UNITS 1 AND 2 i
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l 6l FNP UPRATE LICENSING REPORT TABLE OF CONTENTS SECTION TITLE PAGE 1.0 Executive Summary 1
2.0 BOP Program Descriptiou 2.1 BOP Program Overview 2
2.2 Condensate and Feedwater 5
1 2.3 Circulation Water 9
24 Main Turbine i1 2.5 Main Turbine Auxiliaries 13 2.6 Main Generator & Auxiliaries 16 2.7 Main Steam 18 j
2.8 CCW 20 2.9 Service Water 22 2.10 Spent Fuel Pool 24 2.11 LHSI(RHR)/ HHSI(Charging) 27 2.12 Auxiliary Feedwater 28 2.13 Containment & Subcompartment Analysis 32 2.14 Post LOCA Hydrogen Generation 45 2.15 Electrical Equipment Qualification 52 2.16 Radiological Assessment 54 2.17 Containment Ventilation 67 2.18 Auxiliary Building Ventilation 70 7.19 Misc. Mechanical Reviews 72 2.20 Misc. Electrical Reviews 74 2.21 Misc. I&C Reviews 77 2.22 Environmental impact Evaluations 79 3.0 Conclusion 85 4.0 References 86 O
BOP UPRATE LICENSING REPORT FNP-UNITS 1 AND 2
O 1.0 Executive Summary The purpose of tb Farley power uprate is to increase the electrical output of the Farley Nuclear Plant (FNP). In order to determine the potential impact on major plant design features, a detailed systems and safety analyses review was conducted by Southern Nuclear Operating Company (SNC),
Westinghouse, Bechtel, and Southern Company Services (SCS) to demonstrate the acceptability of increasing the rated thermal power of FNP from 2652 MWt to 2775 MWt.
The NSSS analyses and evaluations prepared to suppon this increase are described in WCAP-14723, "Farley Nuclear Plant Units 1 and 2 Power Uprate Project NSSS Licensing Repon," January 1997.
This repon provides descriptions of the evaluations of the design and licensing aspects of balance of plant (BOP) systems, structures, components, and analyses not included in the NSSS Licensing Report; including programmatic evaluations such as radiological releases and doses, qualification of safety-related electrical equipment, the FNP MOV program, and environmental impact.
The BOP analyses and evaluations used input assumptions and parameter values that are identical to the parameter values used in the NSSS report. The input assumptions and parameter values were developed jointly among involved organizations and are procedurally controlled.
The BOP analyses and evaluations demonstrate that FNP remains in compliance with the applicable design and licensing bases, criteria and requirements at power uprate conditions and, therefore, can operate acceptably at power uprate conditions. This report, in part, provides the technical basis for the conclusions presented in the significant hazards evaluation for the proposed Technical Specifications changes required for the Farley power uprate project.
_.e BOP UPRATE LICENSING REPORT I
FNP - UNITS 1 AND 2
2.0 BOP Program Description 2.1 BOP Program Overview 2.1.1 Introduction / Background Farley Nuclear Plant (FNP) is a two unit site. Each unit is currently licensed for a rated thermal power of 2652 MWt. Southern Nuclear Operating Company proposes to increase the rated thermal power of each unit to 2775 MWt. In order to determine the potential impact on major plant design features, systems and safety analyses, a detailed programmatic review was conducted by Southern Nuclear Operating Company (SNC), Westinghouse, Bechtel, and Southern Company Services (SCS).
2.1.2 Quality Assurance and Code Requirements The BOP analyses and evaluations to support the increase in rated thermal power (uprate) were performed in accordance with quality assurance program requirements and engineering procedures which comply with 10 CFR 50 Appendix B. These analyses and evaluations also conform with industry codes and standards and regulatory requirements applicable to FNP.
2.1.3 Scope of Review i he results of the NSSS systems, components, transient and accident analyses, and nuclear fuel analyses and evaluations are presented in the NSSS Licensing Report. This report provides descriptions of the evaluations of the design and licensing aspects of BOP systems, structures, components and analyses not included in the NSSS Licensing Report; including programmatic evaluations such as radiological releases and doses, qualification of safety-related electrical equipment, the FNP MOV program, and environmental impact. The BOP evaluations and analyses i
include input from the NSSS evaluations and analyses as required, and in some cases, BOP results provide information for the NSSS analyses and evaluations. These interfaces were developed jointly among the involved organizations. The division of responsibility for the BOP evaluations and analyses is described in Table 2.1-1.
Differences between Unit I and Unit 2 were identified. Where differences exist, each unit was evaluated, or a bounding analysis or evaluation for both units was performed. No significant differences between units were identified which would adversely impact uprate. Minor differences are discussed in the individual technical evaluation sections as appropriate.
2.1.4 Plant and Technical Specification Changes The major BOP system modifimtions included in the BOP analyses and evaluations are shown in Table 2.1-2. Technical Specifications changes required as a result of the BOP analyses and evaluations are associated with the containment pressure analyses for LOCA and MSLB cvents. The Containment System Bases sections for internal pressn-- snd structural integrity and Containment Leakage Rate Testing Program Administrative secuon will be revised to reflect the results of the new containment pressure analyses.
O' BOP UPRATE LICENSING REPORT 2
FNP - UNITS 1 AND 2
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i O Table 2.1-1 BOP Division of Responsibility i
BOP Report Lead Section Organization Description i
SCS Introduction / Background 2
SCS Condensate and Feedwater 3
SCS Circulation Water 4
SCS &
Main Turbine Evaluations Westinghouse-Orlando S
SCS &
Main Turbine Auxiliaries Westinghouse-Orlando 6
SCS &
Main Generator & Auxiliaries Westinghouse-Orlando 7
SCS Main Steam 8
SCS & Bechtel CCW 9
SCS Service Water 10 SCS Spent Fuel Pool 11 SCS &
Westinghouse-Pittsburgh 12 Bechtel Auxiliary Feedwater 13 SCS Containment & Subcompartment Analysis 14 Bechtel Post LOCA Hydrogen Generation 15 Bect.tel Equipment Qualification 16 SCS Radiological Assessment 17 SCS Containment Ventilation 18 SCS Auxiliary Building Ventilation 19 SCS Misc. Mechanical Reviews 20 SCS Misc. Electrical Reviews 21 SCS Misc. I&C Reviews 22 SCS & SNC Environmental Evaluations l
lO BOP UPRATE LICENSING REPORT 3
FNP - UNITS 1 AND 2
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Table 2.1-2 Modifications to be Implemented as Part of Power Uprate A.
Main Turbine 1.
Nozzle Blocks 2.
Control Stage Rotating Blades 3.
First Two Stages of Stationary Blades B.
Condensate Pumps 1.
Pump Modifications 2.
Motor Bearing Modifications 3.
SGFP Suction Pressure Trip and Alarm Setpoints C.
Low Pressure Feedwater Heater Channel Side Relief Valves 2.
- 2 Feedwater Heater Drain Valve Modifications D.
Thermal Performance Test Connections l
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BOP UPRATE LICENSING REPORT 4
FNP - UNITS 1 AND 2
0 2.2 Condensate and Feedwater 2.2.1
System Description
The condensate and feedwater system is composed of one twin shell, single pressure condenser; three 50% condensate pumps; two strings oflow pressure feedwater heaters; two 50% variable-speed, turbine-driven feed pumps; two 50% heater drain pumps; two strings of high pressure feedwater heaters; one turbine gland steam condenser; and four 20 SCFM two-stage steam jet air ejectors with associated inter and after condensers. The latter are evaluated in Section 2.3.
2.2.2 Condensate Pumps 2.2.2.1 Scope of Review The purpose of this evaluation was to determine the performance capacity of the modified condensate pumps for normal and abnormal operation at uprate conditions. Parameters evaluated are as follows.
- Total dynamic head and flow
- Condensate pump flow margin
- Condensate pump horsepower
- Condensate pump wear margin
- Condensate pump NPSHA
- SGFP NPSHA
- SGFP suction header alarm and trip setpoints p
- Steam generator feed pump (SGFP) suction V
header pressure The limiting transient is a load rejection of up to 50% of full power with automatic reactor control I
and steam dump, without a reactor trip. To preclude a reactor trip on steam generator low level due to a steam / feed flow mismatch the condensate and feedwater system is designed to be capable of supplying approximately 96% of the feedwater flow at full power witrine steam pressure 100 psi above the full load steam pressure.
2.2.2.2 Summary of Evaluation Approximately 6% increase in condensate flow is necessary to provide sufficient flow to the steam generators to support normal operation at power uprate conditions. To support this increased flow requirement and the associated SGFP suction temperature increase, it is necessary to modify the condensate pumps. With the condensate pump modification, the SGFP suction header pressure setpoints should be modified to 300 psig for alarm and 275 psig for trip. The modified setpoints allow for adequate SGFP NPSHA.
Because of the increased condensate flow required for the 50% load rejection, the evaluation was based on three condensate pumps in service. Two condensate pump operation would not support the required flowrates due to insufficient NPSHA to the condensate pumps.
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v BOP UPRATE LICENSING REPORT 5
FNP - UNITS 1 AND 2
2.2.2.3 Summary of Conclusions NPSHA for the modified condensate pumps was determined to be adequate for uprate conditions.
Assuming three condensate pumps in service, the condensate pumps have enough capacity to pass the required flow during the 50% load reduction condition.
2.2.3 Heater Drain Pumps 2.2.3.1 Scope of Review The purpose of this evaluation was to determine the performance capability of the heater drain pumps for uprate conditions. Parameters evaluated included the following.
- Total dynamic head and flow
- Discharge control valve position
- Net positive suction head available
- Horsepower 2.2.3.2 Summary of Evaluation The heater drain pump performance capability was determined based on evaluation of the following parameters: heater drain pump estimated uprate flow, head from the original design curve at the uprate condition flow, control valve pressure drop and percent open, and brake horsepower.
2.2.3.3 Summary of Conclusions The heater drain pumps were determined to perform adequately in uprate operating conditions when two condensate pumps were in operation at normal operating loads down to approximately 60%.
During part load operation of less than approximately 60% with two condensate pumps in service, the heater drain pumps were not able to pump full heater drain tank flow because of the high SGFP suction header pressure. This result isjudged not to be an operating concern because part load operation under 60% is not normal operation and any unit performance degradation due to dumping of heater drains will be minimal.
2.2.4 Feedwater Heaters 2.2.4.1 Scope of Review The purpose of this evaluation was to oetermine the adequacy of the feedwater heaters for uprate conditions. Parameters addressed include the following.
- Extraction, drain, and feedwater flows
- Extraction steam nozzle velocity
- Feedwater nozzle velocity
- Drain inlet velocity
- Drain outlet velocity
- Shellside relief valve flow
- Tubeside pressure drop
- Shellside pressure drop
- Tubeside design pressure
- Shellside design pressure
- Tubeside velocity
- Heater vibration characteristics Feedwater heater level control is evaluated in section 2.21.
BOP UPRATE LICENSING REPORT 6
FNP - UNITS 1 AND 2
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2.2.4.2 Summary of Evaluation Feedwater heater operating parameters at uprate conditions were compared to Heat Exchanger Institute (HEI) standards. Extraction steam, feedwater, and drain nozzle flows and velocities are within HEl recommendations, with few exceptions which exceed HEl guidelines by a small amount.
Based on the design and materials of construction these exceptions will not cause abnormal erosion or performance degradation and is considered acceptable for uprate.
The largest total shellside pressure drop is for the #3 heater which is within the HEI guideline of <5 psid. The uprate shellside pressure of all of the heaters is below the design rating. Tubeside design pressure is determined by the shutoff head of the condensate pumps, which, based on the proposed shutoff head for the modified pump, meets the tubeside design pressure.
Feedwater heater vibration specifications dictate that the critical flow of the drain cooling zone must be at least 1.35 times the actual cross flow of the drain cooling zone. Each of the six feedwater heaters met this criteria.
2.2.4.3 Summary of Conclusion All uprate condition nozzle flows and velocities, shellside and tubeside pressure ratings, vibration, and pressure drops through the heaters are within acceptable limits. Therefore, the overall
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performance of the feedwater heaters is acceptable for operation at uprate conditions.
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2.2.5 Steam Generator Feed Pumps 2.2.5.1 Scope of Review The purpose of this evaluation was to determine the adequacy of the steam generator feed pumps to support uprate operation. The following parameters were evaluated.
- SGFP total dynamic head, flow, and design pressure
- SGFP power and speed
- Steam generator feedwater control valve (FCV) position.
NPHSA to the SGFPs (provided by the condensate pumps) is evaluated in Section 2.2.2.
In addition to normal uprate operation, the SGFP and FCV were evaluated for their ability to supply the adequate feedwater flow at the required head to the steam generators under the abnormal 50%
load rejection or loss of a single heater drain pump.
2.2.5.2 Summary of Evaluation At normal uprate conditions, it is estimated that the SGFP flow and power increase slightly, and that speed and total dynamic head decrease slightly. The FCV opens approximately an additional 5% of Q
full stroke, reducing the pressure drop across the valve. Uprate operation with a 50% load reduction V
BOP UPRATE LICENSING REPORT 7
FNP - UN 'I 91 AND 2
or during a heater drain pump trip results in a small estimated SGFP speed and power increasc over the current operating condition.
With the modified condensate pump, the design pressure of the SGFP is greater than the maximum condenser backpressure plus the shutoff head of the condensate pump plus the shutoff head of the SGFP; therefore, the SGFP rating is acceptable for power uprate conditions.
2.2.5.3 Summary of Conclusicns The evaluation indicates no significant change to SGFP speed or power with normal uprate operation. The FCVs may open more, but will not exceed the 85% open recommendation. Under abnormal operating conditions, the feedwater requirements do not exceed the SGFP rating or the FCV capacity.
2.2.6 Piping and Valves 2.2.6.1 Scope of Review The piping systems addressed included.
- Condensate and feedwater piping
- Feedwater heater drains
- Condensate pump and SGFP minimum flow
- Feedwater heater tubeside thermal relief valves
- Extraction piping to the L.P. and H.P. feedwater heaters and the moisture separator reheaters 2.2.6.2 Summary of Evaluation The pipe velocities, presures and temperatures were checked against the design requirements. To determine the effect of uprate on the feedwater heater tubeside relief valves, the shutoff head, following condensate pump modifications for uprate, was compared to the current design pressure and current valve setpoints. All extraction steam velocities for uprate conditions are within or well below reasonable design ranges. The present system provides sufficient minimum condensate and feedwater flow protection for the condensate pumps and SOFPs. The present MSR shell drain system does not require modifications for uprate conditions.
Operation at current power levels results in #2 heater drains at design capacity and occasional lifting of the L.P. heater thermal relief valves; thus, modifications to povide increased flow capacity are required to support the increased heater drain flow for power uprate.
2.2.6.3 Summary of Conclusions The evaluation indicates no significant problems with uprate conditiore. With modifications, the #2 heater drain valves and L.P. heater thermal relief valves are acceptable a r uprate conditions.
O BOP UPRATE LICENSING REPORT 8
FNP - UNITS 1 AND 2
rh 2.3 Circulation Water 2.3.1
System Description
The circulating water system (CWS) rejects waste energy from the turbine cycle. The circulating water system for each unit is comprised of a twa shell condenser, three 14 cell cooling towers, two circulating water pumps, associated circulating v>ater piping and valves.
2.3.2 Scope of Review The scope of this review includes an evaluation of the effect of uprate operation on turbine backpressure and circulating water system equipment limitations. A mathematical model based on HEI guidelines and cooling tower algorithms was used to predict condenser backpressure (with and without turbine bypass) and circulating water temperatures where required for the evaluation.
Additional evaluations were performed for:
- Condenser steam side vibration and steam side water impingement, tube erosion, condenser connections, and expansion joint performance;
- Tower performance (turbine backpressure impact, drift, evaporation and makeup rate), station service requirements of the cooling tower fans, and hot water return temperature impact on the life of the cooling tower fill; G
- Cooling cycle performance for normal operation with two circulating water pumps operating and V
abnormal operation with single circulating water pump operating at runout.
l The capacity of the air removal system, and the acceptability ofinter-and after-condenser for the modified condensate pump shutoff head were evaluated in conjunction with the CWS.
2.3.3 Summary of Evaluation Condenser duty increases from approximately 6404 MMBtu/hr to approximately 6785 MMBru/hr under uprate operating conditions with consequent small backpressure, turbine exhaust flow, moisture, and expansion joint temperature increases. The condenser evaluation included the cooling tower impact on condenser backpressure.
The maximum calculated hot water temperature to the tower increased but remains less than the fill temperature limit. Since cooling tower fans are constant volume machines, no change in volumetric flow rate is anticipated. The estimated makeup rate with uprate operation is expected to increase approximately 6% and remain well within the makeup capacity of the service water system.
With two pumps running, the impact on the circulating water pumps is a (less than 1%) decrease in power consumption based on the lower specific gravity of the water at the increased temperature.
For single pump runout operation, the water temperature and the condenser backpressure increase slightly, but remain within system limits.
r BOP UPRATE LICENSING REPORT 9
FNP - UNITS 1 AND 2
Altimugh the condenser duty increases with uprate operation, no detailed evaluation of the circuhting water system is required since the circulating water flow does not change and water temperatures are changing less than 2 F throughout the system.
After uprate, the steam jet hogging ejectors must remove the same air capacity (i.e., evacuate the condenser for startup in the same time constraint) as under the current operating condition, so there is no uprate impact. The design pressure of the inter-and after-condenscrs water chamber is 650 psig, which bounds the shutoff head of the modified condensate pumps, and therefore remains acceptable for power uprate.
With uprate operation, cooling tower fan horsepower will decrease an insignificant amount (less than 1.5%). Makeup to the CWS is estimated to increase approximately 6% with uprate operation.
2.3.4 Summary of Conclusions The results of the analysis for uprated conditions indicate that neither unit will experience operating limitations based on turbine backpressure under normal operating conditions or while experiencing normal equipment degradation.
Steam induced tube vibration and tube erosion, and condenser connection flows and energy levels will not be significantly greater than that experienced with the current operation, and therefore are considered adequate for uprate operation.
The condenser pressures are within the design range of the expansionjoint. Since the spansion joint is shielded and constructed of stainless steel, the slight increase in temperature with uprate is considered to have no impact on the life and suitability of the condenser expansion joint.
