ML20065E384
| ML20065E384 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/30/1994 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20065E385 | List: |
| References | |
| NUDOCS 9404080162 | |
| Download: ML20065E384 (100) | |
Text
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L Hl4 Innessee vaney Aum@ost ONe BOGM Decatur. Mabama 35609 MAR 3 01"4 s
TVA-BFN-TS-318 10 CFE 50.90 U.S. Nuclear Regulatory Commission 'l ATTN: Document Control Desk -!
Washington, D.C. 20555 j i
Gentlemen:
In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN).'- UNITS 1, 2, AND 3 -
p TECHNICAL SPECIFICATION (TS) NO. 318 - ANALOG L TRANSMITTER / TRIP SYSTEM, LEVEL : RL' ACTOR WATER LEVEL SETPOINTS, AND VARIOUS-CALIBRATION FREQUENCIES In accordance with the provisions of 10 CFR 50.4 and 50.90, TVA is submitting a request for an amendment (TS-318) to
- l. licenses DPR-33, DPR-52, and DPR-68 to change the TSs for Units 1, 2, and 3. The proposed change:
b 1. Reflects the installation of an Lnalog Transmitter / Trip l
System (ATTS) on Unit 3, which is similar to the system previously installed on Unit 2.
- 2. Revises the Units 1 and 3 Reactor Vessel Water Level Safety Limit to reflect the analytical limit provided by General Electric. In addition, he Level ' Low Reactor Vessel Water Level setpoint is being revised to provide a more conservative limit. These changes were previously approved to the Unit 2 TSs.
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U.S. Nuclear Regulatory Commission Page 2 HAR 3 01994
- 3. Adds or corrects Unit 2 instrument identifiers to enhance the useability of the TSs.
- 4. Revises calibration frequencies and funct )nal test descriptions for the Unit 2 Reactor High ater Level, Reactor Core Isolation Cooling and High Presrure Coolant Injection Turbine Steam Line High Flow, and Drywell Pressure instrument channels.
- 5. Revises the calibration frequency for the differential .
pressure instrumentation, which actuates.the pressure suppression chamber-reactor building vacuum breakers, in the Units 1, 2, and 3 TSs to reflect current calculations. In addition, tables that specify the minimum number of instrument channels per trip system, function, trip level setting, actions required, remarks, functional test, and instrument check are being added.
- 6. Corrects the capitalization of terms used on the affected Units 1, 2, and 3 TS pages in order to conform with the current TS Definitions section. This part also corrects spelling and capitalization of other words on the same pages.
TVA has determined that there are no significant hazards considerations associated with the proposed change and that the change is exempt from environmental review pursuant to the provisions of 10 CFR 51.22 (c) (9) . The BFN Plant Operations Review Committee and the BFN Nuclear Safety Review Board have reviewed this proposed change and determined that operation of BFN Units 1, 2, and 3 in accordance with the proposed change will not endanger the health and safety of l the public. Additionally, in accordance with 10 CFR 50.91(b) (1) , TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health.
Enclosure 1 to this letter provides the description and evaluation of the proposed change. This includes TVA's evaluation that the proposed change does not involve a significant hazards consideration, and is exempt from environmental review pursuant to the provisions of 10 CFR 51. 2 2 (c) (9) . Enclosure 2 contains copies of the appropriate TS pages from Units 1, 2, and 3 marked-up to show the proposed change. Enclosure 3 forwards the revised TS pages for Units 1, 2, and 3 shich incorporate the proposed change. Enclosure 4 contains the commitment associated with the proposed TS change.
(
l U.S. Nuclear Regulatory Commission Page 3 MAR 3 01994 ,
')
The Unit 3 ATTS modification, reactor vessel water level safety limit change, and Level 1 Low Reactor Vessel Water Level setpoint revision portions of this amendment are needed to support restart of BFN Unit 3. By letter dated ,
December 23, 1993, TVA submitted thr current schedule showing '
TS amendment need dates for.' support of Unit 3 restart. In order to support the restart schedule, TVA requests approval of the enclosed amendment by April 20, 1995. TVA also i requests that.the revised TS be made effective within 30 days 1 of NRC approval. If you have any questions about this change, please telephone me at (205) 729-2636.
Sincern hp--
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1 Pedro Salas Manager of Site Licensing i l
Enclosures cc: See page 4 Subscribed and sworn (q be:? ore me em this 23 l rat day of mRRC6 1994. i f D rm ;
CL h et/ b bh Notary Public My Commission Expires (O 3d \
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i U.S. Nuclear Regulatory Commission l Page 4 MAR 3 0199j Enclosures cc (Enclosures):
American Nuclear Insurers
! Town Center, Suite 300S ,
29 South Main Street !
West Hartford, Connecticut 06107-2445 Mr. W. D. Arndt General Electric Company !
735 Broad Street i Suite 804, James Building Chattanooga, Tennessee 37402 Mr. Johnny Black, Chairman Limestone County Commission 310 Washington Street Athens, Alabama 35611 Mr. R. V. Crlenjak, Project Chief U.S. Nuclear Regulatory Commission Region II 1 101 Marietta Street, NW, Cuite 2000 1 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant ,
Route 12, Box 637 l Athens, Alabama 35611 ]
I Mr. David C. Trimble, Project Manager U.S. Nuclear Regulatory Commission ,
one White Flint, North I 11555 Rockville Pike Rockville, Maryland 20852 Mr. Joseph F. Williams, Project Manager U.S. Nuclear Regulatory Commission .
One White Flint, North -
11555 Rockville Pike Rockville, Maryland 20852 Dr. Donald E. Williamson State Health Officer l
State Department of Public Health State. Office Building .i Montgomery, Alabama 36194 t
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS PERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-318 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE IMP 3X I. DESCRIPTION OF THE PROPOSED CHANGE . . . El-2 II. REASON FOR THE PROPOSED CHANGE . . . . . El-57 III. SAFETY ANALYSIS . . . . . . . . . . . . El-60 IV. NO SIGNIFICANT HAZARDS CONSIDERATION ;
DETERMINATION . . . . . . . . . . . . El-89 ,
V. ENVIRONMENTAL IMPACT CONSIDERATION . . . El-97 VI. REFERENCES . . . . . . . . . . . . . . . El-98 i
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I. DESCRIPTION OF THE-PROPOSED CHANGE-In general, this proposed change to BFN Technical Specifications. consists of six parts. ;
l Part A:. The Unit 3 mechanical pressure and differential .
pressure indicating switches in the Reactor Protection System (RPS) and Emergency Core Cooling System (ECCS) are i
being replaced with an Analog Transmitter / Trip :
System (ATTS). l The ATTS modification includes the replacement of power.
supplies and associated electrical cabling, breakers'and -
fuses. As a result, instrument identifiers ~are being added c and/or revised, the descriptions of the required functional '
testing are being updated, group designators are being~ ;
corrected, notes pertaining to the minimum functional test .
frequency are being amended, the minimum calibration frequencies are being adjusted, and an indicator range is being changed to properly reflect the_new equipment.
The ATTS provides the following system upgrades:
- Reduces the functional tests and calibration frequencies for the primary sensors. f a Decreases the duration and complexity of required testing and calibration of the inputs for safety related parameters.
- Reduces testing and maintenance related scrams.
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- Reduces the number of reportable events related to ,
setpoint drift.
4 In addition, the revised design implements the diversity requirements associated with the Anticipated Transient Without Scram (ATWS) system required by110 CFR 50.62. -
The RPS provides timely protection against the onset and consequences of conditions that threaten the integrities of ,
the fuel barrier (uranium dioxide sealed in cladding) and the nuclear system process barrier. Excessive temperature threatens to perforate the cladding or melt the uranium! a dioxide. Excessive pressure threatens to rupture the nuclear system process barrier. The RPS limits the .
uncontrolled release of radioactive material by terminating-excessive temperature'and pressure increases through the
. initiation of an automatic scram.
