ML20043C178
| ML20043C178 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/29/1990 |
| From: | Hairston W GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20043C179 | List: |
| References | |
| ELV-01536, NUDOCS 9006040182 | |
| Download: ML20043C178 (26) | |
Text
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George Power Company
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May 29, 1990 vee wmme" W. G. Hairston, hl Senior Vice President Ndear Operatons ELV-01536 0339 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gontlemen:
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION in accordance with the provisions of 10 CTR 50.90 and 10 CFR 50.59, Georgia
. Power Company (GPC) hereby pro)oses to amend the Vogtle Electric Generating Plant (VEGP) Units 1 and 2 Tecanical Specifications, Appendix A to Operating Licenses NPF-68 and NPF-81.
The proposed revision to Technical Specification tables 2.2-1 and 3.3-3 will revise the steam generator water level instrumentation setpoints in accordance with modifications to be implemented during the next refueling outage for each unit. The modifications will relocate the level of the instrument itne lower tap from 438" to 333" above the top of the tubesheet of each steam generator.
The upper taps will remain at 566". This will increase the span of the level instrumentation from 128" to 233". The wider range for the level instruments will allow low-low level reactor trip and Engineered Safety Features Actuation System setpoints to be lowered. This will provide additional operating margin for avoidance of inadvertent reactor trips during normally expected changes in steam generator level.
The changes will be implemented first on VEGP Unit 2 during the refueling outage scheduled to begin in September, 1990. The changes will be implemented on VEGP Unit 1 at the refueling outage scheduled in the fall of 1991. The Technical Specification change will provide the appropriate setpoints, allowable values, and data for instrumentation utilizing either a lower tap at 333" or a lower tap at 438". Following completion of the modification on Unit 1, a separate Technical Specification change will be requested for removal of the values associated with the lower tap elevation of 438".
Since the initial modification for VEGP Unit 2 will be made during the refueling outage which is scheduled to begin in September of 1990, GPC requests that this Technical Specification change be approved by September 1,1990.
Enclosure 1 provides a description of the proposed change and the basis for the change request, i i
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U. S. Nuclear Regulatory Commission ELV-01536 Paae Two ,
Enclosure 2 provides the basis for a determination that the proposed change does not involve a significant hazards consideration. This also includes a brief discussion of the effects of the revised setpoints on LOCA and non-LOCA analyses described in the Final Safety Analysis Report.
Enclosure 3 provides instructions for incorporating the proposed change into the Technical Specifications. The proposed revised pages are also provided in Enclosure 3.
In accordance with 10 CFR 50.91, the designated state official will be sent a copy of this letter and all enclosures.
Mr. W. G. Hairston, 111 states that he is a Senior Vice President of Georgia Power Company and is authorized to execute this oath on behalf of Georgia Power Company and that, to the best of his knowledge and belief, the facts set forth in this letter and enclosuics are true.
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GEORGIA POWER COMPANY Py: id Mr<<08%
W. G. Hairston, 111 Sworn to and subscribed before me this day of N in d , 1990.
d' rYkwA.M&h Notar.) Public "waxuamen,,,
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Enclosures:
- 1. Basis for Proposed Change
- 2. 10 CFR 50.92 Evaluation
- 3. Instructions for Incorporation and Revised Pages xc (see next page) l l
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V. S. Nuclear Regulatory Commission ELV-01536 Paae Three -
I xc: Georoin Power Compan_y Mr. C. K. McCoy Mr. G. Bockhold, Jr.
Mr. R. M. Odom Mr. P. D. Rushton NORMS Southern Company Services Mr. L. B. Long U. S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Adminstrator Mr. T. A. Reed, Licensing Project Manager, NRR Mr. R. F. Aiello, Senior Resident Inspector, Vogtle State of Georaia l-Mr. J. L. Ledbetter, Commissioner, Departmtnt of Natural Resources 1
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ENCLOSVRE 1 I V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION BASIS FOR PROPOSED _. CHANGE Proposed Chance 1
This proposed change will add a second set of values for steam generator low-low and high-high instrumentation in Tables 2.2-1 and 3.3-3. The changed values are given on the revised Technical Specification pages, which are included with enclosure 3. The low-low setpoint will change from 18.5% of the old instrument span to 37.8% of the new instrument span, and the high-high setpoint will change ;
from 78% of the old instrument span to 86% of the new instrument span. This represents an increase in the margin between tt.e low-low setpoint and the
! nominal o>erating level of about 20 inches, and an increase in the margin between tie high-high setpoint and the nominal operating level of about 9 inches. The second set of values will be inserted in parenthesis following the current values for Total Allowance, Z, Trip Setpoint, and Allowable Value. The following footnote will be associated with each of these entries in the table:
"The value stated inside parenthesis is for instrumentation that has the lower tap at elevation 333"; the value stated outside the parenthesis is for instrumentation that has the lower tap at elevation 438"."