Uprated CWS operation is acceptable with current cooling tower fill. With two circulating water pumps operating, the hot water temperature limitation of the tower fill will not be exceeded. Under single pump runout operation, the hot water temperature to the tower may exceed the limit for continuous operation. Since the 1% wetbulb will be experienced an estimated 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> per year and the likelihood that a circulating water pump wi!! concurrently be out of service, the increase in hot water temperature has an insignificant impact on the life of the tower fill and remains lower than the allowable excursion temperature. The increase in CWS makeup flow is well within the service water system capacity.
The condenser air removal system is presently operating satisfactorily, and performance will not change at uprated conditions. The inter-and after-condensers design pressure is rated adequately for the modified condensate pump shutoff head.
O BOP UPRATE LICENSING REPORT 10 FNP - UNITS 1 AND 2
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2.4 Main Turbine Evaluations The turbine evaluations are based on a best estimate steam generator discharge pressure of 787 psia that was selected as the design pressure for power uprate. The latest secondary plant thermal 4
performance test data was used to produce preliminary heat balances for Unit I and 2 at current and uprated conditions. Modifications to the high pressure (HP) turbines are required to pass the i
increased volumetric flow rates required for operation at the increased power level at 787 psia steam generator discharge pressure. In order to allow for design and manufacturing uncertainties of the HP turbine modifications and to provide for future operating flexibility a 2% flow margin was selected for the design of the turbine modifications.
The original turbine included a significant amount of flow margin that caused the turbine cycle efficiency to be less than desired. Modifications to the HP turbine were performed during the 1980s to reduce the available flow capacity of the turbine and therefore reduce the flow margin. These modifications improved the turbine efficiency and optimized the MWe output.
2.4.1 Scope of Review i
This evaluation verifies the mechanical capability, including potential operational impacts, of the turbine hardware modifications (nozzle blocks, control stage blades, and two rows of stationary blades per end) to accomplish the Farley power uprate. The turbine evaluation includes the following.
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- Stall / flutter, vibration, and blade, disc,
- Overspeed sensitivity and protection rotor keyway and coupling loads
- Feedwater heaters out of service
- Stop and control valves
- Number of valve admissions
- Erosion and corrosion
- DEH control system
. Valve test interval
- Turbine trip setpoints
- Partial arc operation 2.4.2 Summary of Evaluation The replacement of the control stage blades is acceptable, although these blades have a smaller design margin, especially at the minimum are of admission of 50%. The uprate does not change the recommended stall / flutter and high back pressure limits for the style of L-0 LP blade used on the Farley rotors, and the recommendations concerning the appropriate use of the blade vibration monitor remain valid. The evaluation of blades not being replaced as part of the uprating determined that there is sufficient design margin to pass the increase in steam flow. No significant decrease in the life of the current blades b expected due to uprate. The uprating does not impact the cracking mechanisms, and hence. the probability of blade faiiere remains the same. If a 75%
minimum arc of admission is determined to be required, ti. re will not he an impact to the IR blade root cracking.
3 BOP UPRATE LICENSING REPORT I1 FNP - UNITS 1 AND 2
Susceptibility of the LP discs to stress corrosion cracking (SSC) was evaluated by comparison of the LP steam temperature and moisture levels at the current and uprated conditions, which found that the conditions in the wet rows are virtually identical, thus uprating has no measurable effect. Vibration levels do not change based on the balancing criteria required in the turbine contract. The thrust loading on a double now machine such as Farley Units 1 and 2 is slight, and any effect due to uprating is not significant. The proposed uprating does not involve changing rotors or large changes in blading. The increase in electrical load (torque)is 5% or less. Therefore, changes in torsional response are not significant.
Comparison of the steam coriditions at the last row blade inlet for the current and uprated conditions found moisture levels are essentially the same, and there is no quantifiable effect on last row blade tip erosion rate. The uprating etTect on volumetric flow, moisture content, and temperature (the parameters that most directly affect erosion and/or corrosion wear) is relatively small. Farley has successfully dealt with the issue of erosion and/or corrosion as part of routine maintenance. Wear does not significantly increase as a consequence of the uprate. Since uprate results in the main steam operating pressure decreasing and does not impact the no load pressure, the pressure rating of the valves and piping is suitable for uprate. The turbine-generator couplings have been reviewed with respect to torque and power and found to be suitable for operation under the increased loads.
Overspeed trip points are set such that the unit will not achieve a final overspeed greater than the design overspeed of 120%. For the Farley units, it has been calculated that a mechanical overspeed trip at 111% will not allow the unit to achieve a final overspeed greater than 120%. The 111%
calculated trip point for the mechanical trip device is still valid at the uprated conditions. The uprating has no effect on the turbine trip setroints. A change in the turbine valve control settings is anticipated with the changes in steam parameters associated with the uprating. Changes in the control settings may be accomplished with routine software changes in the DEH control system.
Uprating the turbine has no significant effect on the destructive overspeed design limit. Quarterly valve testing continues to be acceptable.
The method of operating the turbine generator does not change with the uprating. Ramp rates, minimum and maximum temperatures, and blade resonance avoidance points remain the same.
Continued partial arc operation is expected following uprate.
2.4.3 Summary of Conclusions Based on the evaluation of the modifications to be made to the turbine (nozzle blocks, the control stage blades, and two rows of stationary blades per end), the turbine will be capable of operating at the increased NSSS power.
Stall / flutter and high back pressure limits, blade life, rotor vibration and thrust loading, and LP disc SCC are unchanged due to power uprate conditions. L-0 rotating blades and HP and LP stationary component erosion may increase slightly due to uprating, but will cause no significant decrease in blade life, and steam valve velocities and noise increase due to uprate but remain at acceptable levels.
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BOP UPRATE LICENSING REPORT 12 FNP - UNITS 1 AN'D 2
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Uprating does not require a change in overspeed trip settings. A change in turbine control valve i
setting is expected due to uprate; this change can be accomplished by routine changes to DEH software.
4 The method of operating the turbine generator, valve testing intervals, and turbine trip setpoints are not affected by the uprating. Control blading may dictate a change in minimum valve admission from 50% to 75%, and adjustments may be required to the turbine valve control settings. These i
changes are within the normal design capabilities of the equipment and normal maintenance practices.
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i BOP UPRATE LICENSING REPORT 13 FNP - UNITS I AND 2
O 2.5 Main Turbine Auxiliaries 2.5.1 Scope of Review This evaluation assessed the adequacy of the moisture separator reheaters (MSRs), gland sealing steam system (GSS), and the steam generator feed pump turbines (SGFPTs) for the uprate cond.: tion.
Main turbine auxiliary system and equipment parameters included in the evaluation are as follows.
- Cycle and heating steam flows
- Drain outlet velocity
- Cycle steam inlet nozzle velocity (cold reheat)
- MSR pressure drop
- Cycle steam outlet nozzle velocity (hot reheat)
- Vibration characteristics
- Low pressure heating steam nozzle velocity
- Tubeside design pressure
- High pressure heating steam nozzle velocity
- Shellside design pressure
- Shellside (reheater bundles) pressure drop
- HP turbine exhaust pressure
- GSS valve position
- SGFPT operating conditions 2.5.2 Summary c,f Evaluation HEI recommendations are used for evaluation of the MSRs with the following results. Cycle steam flow is below the maximum allowable chevron separator flow. Nozzle flows and velocities are essentially unchanged (small increase or decrease) from current conditions. Operating pressures, both shellside and tubeside, are less than the design ratings of the original MSRs.
The only impact of uprate to the GSS is an increase in HP turbine exhaust pressure of approximately 6%.
The required SGFPT operating conditions to sepport each SGFP were evaluated. De rise in inlet pressure due to uprate is within the design limits of the LP admission inlet stop valve steam chest and nozzle chamber so no new operating limit has been imposed. Although SGFPT mass flow rate increases for the uprated full power condition, changes ia inlet steam conditions result in an increase in available flow passing capability of the LP admission valve. This is results in an increase in flow margin for the LP admission valve. Gland leakage does not increase since the SGFP turbine internal pressures remain basically unchanged. The SGFPT speed control system will operate well within its normal control range. The slightly higher inlet pressure does increase governor valve load and increase system " gain"; however, the increase is small, and no control system recalibration is anticipated. The turbine lube oil and control oil systems have no significant increase in heat load.
Notable changes in cooling water demand are not anticipated.
For the 50% load rejection case at uprated conditions, the steam supply is automatically switched to main steam. This is consistent with current operations; therefore, there is no impact on LP steam admission valves.
BOP UPRATE LICENSING REPORT 14 FNP-UNITS 1 AND 2 l
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2.5.3 Summary of Conclusions All MSR nonles except the drain outlet and the cycle steam outlet (hot reheat) are below HEI limits.
I The drain outlet nonle' flow exceeded HEI guidelines, but the increase is considered insignificant
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andjudged to be adequate. The cycle steam outlet nonle velocity exceeded HEI guidelines, but the j
velocity is essentially unchanged from the current velocity. Therefore, uprate does not cause i
additional erosion or performance degradation, and it is concluded that the cycle steam outlet nonles
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are acceptable for uprate conditions.
There is no calculable loss in margin for MSR tube vibration and no observable increase in tube wear due to the increased heating steam flow. The increase in cycle steam flow over the original maximum calculated performance does not significantly affect MSR tube bundle performance.
Shellside and tubeside uprate operating pressures and pressure drop are below the design pressure rating and thus acceptable.
The GSS system requires no changes to the system due to operation at uprate conditions. Individual valve positions may be adjusted slightly for operation at uprated conditions.
SGFPT inlet design pressure, governor valve position, and flow passing capability of the SGFPTs were each evaluated and found to be acceptable for uprate operation.- Power and speed requirements are within the maximum rated capability of the turbines. There is no impact on the lube oil, overspeed protection system, or on cooling water demand. Therefore, the SGFPT can be operated at the uprated conditions without restriction.
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i BOP UPRATE LICENSING REPORT.
15 FNP-UNITS 1 AND 2
2.6 Main Generator & Auxiliaries 2.6.1 Scope of Review The performance of the following equipment and auxiliary systems was reviewed for impact due to the uprate conditions.
- Generator stator and rotor (generator capability curve, unit steady-state and transient stability, end winding vibration, and phase imbalance)
- Exciter and voltage regulator
- Hydrogen cooling system including gas pressure, dew point, temperature, etc.
- Seal oil system
- Protective relaying 2.6.2 Summary of Evaluation Generator cooling hydrogen pressure will be increased for uprate, thus the only significant effect on the main generator is a reduction in the under-excited capability by 21 MVAR. This slight reduction has little affect on power system operations (see section 2.20). The existing imbalance in the phase currents at the generator is less than 3%. Uprate is not expected to cause an increase in this imbalance. Uprating does not affect the calculated number of speed cycles before tooth top cracking is initiated. There will be some increase in potential for remedial actions due to winding wear and tear which are expected to occur even without uprate. The fiber optics vibration sensing system alarms do not require new setpoints due to uprate.
The maximum exciter cooler heat duty increases; however, the cooler design has sufficient margin and can easily accommodate this increase. The exciter voltage is acceptable.
Uprating does not increase the maximum heat load of the hydrogen cooler beyond its design limits.
The increased load due to uprate has some minimal effect on the difference between RTD readings in the gas discharge path (hot gas AT). All readings should increase incrementally, but some may increase at different rates than others. Increased hydrogen pressure causes an incremental increase in windage losses. Since the effect of power uprate is an increased heat load on the hydrogen coolers, uprate does not exacerbate dewpoint temperature. Performance of the gas dryer (s)is not affected, and if the gas temperature changes at all, it will increase, rather than decrease.
The hydrogen-side and air-side seal oil coolers are load independent and will see no significant effects from uprating. An increase in hydrogen pressure will require a corresponding equal increase in seal oil pressure on both the hydrogen-side and the air-side systems. Both the pumps and the i
regulators are designed to operate in this range.
All protective relaying is based on the maximum capability of the generator. No changes are required due to uprating.
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O; BOP UPRATE LICENSING REPORT 16 FNP - UNITS 1 AND 2 l
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2.63 Summary of Conclusions The hydrogen pressure should be increased from 72 psig maximum to 75 psig maximum, to increase existing '-MVAR capability; otherwise, the MVAR capability would be reduced, though the FNP Unit I and Unit 2 steady-state and transient stability are not significantly affected.
The existing hydrogea cooling system, exciter cooling system, and seal oil system, including the service water supplies to them, are adequate for uprate.
The generator protective relaying is unaffected.
in general, uprating will not affect the expected generator life. Some generator maintenance is expected during its lifetime regardless ofloading.
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O BOP UPRATE LICENSING REPORT 17 FNP-UNITS 1 AND 2
4 2.7 Main Steam 2.7.1 Scope of Review This section describes the evaluation of the Farley Units 1 and 2 main steam system (MSS) for power uprate. The portion of the main steam piping from the outlet of the steam generator nozzle to the high pressure turbine and the steam dump system is addressed. The auxiliary steam generator (i.e., boiler) is not used, and the opposite unit is relied upon to supply auxiliary steam until main steam is suf6cient to handle the required steam demand.
Changes to plant operating conditions and unit-to-unit variations have been considered as required for uprate. Main steam piping stresses are addressed in section 2.19, and environmental quali6 cation of safety-related electrical equipment is addressed in section 2.15 of this report. The review of the MSS includes the following.
- Main steam safety valves (MSSVs)
- MSIV bypass piping / valves
- Main steam atmospheric relief valves (ARVs)
- Main steam drains
- Main steam isolation valves (MSIVs)
- Auxiliary steam systems
- Turbine bypass (steam dump valves) 2.7.2 Summary of Evaluation The evaluation is based primarily on changea to the following key parameters.
6 6
Steam flow:
increasing from 3.87 x 10 to 4.09 x 10 lb/hr/SG Moisture carryover:
unchanged at 0.25%
Steam temperature:
decreasing approximately 4
- F Steam pressure:
decreasing approximately 25 psi The MSS safety valves have an individual mexi: rum and a total minimum design capacity which continue to meet the criteria of sections 4.2 and 6.2.13 of the NSSS Licensing Report. Acceptable MSSV performance is demonstrated by the results of the limiting FSAR Chapter 15 events presented in section 6.2 of the NSSS Licensing Report. MSSV setpoints are not changing for uprate. Current tolerances are acceptable for uprate since ambient conditions in the main steam valve room are not expected to change signincantly (refer to section 2.18) and steamline temperature and pressure are decreasing. Accumulation and blowdown for the MSSVs are supported by current test data (reference 7.1) and do not change due to uprate.
An ARV capacity of 610,000 lb/hr at a saturation pressure of 1085 psig was veri 6ed by the valve manufacturer for use in the RCS cooldown analyses for uprate presented in the NSSS Licensing Report.
The current MSSV and ARV vent stack geometry and the design maximum flow rate are adequate for uprated conditions.
O BOP UPRATE LICENSING REPORT 18 FNP - UNITS 1 AND 2
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l Main Steana Isolation Valves (MSIVs) are typically designed to operate at steam flows much higher l
~ than current conditions with no detriment to valve function. Confirmation of valve capability at increased steam flow required for uprated conditions was obtained from the valve manufacturer.
2
- Steam pressure, temperature, and density are all decreasing for uprate. It is therefore concluded that the MSIVs remain capable of performing as designed under normal and accident conditions.
i Uprated conditions do not require additional steam flow through the bypass line for steam line l
warming or pressure equalization. Therefore, the bypass valve capacity is sufTicient.
f Steam dump capacity calculations were revised for uprate conditions, and results were provided to Westinghouse for analysis. Acceptable results were obtained for the NSSS control system transient analyses that model the steau dump system as described in section 4.3 of the NSSS Licensing Report. Since steamline pres;ure is decreasing for uprate, valve opening stroke time is not j
adversely affected. Therefore, it is concluded that steam dump system is not adversely impacted due 1
to uprate.
L f
Main steam drain capacity was evaluated for the increase in main steam flow, and each auxiliary steam subsystem was evaluated for changes in operation due to, or required for, uprate. The j
i _
auxiliary steam subsystems require a small percent of overall steam flow and the steam demand for the subsystems (GSS, steam jet hogging and air ejectors, auxiliary building evaporators, and turbine
}
building HVAC) is not appreciably affected for uprate. It can therefore be concluded that the auxiliary steam subsystems will continue to perform adequately at uprate conditions.
l l
2.7.3 - Summary of Conclusions i
Changes to the key plant operating conditions (steam flow, pressure, and temperature) affecting MSS j
performance characteristics can be accommodated by the current plant hardware. Continued acceptability of these parameters is demonstrated by meeting the requirements of the applicable NSSS analyses.
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I BOP UPRATE LICENSING REPORT 19 FNP - UNITS 1 AND 2
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l 2.8 CCW System 1
2.8.1
System Description
The component cooling water (CCW) system for each unit is an independent closed-loop cooling water system. The CCWS is comprised of three pumps, three heat exchangers, one two-section surge tank, and the associated piping, valves, and instrumentation. The system is divided into two safety trains with one pump and heat exchanger dedicated to each train. Safety related loads include RHR pumps and heat exchangers, SFP heat exchangers, and charging pumps. Non-essential heat loads are supplied CCW from a common misellanmus header. CCW cooling is provided by station service water.
2.8.2 Scope of Review During normal and accident conditions, the CCWS functions as a closed cooling system transferring heat to the SWS from various NSSS components. The components cooled by CCW include the following.
- Reactor coolant pump oil coolers and thermal barriers
- RHR heat exchangers
- Letdown heat exchanger
- RHR pump seal coolers
- Excess letdown heat exchanger
- Spent fuel pool heat exchangers
- Seal water heat exchanger
- Safety injection / charging pump
- Reactor coolant drain tank heat exchanger lube and gear oil coolers
- Waste gas compressors
- Hydrogen recombiners
- Sample system coolers 2.83 Summary of Evaluation No physical changes to the CCW system or to equipment served by the CCW system are being made. The CCW system heat load calculations were revised for uprate conditions to verify that the current CCW system flows are sufficient to maintain served system and component heat removal duty and cooling water temperatures within their acceptance limits.
The RHR and SI pump cooler heat loads do not change. The spent fuel pool heat exchangers and the RHR heat exchangers see an increase in heat load proportional to the increase in decay heat for the i
uprated core thermal power. As a result of the increased heat load, the SFP temperature increases (see section 2.10) but remains within the design requirements of the SFP cooling system.