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i The RPS includes the motor-generator power supplies with associated control and indicating equipment, sensors, !
relays, bypass circuitry, and switches that cause rapid !
insertion of control rods (scram) to shut down the reactor. ;
It also includes outputs to the process computer system and annunciators. The Reactor Protection System is des 3gned to meet the intent of the IEEE proposed criteria for nuclear power plant protection systems (IEEE-279-1971). A detailed description of the RPS is included in Section 7.2.3 of the Browns Ferry Final Safety Analysis Report (FSAR).
The controls and instrumentation for the ECCS initiate appropriate responses from the various cooling systems so that the fuel is adequately cooled under abnormal or accident conditions. The cooling provided by the systems restricts the release of radioactive materials from the fuel by limiting the extent of fuel damage following situations in which reactor coolant is lost from the nuclear system.
Even after the reactor is shut down from power operation by.
the full insertion of all control rods, heat continues to be generated in the fuel as radioactive fission products decay. An excessive loss of reactor coolant allows the fuel temperature to rise, cladding to melt, and fission products in the fuel to be released. If the temperatures in the reactor rise to a sufficiently high value, a metal (zirconium) water reaction occurs which releases energy.
Such a reaction increases the pressure inside the nuclear i
system and the primary containment. This threatens the integrity of the barriers, which are relied upon to prevent the uncontrolled release of radioactive materials. The controls and instrumentation for the ECCS prevent such a sequence of events by actuating core cooling systems in time to limit fuel temperatures to acceptable levels (less than 2200*F) . A detailed description of the ECCS is included in Section 7.4.3 of the Browns Ferry FSAR.
The ATTS has been installed and successfully operated on BFN Unit 2. The specific differences between the Unit 3 and Unit 2 instrumentation are described in the Safety Analysis section. Similar Technical Specification changes were previously approved for Unit 2 (References 1 ,
through 8). '
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Part B: The Units 1 and 3 reactor vessel water level safety limit is being revised to reflect the analytical limit provided by General Electric and the Level 1 Low Reactor Vessel Water Level setpoint is being revised to provide a more conservative limit.
The safety limits are limits below which the reasonable maintenance of the cladding and primary systems are assured. Exceeding such a limit requires unit shutdown and review by the Nuclear Regulatory Commission before resumption of the unit operation. Operation beyond such a limit, the reactor vessel water level safety limit in this case, may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.
The limiting safety system setting is a setting on instrumentation which initiates the automatic protective action at a level such that the safety limits will not be '
exceeded. The Level 1 Low Reactor Vessel Water Level setpoint is the limiting safety system' setting which is i being revised by this proposed Technical Specification amendment. The region between the safety limit and these settings represent margin with normal operation lying below these settings. The margin has been established'so that with proper operation of the instrumentation the safety limits will never be exceeded.
TVA committed, in the Browns Ferry Nuclear Performance Plan, to ensure that calculations exist to support the safe shutdown basis of Unit 2. During the process of generating these setpoint and accuracy calculations for plant parameters for which no calculational basis could be found, it was determined that the Unit 2 Level 1 Low Reactor Vessel Water Level trip setting was not conservative. based on the new calculation methodology. The Reactor Vessel Water Level Safety Limit and the Level 1 Low Reactor Vessel Water Level setpoint have previously been changed in the BFN Unit 2 Technical Specifications (References 9 through 11). Similar revisions to the Units 1 and 3 Technical Specification are required to reflect the Unit 3 specific calculations and to make the Technical Specifications consistent for all three units.
Part C: For Unit 2, RPS and ECCS instrument identifiers are being added or corrected to enhance useability of the Technical Specifications. Taese changes do not reflect a change in equipment, operation of the associated system, or the safety function of that system.
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Eart D: For Unit'2, Reactor High Water Level, Reactor Core Isolation Cooling (RCIC) and High' Pressure Coolant Injection (HPCI) Turbine Steam Line.High Flow, and Drywell Pressure instrumentation calibration frequencies and functional test descriptions are being revised to-reflect current calculations and-tast methods. These. changes do not reflect a change in equipment, operation of the associated system, or the safety function of'that system.
Part E: For Units 1, 2, and 3, the differential pressure instrumentation, which actuates the pressure suppression chamber-reactor building vacuum breakers, calibration-frequency is being revised. In addition, tables that specify the minimum number of instrument channels per trip system, function, trip level setting, actions required, remarks, functional test, and instrument check are being added.
Automatic vacuum relief devices are used to prevent the primary containment from exceeding the external design pressure. The drywell vacuum relief valves draw air from the pressure suppression chamber. The pressure suppression chamber vacuum relief device draws air from the Reactor Building.
The pressure suppression chamber vacuum relief system consists of two vacuum breakers in series in~each of two lines to atmosphere. One valve is air-operated and is ,
actuated by a differential pressure signal. This valve fails open upon a loss of power. These are the valves being addressed by this Technical Specification change.
The second valve is self-actuating.
Part F: Corrects the capitalization of terms used on the affected Units 1, 2, and 3 TS pages in order to conform with the current TS Definitions section. This part also corrects spelling and capitalization of other words on the same pages.
El-5
The specific proposed' changes to the BFN Technical Specifications are delineated below. The applicability to which unit is specified for every change.
- 1. For Units 1 and 3. Proposed change to Safety Limit 1.1.B, Power Transient. Capitalize the words safety limit in two places as shown below:
Existing Technical Specifications:
"To ensure that the Safety Limits established in Specification 1.1.A are'not exceeded, each required scram shall be initiated by its expected scram signal. The Safety Limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal."
Proposed Technical Specifications:
"To ensure that the SAFETY LIMITS established in Specification 1.1.A are not exceeded, each required scram shall be initiated by its expected scram signal. The SAFETY LIMIT shall be assumed to be exceeded when scram-is accomplished by means other than the' expected scram signal."
- 2. For Units 1 and 3. Proposed change to Safety ,
Limit 1.1.C, Reactor Vessel Water Level. Change the Reactor Vessel Water Level Safety Limit from greater than or equal to 378 inches to greater than or equal to 372.5 inches as shown below:
Existing Technical Specifications:
"Whenever there is irradiated fuel in the reactor vessel, the water level shall be greater than or equal to 378 inches above vesstl zero."
Proposed Technical Specifications:
Whenever there is irradiated fuel in the reactor vessel, the water level shall be greater than or equal to 372.5 inches above vessel zero.
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- 3. For Units 1 and 3. Proposed change to Limiting Safety System Setting 2.1.C.1, Water Level Trip Settings.
Change the Core Spray and Low Pressure Coolant Injection (LPCI) setpoint from greater than or equal to 378 inches to greater than or equal to 398 inches as shown below:
Existing Technical Specifications: .
" Core spray and LPCI actuation -- reactor low water level 2 378 in. above vessel zero" Proposed Technical Specifications:
Core spray and LPCI actuation -- reactor low water level 2 398 in, above vessel zero
- 4. For Units 1 and 3. Proposed change to Limiting Safety System Setting 2.1.C.3, Water Level Trip Settings. i Change the main steam isolation valve closure setpoint from greater than or equal to 378 inches to greater than or equal to 398 inches as shown below:
Existing Technical Specifications:
" Main steam isolation valve closure -- reactor low water level 2 378 in. above vessel zero" Proposed Technical Specifications:
Main steam isolation valve closure -- reactor low water level 2 398 in, above vessel zero El-7
- 5. For Units 1 and 3. Change to Bases 1.1. Revise the Reactor Vessel Water Level Safety Limit from 378 inches to 372.5 inches and delete the reference to the lower reactor low water level trip as shown below:
Existing Bases:
"The safety limit has been established at 378 inches above vessel zero to provide a point which can be monitored and also provide adequate margin to assure sufficient cooling. This point is the lower reactor low water level trip."