Basis The current design of the steam generators for the Vogtle Electric Generating Plant (VEGP) provides taps for these instruments at elevations 566" and 438".
Therefore, the associated instruments can only measure water levels in the 128" span between the two taps. The low-low signal is used to initiate a reactor I trip signal and to initiate the motor driven and steam driven auxiliary l feedwater pumps. The high-high signal initiates a turbine trip and feedwater l isolation. The Technical Specifications express these setpoints as a percentage L of instrument s)an. The narrow span associated with the tap locations in I conjunction wit 1 allowances for instrument error and uncertainty results in a narrow range of allowable level operation. . In the past, changes in steam generator water level associated with norW plant operation have resulted in unnecessary trip signals. In order to reduce the possibility of future trips, j Georgia Power Company has elected to extend the span of the level instruments by lowering the lower tap from 438" to 333". This allows a wider margin to both l the low-low setpoint and the high-high setpoint.
Westinghouse has reviewed each of the accident and transient analyses described in the FSAR that could be affected by the steam generator high-high or low-low instrument setpoints. The analyses that could be affected by these changes have been either evaluated or reanalyzed. The results of these studies are described in Enclosure 2.
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ENCLOSURE 1 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION BASIS FOR PROPOSED CHANGE Basis (continued)
GPC expects to install the lower taps and associated instruments for all of the VEGP Unit 2 steam generators at its first refueling outage which is scheduled to begin in September, 1990. The modification will be implemented for VEGP Unit I at its next refueling outage which is scheduled for the Fall of 1991. During the time between these two refueling outages, the Technical Specifications will be required to include different values for the two units. To account for this, the proposed Technical Specification change has been structured such that it will provide the values associated with either instrument tap location.
Following the completion of the modification on both units, a change to the Technical Specifications will be made to delete the footnote and the values associated with the previous tap location.
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i ENCLOSURE 2 V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Pursuant to 10 CFR 50.92, Georgia Power Company (GPC) has evaluated the attached proposed amendment and has determined that operation of the facility in accordance with the proposed amendment would not involve significant hazards considerations.
Backaround Georgia Power Company (GPC) will move the lower tap for the steam generator level instrumentation from elevation 438" to 333". The current span of the instrument restricts the low-low setpoint to no lower than elevation 438". The actual setpoint is currently set slightly higher (18.5% of instrument span) to account for instrument uncertainty. The lower tap will allow the setpoint to be revised to a lower level. The lower level will correspond to a tri) set)oint of 37.8% of the revised instrument span. In a similar manner, the hig1-hig1 setpoint will be revised to 86% of the new span. Although the physical location of the upper tap is not being changed, the setpoint will be modified to account for flow effects on the level system due to the new location of the lower tap in a high flow velocity region. The flow effects due to the new location of the lower tap were confirmed by tests conducted on a VEGP steam generator during the second refueling outage of VEGP unit 1.
In order to evaluate the effects of revised low-low and high-high steam generator level setpointe, Westinghouse reanalyzed or evaluated the necessary transients and accidents using the revised levels associated with the new setpoints and tap locations. The results of the Westinghouse analyses are described in the following section.
Analysis The steam generator lower level tap relocation program at Vogtle Units 1 and 2 involves moving the low level tap from above the transition cone (elevation 438" above the top of the tubesheet) to below the transition cone (elevation 333"),
thus increasing the narrow range (NR) operating span from 128" to 233". Refer to Figure 1 for a schematic presentation of the modification. This expansion of the acceptable water level operating range will minimize the potential for inadvertent steam generator level-related trips.
Relocation of the low level tap modifies the plant response to any transient which takes credit for the steam generator low-low level reactor trip as analyzed in the FSAR. The reactor trip instrumentation setpoint defined in the Technical Specifications for low-low steam generator level is currently set at greater than or equal to 18.5% of narrow range span, taking into account system and sensor uncertainties. This trip setpoint and allowable value will be changed. The high-high level setpoint and the nominal operating water level will also change as a result of compensation for velocity head characteristics.