For a normal controlled cooldown, the increased RHR heat load will increase the time required to cool down the RCS, and the CCW temperature will increase. As described in the NSSS Licensing Report, the RCS cooldown rate (< 50* F/hr) and the time required to achieve cold shutdown (s 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />) with the higher CCW temperature remain within design and licensing requirements.
Although the CCW temperature increases due to the increase in decay heat, the cooling water temperature supplied to all served systems and components remains within acceptance limits.
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BOP UPRATE LICENSING REPORT 20 FNP - UNITS 1 AND 2
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h' Of the miscellaneous header heat loads, only the reactor coolant pump oil coolers may see a small I
'j increase in heat load at uprated conditions (due to additional work to circulate cooler, more dense reactor coolant). This has an insignificant impact on the total heat load for the CCW system.
'i Following a design basis accident, the CCW/RHR systems in conjunction with the containment spray and fan coolers maintain the containment peak temperature and pressure within design limits, and are capable of rapidly reducing and maintaining them at low values (see section 2.13). CCW supplied to the RHR heat exchangers meets the long term cooling requirements.
l 2.8.4 Summary of Conclusions 1
No increase in CCW system flowrates is required to handle the additional heat load. The additional heat load results in minor temperature increases on the component cooling water system for normal and post accident scenarios. The time required for RCS cooldown for single and dual train RHR system operation is evaluated as part of the NSSS Licensing Report. Based on the revised CCW l
system heat load calculations, the component cooling water system has the capacity to accommodate the additional heat load and resultant temperature increase.
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BOP UPRATE LICENSING REPORT 21 FNP - UNITS 1 AND 2
2.9 Service Water 2.9.1 System Functional Requirements The service water (SW) system removes the heat loads from Engineered Safety Feature (ESF) and non-ESF systems and components during normal and transient / accident operating conditions. A two unit normal shutdown using the SW pond as the ultimate heat sink (UHS) may be postulated.
However, this results in a less uniting heat load on the UHS and is therefore bounded by the postulated case of one unit in post-LOCA cooling with the other unit in normal shutdown /cooldown.
2.9.2 Scope of Review No changes to the equipment served by the SW system are being made due to uprate. Therefore, the removal of heat loads using the service water pond as the UHS during the cooldown following a LOCA in one unit while the second unit is undergoing a normal shutdown /cooldown is the limiting case reviewed for power uprate. The heat loads from the following systems and components were reviewed.
- Turbine / generator and auxiliary systems
- Containment coolers
- Condensate and feedwater pump oil coolers
- ESF room coolers
- CCW heat exchangers
- Diesel generators
- SW pump / motor coolers
- Air compressors and coolers
- RCP motor air coolers
- Other non safety-related loads 2.9.3 Summary of Evaluation The containment cooler heat loads on the LOCA unit decrease slightly as the containment pressure and temperature response decrease for uprate (see section 2.13). The CCW heat exchanger heat loads on both units increase due to higher RHR and spent fuel pool heat loads. The heat load from individual diesel generators is not impacted by uprate; however, the number of diesels required to be operating has decreased since the original UHS analysis due to the assignment ofone diesel as the station blackout (SBO) alternate AC source. ESF room cooler and SW pump / motor heat loads are not significantly impacted by uprate.
Turbine / generator and auxiliary system loads, RCP motor air cooler loads, and condensate and feedwater pump oil cooler loads on the non-LOCA, normal shutdown unit are expected to see small increases in heat loads. Containment cooler duty on the non-LOCA unit increases slightly (see section 2.17). Other non safety-related loads on the non-LOCA unit are not expected to change significantly.
No changes to pond or dam design, atmospheric conditions, or service water system components occur for power uprate; so seepage, leakage and evaporation (for the same pond temperature conditions) and service water piping, valves, supports and instrumentation are not impacted. As described in section 2.3, makeup flow to the circulating water system is not expected to change significantly.
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BOP UPRATE LICENSING REPORT 22 FNP-UNITS 1 AND 2
i' 2.9.4 Summary of Conclusions Significant margins were previously included in the UHS calculations, so that the total heat load on the service water system at uprate conditions is bounded by current analyses and the current UHS analyses remain valid. The temperature profiles used to evaluate CCW, diesel generator, auxiliary building ESF room cooler, and service water pump / motor performance; UHS inventory, surface area, seepage, and evaporation; and containment accident responses remain bounding. Normal service water flows are sufficient to supply additional cooling tower makeup described in section 2.3.
Therefore, the service water system is adequate for power uprate.
i BOP UPRATE LICENSING REPORT 23 FNP - UNITS 1 AND 2
2.10 Spent Fuel Pool 2.10.1 System Description The fuel storage and handling facilities consist of the spent fuel pool (SFP), spent fuel racks, SFP cooling ar.d makeup system, new fuel pit, new fuel storage racks, and the fuel handling equipment.
2.10.2 Scope of Review This review evaluates the effect of uprate on the spent fuel pool and the SFP cooling and cleanup system, including the new and spent fuel storage racks. The major areas of evaluation for uprate are the impact of the additional decay heat load and resultant temperatures on the pool and system components. The current criticality, thermal, and radiation exposure analyses for the new and spent fuel storage racks address fuel enrichments of up to 5.0 wt % U-235. Since the physical fuel design parameters that affect handling are unchanged, the existing fuel handling system is unaffected by power uprate. Therefore, a specific evaluation was not perfortned for the fuel handling system hardware.
2.10.3 Summary of Evaluation New and snent Fuel Rack Criticality i
The new and spent fuel racks were previously evaic:':d for enrichments up to 5 wt % and lead assembly burnup of 60,000 mwd /MTU to support implementation of VANTAGE 5 fuel at FNP. As documented in the NRC SER for Technical Specifications Amendments 91 and 84 (reference 10.1),
new fuel racks maintain a K,g s 0.98 and the spent fuel racks maintain a keg s 0.95, when subjected to normal and transient (water or water fog) conditions.
SFP Decay Heat Loads /SFP Cooling The decay heat load on the SFP cooling system is based on actual fuel discharge for cycles prior to 1997, plus a fuel discharge schedule of 79 assemblies (='/ core) per 18 month cycle with a total 2
assembly burnup of 60,000 MWD /MTU for cycles after 1996. The spent fuel decay heat loads were determined for the following cases using the methodology of BTP ASB 9-2 (with any exceptions noted).
1.
Partial-core offload refueling case - 150 hrs decay time
- 2. Beginning of cycle (BOC) full-core ofiload refueling case - 150 hrs decay time
- 3. End of cycle (EOC) full core offload refueling case - 150 hrs decay time 4.
"Best estimate" full-core offload case -150 hrs decay time
{
5.
Post-refueling spent fuel pool heat load case - 25,40 and 65 days decay time Cases 1 and 2 are equivalent to the " normal maximum heat load" and " abnormal maximum heat load" scenarios described in SRP Section 9.1.3. Case 3 was developed to evaluate the heat load on the spent fuel pool cooling systems for a full core offload at the end of an 18 month operating cycle.
Case 4 evaluates a "best estimate" full core offload refueling case with no uncerainty facter applied BOP UPRATE LICENSING REPORT 24 FNP - UNITS 1 AND 2
t 3
j to the decay heat results. Case 5 estimates the maximum spent fuel pool heat load after an outage for use in evaluating the component cooling water system and ultimate heat sink.
i Accentability of Soent Fuel Pool Bulk Water Temneratures I
~ For the partial-core offload case, the pool bulk water temperature is maintained sl30*F for two train i
operation and sl50*F assuming a failure of one SFP cooling train. For the full-core offload cases, pool bulk water temperature remains s 140'F with both cooling trains in operation and s180'F with i
one SFP cooling train in operation. For the "best estimate" full-core omond case with no uncertainty factors applied, the pool bulk water temperatures are bounded by the BOC and EOC full core offload
)
cases due to the lower total heat load on the spent fuel pool.
a i
The results for the partial core offload case with one (1) cooling train in operation (pool water i
3 temperature sl50*F) is within the maximum normal operating temperature limit for the pool, rack, j
and cooling equipment, but above the demineralizer resins maximum operating temperature liinit af 140*F. A temperature switch, which activates a high temperature alarm whenever the pool
{
temperature reaches 130 F, can be used to alert the operators to manually initiate action to maintain j
the integrity of the cleanup system.
s For the full-core offload cases, pool bulk water temperature remains below 140 F with both cooling i
trains in operation, and below 180 F with one SFP cooling train in operation. The spent fuel pool bulk water temperature with two trains operating remains below the maximum normal operating temperature limit of 150 F for the pool, r6ck, and cooling equipment, and below the maximum operating temperature limit of 140 F for the demineralizer resins. For single train operation, the N
pool temperature remains below the design maximum temperature for SFP cooling pump ' peration o
i at rated flow. Thus no changes to the SFP cooling or cleanup systems are required to support uprate.
i
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SFP Makeun/ Time Available Before Bulk Pool Boiline on Imss of SFP Coolino Although a complete loss of cooling is a beyond-design-basis-event (since the cooling system is seismic Category I and meets the single failure criteria), the time available from the high temperature alarm before bulk pool boiling occurs exceeds 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. For the maximum decay heat load case (BOC full core offload), a maximum makeup rate of 76.4 gal / min was calculated to be required to maintain pool level if boiling occurs. Makeup can be provided directly from the RWST by the refueling water purification pump, by the demineralized water system, or by the reactor makeup water hose station located on the operating area adjacent to the pool, which requires only a minimal hookup time and can be supplied by either one, or both reactor makeup water pumps.
l Fuel Auembiv Clad Temneratures With Loss of SFP Coolino For assembly exposure of 60,000 MWD /MTU with a peaking factor of 1.7, the maximum fuel assembly clad temperature at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown was calculated to be 257.I'F. This temperature is well below a temperature at which damage to the fuel cladding occurs. The maximum heat flux in the fuel assembly was determined to be approximately 1600 BTU /hr-ff, which is considerably less than the critical heat flux (onset of film boiling) for these conditions.
BOP UPRATE LICENSING REPORT 25 FNP - UNITS 1 AND 2
SFP Cleanun The amount of fission products released from the fuel does not increase appreciably for uprate, and the chemical and radionuclide composition of the SFP water also do not change appreciably due to the uprate. Therefore, the existing SFP cleanup system is adequate for maintaining spent fuel pool water purity and clarity for the uprate conditions.
2.10.4 Summary of Conclusions No changes or equipment additions are necessary for the spent fuel pool, the SFP cooling and cleanup system, or to the SFP makeup systems to support the uprate. The existing structures, systems, and components can safely handle the increase in decay heat generation from the uprate conditions and resulting maximum pool temperatures.
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BOP UPRATE LICENSING REPORT 26 FNP - UNITS 1 AND 2
. _. _, __ _ _ _ _ _ _ _ _ - _ _ _ _ _. ~. _ _ _ _ _ _ _.... _ _. _ _ _ _ _
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2.11 BOP Evaluations of LHSI (RHR) / HHS (Charging)
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2.11.1 Scope of Review The major components (pumps, heat exchangers, large bore valves) of the safety injection systems s
are evaluated in the NSSS Licensing Report. The remaining functions and equipment of the low head (RHR) and high head (Charging Pumps) safety injection systems reviewed include the following.
l
- RWST drain down
- Piping and supports
)
- RHR pump seal cooling
- Instrumentation and valves l
7 2.11.2 Summary of Evaluation The minimum and maximum engineered safeguards flows for RHR, containment spray, and charging pumps were used to calculate the time for draindown of the RWST until initiation of manual j
switchover from RCS cold leg injection from the RWST to RCS cold leg recirculation from the containment sump. Based on the minimum RWST volume (specified by the Technical i
i Specifications), the RWST low level alarm setpoint will be reached in approximately 22 minutes.
This time was then used in revised LOCA analyses as discussed in the NSSS Licensing Report.
t' Based on the post accident containment sump temperatures and pressures calculated for uprated l (/'
conditions, the existing piping and support stress calculations remain bounding. Current design i
conditions for RHR pump seal coolers bound the uprated service conditions calculated for post accident containment recirculation sump; therefore, seal cooler performance is not degraded at l
uprated conditions.
l The instrumentation which automatically initiates LHSI/HHSI is considered a part of the ESFAS and i
is addressed in the NSSS Licensing Report. The remaining instrumentation in these systems is unaffected since it is designed to operate at the process line design service conditions which have not i
changed for uprate.
1 i
j Since the current LHSI and HHSI system valve design temperatures and pressures exceed the temperatures and pressures calculated during recirculation for uprated conditions, valve performance is not degraded by power uprate.
1 i
2.11.3 Summary of Conclusions Current design parameters for the low head and high head safety injection systems bound uprate j
conditions. Therefore, the individual components remain capable of performing their functions without being adversely impacted by uprated condition.~
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i k-BOP UPRATE LICENSING REPORT 27 FNP - UNITS 1 AND 2
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7
i 2.12 Auxiliary Feedwater Systern (AFW) 2.12.1 System Description The AFW system is designed to supply high pressure feedwater to the steam generators for removal of residual heat from the core during normal and emergency conditions. The major mechanical j
components supporting this action include two motor-driven AFW pumps and one turbine-driven AFW (TDAFW) pump, control valves, flow orifices, motor-operated isolation valves, and the condensate storage tank.
2.12.2 Scope of Review Uprate does not impact the abili3 of the AFW system to perform normal low power startup operations. Therefore this sec< vu addresses the impact of uprate on the ability of the AFW system i
to respond to normal cooldown and emergency conditions and the adequacy of condensate storage tank (CST) inventory for uprated power conditions. The TDAFW pump turbine design interface with the main steam supply was also reviewed for uprate changes to steam pressure.
2.12.3 Design Requirements The design requirements for AFW flow are shown on Table 2.12-1 for the following events.
- Main Feedwater Line Break
- Loss of Normal Feedwater
- Small Break LOCA
- Main Steam Line Break
- Steam Generator Tube Rupture (SGTR)
- Station Blackout (SBO) 1
- Accidental Depressurization of SG by inadvertent opening of one ARV or MSSV
\\
w 2.12.4 Methbd of Evaluation A review of the AFW design bases f%w requirements and uprate emergency condition requirements has been perfonned. These requirements have been evaluated to verify the AFW system performance during the limiting emergency conditions shown on Table 2.12-1.
A review of the TDAFW pump turbine design interface based on the potential for a lower main steam header supply pressure to the TDAFW pump turbine was performed.
l BOP UPRATE LICENSING REPORT 28 FNP - UNIT; i AND 2 1
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The CST volume necessary to meet the design requirements and Technical Specification Bases was compared to the volume within the protected portion on the tank.
2.12.5 Summary of Evaluation Changes in the AFW flow requirements for main steamline break (MSLB) and small break loss of coolant accident (SBLOCA) have been identified.
The MSLB has been evaluated for both inside and outside containment. Both minimum and maximum flow rate requirements shown in Table 12.2-1 are fulfilled by the available AFW pumps for each case examined, including the effects of single active failure and LOSP as applicable.
The minimum SBLOCA flow requirement as shown in Table 12.2-1 is satisfied assuming a single t.ctive failure and LOSP.
The AFW pump turbine operating parameters remain unchanged. A decrease in the normal operation main steam system pressure to 787 psig as a result of uprate does not impact the TDAFW pump operation. The steam to the turbine is throttled to less than 450 psig at rated steam flow conditions.
To meet the design requirements, the CST must contain 131,000 gallons of water. To meet the
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Technical Specification Bases, 115,100 gallons of water is needed. Both conditions are bounded by the current protected volume (150,000 gallons) of the CST.
2.12.7 SummaryofConclusions As summarized on Table 2.12-1, the results of the AFW evaluations indicate that the AFW design requirements are fulfilled by the minimum available AFW pumps under the limiting conditions. The missile protected portion of the CST contains adequate water to meet design and Technical Specification Bases requirements.
rs m
BOP UPkATE LICENSING REPORT 29 FNP - UNITS 1 AND 2
TABLE 2.12-1 (Sheet 1 of 2)
AFW SYSTEM FLOW PARAMETERS DESIGN BASIS DESIGN BASIS REQUIREMENTS SG CONDITION RESULTS EVENT DESCRIPTION AFW FLOW / SG PRESSURE LOSS OF NORMAL 2350 GPM TOTAL TO 2 SGs (175 SG 2C ISOLATED ACCEPTABLE FEEDWATER GPM EACH) AT 1137.6 PSIA 2 MDPs (LONF)/LOSP OPERATING
- 1. 5 836 GPM TO FAULTED SG SG 28 FAULTED ACCEPTABLE MAIN STEAM LINE FAULTED SG @ 14 7 PSIA BREAK (MSLB) INSIDE INTACT SGs @ 820 PSIA ALL THREE CONTAINMENT P
- 2. 5 2100 GPM TOTAL TO ALL SGs SG 1B FAULTED OPE TING AT FAULTED SG @ 14.7 PSIA MAXIMUM INTACT SGs @ 300 PSIA PERFORMANCE i
- 3. 5 011 GPM TO FAULTED SG SG 2B FAULTED ALL SGs FAULTED SG @ 80 PSIA EXAMINED INTACT SGs @ 690 PSIA (WORST CASE REPORTED)
^
S INE SG A FAULTED ACCEPTABLE FA LTED SG @ 0C SA A SGs @ 1138 PSIA 2Ws OUTSIDE OPERATING CONTAINMENT ALL PUMPS
- 2. 5 750 GPM TO FAULTED SG SG B FAULTED FAULTED SG @ 325 PSIA OPERATING AT INTACT SGs @ 725 PSIA MAX. PERF.
FEED LINE BREAK 2150 GPM TOTAL TO INTACT SGs SG 2B FAULTED ACCEPTABLE (FLB) W/O ISOLATION AT 1130 PSIA TWO MDPs OPERATING ACCEPTABLE ONE MDP AND TDP OPERATING FLB/MSLB (AFTER 2350 GPM TOTAL TO INTACT SGs SG 2C ISOLATED ACCEPTABLE ISOLATION)
AT 1155 PSIA 2 MDPs OPERATING ACCIDENTAL DEPRZ.