Revised Bases:
The safety limit has been established at 372.5 inches above vessel zero to provide a point.which can be monitored and also provide adequate margin to assure sufficient cooling.
- 6. For Units 1 and 3. Addition to the references for Bases 1.1:
- 2. General Electric Document No. EAS-65-0687, Setpoint Determination for Browns Ferry Nuclear Plant, Revision 2.
- 7. For Unit 3 only. Proposed additions to Table 3.1.A, Reactor Protection System (SCRAM) Instrumentation Requirements:
- a. For the High Reactor Pressure trip, insert instrument identifiers: (PIS-3-22AA,BB,C,D).
- b. For the High Drywell Pressure trip, insert instrument identifiers: (PIS-64-56 A-D)..
- c. For the Reactor Low Water Level trip, insert instrument identifiers: (LIS-3-203 A-D).
- d. For the Turbine First Stage Pressure Permissive, insert instrument identifiers: (PIS-1-81A&B, PIS-1-91A&B).
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- 8. For Unit 3 only._ Proposed changes to Table 4.1.A, Reactor Protection System.(SCRAM) Instrumentation ,
Functional Tests Minimum Functional Test Frequencies j for Safety Instr. and Control Circuits:
- a. For the High Reactor Pressure trip, insert instrument identifiers: (PIS-3-22AA, BB, C, D),
insert a reference to Footnote 7 in the Functional Test Column, remove the reference to Footnote 1 in the Minimum Frequency column, and revise the group designator as'shown below:
Existing Technical Specifications:
"Groun (2)
A" Proposed Technical Specifications:
Group (21 B
- b. For the High Drywell Pressure trip, insert instrument identifiers: (PIS-64-56 A-D), insert a reference to Footnote 7 in the Functional Test Column, remove the reference to Footnote 1 in the Minimum Frequency column, and revise the group designator as shown below:
Existing Technical Specifications:
" Group (2)
A" Proposed Technical Specifications:
Group (2)
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- c. For the Reactor Low Water Level trip, insert instrument identifiers: (LIS-3-203 A-D), insert a' reference to Footnote 7 in the Functional Test Column, remove the reference to Footnote'l in the Minimum Frequency column, and revise the group designator as shown below:
Existing Technical Specifications:
"Groun (2)
A" Proposed Technical Specifications:
Group (2)
B
- d. Proposed changes to Table 4.1.A. For the Turbine First Stage Pressure Permissive, insert instrument identifiers: (PIS-1-81A and B, PIS-1-91A and B), insert a~ reference to Footnote 7 in the Functional Test Column, and revise the group designator as shown below:
Existing Technical Specifications:
"Groun (2)
A" Proposed Technical Specifications:
Group (2)
B El-10
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- 9. For Unit 3 only. Proposed changes to Table 4.1.B, Reactor Protection System (SCRAM) Instrumentation Calibration Minimum Calibration Frequencies for Reactor Protection Instrument Channels:
- a. For the High Reactor Pressure trip, insert instrument identifiers: (PIS-3-22AA, BB, C, D) and revise the group designator and minimum calibration frequency as shown below:
(1) Existing Technical Specifications:
"Groun (1)
A" Proposed Technical Specifications:
Group (1)
B (2) Existing Technical Specifications:
" Minimum Frecuency (2)
Every 3 Months" i Proposed Technical Specifications:
Minimum Frecuency (2)
Once/6 Months (9) u I
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- b. For the High Drywell Pressure trip, insert instrument identifiers: '(PIS-64-56 A-D) and revise the group designator and minimum calibration frequency as shown below:
(1) Existing Technical Specifications: ,
" Group (1)
A" Proposed Technical Specifications:
Group (1)
B ,
(2) Existing Technical Specifications:
" Minimum Frecuency (11 Every 3 Months" Proposed Technical Specifications:
Minimum Frecuency (2)
Once/18 Months (9) a El-12
- c. For the Reactor Low Water Level trip, insert instrument identifiers: (LIS-3-203 A-D) and revise the group designator and minimum calibration frequency as shown below:
(1) Existing Technical Specifications:
" Group (1)
- 1. '
Proposed Technical Specifications:
Group (1)
B (2) Existing Technical Specifications:
" Minimum Frecuency (2)
Every 3 "onths" Proposed Technical Specifications:
Minimum Frecuency (2)
Once/18 Months (9)
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- d. For the Turbine First Stage Pressure Permissive, insert instrument identifiers: (PIS-1-81 A&B, R PIS-1-91 A&B) and revise the group designator and ;
minimum calibration frequency as shown below:
(1) Existing Technical Specifications: !
" Group (1)
A" Proposed Technical Specifications:
Group (1) -j B
(2) Existing Technical Specifications:
" Minimum Frecuency (11 Every 6 Months" '
Proposed Technical Specifications:
Minimum Frecuency (2)
Once/18 Months (9) ,
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- 10. For Unit 3 only. Changes to Bases 3.1.
- a. Capitalize the words reactor protection system in two places as shown below:
Existing Bases:
"The reactor protection system automatically initiates a reactor scram to:
... The reactor protection system is made up of two independent trip systems (refer to Section 7.2, FSAR)."
Revised Bases:
The Reactor Protection System automatically initiates a reactor scram to: ... The Reactor Protection System is made up of two independent trip systems (refer to Section 7.2, FSAR).
- b. Capitalize the words limiting conditions for operation and decapitalize the' word inoperable as shown below:
Existing Bases:
"This specification provides the limiting conditions for operation necessary to preserve ... When necessary, one channel may lua made INOPERABLE for brief intervals to conduct required functional tests and calibrations." ,
Revised Bases:
This specification provides the LIMITING CONDITIONS FOR OPERATION necessary to preserve ... When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
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- 11. For Unit 3 only. Changes to Bases 3.1. . Insert the following paragraph as part of the description of the reactor protection system:
The reactor protection trip system is supplied, via a separate bus, by its own high inertia, ac motor-generator set. Alternate power is available to either Reactor Protection System bus from an electrical bus that can receive. standby electrical power. The RPS monitoring system provides an isolation'between nonclass 1E power supply and the class 1E RPS bus. This will ensure that failure of a nonclass-1E-reactor protection power supply will not cause adverse interaction to the class 1E Reactor Protection System.
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- 12. Proposed changes to Table 3.2.A, Primary Containment and Reactor Building Isolation Instrumentation.
- a. For Unit 3 only. For the Reactor Low Water Level Instrument Channel that-hac a trip level setting of 2 538" above vessel zero, insert. instrument identifiers: (LIS-3-203 A-D) and delete the reference to the isolation groups as shown below:
Existing Technical Specifications:
" Remarks
- 1. Below trip setting does the following:
- b. Initiates Primary Containment Isolation (Groups 2, 3, and 6)"
Proposed Technical Specifications:
Remarks
- 1. Below trip setting does the following:
- b. Initiates Primary Containment Isolation
- b. For Units 1 and 3. For the Reactor Low Water Level Instrument Channel that currently has a Trip Level Setting of 2 378" above vessel zero, revise the trip level setting to 2 398" as shown below:
Current Technical Specifications:
Trio Level Settino "2 378" above vessel zero" Proposed Technical Specifications:
Trio Level Settina 2 398" above vessel zero I
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- c. For Unit 3 only. For the Reactor Low Water Level Instrument Channel that currently has a Trip Level Setting of 2 378" above vessel zero, revise the instrument identifiers as shown below:
Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Water Level (LIS-3-56A-D, SW #1)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low Water Level (LIS-3-56A-D)
- d. For Unit 3 only. For the High Drywell Pressure Instrument Channel, revise the instrument identifiers as shown below:
Current Technical Specifications:
Function
" Instrument Channel -
High-Drywell Pressure'(6)
(PS-64-56A-D)"
Proposed Technical Specifications:
Function Instrument Channel -
High Drywell Pressure (6)
(PIS-64-56A-D)
- e. For Unit 3 only. For the Low Pressure Main Steam Line Instrument Channel, insert instrument identifiers: (PIS-1-72, 76, 82, 86)
- f. For Unit 3 only. For the High Flow Main Steam Line Instrument Channel, insert instrument identifiers: (PdIS-1-13A-D, 25A-D, 36A-D, 50A-D)
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- 13. Proposed changes to Table 3.2.B, Instrumentation that Initiates or Controls the Core and Containment Cooling Systems.