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ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued)
Several design basis and regulatory criteria must be evaluated in order to properly assess the effects of level tap relocation on the safe operation of the Vogtle units. These include assumptions which are conservatively made in the safety analyses presented in the FSAR for such parameters as reactor trip setpoints, nominal steam generator mass, and operator action times.
Although the level tap relocation program involves a change to the Technical Specifications, the following evaluations provide the basis for the determination that no significant hazards considerations are involved and for demonstrating continued safe operation of the Vogtle units.
1.0 Non-LOCA Evaluations Vogtle Units 1 and 2 are similar plants which share the same accident analyses. Both of the units have Series F steam generators, all of which will be modified to increase the steam generator level Narrow Range Span (NRS) by lowering the lower level tap from 438" above the top of the tube sheet to 333" above the top of the tube sheet. The high-high level trip setpoint, the low-low level trip setpoint, and the nominal operating level will also be altered to support the new NRS. Table 1 identifies the analysis limits for the current NRS configuration and the revised configuration for the Vogtle unita' steam generators.
The following evaluations have been prepared for non-LOCA accidents to justify the relocation of the lower level tap on the Vogtle units' steam generators. The non-LOCA safety analyses assumptions affected by the steam generator modification include the initial steam generator water level, the steam generator water level low-low reactor trip, and the steam generator water level high-high turbine trip /feedwater isolation function. The following sections provide discussions for each of the non-LOCA events discussed in the FSAR.
1.1 Transients Not Reouirina Reanalysis The following transients were not reanalyzed since either they are not affected by changes in the above mentioned safety analysis assumptions or any changes will not adversely affect the results of the analyses.
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ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE '
STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION l
Analysis (continued) 1 l
1.1.1 Mass and Energy Release for Postulated Secondary System Pipe ,
Ruptures Inside Containment (FSAR Section 6.2.1.4) and Outside i Containment.
The steamline break mass / energy releases for breaks inside containment are generated to ensure that the peak containment pressure and temperature limits are not exceeded. The mass / energy release data is primarily dependent upon secondary side parameters such as the break size, the initial steam generator inventory, steam pressure, auxiliary feedwater flow and feedwater temperature. An increase in the steam generator inventory will result in more limiting mass / energy data. The mass / energy release data currently presented in the FSAR is based upon the current ;
Series F steam generator level program. The revised steam generator nominal level will result in a slight decrease in the '
initial steam generator mass since the revised nominal level (485" plus, a maximum velocity head error of 26" minus a maximum instrument bias of 11") is less than the current nominal level of 502 inches. Thus, no reanalysis is required and the current mass / energy release data presented in the FSAR remains bounding.
For steamline breaks outside containment, which determine adverse environmental conditions for equipment qualification, revised mass energy release calculations have been reviewed to confirm that the limiting temperature profi'ie will not be exceeded.
1.1.2 Feedwater System Malfunctions that Result in a Decrease in .
Feedwater Temperature (FSAR Section 15.1.1) l-l This ANS Condition Il event is bounded by "Feedwater System l
Malfunctions That Result in an Increase in Feedwater Flow" (15.1.2). The DNB design basis is met and the conclusions of the FSAR remain valid.
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e ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued) 1.1.3 Feedwater System Malfunctions That Result in an Increase in Feedwater Flow (FSAR Section 15.1.2)
For this ANS Condition 11 event, cases are analyzed for both full power and zero power conditions. The full aower case was reanalyzed because of the change in the hig1-high level trip setpoint, and this analysis is presented in Section 1.2.1. The l
current analysis for the zero power case demonstrates that the reactivity insertion rate is less than that for " Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low-Power Startup Condition" (15.4.1). Therefore, no reanalysis is required for the zero power case and the conclusions of the FSAR remain valid.
1.1.4 Excessive Increase in Secondary Steam Flow (FSAR Section 15.1.3) ;
For this ANS Condition Il event, cases are analyzed at beginning and end of life conditions both with and without automatic red control. Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation j functions are not challenged in this event and the steam generator '
mass remains relatively constant throughout the event, it is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.5 Inadvertent Opening of a Steam Generator Relief or Safety Valve j (FSAR Section 15.1.4) !
i This ANS Condition II event is bounded by " Steam System Piping Failure" (15.1.5). The DNB design basis is met and the l l
conclusions of the FSAR remain valid. l L
1.1.6 Steam System Piping Failure (FSAR Section 15.1.5) ;
For this ANS Condition IV event, the ANS Condition II criterion of !
meeting the DNBR limit is applied. The analyses are performed at !
zero power and the results are primarily dependent upon secondary !
side parameters such as the break size, steam pressure, auxiliary l l feedwater flow, and feedwater temperature.