MAX TOTAL FLOW TO ALL SGs N/A ACCEPTABLE OF A SG BY s 2200 GPM. MIN FAULT SG BOUNDED BY INADVERTENT PRESSURE = 247.4 PSIA THE MSLB CASE OPENING OF ARV OR MSSV SMALL BREAK LOCA 2681 GPM TOTAL TO ALL SGs (227 ALL SGs @ 1144 ACCEPTABLE (SBLOCA)
GPM EACH ) AT 1144 PSIA PSIA 1 MDP AND TDP OPERATING O
BOP UPRATE LICENSING REPORT 30 FNP - UNITS I AND 2
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i TABLE 2.12-1 (Sheet 2 of 2) 4 AFW SYSTEM FLOW PARAMETERS
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I DESIGN BASIS DESIGN BASIS REQUIREMENTS SG CONDITION RESULTS 1
EVENT DESCRIPTION AFW FLOW / SG PRESSURE l
l I
STEAM GENERATOR 2450 GPM TOTAL TO ALL SGs (150 ALL SGs @ 1100 ACCEPTABLE l
TUBE RUPTURE GPM EACH) AT1100 PSIA PSIA (SGTR) 2 MDPS OPEMTING j
l
/
HELB (STEAM SUPPLY 2285 GPM TOTAL TO ALL SGs AT SGs AT 1137.6 ACCEPTABLE j
TO TDP) 1137.6 PSIA PSIA ONE MDP j
OPERATING I
l NORMAL PLANT 2350 GPM TOTAL TO 3 SGs AT SGs AT 1020 ACCEPTABLE l
COOLDOWN 3020 PSIA PSIA ONE MDP i
OPERATING i
I STATION BLACK-OUT 2350 GPM TOTAL TO ALL SGs AT CJs AT 1137.6 TDP OP T G (SBO) 1137.6 PSIA PSIA I
i BOP UPRATE LICENSING REPORT 31 FNP-UNITS I AND 2
1 2.13 Containment & Subcompartment Analysis 2.13.1 Containment Structure-Analysis 2.13.1.1 Analysis Description Each Farley containment building is designed to withstand an internal pressure of-3 psig to +54 psig and a temperature of 280 F. Containment analysis has been performed to demonstrate the acceptability of the containment structure and subcompartment design. The power uprate containment analyses were performed using the EPRI GOTillC code. Additionally, the containment analysis is used to generate design pressure and temperature curves for use in the evaluation of safety-related electrical equipment inside the containment in accordance with the provisions of 10 CFR 50.49.
The analyses performed for power uprate have made use of previous work to the maximum extent possible. Models developed for the current Farley design basis analyses have been utilized in the uprate containment analyses. FSAR Tables 6.2-1, " Principal Containment Design Parameters,"
and 6.2-2, "lleat Sink Geometric Data," remain unchanged in the uprate analysis. Table 6.2-3,
" Initial Conditions for Pressure Analysis," remains unchanged except for the initial containment pressure and temperature which are changed as described in Sections 2.13.2 and 2.13.5 of this report.
Table 6.2-4, "lient Sink Thermodynamic Properties," remair.s unchanged except for the containment sump to atmosphere interface. The uprate containment analyses used the GOTlilC internal model for interfacial heat transfer between the sump and containment atmosphere.
The most significant changes associated with power uprate are those associated with changes in the mass and energy blowdown from pipe breaks. Blowdown mass and energy values have changed because of the change in reactor power and also because of refinements in blowdown computational methods as described in the NSSS Licensing Report.
Bounding analysis results are presented in the form of summary tables and graphs later in this section.
In addition to the containment analysis, the main steamline break in the main steam valve room and other hig'n energy line breaks (HELBs) were also assessed for uprate. The results of these assessments are included in this section.
2.13.2 Scepe of Review This review consists of four principal parts.
1.
The first part evaluates the containment response to design basis LOCA events. It has been determined that the hot leg break results in the most limiting pressure during the blowdown phase. It has further been determined that the pump suction break yields the highest energy flow rates during the post-blowdown period. Therefore, the containment pressure and temperat"re transient is analyzed for only the blowdown period for the hot leg break and is analyzed for the entire event for the pump suction break. Since the Technical Specifications BOP UPRATE LICENSING REPORT 32 FNP - UNITS 1 AND 2
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j permit operation at +3 psig initial containment pressure, the LOCA analyses were performed at this pressure as well as at the nonnal containment pressure of 0 psig.
2.
The second part evaluates the containment response to the various steam line breaks.' Since the
]
l Technical Specifications permit operation at +3 psig initial containment pressure, the MSLBs i
i which are limiting for pressure (cases 9 and 12) were also run with an initial condition of +3 psig. It was shown that decirasing the containment initial pressure resulted in a higher i
temperature in MSLB cases. Since the Technical Specifications permit operation at a i
containment pressure of-1.5 psig, the MSLBs which are limiting for temperature (cases 1,3 and i 1) were analyzed at this pressure.
i Sixteen MSLB cases plus initial pressure variants were analyzed.
f 2
Case 1:
102% Power,1.069 8 double ended rupture with i'
containment initial pressure = 0 psig and -1.5 psig.
2 i
Case 2:
102% Power,0.700 ft double ended rupture l
Case 3:
102% Power,0.600 ft double ended rupture with 2
containment initial pressure = 0 psig and -1.5 psig.
2 4
Case 4:
102% Power,0.528 A split break 2
l Case 5:
70% Power,1.069 A double ended rupture 2
Case 6:
70% Power,0.600 A double ended rupture 2
i Case 7:
70% Power,0.500 A double ended rupture 2
I Case 8:
70% Power,0.561 A split break j\\
Case 9:
30% Power,1.069 A double ended rupture 2
i containment initial pressure = 0 psig and +3 psig.
2 i
Case 10:
30% Power,0.500 ft double ended rupture 2
Case 11:
30% Power,0.400 A double ended supture containment initial pressure = 0 psig and -1.5 psig.
l Case 12:
30% Power,0.591 A split break with 2
j containment initial pressure = 0 psig and +3 psig.
2 Case 13:
0% Power,1.069 A doubleendedrupture 2
Case 14:
0% Power,0.200 ft double ended rupture 2
i Case 15:
0% Power,0.100 ft double ended rupture 2
l Case 16:
0% Power,0.300 A double ended rupture i
l 3.
The third part evaluates the impact upon containment subcompartments for breaks of the
{
pressurizer spray and surge lines.
4.
The forth part consists of an assessment of the applicability of prior analysis of the main steamline break in the main steam valve room and other high energy line breaks to uprate
]
conditions.
BOP UPRATE LICENSING REPORT 33 FNP - UNITS 1 AND 2
2.13.3 Design Interfaces Containment cooler duty curves and containment spray system performance are interfaces because the cooler and spray system performance are inputs to the containment pressure / temperature (P/F) analysis.
2.13.4 Assumptions 1.
Containment geometry is as described in FSAR Table 6.2-1. Containment structural heat sinks remain the same as those in the FSAR Table 6.2-2 with heat transfer properties as described in FSAR Table 6.2-4. Initial conditions remain consistent with FSAR Table 6.2-3 except as noted in this report. Heat transfer coeflicients used in the FSAR were retained for this analysis; specifically, the Tagami correlation was used for condensing heat transfer for the LOCA, and Uchida was used for condensing heat transfer for the MSLBs. Heat conduction models are essentially the same as those used in the previous analysis except that a more detailed nodalization was used for the heat conductors near the conductor surfaces.
2.
There is little difTerence in the FSAR model and that used in the uprate analysis. The principal difference in the two analyses is the change in the blowdown data (see the NSSS Licensing Report). Other changes were:
(a) RHR heat exchange properties (heat transfer coefficient and CCW flow) were modified to represent actual plant design data.
(b) Containment cooler performance represents highly degraded cooler service conditions.
The model uses a single (degraded) cooler in service.
3.
Initial containment temperature is assumed to be 127 F which corresponds to the containment design operating bulk average temperature plus margin.
4.
Initial containment pressure is assumed to be 14.7 psia for LOCA cases. Atmospheric pressure is used because the containment mini-purge fans are normally run maintaining pressure at essentially atmospheric. Additionally, the LOCA cases are evaluated at the Technical Specifications maximum containment pressure LCO value of +3 psig.
5.
Initial containment pressure is assumed to be 14.7 psia for all MSLB cases. Upon determining the limiting pressure cases, the limiting MSLB cases were analyzed at the Technical Specifications limit of +3 psig initial pressure. Analysis oflimiting pressure cases verifies that the containment integrity is not challenged by these cases. The MSLB is the accident which produces the maximum temperature transient. Cases which produce the highest peak containment temperature are the limiting cases for safety-related electrical equipment qualification. Since the results demonstrated that the temperature transient is reduced when the initial pressure is increased, the limiting temperature MSLB cases were run at the Technical Specaications minimum containment pressure of-1.5 psig. Analysis oflimiting temperature cases provides bounding temperatures for safety-related electrical equipment qualification.
BOP UPRATE LICENSING REPORT 34 FNP-UNITS 1 AND 2 N
i A(v) 6.
Minimum safeguards performance is assumed for all cases because the minimum safeguards j
cases have proven to be limiting in past analyses. Minimum safeguards consist of one containment cooler and one train of containment spray, safety injection, residual heat removal, component cooling water, and service water.
7.
Blowdown, reflood and post reflood mass and energy releases described in Section 6.4 of the NSSS Licensing Report are modeled.
2.13.6 Methods of Evaluation 2.13.6.1 LOCA and MSLB The LOCA and MSLB analyses were performed using the GOTHIC computer code. GOTHIC was 1
developed under EPRI contract from the older NRC code, FATHOMS. GOTHIC was developed under a fully qualified quality assurance program (reference 13.1) and has undergone extensive peer review.
GOTHIC has been validated for safety-related applications at Southem Company Services.
2.13.6.2 Subcompartment Analysis Subcompartment analysis was performed for the pressurizer compartment. The analysis was also performed using the GOTHIC ccmputer code. Results of the analysis showed that the uprate blowdown pressure and temperaf are remain bounded by the previous design basis analysis.
m )
As described in Section 6.4.2 4 of the NSSS Licensing Report, RCS design pressure and temperature condition changes have been evaluated to demonstrate that " current licensing basis subcompartment analyses that consider breaks in the primary loop reactor coolant system piping... remain bounding."
i 2.13.6.3 Main Steam Valve Room Reference 13.2 previously assessed a steam line break in the main steam valve room. Blowdown data for the main steam valve room under uprate conditions were compared with the blowdown data used in the previous analysis and were found to be bounded by the prior analysis at all times.
2.13.6.4 Other High Energy Line Breaks FSAR Table 3K.F-1 lists the temperature and pressure used in previous design basis analysis of other high energy line breaks. Those temperatures and pressures are compared with the uprate operating conditions in the table below. As shown below, prior design basis analyses remain bounding for all of these events.
V BOP UPRATE LICENSING REPORT 35 FNP - UNITS 1 AND 2
O Uprate Line Uprate Press.
Press.
Size Temp.
Temp.
(Po)
(Po)
System (in)
(F)
(F)
(psig)
(psig)
Main steam 32 547 518 1005 798 Bounded by previous data.
Main steam 36 547 518 1005 798 Bounded by previous data.
Main feedwater 14 442 437.3 1055 1007*
Bounded by previous data.
Auxiliary steam 3
547 518 1005 798 Same as MS.
Auxiliary steam 4
547 518 1005 798 Same as MS.
442 437.3 1055 1007*
Same as FW.
442 437.3 1055 1007*
Same as FW.
Auxiliary feedwater 10 442 437.3 1055 1007*
San.e as FW.
CVCS and BTRS 3
380 380 550 550 System not affected by uprate.
547 518 1055 1007*
Temp. same as MS blowdown Press. same as FW.
2.13.7 Summary of Evaluation l
The evaluation of the analysis results demonstrated that the containment design limits for LOCA and MSLB continue to bound the power uprate accident conditions. Results are shown on Table 2.13-1 and Figures 2.13-1 through 2.13-5. Figures 2.13-1 through 2.13-3 illustrate the containment temperature response for various initial pressures. From Figure 2.13-1, it is observed that the temperature is essentially identical for the double ended pump suction guillotine (DEPSG) and hot leg breaks for any initial pressure.
Figure 2.13-2 illustrates the temperature impact of an increase in initial containment pressure from 0 psig to +3 psig. Comparison of the curves in Figure 2.13-2 to the corresponding curves in Figure 2.13 1 illustrates that an increase in containment initial pressure produces a decrease in peak transient temperature. The LOCA case resultant temperature is not as sensitive to initial pressure as are the MSLB cases.
Figure 2.13-3 illustrates the temperature increase due to a decrease in initial containment pressure from 0 psig to -1.5 psig.
Figures 2.13-4 and 2.13-5 illustrate the containment pressure for limiting cases. The limiting cases were run at the +3 psig initial pressure, and the results are illustrated on Figure 2.13-5. Since lowering BOP UPRATE LICENSING REPORT 36 FNP-UNITS 1 AND 2
k i
j the initial containment pressure clearly decreases the peak accident pressure, the -1.5 psig cases were l
not illustrated.
It was demonstrated that the pressurizer subcompartment pressure is bounded by the prior analysis.
i Results are shown in Table 2.13-2.
?
I Further, it was demonstrated that the blowdown data for an MSLB outside containment at uprated conditions is bounded by the data in reference 13.2; therefore, the reference 13.2 analysis is a
considered bounding for uprate.
Comparison of system conditions used in analysis ofline breaks outside containment with the uprate system conditions demonstrated that the previous design basis analyses remain bounding for these events.
2.13.8 Summary of Conclusions The containment analyses demonstrated that the containment pressure and temperature transient results are bounded by the previous analyses for the LOCA and MSLB cases initiated at 0 psig.
MSLB cases which start at +3 psig yield pressures which exceed the previous analyses values but remain below the containment design pressure. LOCA cases initiated at +3 psig produced peak containment pressures below the previous analyses values. LOCA and MSLB cases which initiate at
-1.5 psig produce higher peak temperatures than the 0 psig cases, but remain bounded by previous design basis analyses.
Subcompartment analysis demonstrated that the subcompartment wall differential pressure and uplift iorces are also bounded by the previous design basis analyses; therefore, containment subcompartments are not impacted by uprate conditions. The blowdown data for an MSLB outside containment under uprate conditions are also bounded, so that previous design basis main steam valve room conditions are bounding for uprate.
i BOP UPRATE LICENSING REPORT 37 FNP-UNITS 1 AND 2 s
Table 2.13-1 Containment High Energy Line Break Results Peak Peak HELB Case Pressure Temperature (psia)
(F)
LOCA - Pump Suction 52.7 260 LOCA - Pump Suct,3 psig 56.2 261 LOCA - Hot Leg 54.4 263 LOCA - Hot Leg,3 psig 57.7 263 MSLB case 1 58.3 368 MSLB case 1. -1.5 psig 56.6 383 MSLB case 2 55.7 355 MSLB case 3 55.4 362 MSLB case 3. -1.5 psig 53.7 370 MSLB case 4 59.0 365 MSLB case 5 59.9 324 MSLB case 6 57.3 331 MSLB case 7 56.7 354 MSLB case 8 61.6 363 MSLB case 9 63.0 294 MSLB case 9,3 psig 67.0 288 MSLB case 10 58.6 313 MSLB case i1 57.3 342 MSLB case i1,-1.5 psig 56.0 347 MSLB case 12 63.3 359 MSLB case 12,3 psig 67.1 347 MSLB case 13 61.0 273 MSLB case 14 43.1 262 MSLB case 15 33.9 302 MSLB case 16 45.3 324 Pre-Uprate LOCA 62.7 302 Pre-Uprate MSLB (case 12) 63.0 378 O
BOP UPRATE LICENSING REPORT 38 FNP - UNITS 1 AND 2
i 4
i i -
4 LO e
- l Table 2.13-2 s
Pressurizer Subcompartment Results 1
P;essurizer Subcompartment Maximum Maximum j
Cases Differential UpliR
- j Pressure Force j.
(psid)
(Ib)
Uprate Spray Line Break 9.4 N/A j
Design Value 20.7 N/A Uprate Surge Line Break N/A 4.5x10" I
Design Value N/A 6.5x10' i
i i
l 4,
t d
a 1
1 i
P BOP UPRATE LICENSING REPORT 39 FNP - UNITS 1 AND 2
4 O
Figure 2.13-1 Temperature Limiting Cases InitialL 2:1;..at Pressure = 14.7 psia i
f ItziptsidLae(F) i j
l'!
11 l
l
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od; Hi i
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=~
o 1m l!I i h'i i hi b-N llIl!
~T-J
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m
$i l
M h
b 1E-2 1E-1 180 191 192 193 184 185 186 Tnn(sec)
EPS3 FOTLIG M1B1 M1B3 MLB11
_4_
_ g _ _, _ _ 4_.
"O BOP UPRATE LICENSING REPORT 40 FNP - UNITS I AND 2
O Figure 2.13-2 Temperature Limiting Cases Initial Containment Pressure - 17.7 psia (b
ll _hl
!f l
!lll!
4 II I
40 s
i j
4 "I
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ll 14 l
llli u
n
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a
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t l'l"ll I"!s
!g
! n" Q!
in 1 :
h 0
152 151 1E+0 1E+1 1E+2 193 1E44 1E+5 1E+6 Tne(sec)
EFEG FOTLB3 IVElB1 MlB11 NELB3
__4_
_ g _._4._
__ e __
0 DOP UPRATE LICENSING REPORT 41 FNP-UNITS 1 AND 2
n U
Figure 2.13-3 Temperature Limiting Cases initial Containment Pressure = 13.2 psia IW W (b
j i
C
'~
i k'
h i, 48 I d!
h
!f I'
9
- * ; -- 4
---e
+
k 2u b!!L b.
p
~ &))d]9@
300
-!l U f'U'll W
f
~
O 200 h
.1 100 j
ly y
pll h-bp-- y$ i pI llh f
l I
I Iqu
- j iHj q
- %l0 ii i
l 0
h l~
l l -l-l l
i "i
0 152 151 1E40 1E+1 1E+2 1E+3 1E44 1E+5 1E+6 TirTe(Sec)
DEPS3 FOTIB3 NELB1 NELB11 M1B3
+
Note:
The DEPSG and hot leg breaks are included on this graph for iriformation. Since LOCA temperature is never limiting and is not significantly affected by initial containment pressure, the LOCA cases were run for initial pressures of 0 psig and +3 psig only.
j O
1 BOP UPRATE LICENSING REPORT 42 FNP-UNITS 1 AND 2 j
1 4
4
' O Figure 2.13-4 Pressure Limiting Cases Initial Containment Pressure = 14.7 psia i
f
(
70
.lif
-"j lli p
4]
.f n'
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d N
l 9
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l' 1
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o t
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li
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i ll
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e i
1EO 1E+1 1E+2 1E+3 1EM 1E+5 1E46 Tme(Sec)
CEEG lorLID MIB9 M1B12
)
-+-
O c
1 f
- 0
't BOP UPRATE LICENSING REPORT 43 FNP - UNITS 1 AND 2
)
\\
l i
O
)
I i
Figure 2.13-5 Pressure Limiting Cases i
Initial Containment Pressure = 17.7 psia
(
70 lll l i i!