- a. For Unit 3 only. For the two Reactor Low Water Level-Instrument Channel entries that have a trip level setting of 2 470" above vessel zero, insert instrument identifiers: (LIS-3-58A-D).
- b. For Units 1 and 3. For the first listing of the Reactor Low Water Level Instrument Channel that has a trip level setting of 2 378" above vessel zero, revise the trip level setting as shown below:
Current Technical Specifications:-
Trio Level Settina "2 378" above vessel zero."
Proposed Technical Specifications:
Trio Level Settina 2 398" above vessel zero.
- c. For Unit 3 only. For the first listing of the' Reactor Low Water Level Instrument Channel that has a trip level setting of 2 378" above vessel zero, revise the instrument identifiers as shown ,
below:
Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Water Level (LIS-3-58A-D, SW #1)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low Water Level (LS-3-58A-D)
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- d. For Units 1 and 3. For the next listing'of.the Reactor Low Water Level Instrument Channel that has a trip level setting of 2 378" above vessel zero,-revise the trip level setting as-shown below:
Current Technical Specifications:
Trio Level Settina "2 378" above vessel zero."
Proposed Technical Specifications:
Trio Level Settina 2 398" above vessel zero.
- e. For Unit 3 only. For the next listing of the Reactor Low Water Level Instrument Channel that-has a trip level setting of 2 378" above vessel zero, revise the instrument identifiers as shown >
below:
Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Water Level (LIS-3-58A-D, SW #2)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low Water Level (LS-3-58A-D) 1 l
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- f. For Unit 3 only. For the Reactor Low Water Level Permissive Instrument Channel, revise the instrument identifiers as shown below:
Current' Technical Specifications:
__ Function
" Instrument Channel -
Reactor Low Water Level Permissive (LIS-3-184 &
185, SW #1)"
Proposed Technical Specifications:
Function Instrum2nt Channel -
Reactor Low Water Level Permissive (LIS-3-184, 185)
- g. For Unit 3 only. For the Reactor Low Water Level Instrument Channel that has a trip level setting of 2 312 5/16" above vessel zero, revise the instrument identifiers as shown'below:
Current Technical Specifications: ,
Function
" Instrument Channel'-
Reactor Low Water Level (LITS-3-52 and 62, SW #1)"
Proposed Technical. Specifications:
Function Instrument Channel - i Reactor Low Water Level (LIS-3-52 and LIS-3-62A) 1
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- h. For Unit 3oonly. For the Drywell High Pressure Instrument Channel with a trip level setting of'l s p s 2.5 psig, revise the instrument identifiers as shown below:
Current Technical Specifications:
Function
" Instrument-Channel -
Drywell High Pressure (PS-64-58 E-H)"
Proposed Technical Specifications:
Function Instrument Channel -
Drywell High Pressure (PIS-64-58 E-H)
- 1. For Unit 3'only. For the first Drywell High Pressure Instrument Channel with a trip level setting of s 2.5 psig, revise the instrument identifiers as shown below:
Current Technical Specifications: ,
Function
" Instrument Channel -
Drywell High Pressure (PS-64-58 A-D, SW #2)"
Proposed Technical Specifications:
Function Instrument Channel -
Drywell High Pressure .,
(PIS-64-58 A-D)
El-22
- j. For Unit 3 only. For the second Drywell High Pressure Instrument Channel with a trip level setting of 5 2.5 psig, revise the instrument identifiers as shown below:
Current Technical Specifications:
Function a
" Instrument Channel -
Drywell High Pressure (PS-64-58A-D, SW #1)"
Proposed Technical Specifications:
Function Instrument Channel -
Drywell High Pressure (PIS-64-58A-D)
- k. For Unit 3 only. For the third Drywe71 High Pressure Instrument Channel with a trip level setting of 5 2.5 psig, revise the instrument identifiers as shown below:
Current Technical Specifications:
Function l
" Instrument Channel -
Drywell High Pressure (PS-64-57A-D)" .
Proposed Technical Specifications:
".'unct ion Instrument Crannel -
Drywell High Pressure ,
(PIS-64-57A-D) i El-23
s 1. For Unit 3 only. For the Reactor Low Pressure .
Instrument channel with a trip level setting.of 450 psig 15, revise the instrument identifiers as shown below:
Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Pressure (PS-3-74 A & B, SW #2)
(PS-68-PS, SW #2)
(PS-68-96, DW #2)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low Pressure (PIS-3-74A & B)
(PIS-68-95, 96)
- m. For Unit 3 only. For the Reactor Low Pressure '
Instrument Channel with a trip level setting of 230 psig i 15, revise the instrument identifiers as shown below:
Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Pressure (PS-3-74 A & B, SW #1)
(PS-68-95, SW #1)
(PS-68-96, SW #1)"
Proposed' Technical Specifications:
Function Instrument Channel -
Reactor Low Pressure (PS-3-74A & B)
(PS-68-95, 96)
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- n. For Unit 3 only. For the Reactor High Water Level Instrument Channel that has a trip level setting of 5 583" above vessel zero and above the trip setting trips the RCIC turbine, insert instrument identifiers: (LIS-3-208A'and LIS-3-208C).
- o. For Unit 3 only. For the RCIC Turbine Steam Line High Flow Instrument Channel, insert instrument identifiers: (PDIS-71-1A and 1B).
- p. For Unit 3 only. For the next Reactor High Water Level Instrument Channel that has a trip level setting of 5 583" above vessel zero, and above the trip setting trips the HPCI turbine, insert instrument identifiers: (LIS-3-208B and LIS-3-208D).
- q. For Unit 3 only. For the HPCI Turbine Steam Line High Flow Instrument Channel, insert instIument.
identifiers: (PDIS-73-1A and 1B)
- r. For Units 1 and 3. For Note 15 in Notes for Table 3.2.B section, revised the vessel low water level setpoint from 2 378" to 2 398" as shown below:
Current Technical Specifications:
- 15. "The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (2 378" above vessel zero) originating in the core spray system trip system."
Proposed Technical Specifications:
- 15. The accident signal is the satisfactory completion of a one-out-of-two taken'twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (2 398" above vessel zero) ,
originating in the core spray system trip system.
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L
- 14. For Unit 3 only. Proposed changes to Table 3.2.F, i Surveillance Instrumentation. -l
- a. For the Reactor Water Level Instrument, revise the instrument identifiers as shown below:
Current Technical Specifications:
Instrument #
"LI-3-46 A LI-3-46 B" Proposed Technical Specifications:
Instrument #
LI-3-58A LI-3-58B b
El-26
. . . . . - , . - . - . . - , - . . - . .-- . . . . .- . ~ . ~ - ..
- b. For the Reactor Pressure Instrument, revise'the- l instrument identifiers and range as shown below: j (1) Current Technical Specifications:
Instrument # ,
y "PI-3-54 f PI-3-61"
-Proposed Technical Specifications:'
Instrument #
PI-3-74A PI-3-74B (2) Current Technical Specifications: :
Type Indication and Range .. ;
" Indicator 0-1500 psig" Proposed-Technical Specifications: <
Type Indication and Ranae ;
Indicator 0-1200 psig *
.s
- 15. For Units 2 and 3.' Proposed correction:to the-spelling of the word instrumentation in Table 3.2.L:
Current
Title:
" Anticipated Transient Without Scram'(ATWS) -
Recirculation Pump Test (RPT) Surveillance.