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9 ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued)
Since the core response is not sensitive to the initial steam generator mass, the transient is not adversely affected by the lower initial steam generator mass resulting from the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.7 Steam Pressure Regulator Malfunction or Failure That Results in Decreasing Steam Flow (FSAR Section 15.2.1)
There are no steam pressure regulators in the Vogtle units whose failure 'or malfunction could cause a steam flow transient.
Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.8 Loss of External Electrical Load (FSAR Section 15.2.2)
This ANS Condition 11 event is bounded by " Turbine Trip" (15.2.3).
The DNB design basis is met and the conclusions of the FSAR remain valid.
1.1.9 Turbine Trip (FSAR .iection 15.2.3)
For this ANS Condition 11 event, cases are analyzed at beginning and end of life conditions both with and without pressurizer control. All four cases are terminated by a high pressurizer pressure reactor trip. Since the low-low water level reactor trip and high-high water level turbine trip / isolation functions are not challenged, the analysis is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.10 Inadvertent Closure of Main Steam Isolation Valves (FSAR Section 15.2.4)
This ANS Condition 11 event is bounded by " Turbine Trip" (15.2.3).
The DNB design basis is met and the conclusions of the FSAR remain valid.
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ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATIOR Analysis (continuedl 1.1.11 Loss of Condenser Vacuum and Other Events Resulting in Turbine Trip (FSAR Section 15.2.5)
This ANS Condition 11 event is bounded by " Turbine Trip" (15.2.3).
The DNB design basis is met and the conclusions of the FSAR remain valid.
1.1.12 Partial loss of Forced Reactor Coolant Flow (FSAR Section 15.3.1)
For this ANS Condition 11 event, the transient is terminated by a low RCS loop flow reactor tri). Since the steam generator low-low water level reactor trip and ligh-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.13 Complete loss of Forced Reactor Coolant Flow (FSAR Section 15.3.2)
For this ANS Condition 11 event, the transient is terminated by an undervoltage or underfrequency reactor trip. Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.14 Reactor Coolant Pump Shaft Seizure and Locked Rotor, Rods-in-DNB (FSAR Section 15.3.3)
For reactor coolant pump shaft seizure, the ANS Condition IV event criteria include showing that peak design pressures are not exceeded and that the cladding at the " hot spot" in the core remains intact. Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by the steam generator modification.
Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
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4 ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued) 1 For locked rotor, rods-in-DNB, the event is analyzed to determine '
the percentage of fuel rods that experience DNB. Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.15 Reactor Coolant Pump Shaft Break (FSAR Section 15.3.4)
This ANS Condition IV event is bounded by " Reactor Coolant Pump Shaft Seizure" (FSAR Section 15.3.3). Therefore, the conclusions of the FSAR remain valid.
1.1.16 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Subcritical or Low-Power Startup Condition (FSAR Section 15.4.1)
For this ANS Condition 11 event, the analysis is performed at zero power conditions. A rapid reactivity addition results from the withdrawal of a bank of rods. Because of the fast nature of this event, the secondary side is not modeled. Therefore, no i reanalysis is required and the conclusions of the FSAR remain valid.
1.1.17 Uncontrclied Rod Cluster Control Assembly Bank Withdrawal at Power
, (FSAR Section 15.4.2)
For this ANS Condition 11 event, various power levels and reactivity insertion rates for both minimum and maximum reactivity feedback are analyzed. The transients are terminated by an overtemperature delta-T or high neutron flux reactor trip. Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively const tnt throughout the event, it is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
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ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued) 1.1.18 Rod Cluster Control Assembly Misalignment (FSAR Section 15.4.3)
For the events presented in this section of the FSAR, the DNBR criterion is applied. Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant throughout the event, they are not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.19 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (FSAR Section 15.4.4)
For this ANS Condition II event, the DNBR criterion is applied.
Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain i valid.