4 if l
m d
l
]l klll b
b
_._ 'il lbj h
f IH J%
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[
+1'b L k 2
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l 10 1E+0 1E+1 1E+2.
1E+3 1EM 1E+5 1E+6 IN(M ERG HJrLID MIB9 MLB12
-+-
L O
l BOP UPRATE LICENSING REPORT 44 FNP - UNITS 1 AWD 2
(D V#
2.14 Post LOCA Hydrogen Generation 2.14.1 System Description i
Hydrogen is generated following a loss of coolant accident (LOCA) inside containment. liydrogen is generated due to the following reactions:
Zirconium-water reaction; Corrosion of construction materials in the containment; and, Radiolytic decomposition of core and sump solution.
In addition, hydrogen present in the reactor coolant and pressurizer vapor space is released in to the containment during a LOCA.
In order to maintain the hydrogen concentration at a safe level, the following systems are provided:
l a) Hydrogen Recombiners; b) Post accident containment venting system; c) Post accident containment mixing system; and, d) Post accident combustible gas sampling system i
2.14.2 Scope of Review
(-
(
The effect of power uprate is reviewed for the above three modes of post LOCA hydrogen production and for the capability of the combustible gas control system (containment mixing fans, recombiners, and venting system) to maintain acceptable hydrogen concentration inside the j
containment.
2.14.3 Summary of Evaluation The physical design of the fuel is not changed, and the NRC hydrogen generation model continues to be used; therefore, the hydroge generated due to the zirconium water reaction is not affected by power uprate.
The analysis, which is based on the Regulatory Guide 1.7 hydrogen yield value of 0.5 molecule per 100 electron volt (Ev) of decay energy produced post-LOCA, results in the hydrogen generated due to core and sump radiolysis being increased by approximately 5%.
The post-LOCA containment temperature profiles used to determine the corrosion rate for aluminum and zine are unchanged. The corrosion rate is conservatively used as a reducing step function to 4000 seconds after the accident and a constant rate is used thereafter. The constant rates used after 4000 seconds are 200 and 5 mils / year, for aluminum and zine respectively. The containment temperature profile discussed in section 2.13 is lower for power uprate than that used in previous design basis analyses; therefore, using the existing containment temperature profile provides conservative corrosion rates.
(]
(
BOP UPRATE LICENSING REPORT 45 FNP - UNITS 1 AND 2
i i
4 f
O1 The results of the post accident hydrogen generation analysis are provided in Tables 2.14-1 and 2.14-2 and Figures 2.14-1 and 2.14-2. In the unlikely event that both of the redundant hydrogen recombiners fail to start, containment venting will be initiated 18 days after the start of the accident at a rate of 35 SCFM. The consequent doses at the low population zone due to containment venting are provided in Table 2.14-3.
The increase in hydrogen generation rate due to power uprate is determined to have a negligible effect on the post accident hydrogen mixing system.
2.14.4 Summary of Conclusions The hydrogen generated due to Zr-water reaction is not affected by power uprate since total quantity of Zr (in the fuel cladding) remains unchanged. The hydrogen generated post-LOCA due to sump and core radiolysis increases proportional to the increase in reactor power level.
The doses due to containment venting added to dose from other post accident consequences are within the limits of 10 CFR 100, and the control room doses remain within the 10 CFR 50, Appendix A, GDC 19 limits.
Based on the above, the combustible gas control system remains capable of maintaining the post-LOCA hydrogen concentration inside containment below the lower flammability limit of 4 volume percent.
l l
O BOP UPRATE LICENSING REPORT 46 FNP - UNITS 1 AND 2
TABLE 2.14-1 i
POST ACCIDENT HYDROGEN GENERATION WITH HYDROGEN RECOMBINER IN SERVICE TIME SUMP CORE AL & ZN HYDROGEN GRAND CONTAINMENT (DAYS)
RADIOLYSIS RADIOLYSIS CORROSION RECOMBINER TOTAL
- CONCENTRATION (%
(CU.FT.)
(CU.FT.)
(CU.FT.)
(CU.FT.)
(CU. FT.)
VOL) 0 0
1 4956 2793 12044 0
36233 2.07 10 11025 15330 21191
<26987>
36999 2.1 20 14385 25410 27749
<54453>
29531 1.69 30 16485 33705 33328
<76273>
23685 1.36 40 18165 40845 38907
<95811>
18546 1.07 50 19530 47145 44487
<111155>
16447 0.946 l
60 20790 52815 50066
<124778>
15333 0.883 70 21840 58170 55645
<137487>
14608 0.841 80 22785 63105 61224
<149600>
13954 0.804 90 23730 67620 66804
<161175>
13419 0.773 100 24675 72030 72383
<172310>
13218 0.762 l
t
- GRAND TOTAL INCLUDES THE HYDROGEN GENERATED DUE TO ZR-WATER REACTION (15,410 FT*) AND HYDROGEN PRESENTIN THE RCS AND PRESSURIZER VAPOR SPACE (1030 FT*)
BOP UPRATE LICENSING REPORT 47 FNP-UNITS 1 AND 2
Table 2.14-2 POST ACCIDENT HYDROGEN GENERATION WITH CONTAINMENT VENTING SUMP RADIOLYSIS CORE AL & ZN CONTAINMENT GRAND TOTAL
- CONTAINMENT TIME (CU.FT.)
RADIOLYSIS CORROSION PURGE (CU.FT.)
(CU. FT.)
CONCENTRATION (%
(DAYS)
(CU.FT.)
(CU.FT.)
VOL) 0 0
1 4956 2793 12044 0
36233 2.07 10 11025 15330 21191 0
63986 3.58 12 11379 15708 23223 0
68706 3.84 13 11547 16018 23843 0
70670 3.48 18 14067 24568 26633 0
80180 3.93 20 14385 25410 27749
<3964>
80020 3.92 30 16485 33705 33328
<23268>
76690 3.73 40 18165 40845 38907
<42060>
72297 3.49 50 19530 47145 44487
<59653>
67949 3.26 60 20790 52815 50066
<76083>
64028 3.05 70 21840 58170 55645
<91472>
60623 2.87 80 22785 63105 61224
<105960>
57594 2.71 90 23730 67620 66804
<119650>
54944 2.58 100 24675 72030 72383
<132642>
52886 2.47
- GRAND TOTAL INCLUDES THE HYDROGEN GENERATED DUE TO ZR-WATER REACTION (15,410 CU.FT.) AND HYDROGEN PRESENT IN RCS AND PRESSURIZER VAPOR SPACE (1030 CU.FT.)
BOP UPRATE LICENSING REPORT 48 FNP-UNITS 1 AND 2 O
O O
O O
O TABLE 2.14-3 OFFSITE DOSES FROM CONTAINMENT VENTING TO CONTROL POST ACCIDENT HYDROGEN CONCENTRATION IH(ROlD DOSE DOSE DUE TO LOW POPULATION ZONE (rem)
POST-LOCA CONTAINMENT VENTING 21 LOCA W/O MINI-PURGE' 79 2.1 MINI PURGE INCREMENT 102.1 TOTAL 10 CFR 100 LIMIT 300 WHOLE BODY DOSE DOSE DUE TO LOW POPULATION ZONE (rem)
POST-LOCA CONTAINMENT VENTING 0.066 1
LOCA W/O MINI-PURGE' 1.5 MINI PURGE INCREMENT 0.022 TOTAL 1.6 10 CFR 100 LIMIT 25
' Values from Table 2.16-4 l
i V
BOP UPRATE LICENSING REPORT 49 FNP-UNITS I AND 2
]
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0 1
2 DNA I
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2.15 Safety-Related Electrical Equipment Qualification 2.15.1 Scope of Review As pan of the safety-related electrical equipment evaluation, the dose limit for safety-related electrical equipment located in a harsh environment (containment, main steam valve room, and auxiliary building subject to high radiation due to post-LOCA recirculatory fluids) was reviewed based on the new source terms developed for power uprate. Due to changes in the mass and energy releases following a design basis accident under uprate conditions, the containment pressure and temperature evaluation necessitated a review of the composite pressure and temperature curves for safety-related electrical equipment qualification. Components which were qualified based on calculated surface temperatures were also reviewed as part of the evaluation.
2.15.2 Summary of Evaluation Review of the radiological doses due to power uprate showed that many of the original design basis doses were bounding for uprate. For safety-related electrical equipment with uprate doses not bounded by the original design basis, radiological doses at uprate conditions were compared against the dose threshold limits used for the individual component or equipment. Comparing the dose threshold limit against the uprate dose showed sufficient margin is available to accommodate the increased uprate dose without compromising the equipment qualification.
Comparison of the composite uprate temperature profile to the existing composite temperature profile indicates that the uprate maximum accident temperature is approximately 10 'F less than the existing maximum design basis accident temperature. The composite uprate temperature profile is enveloped by the existing design basis composite temperature profile except for the first 70 seconds and after 30,000 seconds. These two areas are discussed below.
During the first 70 seconds the uprate MSLB temperatures are higher than the existing MSLB. The MSLB temperatures are for a relatively short duration and considering the thermal lag time associated with increasing the temperature of the containment the initial higher temperatures would not have an impact on safety-related electrical equipment qualification. Funhermore, the steam saturation temperature at 70 seconds is 280 'F at approximately 55 psia, which is well below the existing design basis composite temperature.
Towards the end of the composite temperature profiles (greater than 30,000 seconds), the uprate temperatures exceed the existing design basis profile by a few degrees (approximately 5 F). In general, the safety-related electrical equipment qualification test temperatures have enough margin to envelop the higher temperatures. In addition, there is adequate margin between the uprate composite profile and the existing design basis composite profile between 70 seconds to 10,000 seconds to compensate for the time when the uprate composite is slightly higher than the existing design basis profile.
The environmental qualification testing of safety-related electrical equipment that envelops the existing composite temperature profile adequately envelops the composite uprate temperature profile.
BOP UPRATE LICENSING REPORT 52 FNP-UNITS 1 AND 2
i
)
The composite uprate pressure profile is enveloped by the existing design basis composite pressure
}
profile.' Therefore, the environmental qualification of the safety-related electrical equipment is not affected by the accident pressure level as a result of power uprate.
?
Review of the MSLB cases used in the current design basis surface temperature analyses against the corresponding uprate MSLB cases showed comparable resuits especially in time duration above a high temperature such as 300 'F. Maximum temperature for the MSLB cases used in the design j
basis surface temperature analyses were greater than or compara' ole with the corresponding uprate j
MSLB cases. Original MSLB cases not selected for further consideration were either of a short l-duration at a high temperature or of a lower maximum temperature and are comparable with the
{
similar uprate MSLB cases which need not be considered in surface temperature analysis.
j in the main steam valve room accident analysis, the new mass and energy releases due to power uprate did not affect the original design basis pressure and temperature profiles for this room, and therefore, the original design is considered bounding. As such, no equipment evaluations were j
necessary.
2.15.3 Summary of Conclusions l
From the uprate evaluation,the radiological cumulative dose or dose rate was either enveloped by the i
results of previous design basis radiological analysis or was within the threshold limit for which the
/
individual component or equipment was qualified.
4 k
i The current design basis environmental qualification profiles for safety-related electrical equipment i
inside containment contain sufficient margin so that the impact of temperature and pressure conditions as a result of MSLBs or LOCA at power uprate conditions remains bounded.
l l
Comparing the original surface temperature design basis MSLB cases against the comparable uprate i
MSLB cases as well as the remaining MSLB cases shows the surface temperature results are bounding and, therefore, are not impacted by power uprate. Comparing the design basis post-LOCA containment temperature profile against the uprate post-LOCA containment temperature profile j
shows the original temperature curve as bounding and does not affect the surface temperature results I
BOP UPRATE LICENSING REPORT 53 FNP - UNITS I AND 2
2.16 Radiological Assessment 2.16.1 Description The impact of power uprate on radiological assessment is comprised of three main topics.1) The increased power may result in higher releases as compared to the presen' design basis due to the assumed increase in RCS activity levels associated with fuel defects, which can result in higher in-plant source terms with resulting higher on-site doses to operators and equipment. 2) The same increased source terms may result in increased releases to offsite locations during normal operations via liquid and/or gaseous release pathways which can result in higher offsite normal operations doses to the public. 3) liigher power may lead to increased inventories of contained radioactive products (e.g., core inventory, spent fuel, waste gas decay tanks); changes to accident consequences (e.g.,
changes to radiation release barrier-clad, RCS, containment-postulated damage levels and/or higher mass release rates for postulated accidents) which can result in higher post accident doses to the public or control room operators.
l 2.16.2 Scope of Review For power uprate, revised source terms were prepar-i for the core inventory, spent fuel, RCS, and
)
waste gas decay tanks. In addition, revised mass re. ease data were generated for the following accidents.
l
- Steam Generator Tube Rupture (SGTR)
- Main Steam Line Break (MSLB)
- Loss of Offsite Power (LOSP)
- Turbine Trip / Loss of Load (TT/LOL)
- Control Rod Ejection (CRE)
- RCP Locked Rotor (LR)
The impact of the new source terms and mass release data on the three areas described above was evaluated as follows.
- 1. For increased RCS source terms, the review considered normal shielding and operational dose rates. Doses to equipment are considered in the safety-related electrical equipment qualification evaluation in section 2.15.
- 2. Increased RCS source terms may also be transported via liquid and airborne pathways throughout the plant and ultimately be released offsite through the normal plant liquid and gaseous discharge points. This evaluation considered the increased power level impact on the design basis normal operation releases. In addition, revised source terms and releases from potential operation of the gaseous waste processing system in a compressed gas (high pressure) mode were considered.
Revised design basis normal operation offsite releases and the consequential offsite dose to the public were calculated.
)
i
- 3. Post accident offsite doses to the public and doses to control room operators were evaluated for the above accidents as well as for waste gas decay tank (WGDT) rupture, fuel handling accident l
BOP UPRATE LICENSING REPORT 54 FNP - UNITS 1 AND 2
... - - - - -. ~.. -. - - -. -. -
E e
f f
V (FHA), and the Loss of Coolant Accident (LOCA). Changes to source terms (e.g., core inventory) a potentially impacting post accident areas vital to operator actions were also evaluated. Post accident impact on radiolytic hydrogen generation is evaluated in Section 2.14.
2.16.3 Design Requirements Shielding for normal operations must meet the requirements of 10 CFR 20 related to operator dose and access control. Additional guidance for shielding is provided by USNRC Regulatory Guide 8.8 as described in FSAR Sections 12.1 and 12.3. The design of radwaste equipment must be such that the plant is capable of maintaining offsite releases and resulting doses within the requirements of 10 CFR 20 'and 10 CFR 50, Appendix 1 (and the Concluding Statement of Position of the Regulatory Staff (Docket-RM-50-2)). Additional guidance for evaluating compliance with these requirements is taken from USNRC Regulatory Guides 1.109 through 1.113, as discussed in FSAR Section 11.2.9.
Actual performance and operation ofinstalled equipment, and reponing of actual offsite releases and doses continues to be controlled by the.equirements of the Offsite Dose Calculation Manual.
Offsite and control room doses must meet the guidelines of 10 CFR 100 and requirements of 10 CFR 50, Appendix A, General Design Criterion 19 respectively. Further acceptance criteria for specific postulated accidents are provided by the NRC in the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," NUREG-0800, which indicates each accident should be "within"( < 100%), "well within"( < 25%), or a "small fraction of" (s 10%) the 10 CFR 100 guidelines.
Additional input assumption guidance for specific accidents is taken from USNRC Regulatory i
Guides, as discussed in FSAR Appendix 3A.
- 1.4 Loss of Coolant Accidents
- 1.24 Waste Gas Decay Tank Failure
- 1.25 Fuel Handling Accident
- 1.52 ESF Filter Systems
- 1.77 Control Rod Ejection
- 1.78 Control Room Habitability 2.16.4' Design Interfaces Source term and mass releases were developed in the NSSS sections of the licensing report.
RCS activities and the core inventory did not change significantly (see Section 2.16.8), so the shielding and onsite dose (except control room) evaluation was terminated after review of the source terms.
For normal offsite releases and doses, interfaces include all potentially radioactive systems, structures and components, as well as the environs around the plant (e.g., atmospheric parameters, ground and river water, aquatic and terrestrial plant and animal life, etc.).
O BOP UPRATE LICENSING REPORT 55 FNP - UNITS 1 AND 2
i Interfaces for the control room and individual accident offsite analyses vary with the accident. Each accident requires input of source terms and mass release data as described above. Additional interfaces are with radioactivity transport pathways, fission product cleanup and removal systems and components, control room structure:: and habitability systems, and atmospheric transport.
2.16.5 Assumptions Normal shielding reviews are based on 1% defective fuel. Accident shielding is based on 100% core melt source terms.
Normal offsite releases and dose assumptions for this evaluation are consistent with the original licensing basis analyses as described in FSAR Chapter 11, except for the increase in power.
The major assumptions used in calculation of offsite and control room doses are consistent with USNRC Regulatory Guides noted above except as follows.
- Dose conversion factors are taken from International Committee on Radiation Protection Publication 30 in lieu of the Regulatory Guides or TID-14844.
- lodine spike models as described in the appropriate sections of NUREG-0800 are considered in those FSAR accidents which currently do not include them. Although not specifically required by NUREG-0800, a pre-existing iodine spike is also modeled for the LOSP to provide a consistent treatment for a!! of the analyses.
- Elemental iodine spray remove' and plateout for LOCA analyses are modeled as time dependent phenomena based an deletion of the spray additive tank in lieu ofinstantaneous plateout of 50% of the airborne iodine as described in Regulatory Guide 1.4.
2.16.6 Method of Evaluation A shielding design review in conjunction with a review of the RCS, core inventory and waste gas decay tank source terms provided in the NSSS Licensing Report due to the FNP power uprate was completed. Since the shielding design methods contain significant conservatism, the review was terminated after a comparison of source terms.