Instrumenation" Proposed.
Title:
I Anticipated' Transient Without Scram (ATWS) --
Recirculation Pump Test (RPT) Surveillance: i
-Instrumentation l
El-27 2
, *6 r v v
- 16. For Units 2 and 3. Proposed addition to Table 3.2.L, Anticipated Transient Without Scram (ATWS) -
Recirculation Pump Test (RPT) Surveillance Instrumentation.
- a. For the ATWS/RPT Logic Reactor Dome Pressure High Function, insert instrument identifiers:
(PIS-3-204 A-D).
- b. For the Reactor Vessel Level Low Function, insert instrument identifiers: (LS-3-58 Al-D1).
El-28 s: .-
- 17. For Unit 3 only. Proposed changes to Table 4.2.A, Surveillance Requirements for Primary Containment and Reactor Building' Isolation Instrumentation,
- a. For the first listing of the Reactor Low Water Level Instrument Channel, revise the instrument identifiers, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
(1) Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Water Level (LIS-3-203A-D, SW 2-3)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low. Water Level (LIS-3-203A-D)
(2) Current Technical Specifications:
Calibration Frecuency
"(5)"
Proposed Technical Specifications:
Calibration Frecuency once/18 Months (29)
El-29
- b. For the second listing of the Reactor Low Water Level Instrument Channel, revise the instrument identifiers,. insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
(1) Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Water Level (LIS-3-56A-D, SW #1)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low Water Level (LIS-3-56A-D)
(2) Current technical Specifications:
Calibration Freauency "once/3 month" Proposed Technical Specifications:
Calibration Frecuency once/18 months (29)
El-30
i c. For the High Drywell Pressure Instrument Channel, revise the instrument identifiers, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
(1) Current Technical Specifications:
Function
" Instrument Channel -
High Drywell Pressure (PS-64-56A-D)"
Proposed Technical Specifications:
Function Instrument Channel -
High Drywell Pressure (PIS-64-56A-D)
(2) Current Technical Specifications:
Calibration Frecuency
"(5)"
Proposed Technical Specifications:
Calibration Frecuency once/18 Months (29)
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- d. For the Low Pressure Main Steam Line Instrument Channel, insert instrument identifiers:-
(PIS-1-72, 76, 82, 86) and revise the functional test and calibration frequency as shown below:
(1) Current Technical Specifications:
Functional Test "once/3 months (27)"
Proposed Technical Specifications:
Functional Test (28) (27)
(2) Current Technical Specifications:
Calibration Frecuency "once/3 months" Proposed Technical Specifications:
Calibration Frecuency once/18 Months (29) 1 9
El-32
- e. For the High Flow Main Steam Line Instrument Channel, insert instrument identifiers:
(PdIS-1-13A-D, 25A-D, 36A-D, 50A-D) and_ revise the functional test and calibration frequency as shown below:
(1) Current Technical Specifications:
Functional Test "once/3 months (27)"
Proposed Technical Specifications:
Functional Test (28) (27)
(2) Current Technical Specifications:
Calibration Frecuency "once/3 months" Proposed Technical Specifications:
Calibration Frecuency once/18 Months (29)
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e
- 18. Proposed changes to Table'4.2.B, Surveillance Requirements for Instrumentation that Initiate or Control the CSCS.
- a. For Units 2 and 3. For the first listing of the Reactor Low Water Level Instrument Channel, revise the instrument identifiers as shown below:
Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Water Level (LIS-3-58A-D)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low Water Level (LS-3-58A-D, LIS-3-58A-D)
- b. For Unit 3 only. For the first listing of the Reactor Low Water Level Instrument Channel, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
Current Technical Specifications:
Calibration Freauency "once/3 months" Proposed Technical Specifications:
Calibration Precuency once/18 months (29)
El-34 i
- c. For Unit 3 only.- For the second listing of the Reactor Low Water Level Instrument Channel (LIS-3-184 & 185), insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
Current Technical Specifications:
Calibration Frequgncy "once/3 months" Proposed Technical Specifications:
Calibration Frecuency once/18 months (29)
El-35
- d. For Unit 3 only. For.the third listing of the Reactor Low Water Level Instrument Channel, revise the instrument identifiers, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
(1) Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Water. Level .
(LITS-3-52 & 62)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low Water Level (LIS-3-52 & 62A)
(2) Current Technical Specifications:
Calibration Frecuency "once/3 months" Proposed Technical Specifications:
Calibration Freauency once/18 months (29)
El-36
- e. For Unit 3 only. For the first Drywell High Pressure Instrument Channel, revise the instrument identifiers, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
(1) Current Technical Specifications:
Function
" Instrument Channel -
Drywell High Pressure (PS-64-58E-H)"
Proposed Technical Specifications:
Function Instrument Channel -
Drywell High Pressure (PIS-64-58E-H)
(2) Current Technical Specifications:
Calibration Frecuency "once/3 months" Proposed Technical Specifications:
Calibration Freauency once/18 months (29) l l
I El-37 l
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- f. For Unit 3 only. For the second Drywell High Pressure Instrument Channel, revise the instrument identifiers, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
(1) Current Technical Specifications:
Function
" Instrument Channel -
Drywell High Pressure (PS-64-58A-D)"
Proposed Technical Specifications:
Function Instrument Channel -
Drywell High Pressure (PIS-64-58A-D)
(2) Current Technical Specifications:
Calibration Freauency "once/3 months" Proposed Technical Specifications:
Ca_libration Freauency once/18 months (29)
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)
l 1
- g. For Unit 3 only. For the third Drywell High l Pressure Instrument Channel, revise the instrument identifiers, insert a reference'to Footnote 28 to the Functional Test column and !
revise the calibration-frequency as shown below:
(1) Current Technical Specifications:
Function
" Instrument Channel -
Drywell High Pressure (PS-64 -57 A-D) "
Proposed Technical Specifications:
Function _
r Instrument Channel -
Drywell High Pressure (PIS-64-57A-D)
(2) Current Technical Specifications:
Calibration Frecuency "once/3 months" Proposed Technical Specifications:
Calibration Freauency once/18 months (29) f El-39 I
+- . - - - - _ - - - - _ - _ _ _ _ .
- h. For Unit 3 only. For the Reactor Low Pressure Instrument Channel, revise the instrument identifiers, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
(1) Current Technical Specifications:
Function
" Instrument Channel -
Reactor Low Pressure (PS-3-74A & B)
(PS-68-95) '
(PS-68-96)"
Proposed Technical Specifications:
Function Instrument Channel -
Reactor Low Pressure (PIS-3-74A&B, PS-3-74A&B)
(PIS-68-95, PS-68-95) ,
(PIS-68-96, PS-68-96)
(2) Current Technical Specifications:
Calibration Freaupncy "once/3 months" Proposed Technical Specifications:
Calibration Frecuency once/6 months (29)
- 1. For Units 2 and 3. .For-the Reactor High-Water Level Instrument Channel, insert instrument identifiers: (LIS-3-208A-D).
El-40
- j. For Unit 2 only. For the Reactor _High Water Level Instrument Channel,' insert a reference to Footnote 27 to the Functional Test column and revise the calibration frequency as shown below:
Current Technical Specifications:
Calibration Freauency "Once/3 months" Proposed Technical Specifications:
Calibration Frecuency Once/18 months (28)
- k. For Unit 3 only. For the Reactor High Water Level Instrument Channel, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
Current Technical Specifications:
Calibration Frecuency "once/3 months" Proposed Technical Specifications:
Calibration Freauency once/18-months (29)
El-41
J 4
. 1. For Unit.2 only. For the RCIC Turbine Steam Line High Flow Instrument Channel, insert a reference ,
to Footnote 27 to the Functional Test. column ~and '
revise the calibration frequency as shown below: i Current Technical Specifications:
Calibration Frecuency '
"Once/3 months" Proposed Tec,hnical Specifications:
Calibration Frecuency Once/18 months (28) '
- m. For Unit 3 only. For the RCIC Turbine 1 Steam Line '
High Flow Instrument Channel,-insert a-reference-to Footnote 28 to the Functional Test columniand revise the calibration frequency as shown below:
Current Technical Specifications:
Calibration Frecuency "once/3 months" .