1.1.20 Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant (FSAR Section15.4.6)
This ANS Condition II event is analyzed to show that adequate time exists for operator action to terminate a dilution event prior to a loss of shutdown margin. Any changes to the secondary side l steam generator level setpoints do not affect the determination of the time available for operator action. With respect to the DNBR criterion, the at power cases (Modes 1 and 2) are bounded by the
" Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power" (15.4.2). Therefore, no reanalysis is required and the conclusions of the FSAR remain valid, t
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ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued) 1.1.21 Inadvertent Loading and Operation of a Fuel ?J46:bly in an Improper Position (FSAR Section 15.4.7)
This ANS Condition III event is a stead.$-state power distribution calculation which does not model the secondary side. Therefore, ,
the changes associated with the steam generator modification do not affect this transient. ;
1.1.22 Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSAR Section 15.4.8)
This is a Condition IV event and is analyzed to show that the fuel melt is less than 10% at the " hot spot" in the core and that peak clad temperature limits are not exceeded. A rapid reactivity addition results from the ejection of an RCCA. Because of the fast nature of this event, the secondary side is not modeled.
Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.23 Inadvertent Operation of Emergency Core Cooling System During Power Operation (FSAR Section 15.5.1) for this ANS Ccadition II event, the transient is initiated by a spurious safety injection signal. The injection of borated water drives the nuclear power and RCS temperature down. Since the J steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challeged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.1.24 Chemical and Volume Control System Malfunction That Increases Reactor Coolant Inventory (FSAR Section 15.5.2)
I This ANS Condition 11 event is bounded by " Chemical and Volume Control System Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant" (15.4.6). Therefore, no reanalysis is required and the conclusions of the FSAR remain valid. :
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I ENCLOSURE 2 (CONT'D)
VDGTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE
, STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued) 1.1.25 Inadvertent Opening of a Pressurizer Safety or Relief Valve (FSAR Section 15.6.1)
For this ANS Condition 11 event, the transient is terminated by an overtemperature delta-T reactor trip. Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant i
throughout the event, it is not affected by the steam generator modification. Therefore, no reanalysis is required and the conclusions of the FSAR remain valid.
1.2 Transients Reauirina Reanalysis The following transients were reanalyzed since they are affected by changes in the steam generator low-low water level reactor trip, high-high water level reactor trip, and nominal water level.
Feedwater System Malfunctions That Result in an increase in 1.2.1 Feedwater Flow (FSAR Section 15.1.2)
This ANS Condition 11 event is analyzed at full power and at zero !
l power. The zero power case is examined to demonstrate that the reactivity insertion rate is less than that for " Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Suberitical or Low-Power Startup Condition" (15.4.1). The reactivity insertion for that event is greater than for the increase in feedwater flow, therefore, the DNB criterion is met for the zero power case. The full power case was reanalyzed to demonstrate that the DNB limit is not exceeded. The addition of excessive feedwater will cause an increase in core power by decreasing reactor coolant temperature. The overpower /overtemperature protection prevents any power increase that could lead to DNB. The peak fuel cladding temperature is also prevented from increasing significantly above its initial value by the high neutron flux reactor trip. The continuous addition of excessive feedwater is prevented by the steam generator high-high water level.
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ENCLOSURE 2 (CONT'D)
L V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued)
Method of Analysis Only the full power case needed to be reanalyzed to support the steam generator modification (see Section 1.1.3). All the analysis assumptions are in accordance with the FSAR with the exception of the initial steam generator level and the high-high level trip setpoint which was assumed to be 100% of the Narrow Range Span.
Results and Conclusions The transient results show that with the change in the initial steam generator level and the high-high level trip setpoint, the DNBR does not fall below the limit value and the peak fuel cladding temperature does not increase significantly above its initial value during the transient.
1.2.2 Loss of Non-Emergency AC Power to the Station Auxiliary / Loss of Normal Feedwater (FSAR Section 15.2.6/15.2.7) l These ANS Condition 11 events are analyzed to demonstrate that adequate heat removal capability exists to remove core decay heat and stored energy following reactor trip. This is ensured by showing that there is no overpressurization of the primary or secondary side and that pressurizer filling does not occur. A reduction in the steam generator low-low level reactor trip setpoint due to the level tap relocation minimizes the amount of mass available to remove the core decay heat and stored energy following reactor trip, resulting in a more severe transient.
Method of Analysis The analysis assumptions included auxiliary feedwater flow of 510 gpm delivered to two steam generators at a temperature of 130 0F.
To maximize the pressurizer filling, the pressurizer power-operated relief valves and pressurizer spray were assumed to operate. The steam generator low-low water level setpoint was assumed to be at 30.2% of the revised Narrow Range Span. In addition, a 10% steam generator tube plugging level was assumed.