Normal offsite doses and releases for uprated power were recalculated using the GALE, GASPAR, and LADTAP computer codes and Farley specific input for upraud power and plant system models.
A comparison of gaseous source terms was then made to the waste gas processing system source terms provided for high pressure operation and a composite bounding set of gaseous releases was input to GASPAR. Offsite doses for both the liquid and gaseous release pathways were calculated and comparisons to the acceptance criteria were made.
O BOP UPRATE LICENSING REPORT
's FNP - UNITS 1 AND 2
O Offsite dose calculations for accident releases were prepared using the multi-node TACTS computer code. Initial activity and release transport along the applicable pathway nodes (e.g., core to RCS to steam generator) to the environment are modeled for each accident.
2.16.7 Summary of Evaluation 2.16.7.1 Shielding The original shielding design was based on radiation source terms developed from a core thermal power of 2774 MWt and the equivalent of 1 percent fuel cladding defects. For power uprate, the RCS, core and WGDT activities are based on 102 percent of the uprate power of 2775 MWt. This is approximately 2 percent higher than the original basis provided in the FSAR Table 11.1-1.
Significant conservatism was included in the originally calculated dose rates for shielding design.
Use of 1 percent fuel cladding defects resulted due to uncertainty and limited experience with fuel performance and provides a significant amount of conservatism in the shielding design basis.
Current NRC guidance provided in NUREG-0800 suggests the use of 0.25 percent fuel cladding defects. For accident conditions, based on the revised source terms, the shielding design has a significant amount of conservatism to prevent creating additional inaccessible areas in the plant based on a power uprate core of 102% of 2775 MW:
conservatism compensates for any increase in dose rate due to power uprate.
2.16.7.2 Normal Offsite Releases and Doses J
The original FSAR calculations, prepared at a power of 2766 MWt, were revised to the uprated power of 2775 MWt, and include the impact of operation of the gaseous radwaste system at high pressure. These bounding calculations prepared to evaluate conformance to 10 CFR 20 and 10 CFR 50, Appendix 1 demonstrate that sufficient radwaste equipment is provided in the Farley design to maintain releases within the limits of 10 CFR 20, Appendix B, and the resulting ofTsite dose to the most exposed individual meets the limits of 10 CFR 50, Appendix 1 (and docket RM-50-2). No changes to radwaste system design or operation are required for uprate, and no significant changes to actual offsite gaseous and liquid releases and doses are expected.
2.16.7.3 Accident Doses 2.16.7.3.1 Evaluation of the Radiological Consequences of a Steam Generator Tube Rupture The radiological consequences of the SGTR were evaluated utilizing the assumptions of Standard Review Plan Section 15.6.3, except that partition factors are assumed to be limited to 10; i.e., there is no tube uncovery nor immediate flashing of primary to secondary leakage consistent with the current FSAR, Section 15.4.3. Two cases were evaluated.
/
BOP UPRATE LICENSING REPORT 57 FNP - UNITS 1 AND 2
Case 1. An SGTR with a preaccident iodine spike of 30 Ci/ gm results in offsite doses less than 10 CFR 100 guidelines, which meets the acceptance criteria.
Thyroid Dose Whole Body Beta Skin (Rem)
Dose (Rem)
Dose (Rem)
EAB 19 0.15 0.20 LPZ 7.7 0.06 0.07 Case 2. An SGTR with an accident initiated iodine spike with an appearance rate 500 times the equilibrium rate, corresponding to the technical specification limit of 0.5 Ci/gm results in offsite doses that are a small fraction (10%) of the 10 CFR 100 guidelines, which also meets the acceptance criteria.
Thyroid Dose Whole Body Beta Skin (Rem)
Dose (Rem)
Dose (Rem)
EAB 2.8 0.13 0.19 LPZ l.7 0.05 0.07 The potential for uncovery of the steam generator tubes during the event was also evaluated for uprated conditions. Assuming technical specification limit for RCS activity (0.5 pCi/gm) and leak rate (150 gpd to each intact generator) and release directly to the environment ( i.e., no mixing with the secondary side water) for the first 30 minutes, the offsite doses remain a small fraction of the 10 CFR 100 guidelines.
2.16.7.3.2 Evaluation of the Radiological Consequences of a Main Steam Line Break The radiological consequences of the MSLB were evaluated utilizing the assumptions of Standard Review Plan Section 15.1.5, except that partition factors in the intact generators are assumed to be limited to 10; l.c., there is no tube uncovery nor immediate flashing of primary to secondary leakage consistent with the current FSAR, section 15.4.2. Two cases were evaluated.
Case 1. An MSLB with a preaccident iodine spike of 30 pCi/ gm results in offsite doses less than 10 CFR 100 guidelines, which meets the acceptance criteria.
Thyroid Dose Whole Body Beta Skin (Rem)
Dose (Rem)
Dose (Rem)
EAB 7.0 8.9 x 10
4.6 x 10
LPZ 3.0 3.7x10
l.9 x 10
Case 2. An MSLB with an accident initiated iodine spike with an appearance rate 500 times the equilibrium rate, corresponding to the technical specification limit of 0.5 pCi/gm results O
BOP UPRATE LICENSING REPORT 58 FNP - UNITS 1 AND 2
O in offsite doses that are a small fraction (10%) of the 10 CFR 100 guidelines, which also meets the acceptance criteria.
i Thyroid Dose Whole Body Beta Skin (Rem)
Dose (Rem)
Dose (Rem)
EAB 6.9 8.8 x 10
4.5 x 10
]
LPZ 3.0 3.6 x 10
l.9 x 10
In addition, the impact ofimplementing steam generator tube alternate repair criteria in accordance with Units 1 and 2 Technical Specifications (Amendments 117 and 106, respectively) was reviewed.
Both the pre-existing and accident initiated iodine spike were evaluated based on the same criteria described above. The accident initiated iodine spike case is the more limiting, yielding a maximum J
allowable primary-to-secondary leak in the affected generator of 20 gpm. For this case the control room thyroid dose was 20.4 Rem.
The potential for uncovery of the steam generator tubes during the event was also evaluated for i
uprated conditions. Assuming technical specification limits for RCS activity (0.5 Ci/gm) and leakage (150 gpd to each intact generator) and release directly to the environment, ( i.e. no mixing with the secondary side water) for the first 30 minutes, the offsite doses remain a small fraction of the 10 CFR 100 guidelines.
2.16.7.3.3 Evaluation of the Radiological Consequences of a Loss of Offsite Power, Loss of
,k Load, Turbine Trip The radiological consequences of the bounding steam releases from a loss of offsite power, loss of load, and turbine trip were evaluated utilizing the assumptions of Standard Review Plan Sections 15.2.1-15.2-6, except that partition factors are assumed to be limited to 10; i.e., there is no tube uncovery nor immediate flashing of primary to secondary leakage consistent with the current FSAR, Section 15.2.9. These releases result in offsite doses which meet the acceptance criteria assuming no modine spike.
1 Thyroid Dose Whole Body Beta Skin (Rem)
Dose (Rem)
Dose (Rem)
EAB 8.9 x 10 '
l.2 x 10
7.9 x 10" LPZ 7.6 x 10
8.1 x 10" 5.8 x 10" A pre-existing spike which increases the RCS iodine concentrations to 30 Ci/gm deli3i results in 1
an increase in thyroid dose to 0.93 Rem and 0.82 Rem at the EAB and LPZ respectively.
i
i i
The potential impact of uncovery of the steam generator tubes during the event (s) was also evaluated for uprated conditions. Assuming technical specification limits for RCS activity (0.5 pCi/gm) and leak rate (150 gpd per generator) and release directly to the environment ( i.e., no mixing with the secondary side water) for the first 30 minutes, the offsite doses remain a small fraction of the 10 CFR 100 guidelines.
2.16.7.3.4 Evaluation of the Radiological Consequences of a Loss of Coolant Accident The radiological consequences of the large break LOCA were evaluated for the uprated core inventory utilizing the methods described in WCAP-ll611 (reference 16.1) for the containment model and Farley specific assumptions as shown in Tables 2.16-1 through 2.16-3. The results shown in Tables 2.16-4 and 2.16-5 include the contribution from ECCS recirculation loop leakage outside containment as described in FSAR Section 6.3 Evaluation of the offsite doses from hydrogen venting of 35 cfm starting 18 days after the LOCA (see Section 2.14) indicate that the total offsite doses and control room doses from all sources continue to meet 10 CFR 100 guidelines and GDC 19 requirements respectively.
A small break LOCA which results in 100% failure of the fuel clad and release of 100% of the gap activity, but does not result in a containment pressure high enough to initiate containment spray, was also evaluated. The assumptions shown in Tables 18.12-1 through 18.12-3, except the release source term and spray removal constants for iodine cleanup in the containment, were used. The resultant offsite doses (from containment leakage without minipurge or hydrogen venting contributions) are:
Thyroid Dose Whole Body Beta Skin (Rem)
Dose (Rem)
Dose (Rem)
EAB 41.2 0.4 0.3 LPZ 29.0 0.2 0.1 2.16.7.3.5 Evaluation of the Radiological Consequences of a Fuel Handling Accident The radiological consequences of the FHA were evaluated for the uprated core inventory. Two cases were considered, an accident in the Auxiliary Building and one in the Containment. The accident in the Auxiliary Building conforms to the guidelines of Regulatory Guide 1.25 and Standard Review Plan Section 15.7.4, except dose conversion factors from ICRP 30 are used in lieu of those from Regulatory Guide 1.25. The accident in the containment is evaluated assuming the entire containment airborne source is exhausted via the containment purge filter. The releases result in ofTsite doses that are well within the 10 CFR 100 guidelines, which meets the acceptance criteria.
O BOP UPRATE LICENSING REPORT 60 FNP - UNITS 1 AND 2
O k
Case 1. FHA in the Auxiliary Building Thyroid Dose Whole Body (Rem)
Dose (Rem)
EAB 30.0 0.38 LPZ l 1.4 0.14 Case 2. FHA in the Containment:
Thyroid Dose Whole Body (Rem)
Dose (Rem)
EAB 51.4 0.38 LPZ l 8.9 0.14 2.16.7.3.6 Evaluation of the Radiological Consequences of a Waste Gas Decay Tank Rupture The radiological consequences of the WGDT rupture were evaluated for a tr.nk containing the maximum inventory in accordance with Technical Specification 3/4.11.2.6. The accident evaluation conforms to the guidelines of Regulatory Guide 1.24. The releases result in offsite doses are a small fraction of the 10 CFR 100 guidelines, which meets the acceptance criteria.
O Skin Dose Whole Body (Rem)
Dose (Rem)
EAB 0.57 0.30 LPZ 0.21 0.11 2.16.7.3.7 Evaluation of the Radiological Consequences of a Control Rod Ejection The radiological consequences of the releases (assuming 10% of the gap inventory and 0.25% of the core inventory) to containment and secondary side steam releases from a control rod ejection were evaluated utilizing the assumptions of Standard Review Plan Section 15.4.8 and Regulatory Guide 1.77, except that typical secondary side partition factors of 100 discussed in other sections of the Standard Review Plan are assumed to be limited to 10; i.e., there is no tube uncovery nor immediate flashing of primary to secondary leakage consistent with the current FSAR, Section 15.4.6. Rese releases result in offsite doses well within the guidelines of 10 CFR 100, which meets the acceptance criteria.
Thyroid Dose Whole Body Beta Skin (Rem)
Dose (Rem)
Dose (Rem)
EAB 33.5 0.19 0.13 LPZ 27.4 0.10 0.07 ed 1
BOP UPRATE LICENSING REPORT 61 FNP - UNITS 1 AND 2
The potential impact of uncovery of the steam generator tubes during the event was also evaluated for uprated conditions. Assuming technical speciGcation limits for RCS activity (0.5 Ci/gm) and leak rate (150 gpd per generator) and release directly to the environment ( i.e. no mixing with the secondary side water) the offsite doses remain well within the 10 CFR 100 guidelines.
2.16.7.3.8 Evaluation of the Radiological Consequence af an RCP Locked Rotor The radiological consequences of RCP locked rotor releases assuming 20% of the fuel clad / pellet gas gap is released to the RCS with subsequent leakage to the steam generators and secondary side steam releases were evaluated utilizing the assumptions of Standard Review Plan Section 15.3.3. These releases result in offsite doses that are a small fraction of the guidelines of 10 CFR 100, which meets the acceptance criteria.
Thyroid Dose Whole Body Beta Skin (Rem)
Dose (Rem)
Dose (Rem)
EAB 15.3 2.03 1.77 LPZ 29.4 1.06 0.90 j
2.16.8 Summary of Conclusions G,
No changes or additions to structures, equipment, or procedures are necessary to provide adequate radiation protection for the operators or the public during normal or post-accident operations to support the uprate. The existing structures, systems, and components can safely handle the changes j
in post accident source terms and releases from the uprate conditions, and resulting onsite and offsite doses are less than the 10 CFR 100.11 guidelines and are within the Standard Review Plan guidelines. Therefore the radiological consequences acceptance criteria for postulated Condition 11, 111, and IV events are satis 6ed.
1 4
O BOP UPRATE LICENSING REPORT 62 FNP - UNITS 1 AND 2
i
,_. _. _ _ _ _ _ _..__.~._._..._._
1 1
2 a
TABLE 2.16-1 i
4
~
(REVISED FSAR TABLE 6.2-29)
SPRAY EVALUAT!ON PARAMETERS Power (MWt) 2831 i
l
^
Containment pressure (psia) 68.7 Containment temperature ('F) 276 Spray flowrate(gal / min) 2175 pi-! (spray injection) 4.5 (spray recirculation) 7.5 Containment sprayed volume (ft')
1.62 x 10' i
Minimum spray fallheight(ft) 110 A, (h)
Elemental 1.4 (DF < 14) 0.0 (DF 214) organic 0.0
?
Particulate 4.8 (DF < 100) 0.48 (100 s DF <1000) 0.0 (DF 21000) i i
"s 1
i i
I i
1 i
,o l
l J
4 I
BOP UPRATE LICENSING REPORT 63 FNP - UNITS 1 AND 2
TABLE 2.16-2 (REVISED FSAR TABLE 15.4-14)
PARAMETERS USED IN THE LOCA ANALYSIS Core thermal power 2831 MWt (2775 x 1.02)
Containment free volume 2.03 x 10' ff Volume fractions Sprayed 0.822 Unsprayed 0.178 Mixing rate between sprayed and 40,500 ff/ min unsn syed containment volumes Coro fission product inventories See Table 15.1-4 Activity released to containment Noble gases 100% of core inventory lodines 50% ofcoreinventory Plateout of elemental iodine activity 12.5 h (DF < 100) released to containment 0.5 h (100 s DF s 1000) 0.0 h (DF > 1000)
Form ofiodine activity in containment available for release Elemental 95.5 %
O <tanic 2.0 %
Partics kw 2.5 %
Spray removal constants Elemental 1.4 h (DF <l4) 0.0 h (DF 214)
Methyl 0.0 h
Partic late 4.8 h (DF < 100) 0.48 h (100 s DF <1000) 0.0 h (DF 2 1000)
Time to reach decontamination factor Elemental 21 min Methyl N/A Particulate 7.3h Containment leak rate 0-24 h 0.15%' day 1-30 days 0.075%/ day Atmospheric dilution estimates See Table 15.B-2 O
BOP UPRATE LICENSING REPORT 64 FNP-UNITS 1 AND 2
. -... _. _ _ _ -. _ _ _ _. _. - _... ~. - _. _... -
\\
TABLE 2.16-3 (REVISED FSAR TABLE 15.4-16)
PARAMETERS USED IN ANALYSIS OF POST-LOCA CONTROL ROOM DOSES Filtered pressurization rate (ft'/ min) 450 Filtered recirculation rate (ft'/ min) 2700 Unfiltered inleakage rate (ft'/ min) 10 Filter efficiencies (all forms ofiodine)(%)
Pressurization air 99 Recirculation air 95 Volume (ft')
114,000 Operator breathing rate (m'/s) 3.47 x 10" Percent of time operator is in control room following LOCA 01 day 100 1-4 days 60 4 30 days 40 Atmospheric dilution estimates (s/m')
i 0-2h 3.28 x !0
2-8 h 2.65 x 10
8-24 h 2.19 x 10
l-4 days 1.64 x 10" 4-30 days 1.08 x 10" 1
I i
O BOP UPRATE LICENSING REPORT 65 FNP-UNITS 1 AND 2
~__
l 1
l TABLE 2.16-4
~
(REVISED FSAR TABLE 15.4-15)
OFFSITE DOSES FROM LOCA Thyroid Dose. (REM)
LOCA w/o Mini-Mini-Purge 10 CFR 100 Pmsc Incremental
_ Intal Limit Site boundary 131 5.7 137 300 Low-population zone 79 2.1 81 300 Whole Body. (REM)
LOCA w/o Mini-Mini-Purge 10 CFR 100 Pmgs incremental
_Tmal Limit Site boundary 3.0 8.7 x 10
3.0 25 Low-population zone 1.5 2.2 x 10
l.5 25 Offsite doses are based on a LOCA plus the incremental dose resulting from purging prior to isolation of the mini-purge system 6 seconds following a LOCA.
TABLE 2.16-5 (REVISED FSAR TABLE 15.4-17)
CONTROL ROOM DOSES FOLLOWING A LOCA (Dose (rem)
Thyroid 13.2 Whole body 0.7 Beta-Skin 17.2 O
BOP UPRATE LICENSING REPORT 66 FNP - UNITS 1 AND 2
i l
f i
2.17 Containment Ventilation 2.17.1 System Description The containment ventilation system (CVS) is designed to remove heat and/or ventilate specific locations in the containment. It also filters the containment air in order to reduce personnel exposure i
and minimize radioactivity releases to the outside environment. The CVS is comprised of the following subsystems.
i i
- Containment fan coolers
- Containment purge / mini-purge systems
- Containment dome recirculation fans
- Containment preaccess filter system i
- Reactor cavity cooling fans
- Refueling water surface ventilation fans
- Control rod drive mechanism cooling fans
- Post-accident containment mixing fans The post-accident containment mixing fans and vent system are discussed in section 2.13 as part of the post-LOCA hydrogen generation evaluation.
2.17.2 Scope Of Review The review of the CVS is limited to those subsystems identified above.
2.17.3 Design Requirements For normal operation, the containment fan coolers are required to keep containment air temperature below the value (120*F) allowed by the Technical Specifications. For accident conditions, the fan coolers are required for the long-term energy removal from the containment atmosphere.