Proposed Technical Specifications: _
Calibration-Frecuency once/18. months (29) q I
El-42
- n. For Unit 2 only. For the HPCI Turbine Steam Line High Flow Instrument Channel, insert a reference to Footnote 27 to the Functional Test column and revise the calibration frequency as shown below:
Current Technical Specifications:
Calibration Frecuency "Once/3 months" Proposed Technical Specifications:
Calibration Frecuency Once/18 months (28)
- o. For Unit 3 only. For the HPCI Turbine Steam Line High Flow Instrument Channel, insert a reference to Footnote 28 to the Functional Test column and revise the calibration frequency as shown below:
Current Technical Specifications:
Calibration Freauency "once/3 months" Proposed Technical Specifications:
Calibration Freauency once/18 months (29)
'l H
~
q El-43 L
- 19. Proposed changes to Table 4.2.F, Minimum Test and Calibration Frequency'for Surveillance Instrumentation.
- a. For. Unit 3 only. For the Reactor Water Level Instrument Channel, insert instrument identifiers: (LI-3-58A&B).
- b. For Units 2 and 3. For the Reactor Water Level Instrument Channel, revise the calibration frequency as shown below:
Current Technical Specifications:
___ Calibration Frecuency "Once/6 months" Proposed Technical Specifications:
Calibration Frecuency once/18 months
- c. For Unit 3 only. For the Reactor Pressure Instrument Channel, insert instrument identifiers: (PI-3-74A&B).
- d. For Unit 3 only. For the third Drywell Pressure-Instrument Channel, revise the instrument identifiers as shown below:
Current Technical Specifications:
Instrument Channel "11) Drywell Pressure (PS-64-58B)"
Proposed Technical Specifications:
Instrument Channel
- 11) Drywell Pressure (PIS-64-58A)
El-44
l
\
- e. For Units 2 and.3. For the third Drywell l Pressure Instrument Channel, revise the calibration frequency as shown below:
Current Technical Specifications:
Calibration F_tecuency "Once/6 months" Proposed Technical Specifications:
Calibration Frecuency Once/18 months
- 20. For Units 1 and 3. Changes to Bases 3.2.
- a. Capitalize the words primary containment integrity as shown below:
Existing Bases:
"Such instrumentation must be available whenever primary containment integrity is required."
Revised Bases:
.Such instrumentation must be available whenever PRIMARY CONTAINMENT INTEGRITY is required.
El-45 l
1 l
- b. Revise the react or vessel- water low level trip ' l that closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor ,
Water Sample Valves from 378 inches _to greater than or equal to 398 inches as shown below: >
Existing Bases: ,
"The low water level. instrumentation set to trip at 378 inches above vessel zero.
(Table 3.2.B) closes the Main Steam-- :
1 solation Valves, the Main-Steam Line Drain.
Valves, and the Reactor. Water Sample Valves i (Group 1)."
Revised Bases: ,
The low water level instrumentation set to f trip at 2 398 inches above vessel zero (Table 3.2.B) closes the Main Steam-Isolation Valves, the Main Steam Line Drain' valves, and the Reactor Water Sample Valves (Group 1).
- c. Revise the reactor vessel water low level trip ,
that Initiates the LPCI, Cores Spray Pumps, contributes to ADS initiation, and starts the diesel generators from 378 inches to greater than or equal'to 398 inches as shown below:-
Existing Bases:
"The low reactor water level instrumentation that is set'to trip when .'
reactor water level is 378 inches above vessel zero (Table 3.2.B)_ initiates the LPCI, Core Spray Pumps,. contributes to ADS initiation, and starts the diesel ,
generators." l
.i Revised Bases: ;
The low reactor. water level'in'strumentationL that is set to trip when' reactor-water- '
level is 2 398' inches'above: vessel zero ,
(Table 3.2.B) initiates the LPCI, Core-Spray Pumps, contributes to' ADS' initiation, and-starts'the diesel generators.
'l
-1 El-46 l
d
- 21. For Units 1, 2, and 3. Proposed change to Limiting Conditions For Operation (LCO) 3.7.A.3.a, Primary Containment - Pressure Suppression Chamber - Reactor Building Vacuum Breakers. Capitalize the words primary containment integrity and revise the LCO to refer to Table 3.7.A for the pressure suppression chamber-reactor building vacuum breakers actuation setpoint as shown below:
Existing LCO:
"Except as specified in 3.7.A.3.b below, two pressure suppression chamber-reactor building vacuum breakers shall be OPERABLE at all times when primary containment integrity is required.
The setpoint of the differential pressure instrumentation which actuates the pressure suppression chamber-reactor building vacuum breakers shall be 0.5 psid."
Revised LCO:
Except as specified in 3.7.A.3.b below, two pressure suppression chamber-reactor building vacuum breakers shall be OPERABLE at all times when PRIMARY CONTAINMENT INTEGRICY is required.
The setpoint of the differential pressure instrumentation which actuates the pressure suppression chamber-reactor building vacuum breakers shall be per Table 3.7.A.
El-47
1
- 22. For Units 1, 2, and 3. Proposed change to Surveillance Requirement 4.7.A.3.a, Primary Containment - Pressure Suppression Chamber - Reactor i Building Vacuum Breakers. Revise the Surveillance Requirement to refer to Table 4.7.A for the pressure suppression chamber-reactor building vacuum breakers calibration frequency as shown below:
Existing Surveillance Requirement:
"The pressure suppression chamber-reactor building vacuum breakers shall be exercised.in accordance with Specification 1.0.MM, and the associated instrumentation including.setpoint shall be functionally tested for proper operation each three months."
Revised Surveillance Requirement:
The pressure suppression chamber-reactor building vacuum breakers shall be exercised in accordance with Specification 1.0.MM, and the associated instrumentation including setpoint shall be functionally tested for proper operation per Table 4.7.A.
- 23. For Units 1, 2, and 3, Insert a new Table 3.7.A, Instrumentation for Containment System, as shown below:
b t
El-48
TABLE 3.7.A INSTRUMENTATION FOR CONTAINMENT SYSTEMS Minimum No.
Operable Per Trio System Function Trio Level Setting Action Remarks 2 Instrument Channel - 0.5 psid
- Actuates the pressure Pressure suppression suppression chamber-reactor chamber-reactor building building vacuum breakers.
vacuum breakers (PdIS-64-20, 21)
Footnote:
- - Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, declare the system or component inoperable.
El-49 i
. . . - . . . . . - . . _ . m __. , . _ ~- . . ___ _ _ _ _ _ __
- 24. For Units 1, 2, and 3, Insert a new Table 4.7.A, Containment System Instrumentation Surveillance Requirements, as shown below:
El-50
TABLE 4.7.A CONTAINMENT SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Function Functional Test Calibration Instrument Check Instrument Channel - Once/ month
- Once/18 months
- None.
Pressure suppression chamber-reactor building vacuum breakers (PdIS-64-20, 21)
Footnotes:
- - Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify OPERABILITY of the trip and alarm functions.
- - Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable-range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the level setting.
L i El-51 l
L ._ -._.m _-
- 25. Change to LCO'3.7.A.3.b, Primary Containment -
Pressure Suppression Chamber - Reactor Building Vacuum Breakers,
- a. For Units 1 and 3. Decapitalize the_ word inoperable and capitalize the words primary containment integrity as shown below: ,
Existing LCO:
"From and after the date that one of the prassure suppression chamber-reactor building vacuum breakers is made or found to be INOPERABLE for any reason, reactor-operation is permissible only during the-succeeding seven days,.provided that the repair procedure does not violate primary containment integrity."