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ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued)
Results and Conclusions The transient results show that with the changes discussed above, including the initial steam generator level and the low-low level setpoint, the capacity of the auxiliary feedwater system is adequate to provide sufficient heat removal from the RCS following reactor trip. The criterion that the pressurizer does not fill is met. For the case without power, the results verify the natural '
circulation capacity of the RCS to provide sufficient heat removal capability to prevent fuel or clad damage following reactor coolant pump coastdown.
L2.3 Feedwater System Pipe Break (FSAR Section 15.2.8)
This ANS Condition IV event is analyzed to demonstrate that the peak primary and secondary side pressures do not exceed allowable limits and that the core decay heat removal is adequate. A reduction in the steam generator low-low level associated with the level tap relocation minimizes the amount of mass available following reactor trip to remove the core decay heat and stored l energy, resulting in a more severe transient.
Method of Analysis An auxiliary feedwater flow of 510 gpm delivered to three steam l generators at a temperature of 130 0F was assumed. The initial l
power level was assumed to be 3636 MWT. A conservative core >
residual heat generation model based on the 1979 version of l ANS-5.1 was used. The steam generator low-low water level was ;
L assumed to be at 16% NRS. This value is lower than that used in :
" Loss of Non-Emergency AC Power to the Station Auxiliaries" (15.2.6) and " Loss of Normal Feedwater" (15.2.7) to account for adverse environmental errors. In addition, a 10% steam generator tube plugging level was assumed.
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ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued)
Results and Conclusions The transient results show that with the changes discussed above, including the initial steam generator level and the low-low level setpoint, the capacity of the auxiliary feedwater system is adequate to provide sufficient heat removal from the RCS to prevent overpressurization of the RCS or the secondary side.
The reactor coolant remains subcooled, assuring that the core remains covered with water. For the case without offsite power, the results verify the natural circulation capacity of the RCS to provide sufficient heat removal capability to prevent RCS or secondary overpressurization and fuel or clad damage following reactor coolant pump coastdown.
1.3 ,Non-LOCA Conclusion Based upon the analyses and evaluations presented, the Vogtle Unit 1 and Unit 2 steam generator level tap relocation program can be implemented without violating any of the conclusions of the FSAR non-LOCA analyses.
The technical specification changes required to support the revised analyses are presented in Enclosure 3.
2.0 LOCA Evaluation The effects of the relocation of the steam generator low level tap have been considered for the current LOCA safety analyses presented in the Vogtle FSAR. An evaluation of the effects on the LOCA and LOCA-related analyses is presented below.
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a ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued) 2.1 Large Break LOCA (FSAR Chapter 15.6.5)
Large break LOCA analyses performed to satisfy the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50, including 1981 Evaluation Model analyses such as the one presented in the FSAR, do not model trip setpoints associated with steam generator level, nor do they explicitly model the secondary side volume / mass, etc. Trips modeled in the Westinghouse large break LOCA model are the pressurizer pressure and containment pressure trips. Thus, a change in the steam generator level tap and associated trip setpoints and initial water level will not have any effect on the large break LOCA analysis currently in the FSAR.
2.2 Small Break LOCA (FSAR Chapter 15.6.5)
The Vogtle small break LOCA analysis presented in the FSAR is a WFLASH Evaluation Model analysis. Small break LOCA analyses performed to satisfy the requiremants of 10 CFR 50.46 and Appendix K of 10 CFR 50, including Vogtle's WFLASH analysis, do not model trips associated with steam generator level. Trips modeled in the Westinghouse small break LOCA model are the pressurizer low pressure trips. Thus, a change in the steam generator level trip setpoints will not have any effect on the small break LOCA analysis presented in the FSAR. However, changes in the initial nominal water level may affect the small break LOCA results.
This evaluation was conservatively performed for a decrease of 10" in nominal steam generator level which is sufficient to envelope the expected decrease in actual nominal level and the shift due to the 100% aower velocity head correction. The steam generators act as heat sin (s for a significant portion of a small break transient. Therefore, a lower initial steam generator level will tend to reduce the amount of primary-to-secondary heat transfer during the transient.
An existing sensitivity study of the effects of steam generator level on peak clad temperature as a result of a small break LOCA was reviewed. The study varied only initial steam generator level. The resulting Peak Clad Temperture (PCT) penalty was evaluated to te 110F, The cumulative PCT values for both plants remain well under the 10 CFR 50.46 limit of 22000F. Therefore, the steam generator level tap modification is acceptable for Small Break LOCA.