The function of the purge system is to prevent pressure build-up in containment and limit build-up of noble gases, iodine, and particulates to provide an acceptable working environment inside containment. Mini-purge i, solation is required within 5 seconds of receipt of closure signal as modeled in the safety analyses.
The function of the remaining non-safety-related containment HVAC subsystems is to provide mixing of the containment atmosphere to:
-limit thermal stratification;
- provide reactor cavity concrete and vessel support cooling;
- provide excore Nuclear Instrument detector and CRDM cooling;
- limit localized equipment, personnel or structure high temperatures; and
- help maintain radioactivity levels inside containment consistent with occupancy requirements.
These fans and filtration subsystems are not required for accident mitigation or long-term post accident energy removal from containmen'.t BOP UPRATE LICENSING REPORT 67 FNP - UNITS 1 AND 2
l O
2.17.4 Design Interfaces The Farley units currently operate very near the maximum uprate average reactor coolant temperature of 577.2
- F, and uprate RCS source terms, specifically 1-131, are bounded by evaluations supporting current operation as described in the FSAR.
The ability of the fan coolers to maintain containment temperature is driven by sources of heat gain within the containment and by service water temperature. For normal operation, the maximum service water (SW) temperature of 95 ' F is assumed. For accident conditions, a composite SW temperature profile from the ultimate heat sink (UHS) analysis is assumed. Minimum performance is assumed in the LOCA and MSLB containment pressure / temperature (P/T) analyses used for safety-related electrical equipment qualification.
The horsepower requirements of the CVS fan motors impact the electrical distribution system.
However, no changes result since there are no changes to containment design conditions, and actual conditions during normal operation are not changing significantly for uprate. Additionally, there are no changes to either containment conditions or design requirements of the CVS subsystems associated with uprate. Therefore, there is no impact on the instrumentation and control systems for uprate.
2.17.5 Method of Evaluation The CVS functional design of each subsystem was evaluated with respect to changes to the key parameters affecting system performance for both normal and accident conditions, as applicable.
Both design limits and actual plant operating conditicas were considered.
The primary heat load sources on the containment fan coolers were evaluated for increases resulting from the uprate. Specifically, changes in reactor coolant temperature and main steam and feedwater temperature were considered as well as heat addition from the RCP motor coolers and the CRDM cooling system. The CRDM and RCP motor cooler loads are increasing slightly, and are expected to increase containment temperature insignificantly.
Containment cooling system post-accident performance is verified acceptable as a result of the containment P/T analyses which incorporate the fan cooler duty curves. Data for fan cooler duty are shown below.
CTMT Temp CTMT AIR COOLER DUTY
(*F)
(BTU /HR)
UPRATE CURRENT 120.4 3.024 E+06 3.336 E+06 139.1 5.890 E+06 6.15l E+06 201.0 16.820 E+06 17.210 E+06 265.8 29.130 E+06 29.496 E+06 274.6 30.788 E+06 31.161 E+06 O
BOP UPRATE LICENSING REPORT 68 FNP - UNITS 1 AND 2
- -. ~. ~. -. - ~
d i
i I
D
]
Revised source terms developed for the upate are bounded by current analyses. Uprate does not affect the maximum isolation valve closure time. Therefore, the containment preaccess filter system and purge / mini-purge systems continue to meet the design requirements.
The functions of the remaining non-safety-related containment HVAC subsystems were verified for j
the increase in total containment heat load estimated for uprate conditions. No fluid system changes which might generate new high temperature areas or RCS leakage locations are anticipated for power i
uprate conditions. Therefore, these subsystems continue to meet their design requirements.
2.17.6 Summary of Evaluation The CVS was evaluated against current performance with respect to potential increased demand on 3
each subsystem to perform its design functions. Any changes to the containment environment due to j.
increased heat load and radioactivity were addressed. It was determined that changes to normal operation heat loads are negligible. Further, revised source terms developed for the uprate are 1
bounded by current analyses. Since actual containment conditions change insignificantly and design parameters are bounded by current design basis analyses, it is concluded that no changes to the CVS are required for uprate.
)
2.17.7 SummaryofConclusions l
Changes to the key parameters affecting CVS performance are slight or are bounded by current j
design. A revised fan cooler duty curve was developed for post-accident P/T analyses. It was i N determined that the uprate does not significantly challenge the CVS; however, an insignificant j
containment temperature increase may occur due to slight increases in heat loads from the CRDM cooling fans and RCP motor coolers.
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i BOP IJ/ RATE LICENSING REPORT 69 FNP - UNITS 1 AND 2 i
2.18 Auxiliary Building Ventilation 2.18.1 System Description This section describes the evaluation of the Farley Units 1 and 2 Auxiliary Building Ventilation System (ABVS) for power uprate. The ABVS is designed to provide a suitable environment for equipment and personnel.
2.18.2 Scope of Review The adequacy of the components for current operation is demonstrated by plant operating history and previous evaluations. Changes to plant operating conditions and unit-to-unit variations are considered for uprate. Changes to the auxiliary building operating environment can potentially affect the electrical distribution system, for example by changing the horsepower requirements of fan motors. It was determined that uprate does not result in any significant changes to the auxiliary building ambient conditions. Subsequently, there is no effect on the electrical distribution system.
An SFP temperature increase could impact the ability of the ABVS to maintain area temperatures in the fuel handling area within design limits. While decay heat loads increase for uprate, previous SFP temperature alarm setpoints do not change for uprate. Also, specifically addressed are the control room and TSC, the rooms provided with service water (SW) coolers, the main steam valve room, and the turbine driven AFW pump room since these rooms house safety-related equipment.
Ol 2.18.3 Method of Evaluation The functional requirements of the primary ABVS subsystems were evaluated with respect to changes to operating conditions resulting from uprate. The adequacy of the components for current operation is demonstrated by plant operating history.
2.18.4 Summary of Evaluation 2.18.4.1 Control Room and TSC liVAC Systems j
The control room and TSC air conditioning and filtration systems are designed with sufficient redundancy and separation of components to provide reliable operation under normal and emergency conditions. Uprate does not affect normal ambient conditions inside control room or TSC, and post accident doses meet General Design Criterion 19. Therefore, the control room and TSC HVAC systems capabilities are not impacted by uprate and the safety-related components do not exceed any design limit considering a single active or passive failure of the liVAC system.
2.18.4.2 Fuel liandling Area liv and Filtration System The system provides adequate capacity *.nsure that proper temperatures are maintained during normal and shutdown conditions and to casure that effluent discharges are maintained within acceptable levels. A bounding analysis was performed for the SFP which resulted in an increase in O
BOP UPRATE LICENSING REPORT 70 FNP - UNITS 1 AND 2
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the maximum pool temperature; however, the high temperature alarm setpoint of 130
- F is not being increased. It is therefore concluded that no HVAC design limits are challenged for uprate.
The penetration room filtration system is used to process the fuel pool area atmosphere in the event
<< a fuel handling accident. As noted above the pool temperature is not expected to change; i
therefore, there is nc impact to the penetration room filtration system capability to perform its design function following an FHA.
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For uprate, the existing post LOCA pressure / temperature profiles in rooms served by the ESF pump room coolers remain conservative. It is therefore concluded that the penetration room filtration system remains capable of performing its design function following a LOCA.
2.18.4.3 ESF Pump Room Coolers The pump room coolers are designed to maintain the ambient temperature in each of the charging, RHR, containment spray, CCW and motor driven AFW pump rooms at or below 104 F for normal operation. During normal operation when the equipment in these rooms is in use, uprate conditions will not impact heat loads in these rooms. Therefore, the normal functions of the room coolers are l
not impacted.
Previous evaluations addressed increased heat loads due to a higher post LOCA SW temperature, which continues to bound uprate. Therefore, the ESF pump room coolers remain capable of f) performing their design function at uprate conditions.
a 2.18.4.4 Main Valve Steam Room For uprate, steam temperature is decreasing approximately 4
- F while feedwater temperatures are increasing approximately 6
- F. For normal conditions, these temperature variations are not significant and somewhat offset. Under accident conditions, the uprate mass and energy releases for MSLB outside containment are bounded by the current design basis analysis; therefore, there is no main steam valve room temperature impact due to uprate.
2.18.4.5 Turbine Driven AFW Pump Room The TDAFW pump room does not have safety-grade cooling. Room temperatures are driven by operation of the turbine rather than steam temperatures (which decrease at uprate conditions). It was determined that the maximum post accident room temperature would not exceed 125
- F and that the associated system components remain capable of performing their design functions. It is therefore concluded that the current evaluated room temperatures apply for uprate.
2.18.5 Summary of Conclusions It was determined that the ABVS subsystems remain capable of maintaining operating temperatures within design limits.
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BOP UPRATE LICENSING REPORT 71 FNP - UNITS 1 AND 2
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9i 2.19 Miscellaneous Mechanical Reviews 2.19.1 Scope of Review l
The following miscellaneous mechanical systems were reviewed for potential impacts due to uprating.
- Turbine building HVAC systems
- Secondary side water chemistry
- Steam generator blowdown processing system
- Secondary system piping and valves
- CVCS piping and supports
- Radwaste systems
- Cold overpressure mitigation (COM) 2.19.2 Summary of Evaluation The TB HVAC systems were evaluated against current performance with respect to potential increased demand on the each subsystem to perform the design functions. It was determined that changes to current operation heat loads are negligible.
Secondary side erosion / corrosion may impact iron transport which is expected te increase by about I to 2 %. Other major control parameters not related to erosion / corrosion included in this enluation (sodium, chlorides, sulfates, cation conductivity, feedwater dissolved oxygen, and ETA) are bounded by current operations.
Main steam piping thermal e'.,, arm.un at uprated conditions is bounded by the current design basis evaluation. He slight increase in steam flow rates has no significant impact on the main steam piping, valves, in-line components, interconnected branch lines, or pipe supports as a result of dynamic loads due to steam hammer resulting from turbine stop valve closure. The main steam relief valve setpoints, rated capacities, and corresponding dynamic loads due to valve operation imposed on the piping and adjacent structures do not change as a result of uprate.
Temperatures in the condensate and feedwater systems are lower than the current temperature, except for the # 6 feedwater heater. The latter temperature increase is small and is considered negligible. Condensate and feedwater system dynamic loads due to water hammer were also reviewed and no changes to analyses were required.
The uprate temperature of the SGBD fluid is less than the current normal operating temperature. The CVCS temperature at uprated conditions is bounded by current design. Since the current conditions bound the uprate conditions, the current piping analysis and component design margins are still applicable and bounding.
Valves in the FNP MOV program were reviewed for impact due to uprated conditions. Since pump shutoff pressure was used in the MOV actuator analyses (except charging pump miniflow which is not impacted by uprate) and no significant pump changes were required to support operation at uprated conditions, the pressures used to evaluate MOV actuators do not change. Further, process temperatures do not change, or change only a few degrees. Therefore, uprate has no impact to the FNP MOV program.
BOP UPRATE LICENSING REPORT 72 FNP - UNITS 1 AND 2
%J No significant changes to the solid, liquid, or gaseous radwaste systems or components were identified as a result of power uprate. No design or operating limits are changed. The bounding "high pressure" operating mode for the gaseous radwaste system was considered for impact on offsite doses and releases and is acceptable as described in Section 16.
For cold overpressure protection, power uprate changes to inadvertent charging pump start and the equivalent mass injection rate which may occur when an RCP is started with steam generator temperature higher than the RCS were reviewed for impact on COM. The mass injection flow rates resulting from cold overpressure sources remain within the design capacity of a RHR suction relief valve.
2.19.3 Summary of Conclusions The current TB HVAC systems can accommodate the uprating.
Based on the predicted erosion / corrosion changes resulting from power uprate, increases in the iron transport rate to the steam generators is expected to be negligible. The pH changes that occur after the power uprate are small and do not impact the overall chemistry control of the secondary system.
There is no adverse impact to the main steam, condensate, and feedwater piping as a result of system pressure and temperature changes due to uprate.
/O The current design basis and normal operating conditions of the CVCS and SGBD system bound uprate conditions. Therefore the piping, components, and supports remain capable of performing their design basis functions.
There is no adverse impact on the operation of the secondary system MOVs and no impact on the current FNP MOV program as a result of uprate.
The existing radwaste systems and components are capable of processing radwaste generated at the uprated power level. These systems will adequately perform their design functions of controlling and limiting normal offsite releases and doses at uprated power levels.
The COM RHR relief valves remain capable of protecting the RCS from cold overpressurization events at uprated conditions.
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BOP UPRATE LICENSING REPORT 73 FNP - UNITS 1 AND 2
2.20 Miscellaneous Electrical Reviews The system evaluations for power uprate were reviewed for their impact on the plant electrical distribution system. This section contains a complete evaluation of the individual system interfaces with the plant electrical distribution system and the electrical components where an impnct to the plant electrical distribution system has been identified. The following systems and components have been evaluated.
- Main power transformers (MPT)
- Isophase bus
- Condensate pump motors
- Reactor coolant pump motors
- Diesel generators
- Station service assessment
- Startup transformer
- Unit auxiliary transformer (UAT)
- Station blackout (SBO)
- Grid stability 2.20.1 Scope of Review The purpose of this review is to determine if the MPTs, isophase bus (ampacity and cooling), large motors (RCP and condensate pump) will operate satisfactorily for uprated conditions and to verify that all station auxiliary loads can continue to perform their intended safety-related and non safety-related functions at uprated conditions.
Diesel Generator load profile for the LOSP, SI, or SBO events have been reviewed to verify the continued capability of the emergency diesel generators to perform their safety-related function.
Major SBO program elements and the electrical system grid (voltage and frequency) were reviewed to verify satisfactory performance at uprated conditions.
2.20.2 Summary of Evaluation A comparison of the transformer nameplate MVA ratings against the expected generator MVA indicates that the transformers have sufficient capacity. Four loading cases were evaluated to deternse transformer operating temperatures.
- Uprated generator capability with the UAT load
- Existing generator capability with the UAT toad
- Uprated generator capability less the UAT load
- Existing generator capability less the UAT load Based on a comparison of these cases there is sufficient margin in the MPTs to handle the uprated electrical power requirements.
The isophase bus for Unit I and Unit 2 remains capable of supporting the output of the main generator at uprated conditions without modification. The maximum uprated current of Unit I and Unit 2 is less than the rating of both the main and delta section of the isophase bus. The Unit I and Unit 2 (main bus and at termination box) primary cooling units are designed to remove heat loss at the maximum bus a npacity rating. In addition, the Unit 2 back-up cooling unit can remove the heat BOP UPRATE LICENSING REPORT 74 FNP - UNITS 1 AND 2
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loss at maximum bus ampacity rating. No known hot spots curcently exist, and none are expected to l
result from uprate conditions. The Unit I and Unit 2 cooling units are capable of removing the additional heat during uprated conditiot.s.
Upon examination of the parameter changes for power uprate, no changes which impact the diesel generator loading values were identified. The diesel load during the various LOSP, LOSP with LOCA, and SBO events is within the design and licensed ratings of these machines; therefore, the existing diesel generator loading analysis remains bounding. The response time capability for diesel generating starting and loading is not impacted due to power uprate.
An examination of the SBO coping criteria revealed that coping duration is a function of offsite power design characteristic group, emergency power configuration group, and emergency diesel i
generator target reliability, as discussed in Regulatory Guide 1.155. None of these are affected by power uprate. As concluded in Sections 2.17 and 2.18, the containment, auxiliary building, and main steam valve room operating temperature profiles will be the same at the onset of the SBO scenario as current design basis evaluations. Since the dedicated SBO diesel remains available within 31 minutes of the SBO event initiation, any change in ambient temperature rise in that 31 minute period due to uprate will be insignificant. The condensate storage tank inventory remains much greater than that needed for power uprate.
]
Evaluations were performed which indicated the horsepower load increases of the RCPs and
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condensate pumps do not adversely impact the station auxiliary electrical distribution system. Load, voltage, and short circuit values for uprate conditions at all distribution levels are minimally impacted when compared to current as-built values. The total projected loads on the 4160V buses are within the rated capacities of the buses, breakers, and transformers (Startup and UATs).
Bounding steady state voltages at the 4160V buses decrease by no more than 0.4%. Motor starting voltages remain within acceptable limits. Short circuit currents are within acceptable values. Bus and transformer relay settings are not impacted.
The condensate pump motor load increase due to power uprate has been evaluated. Based on this evaluation, the impact of the load increase to the electrical distribution system is acceptable. The acceleration time for the pump, motor structural and torque loading, motor insulation life, and existing relay setpoints are acceptable for operation at the increased horsepower required to support the condensate flow increase discussed in Section 2.2..
The existing protective relays are adequate for RCP motor protection under uprate conditions.
The electrical system grid was evaluated to ensure stable operation of Farley Units 1 and 2. The evaluation of the stability impact ofincreasing Farley generation to 920 MW per unit indicated there is a small negative impact on the stability of the units. However, as is the case today, stability can be assured by reducing the output of unit 2 to approximately 780 MW if there is an extended outage of the Farley-Snowdoun 230 kv line.
2.20.3 Summary of Conclusions The MPTs are designed to handle the uprated generator output. This was confirmed by a comparison of the expected uprated generator output to the nameplate ratings for the MPTs. In addition, the f
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l BOP UPRATE LICENSING REPORT 75 FNP - UNITS 1 AND 2 l
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evaluation of the transformers for the expected uprated generator output during the assumed worst case conditions indicate that the transformers may reach their present alarm setpoints but will not exceed the transformer temperature design limits.
Both Unit I and Unit 2 isophase buses are capable of handling the uprated generator output without modification.
The diesel generators are not impacted by the uprate of FNP. Examination of all LOSP, LOCA, and SBO diesel generator load profiles reveals the diesel generators continue to operate within their design and licensed ratings. As such, the diesel generators continue to be able to perform their intended safety-related function.
Power uprate does not impact the current Station Blackout Program. It was concluded that:
- SBO coping durations are unaffected;
- diesel generator loading in the SBO scenario remains within diesel ratings;
- current design basis temperature profiles in areas housing SBO required equipment remain bounded for an SBO in an uprated plant; and
- condensate storage tank inventory is much greater than that needed for power uprate.
The station auxiliary electrical distribution system is not adversely impacted by the proposed power uprate of Farley. Plant safe shutdown equipment continues to be able to perform its intended safety-related functions.