Revised LCO:
From and after the date that one of the pressure suppression chamber-reactor building vacuum breakers is made or found to be inoperable for any reason, reactor operation is permissible only-during.the succeeding seven days, provided that the repair procedure does not violate PRIMARY CONTAINMENT INTEGRITY.
l El-52
}_
- b. For Unit 2. Capitalize the words primary containment integrity as shown below:
Existing LCO:
~
"From and after the date that one of the- ,
pressure suppression chamber-reactor building vacuum breakers is made or found :
to be inoperable for any reason, reactor ~ -
operation is permissible only during:the?
succeeding seven days, provided that the repair procedure does not violate primary '
containment integrity."
Revised LCO:
From and after the date that'one of the pressure suppression chamber-reactor-building vacuum breakers is made or found to be inoperable for any reason, reactor operation is permissible only during the ;
succeeding seven days, provided that the i repair procedure does-not violate PRIMARY CONTAINMENT INTEGRITY.
t i
9 El-53
..i
- 26. Change to Surveillance Requirement 4.7.A.4.b, Primary Containment - Drywell-Pressure Suppression Chamber Vacuum Breakers,
- a. For Units 1 and 3. Decapitalize the word inoperable in two places and capitalize the word operability as shown below:
Existing Surveillance Requirement:
"When it is determined that two vacuum breakers are INOPERABLE for opening at a time when operability is required, all other vacuum breaker valves shall be exercised immediately and every 15 days thereafter until the INOPERABLE valve has been returned to normal service."
Revised Surveillance Requirement:
When it is determined that two vacuum breakers are inoperable for opening at a time when OPERABILITY is required, all other vacuum breaker valves shall be exercised immediately and every 15 days thereafter until the inoperable valve has been returned to normal service.
'i I
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El-54
- b. For Unit 2. Capitalize the word operability as shown below:
Existing Surveillance Requirement: ,
"When it is determined that two vacuum breakers are inoperable for opening at a time when operability is required, all '
other vacuum breaker valves shall be exercised immediately and every 15 days thereafter until the inoperable valve has been returned to normal service."
Revised Surveillance Requirement:
When it is determined that two vacuum breakers are inoperable for opening at a time when OPERABILITY is required, all other vacuum breaker valves shall be exercised immediately and every 15 days thereafter until the inoperable valve has been returned to normal service.
El-55 l
1 l
I
- 27. Changes to Bases 3.7.D/4.7.D, Primary Containment Isolation Valves,
- a. For Unit 1. Correct the spelling of the word l specifications as shown below:
Existing Bases:
"The procedures are subject to the change control provisions for plant procedures in the administrative controls section of the Technical Specifictions."
Revised Bases:
The procedures are subject to the change control provisions for plant procedures in the administrative controls section of the Technical Specifications.
- b. For Units 1 and 3. Revise the reactor vessel water low level process line isolation trip from 378 inches to greater than or equal to 398 inches in two places as shown below:
Existing Bases:
" Group 1 - Process lines are isolated by reactor vessel low water level (378") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems. ... The reactor water-sample line valves isolate only_on reactor low water level at 378" or main steam line high radiation."
Revised Bases:
Groun 1 - Process lines are_ isolated by reactor vessel low water level (2 398") in order ,to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems. ... The reactor water sample line valves isolate only on reactor low water level at 2 398" or main steam line high radiation."
El-56
i II. REASON FOR THE PROPOSED CHANGE 1
These changes are proposed for the following reasons: l 1
Part A: The Unit 3 mechanical pressure and differential pressure indicating switches in the Reactor Protection System (RPS) and Emergency Core Cooling System (ECCS) are being replaced with an Analog Transmitter / Trip System (ATTS) similar to that previously installed on Unit 2. This modification also includes the replacement of power supp3.ies and associated electrical cabling,. breakers and fuses. The differences between the Unit 3 and Unit 2 instrumentation are described in the Safety Analysis section. The ATTS modification provides for continuous monitoring of critical parameters in additie to performing basic logic trip operations. The ATTS moditication provides the following system upgrades:
- Reduces the functional tests and calibration frequencies for the primary sensors.
- Decreases the duration and complexity of required testing and calibra? ion of the inputs for safety related parameters.
- Reduces testing and maintenance related scrams.
- Reduces the number of reportable events related to setpoint drift.
In addition, the revised design implements the diversity requirements associated with the Anticipated Transient Without Scram (ATWS) system required by 10 CFR 50.62.
These changes are described in the Description of the Proposed Change section and included as part of Items 7a-d, 8a-d, 9a-d, 11, 12a, 12c-f, 13a, 13c, 13e-r, 14a-b, 16a-b, 17a -e , 18a-i, 18k, 18m, 180, and 19a-d.
El-57
Part B: The Units 1 and 3 Reactor Vessel Water Level Safety Limit is being revised to reflect the analytical limit provided by General' Electric and the Level 1 Low Reactor Vessel Water Level setpoint is being revised to provide a more conservative limit.
The Reactor Vessel Water Level Safety Limit is provided by the Nuclear Steam System Supply vendor, General Electric, and is the design basis limit that should not be exceeded.
A limiting safety system setting, such as the Level 1 low reactor vessel water level, is a setting on instrumentation which initiates the. automatic protective action at a level such that the safety limits will not be exceeded. The.
region between the safety limit and these settings represent margin. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.
These changes are described in the Description of the Proposed Change section and included as part of Items 2, 3, 4, 5, 6, 12b, 13b, 13d, 13r, 20b-c, and 27b.
Part C: For Unit 2, instrument identifiers are being added or corrected to enhance the useability of the Technical >
Specifications. These changes are described in the
~
Description of the Proposed Change section and included as part of Items 16a-b, 18a, and 181.
Part D: For Unit 2, reactor high water level, Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) turbine steam line high flow, and drywell pressure instrumentation calibration frequencies and functional test descriptions are being revised to reflect current calculations and test methods. This instrumentation was upgraded as part of the installation of the ATTS, which was installed prior to Cycle 6 operation.
These changes are described in the Description of the Proposed Change section and included as part of Items 18j, 181, 18n, and 19d.
El-58
.~. . . . . - . - - .
1 .
'4 Part E: For Units 1, 2, and 3, the differential pressure ,
instrumentation, which actuates the pressure suppression ,
chamber-reactor building vacuum breakers, calibration frequency is being revised to reflect current Unit'2 and 3 calculations. -The Unit 1 change is based'on the similarity ~
of this system and. equipment between the three units. In- .
addition, tables that specify the minimum number of j instrument channels per trip system, function, trip. level i
setting, actions required, remarks, functional test, and instrument check are being added in order to be consistent. ,
with the treatment of'other electronic trip circuitry inL the Technical Specifications. These changes are described-in the Description of the Proposed Change section and included as part of Items 21, 22, 23, and 24.
Part F: The capitalization of terms used on the affected Units 1, 2, and 3 Technical Specification pages is being .
corrected in order to conform with the current TS Definitions section. Spelling and capitalization of other words is also being corrected on the same'pages. These ,
changes are described in the Description of the Proposed Change section and included as part of Items 1, 10a-b, 15, 16, 20a, 22, 25a-b, 26a-b, and 27a.