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m u 1,. .:
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ENCLOStM 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE ;
STEAM GENERATOR LEVEL INSTRUMENTATION 1 10 CFR 50.92 EVALUATCii Analysis (continued)
]
'i 2.3 toCA Blowdown Reactor Vessel and RCS Loop Forces (FSAR Chapter 3.9)
Uculations per& sad to determine the hydraulic forcing functions on the ,
reactor vessel, usul internals and the RCS loop piping as the result of a '
hypothetical L6G typically calculate the peak forces to occur within the l first 500 millisuunds. This is well before the generation of typical trip 1 l signals from steam generator low level and much less than the typical- RCS ,
- l. loop transit time. As a result, the steam generator is not modeled in a ;
hydraulic for::ing functions analysis. Therefore, changes to the steam i
generator trip setpoints and initial operating level associated with the steam generator level tap modification will not have any effect on the LOCA l hydraulic forcing functions used to demonstrate the integrity of the reactor vessel, vessel internals or RCS loop piping. !
2.4 Hot Leg Switchover of the ECCS to Prevent Potential Boron Precipation (FSAR Chapter 6.3.2.8,6.3.3.3) j The hot leg switchover time is determined for inclusion in emergency !
procedures to ensure no boron precipitation in the reactor vessel following l-boiling-in the core. The predicted time is dependent on power level, and 1 the RCS, RWST, and accumulator water volumes and boron concentrations. The '
steam generator level tap modification design does not change any of these :
parameters. Thus, the current hot leg switchover time appearing in the l FSAR will not be affected.
2.5 Post-LOCA Long Term Core Cooling (Associated with FSAR Chapter 15.6.5) !
The Westinghnse evaluation model commitment established by WCAP-8339,
" Westinghouse ECCS Evaluation Model - Summary," is that the reactor will <
remain shutdown by borated ECCS water residing in the RCS/ sump after a LOCA. Calculations performed to determine the containment sump post-LOCA mass and boron concentration are dependent upon the initial mass and boron i concentration of sources of water that will reside in the containment sump l- post-LOCA. These calculations are independent of assumptions on secondary L_ side conditions or feedback from secondary side trips. Thus, the steam generator level tap modification design will not affect the containment sump boron concentration post-LOCA.
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- h. A ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALVATIM Analysis (continued) 2.6 Rod Ejection Mass Release (FSAR Table 15.4.8-2)
The current FSAR rod ejection mass release calculation for the Vogtle Units was performed with the WFLASH Evaluation Model.
The rod ejection mass release transient is reasonably similar to a small
' break LOCA transient. The only key difference is the size and location of the break. In the existing steam generator level sensitivity analysis that was described in the small break LOCA section, the break flows and steam generator safety valve flow that were predicted during the reduced steam generator level case transient were equal to or slightly less than those for the base steam generator level case. Based on this observation, it is concluded that the existing Vogtle rod ejection mass releases also remain bounding for the proposed steam generator level tap modification design.
2.7 LOCA Conclusion The effects of the steam generator lower level tap modification and associated changes to the trip setpoints ard nominal operating water levels for Vogtle have been evaluated for the LOCA events identified. The only significant change that affects the LOCA events is the initial nominal steam generator operating' level. A decrease of 10" in level was-evaluated to conservatively bound the revised nominal operating level. In each case the applicable regulatory or design limit remains satisfied. The small break LOCA event was assessed an 110F PCT penalty which was easily absorbed by existing margin to the 22000F 10 CFR 50.46 limit.
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.=
te E.y ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE
-STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Analysis (continued) 3.0 Steam Generator Tube Rupture (SGTR) Evaluation In qualitative terms, the relocation of the steam generator NR lower level tap at the Vogtle units should be beneficial in the recovery from a SGTR event. Specifically, since the tap location will be over 100 inches below the previous location, narrow range level will be expected to stay on scale following reactor trip or return to the narrow range scale shortly after reactor trip. Previous SGTR analyses for Vogtle have shown that the steam generator narrow range level falls outside the narrow range scale and remains offscale low for an extended period of time; thus, the operators have no steam generator narrow range level indication. The availability of a steam generator narrow range level indication should enable the operators to diagnose a SGTR event in a more timely manner since the steam generator with the ruptured tube should exhibit a greater narrow range level than the.
other three intact steam generators. Consequently, primary to secondary break flow through the ruptured tube would be expected to be terminated more quickly following event initiation, thereby resulting in less integrated break flow from the RCS to the secondary side of the ruptured steam generator for the duration of the SGTR event.