Operation of the condensate pump motor at the expected increased load has been evaluated and is within the design capability of the motors and cable. The protective relay setpoints and life of the insulation for the condensate pump motor are acceptable for operation of the motor at the expected increased horsepower. Based on the load study evaluation, the impact of the load increase to the electrical distribution system is acceptable.
Evaluation indicates that the protective relays and cable are adequate for RCP motor operation under power uprate conditions.
Under normal expected transmission system operating conditions, the grid remains stable and safety-related buses continue to be supplied by the off-site preferred power source for single contingency events and faults. There is a slight decrease in the margin of stability for limiting faults during valley load conditions. Normal system load growth offsets the slight decrease in margin of stability within 3 to 5 years.
O BOP UPRATE LICENSING REPORT 76 FNP - UNITS 1 AND 2
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t'v 2.21 Miscellaneous BOP l&C Reviews l
2.21.1 Scope of Review l
The Condensate and Feedwater subsystems reviewed include the following.
- Condensate pump minimum flow control
- Heater drains controls
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- Steam generator feed pump (SGFP) minimum flow control
- SGFP speed control
- SGFP net pump suction pressure alarm and trip
- Hotwell level control
- Steam generator water level controls In addition, the steam jet air ejectors bypass flow control and the radiation monitor RE 0015 were evaluated.
The Circulating Water System instruments and controls were evaluated at uprated conditions.
The Main Steam System review includes the steam dump controls, steamline isolation logic, low steamline pressure /high steamline differential pressure safety injection, main steam drain pot controls, and the main steamline radiation monitors RE 70 A, B, and C.
The current main turbine and auxiliaries instruments and controls were examined to determine if they have adequate design margin for the uprate conditions. In addition to the DEH controls, the
/3 lube oil system controls; bearing vibratior/ thrust, oil, and temperature instruments; tnrbine b
temperature monitors; and autostop oil pressure switches were reviewed.
Miscellaneous radiation monitors credited in accident analyses (containment, control room, and fuel handling area exhaust) were reviewed.
2.21.2 Summary of Evaluation The evaluation of the Condensate and Feedwater subsystems resulted in no controls changes to the SGFP speed controls, hotwell level, heater drains, and minimum flows of the condensate pumps and feedwater pumps. However, changes are being made for the SGFP NPSH alarm and trip serpoints.
No increase in condenser air influx is expected; therefore, no change to the sample rate of the radiation monitor RE 0015 is required. Necessary adjustment spans are available for the condensate system to meet the demand which might be imposed on the steam jet air ejectors bypass flow controls.
The slight increase in main steam flow rates has no adverse impact on the main steam instruments and controls. Rescaling of the radiation monitors RE 70A, B, and C is required for the increase in main steam flow. Rescaling of the 7300 control loops is required as described in the NSSS Licensing Report.
Evaluation of the turbine and turbine auxiliary controls (GSSC, feedwater bypass, DEH controls)
/N resulted in available span for adjustment to the optimum differential pressures evaluated for Section k
BOP UPRATE LICENSING REPORT 77 FNP - UNITS 1 AND 2
i 2.5 (Turbine Auxiliaries) and to accommodate valve position changes evaluated in Section 2.4 (Main Turbine); the lube oil system, bearing instrumentation and other monitoring instrumentation is not affected by the uprating. The MSR reheater controls are not affected by the uprating. Sufficient control ranges exists to maintain LP turbine temperature as desired. The increase in heat loads in the turbine generator does not affect the instrumentation and controls associated with the generator cooling system.
For power uprate, the resultant airborne radioactivity concentrations will be slightly higher and the radiation monitors credited in accident analyses will respond slightly faster for the same design basis accident sequences or RCS leak rate. Since process system parameters (e.g., sample flow rates, HVAC flow rates, building dilution volumes, containment leakage rate, atmospheric dispersion) have not changed, the setpoints for these instruments need not be changed.
2.21.3 Summary of Conclusions Uprated conditions do not have a significant impact on the instrumentation and controls of the condensate and feedwater system, circulating water system, main steam system controls or controls interfaces, turbine trip setpoints or the setpoints of the process parameters for the turbine support 1
systems, or miscellaneous radiation monitors. As such, the current control systems configuration can accommodate the uprating.
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BOP UPRATE LICENSING REPORT 78 FNP - UNITS 1 AND 2
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\\v 2.22 Environmental Impact Evaluations The Farley Nuclear Plant Final Environmental Statement (FES or FES-OL, reference 22.2) evaluates the nonradiological impact of the two units at Farley Nuclear Plant. The conclusions of the Final Environmental Statement are based on review ofinformation contained in the Environmental Report-Operating Licensee Stage (reference 22.1). This environmental evaluation provides an assessment of environmental impact associated with power uprate based on comparisons of the operating parameters established for power uprate with the parameters and conclusions in the above referenced reports.
Section 3.1 of the Farley Nuclear Plant Environmental Protection Plan (EPP), Appendix B to Facility Operating Licenses NPF-2 and NPF-8 (reference 22.3), states that "the licensee may make changes in plant design or operation, or perform tests or experiments affecting the environment provided such activities do not involve an unreviewed environmental question and do not involve a change to the EPP". Section 3.1 requires that an environmental evaluation be prepared and recorded prior to engaging in any activity which may significantly affect the environment. Section 3.1 further states that "A proposed change, test or experiment shall be deemed to constitute an unreviewed environmental question ifit concerns: (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the FES-OL, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board; or (2) a significant change in effluents or power level; or (3) a matter not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact."
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2.22.1 Scope of Review in accordance with the requirements discussed above, an enluation assessing the environmental impact of the proposed NSSS power level uprate from 2660 MWt to 2785 MWt has been performed.
This evaluation documents that the proposed change in power level is not significant relative to adverse environmental impact. The following environmental evaluation specifically considers effects on the following parameters.
River Water / Service Water Intake System Withdrawal rate intake canal velocity Circulating Water System Changes in rate of cooling tower blowdown Changes in temperature of cooling tower blowdown Changes in makeup to the cooling towers Changes in the amount of cooling tower drift Changes in cooling tower chemistry Changes in consumptive water use
/~N Groundwater Withdrawal System t
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BOP UPRATE LICENSING REPORT 79 FNP - UNITS 1 AND 2
I Changes in groundwater withdrawal to supply sanitary water system Changes in groundwater withdrawal to supply fire protection system Radwaste Dilution System Changes in liquid radwaste which would impact dilution flows River Discharge System Changes in discharge flow or velocity Changes in discharge temperature or thermal plume Changes in discharge chemical composition Based on information contained in the evaluation of cooling towei and service water performance parameters, and review ofinformation cantained in the Environmental Report - Operating License Stage and FES relative to impacts associated with the above systems, the following information is provided.
2.22.2 Summary of Evaluation River Water Intake. Service Water. and Circulating Water System The proposed power uprate will result in an increase in cooling tower duty of approximately 381 MMBtu/hr over the current operating condition, with a corresponding increase in evaporation, makeup, and cooling tower blowdown temperature. Cooling tower flowrate does not change as a result of power uprate. Cooling tower drift, which is a function of flowrate, also does not change. A slight increase in heat load to the service water system may also occur. Original design cooling tower performance parameters, upon which the conclusions of the FES are based, have changed since the plant began operation due to inability to achieve expected performance levels. This discrepancy has been evaluated and determined not to be significant relative to environmental impact.
Since the information in the Final Environmental Statement no longer reflects current cooling tower operation, the changes in cooling tower parameters associated with power uprate have been compared to values in the FES and the current operating condition in order to assess emironmental impact. The proposed power uprate will result in an increase in cooling tower blowdown temperature of approximately 0.2 *F over the current operating condition. In order to evaluate the environmental impact of the increased blowdown temperature, thermal balance calculations were performed comparing the current cooling tower operation with the uprated condition at the design 78 'F wet bulb temperature utilized in the FES. The results of this comparison, including the original FES values, are summarized in Table 2.22-1. These calculations were conducted at conditions consistent with the FES calculations (78 *F wet bulb temperature, service water temperature of 94.5 F, normal 2 unit service water flou of 35,800 gpm, cooling tower blowdown flow of 10,200 gpm and river flow of 1,210 cfs) to support comparison with FES conclusions relative to thermal impact. He resulting change in discharge temperature associated with power uprate increases the BOP UPRATE LICENSING REPORT 80 FNP - UNITS 1 AND 2
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service water discharge temperature an additional 0.06 F above the current operating condition with a resulting increase in river temperature ofless than 0.004 'F at these severe conditions of extreme wet bulb temperature, maximum recorded temperature and low river flow. Operation of Farley Nuclear plant at the uprated power condition will produce an increase in river temperature of approximately 0.56 'F above ambient river temperature during extreme temperature and flow l
conditions. The FES concluded that the approximately 0.5 'F increase in river temperature associated with operation of Farley Nuclear Plant at extreme temperature and flow conditions did not result in significant adverse environmental impact.
Based on the above, the additional heat load to the Chattahoochee River associated with power i
uprate does not significantly impact the conclusions of the FES relative to thermal impact. The FES l
j concluded that the original analysis conducted under extreme temperature and flow conditions was conservative and protective of water quality standards. The minimal increase in river temperature j
associated with power uprate is also conservative and protective of water quality standards. As discussed in the FES, adequate mixing occurs such that the size of the thermal plume is acceptably j
small. This conclusion remains valid for power uprate.
i An additional conservatism relative to thermal impacts, beyond those discussed above, exists in the i
operating methodology for the cooling towers. This methodology will also be utilized after power uprate. The above evaluation investigated operation of the cooling towers at 3.5 cycles of concentration which is consistent with the FES conditions and determined that no significant adverse environmental impact occurred. After implementation of power uprate, it is anticipated that the i
cooling towers will continue to be operat:d at 10-12 cycles of concentration since the low hardness j'
characteristic of surface water withdrawn from the Chattahoochee River should not significantly change. The significantly lower blowdown flow (74 kpa rather than 10,200 gpm) will produce a reduction in final discharge temperature with a corresponding reduction ofincrease in river temperature. As such, actual thermal impacts are even less than those discussed above.
In addition to the FES, the thermal impact associated with power uprate was evaluated relative to the 7
Farley Nuclear Plant NPDES permit. The Alabama Department of Environmental Management issued NPDES Permit No. AL0024619 to Farley Nuclear Plant. The permit, which was renewed in 1995, contains no limits for temperature. This is based on previous permit monitoring relative to temperature and a study conducted in 1990 which confirmed, under extreme temperature and flow conditions, that the thermal discharge from Farley Nuclear Plant did not result in significant adverse environmental impact and did not warrant numerical permit limits. No additional monitoring requirements or other changes relative to the NPDES permit are required as, a result of power uprate.
Service water intake flow will increase slightly in response to the necessary increase in cooling tower makeup. Ultimately, this increase will require an increase in withdrawal of service water from the service water pond and a corresponding increase in river water withdrawal to supply the pond. Tins slight increase (of approximately 2,032 gpm over the current operating condition) falls well within the 90,000 gpm maximum withdrawal rate evaluated in the FES. Similarly, impingement and entrainment concerns associated with river water intake canal velocity were previously evaluated at the 90,000 gpm maximum discussed above, and are not affected by power uprate. The amount of water supplied to the cooling towers as makeup will increase slightly over the current value due to increased evaporation associated with power uprate. This small increase in consumptive use (3.3 cfs BOP UPRATE LICENSING REPORT 81 FNP - UNITS 1 AND 2
during worst case conditions) is less that 0.3 % of historical low flow and is not significant relative to environmental impact on the Chattahoochee River. In addition, the slight increase in service water temperature due to increased heat load has also been reviewed. This increase will be bounded by no more than 6 % of the current 8.5 *F heat load added by the service water system and will have negligible impact on final discharge temperature.
With the exception of the parameters discussed above, the operating parameters evaluated with regard to potential for environmental impact associated with power uprate either retain the same values as the original values evaluated in the Final Environmental Statement or are bounded by those values.
Other Systems The evaluation also considered the flowrate required by the liquid radwast; system due to the proposed power level increase. No significant change in liquid radwaste qumities or activity levels which would increase the required radwaste dilution flow are expected.
No significant change in groundwater withdrawal required to supply the sanitary water system or fire protection system will result from power uprate.
2.22.3 Summary ofConclusions Based on the above evaluation, the plant operating parameters impacted by the proposed power uprate do not result in significant adverse environmental impact. The Final Environmental Statement concluded that no significant environmental impact would result from operation of Farley Nuclear Plant. This conclusion remains valid for the proposed power uprate. In accordance with the above evaluation, it can be concluded that no significant environmental impact will result from the proposed NSSS power level increase from 2660 MWt to 2785 MWt.
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TABLE 2.22-1 2
ENVIRONMENTAL EVALUATION PARAMETERS I
i Parameter Existing Uprate Value Change FES Value 4
Value Cooling Tower Makeup 17,077 gpm 18,093 gpm 1,016 gpm 17,800 l
bpm i
See Note 1
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Cooling Tower Flow 692,900 gpm 692,900 gpm 0gpm 635,000 SPm Cooling Tower Evaporation 12,808 gpm 13,570 gpm 762 gpm i1,340 bpm See Note 2 Cooling Tower Drift 1,386 gpm 1,386 gpm 0 gpm 1,270 gpm See Note 3 Consumptive Water Use 14,194 gpm 14,956 gpm 762 gpm 12,700 bpm Cooling Tower Blowdown 3,951 gpm 4,238 gpm 287 gpm 5,200 gpm See Note 4 Cooling Tower Blowdown Temp 96.4 "F 96.6
- F 0.2 *F See Note 5 m
Cooling Tower Chemistry See Note 6 River Water Withdrawal 67,504 gpm 69,536 gpm 2,032 gpm 90,000 SPm i
See Note 7 Intake Canal Velocity See Note 7 Discharge Flow Rate See Note 7 Discharge Temp Change 95.04 *F 95.10 *F 0.06 *F 89*F River Temp Change 0.559 "F 0.561*F 0.004*F 0.5 *F Groundwater Withdrawal See Note 8 NOTES
- 1. The Final Environmental Statement considered a cooling tower flowrate of 635,000 gpm and evaporation and drift losses of 12,700 gpm per unit. The FNP circulating water system currently operates at a flow rate of 692,000 gpm as a result of pump modifications. The current 692,000 gpm flowrate will not change as a result of uprate. No change in drift over the current operating condition will occur. The change in evaporation associated with uprate does produce a slight change in makeup requirements and in consumptive water use. This increase of 2,032 gpm for two unit withdrawal is enveloped by the 90,000 gpm rate considered in the FES ( See Note 7).
Consumptive use is discussed in Note 2 below.
2.
The change in evaporation associated with uprate does produce a small increase in consumptive water use. This small increase of approximately 1,524 gpm (3.4 cfs) for two unit operation is less than 0.3 percent of historical low river flow and is insignificant the relative to the impact on O
consumptive water use discussed in the FES.
BOP UPRATE LICENSING REPORT 83 FNP-UNITS 1 AND 2
O (NOTES to Table 22.2-1 continred) 3.
Since circulating water flow does not change, no change in cooling tower drift will occur as a result of uprate.
4.
Cooling tower blowdown flow is actually less than the blowdown flow value used in the FES.
B!owdown flow will not significantly impact the discharge flow rate. Change in discharge flow rate is enveloped by the criteria upon which the conclusions of the FES are based.
- 5. The approximately 0.2 'F increase in cooling tower blowdown temperature does not result in a
[
significant increase in service water discharge temperature (0.06 'F) with a corresponding 'F increase in river temperature. This increase is not significant relative to the 0.5 'F increase evaluated in the FES.
6.
No changes in cooling tower chemistry will result from uprate. Makeup and blowdown will be adjusted accordingly to maintain acceptable cycles of concentration.
- 7. The slight increase in river water withdrawal to compensate for increase in cooling tower makeup is not significant and is enveloped by the 90,000 gpm maximum withdrawal criteria upon which the conclusions of the FES are based. No significant increase in intake canal velocity will result from the slight increase in withdrawal.
O 8.
There will be na significant increase in groundwater withdrawal associated with power uprate.
O BOP UPRATE LICENSING REPORT 84 FNP - UNITS I AND 2
3.0 Conclusion A detailed review of systems, structures, and components not covered by the NSSS Licensing Report has been completed. Along with the implementation of the plant modifications described in Table 2.1 1, the effects of changing rated thermal power from 2652 MWt to 2775 MWt has been evaluated for BOP systems, structures and components which may be impacted. These systems, structures, and components have been found to be bounded by current design criteria and analyses, or the impacts on their function and operation have been analyzed or evaluated to have no significant impact on the function or operation. Evaluation of programmatic issues such as ALARA, Environmental Qualification, Environmental Protection, etc. has also determined that there is no significant impact due to operation at uprated conditions. Based on the results presented in this report, it is concluded that a rated thermal power of 2775 MWt is achievable with no significant impact on the design or licensing basis of FNP.
BOP UPRATE LICENSING REPORT 85 FNP-UNITS 1 AND 2
4.0 References 2.1
_WCAP-14223, Farley Nuclear Plant Units 1 and 2 Power Uprate Project NSSS Licensing Report, January 1997.
7.1 Wyle Laboratory MSSV test reports 48516/QP-ll69 (Unit 1/1987) and 42539/QP-7294 (Unit 2/1992), Alabama Power, Farley Nuclear Plant, Dothan, AL.
10.1 Letter to the NRC dated 7/15/91,
Subject:
VANTAGE-5 Fuel Design Amendment; and NRC letter dated 12/30/91,
Subject:
Tech. Spec. Amendments 91 and 84 13.1 EPRI/NAl 8907-09, Rev. 3, " GOTHIC Containment Analysis Qualification Report," Version 5.
13.2 WCAP-14013. " Joseph M. Farley Nuclear Station, Units I and 2 Main Steam Valve Room Temperature Response to Superheated Steam," March,1994.
16.1 WCAP-11611," Methodology for Elimination of the Containment Spray Additive," March 1988.
22.1 Joseph M. Farley Nuclear Plant Environmental Repon - Operating License Stage 22.2 Joseph M. Farley Nuclear Plant Final Environmental Statement (NUREG 0727) 22.3 Joseph M. Farley Nuclear Plant Environmental Protection Plan (EPP), Appendix B to Facility Operating Licenses NPF -2 and NPF-8 l
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