It is noted that the steam generator narrow range lower level tap relocation is expected to aid plant operators in diagnosing a SGTR event sooner via steam generator narrow range level indication. With respect to the SGTR margin to overfill and offsite dose analyses, an earlier diagnosis of a SGTR event should result in earlier break flow termination and ultimately an increase in the margin to steam generator overfill al.d reduction in the offsite radiation doses.
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j..,
ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALVATION Analysis (continued) 4.0 Radioloaical Assessment Radiological consequences for offsite doses are based on mass release data for such transients as rod ejection, steam generator tube rupture and large break LOCA. The evaluations presented above have determined that no additional mass release from that which has previously been calculated in the FSAR will result from the level tap relocation. Therefore, the parameters, assumptions and subsequent results predicted in the FSAR for dose consequences remain bounding for plant operation with the level tap relocation.
5.0 Containment Intearity The level tap relocation does not affect the limiting conditions assumed in the containment integrity analyses and does not affect the normal plant operating parameters, system actuations, accident mitigating capabilities or assumptions important to the containment analyses. Therefore, the conclusions presented in the FSAR remain valid with respect to the containment.
Results The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The replacement transmitters are environmentally qualified and have reliability equivalent to the current transmitters such that a feedwater malfunction event resulting from transmitter error will be no more prnbable than already assumed in the FSAR.
The consequences of previously e:Cuated accidents have either been re-evaluated or re-analyzed assuming the revuod instrumentation and setpoints. The results were either bounded by previous analyses or insignificantly changed. In all cases, the results were within the applicable design and safety criteria and the conclusion of the FSAR remain valid.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The level instrumentation will utilize lower taps; otherwise, its functions remain unchanged. Since the change does not introduce a new failure mode, a new or different kind of accident is not introduced.
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3r . := y, ENCLOSURE 2 (CONT'D)
V0GTLE ELECTRIC GENERATING PLANT TECHNICAL SPECIFICATION CHANGE STEAM GENERATOR LEVEL INSTRUMENTATION 10 CFR 50.92 EVALUATION Results (continued)
The proposed change does not involve a significant reduction in a margin of safety. The effects of the change on transient and accident analyses have been evaluated and it has been determined that there are no significant changes in the results of previous analyses presented in the FSAR. The applicable design and safety criteria continue to be met, the conclusions reached in the FSAR remain valid and safety margins are not changed.
Conclusion Based on-the preceding analysis, GPC has determined that the proposed change to the Technical Specifications does not involve-a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident fror, any previously evaluated or involve a significant reduction in a margin of safety. Therefore, GPC concludes that the proposed change meets the requirements of 10 CFR 50.92(c) and does not involve significant hazards considerations.
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Table 1 Current and Modified NRS Configuration for Vogtle Units 1 and 2 4 (Non-LOCA Analysis Limits)
Current Analysis Modified Analysis
-limits limits Narrow Range Span 438 to'566 inches 333 to 566 inches
~ Nominal Level- 502 inches 485 inchesl (at full power) (50% NRS) (65.0% NRS)
Low-Low Level 438 inches 370 inches 2 (0% NRS) (16% NRS) 403 inches 3 (30% NRS)
High-High-Level 544.4 inches 566 inches (90.7% NRS) (100% NRS) 1This nominal level analysis limit does not include the velocity head effect.
2Low Low level analysis limit assumed for "Feedwater System Pipe Break" (15.2.8); environmental errors are included.
3Low-Low level analysis limit assumed for " Loss of Non-Emergency AC Power to the Station Auxiliaries" (15.2.6) and " Loss of Normal Feedwater" (15.2.7); no environmental errors are included.
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- C " durrent Design Propossd -
Steam O'utlet Modification
- 7. To Turbine
. Secondary g Moisture Separator
/
Primary Moisture Upper Tap 566 in. 4 Separator Feedwater Rings Manway ,
4 2, 4, ' ,
3 8 +
8 8 ,
"J" Nozzles M
3 '. : $' . ' . .: "Jd Anti Vibration
, i i 3 Bars Feedwater intet j- -
Lower Tap 438 in.
i 5 [ Tube Bundle Lower Tap 333 in.
,/
" Quatrefoil" Tube Support
.- Plates 1
1 Tubesheet
,l'M >T Top of Tubesheet 0 in. grj y..
Manway Coo: ant intet Figure 1 Steam Generator Low Level Tan Relocation
. . - - _ -