The Unit 3 ATTS modification, reactor vessel water level safety limit change, and Level 1-Low Reactor Vessel' Water:
Level setpoint revision portions of this amendment are
~
needed to support restart of BFN Unit 3. By letter-dated December 23, 1993, TVA submitted.the current schedule'
- showing TS amendment need dates for support of Unit 3 restart. In order to support the restart schedule, TVA' requests approval of the enclosed amendment by April 20, 1995.
t t
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III. SAFETY ANALYSIS These changes are justified for the following reasons:
Part A: The Unit 3 Barton, Barksdale, Static-O-Ring, and Yarway instruments in the Reactor Protection System (RPS) and Emergency Core Cooling System (ECCS) are being replaced with an environmentally qualified Analog Transmitter / Trip System (ATTS). The ATTS modifications provide for continuous monitoring of critical parameters in addition to performing basic logic trip operations. This system, including the new instrumentation, was designed to meet or exceed the requirements in General Electric NEDO-21617-A, Analog Transmitter / Trip Unit System for Engineered Safeguard Sensor Trip Units. Overall, the ATTS provides the following system upgrades:
- Reduces the functional tests and calibration frequencies for the primary sensors.
- Decreases the duration and complexity or required testing and calibration of the inputs for safety related parameters.
- Reduces testing and maintenance related scrams.
- Reduces the number of reportable events related to setpoint drift. '
I El-60 j l
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The generic NRC approval of the ATTS is included in General Electric NEDO-21617-A, Analog Transmitter / Trip Unit System for Engineered Safeguard Sensor Trip Units. This Licensing Topical Report was approved by NRC letter, dated June 27, 1978, to the General Electric Company. This letter states that: "The staff does not intendito repeat its review of this topical report when it appears as a ;
reference in specific license applications, except to assure that the report is applicable to the specific plants involved." In Section 5.4 of NEDO-21617-A, each applicant that uses this topical report as licensing basis was required to provide the following plant specific information to NRC:
Plant SDecific Information Recuired Section 5.4.1 - Specific Instrument Loops Supply information for each instrument loop that will be converted to the analog sensor system as identified below:
- 1. Variable name >
- 2. Part number of device being deleted
- 3. System involved
- 4. The engineered safeguards division
- 5. Model number and vendor of the transmitter or RTD TVA Resoonse The requested information is provided below.
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BECTION 5.4.1 - SPECIFIC INSTRUMENT LOOPS '
DELETED NEW TRANSMITTER NEW ATU INSTRUMENT TVA LOOP NO. TVA LOOP NO. SYSTEM ;
VARIABLE VENDOR / NO. VENDOR / NO. VENDOR /NO. INPUTS l NAME (LOCATION) (LOCATION) (LOCATION) DIVISION (LOCATION)
Main Steam Line PDIS-1-13A PDT-1-13 A PDIS-1-13 A IA PCIS High Flow Barton 278 Rosemount 1153 Rosemount 710DU (9-15)
(25-56A) (25-56A) (9-83)
PDIS-1-13B PDT-1-138 PDIS 1-13B IB PCIS Barton 278 Rosemount 1153 Rosemount 710DU (9-17)
(25-56A) (25-56A) (9-84)
PDIS-1-13C PDT-1-13C PDIS-1-13C IIA PCIS Barton 278 Rosemount 1153 Rosemount 710DU (9-15)
(25-56A) (25-56A) (9-85)
PDIS-1-13D PDT-1-13D PDIS-1-13D llB PCIS Banon 278 Rosemount i153 Rosemount 710DU (9-17)
(25-56A) (25-56A) (9-86)
Main Steam Line PDIS-1-25 A PDT-1-25A PDIS-1-25A IA PCIS High Flow Barton 278 Rosemount 1153 Rosemount 710DU (9-15)
(25-56A) (25-56A) 9-83)
PDIS-1-25B PDT-1-25 B PDIS-1-25B IB PCIS Barton 278 Rosemount 1153 Ronemount 710DU (9-17)
(25-56A) (25-56A) (9-84)
PDIS-1-25C PDT-1-25C PDIS 1-25C !!A PCIS Barton 278 Rosemount 1153 Rosemount 710DU (9-15)
(25-56A) (25-56A) (985)
PDIS-125D PDT-t-25D PDIS-1-25D IB PCIS Barton 278 Rosemount 1153 Rosemount 710DU (9-17)
(25 56A) (25-56A) (9-86)
Main Steam Line PDIS-1-36A PDT-1-36A PDIS-1-36A lA PCIS High Flow Barton 278 Rosemount 1153 Rosemount 710DU (9 15)
(25-568) (25-568) (9-83)
PDIS-1-36B PDT-t-36B PDIS-1-36B IB PCIS Barton 278 Rosemount !153 Rosemount 710D0 (9-17)
(25-560) (25-56B) (9-84)
PDIS-1-36C PDT-1-36C PDIS-1-36C llA PCIS Barton 278 Rosemount 1153 Rosemount 710DU (9-15)
(25-56B) (25-568) (9 85)
PDIS-1-36D PDT-1-36D PDIS-1-36D IIB PCIS isarton 278 Rosemount 1153 Rosemount 710DU (9-17)
(25-56B) (25-568) (9-86)
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l l DELETED NEW TRANSMITTER NEW ATU l
INSTRUMENT TVA LOOP NO. TVA LOOP NO. SYSTEM VARIABLE VENDOR /NO. VENDOR /NO. VENDOR /NO. INPUTS NAME (LOCATION) (LOCATION) (LOCATION) DIVISION (IX) CATION) i l
Main Steam Line PDIS-1-50 A PDT-1-50A PDIS-1-50A IA PCIS High Flow Barton 278 Rosemount 1153 Rosemount 710DU (9-15)
(25-568) (25-568) (9-83)
PDIS-1-50B PDT-1 -50B PDIS-1-50B IB PCIS Barton 278 Rosemount 1153 Rosemount 710DU (9-17)
(25-56B) (25-56B) (9-84)
PDIS-150C PDT-1-50C PDIS-1-50C llA PCIS l Barton 278 Rosemount 1153 Rosemount 710DU (9-15)
! (25-56B) (25-56B) (9-85)
PDIS-150D PDT-1-50D PDIS-1-50D llB PCIS Barton 278 Rosemount 1153 Rosemount 710DU (9-17)
(25-56B) (25-56B) (9-86)
Main Steam Line PS-1-72 PT-1-72 PIS-1-72 IA PCIS j Low Pressure Barksdale Rosemount 1153 Rosemount 710DU (9-15)
B2T-A12SS (25-112) (9-83)
(25-112)
PS-1-76 FT-1-76 PIS-1-76 IB PCIS Barksdsle Rosemount 1153 Rosemount 710DU (9-17)
B2T-A12SS (25-112) (9-84)
- (25-112) l l
Turbine First PS-1-81 A 17T-1-81 A PIS-1-81 A IIB RPS/RPT Stage Pressure Barksdale Rosemount 1153 Rosemount 710DU (9-17) l Permissive B2T-A12SS (25-111) (9 86)
(25-111)
PS-1-81 B E7T-1-81 B PIS-1-81 B llA RPS/R17T Barksdale Rosemount 1153 Rosemount 710DU (915) l B2T-A12SS (25-111) (9-85) l (25-111) l Main Steam Line PS-1-82 FT-1 82 PIS-I-82 11A PCIS l Low Pressure Parksdale Rosemount 1153 Rosemount 710DU (9-15)
B2T-A12SS (25-l l3C) (9-85) l I (25113C)
PS-1-86 FT-1-86 PIS-1-86 IIB PCIS Barksdale Rosemount !!$3 Rosemount 710DU (9-17)
E2T-A12SS (25-113C) (9-86)
(25-113C) l l
El-63
DELETED NEW TRANSMITTER NEW ATU INSTRUMENT TVA LOOP NO. TVA LOOP iO, SY'""EM VARIABLE VENDOR / NO. . VENDOR / NO. VENDOR /NO. INPbTS NAME (LOCATION) (LOCATION) (LOCATION) DIVISION (LOCATION)
Turbine First PS-1-91 A PT-1-91 A PIS-1-91 A IB RPS/RPT Stage Pressure Barksdale Rosemount 1154 Rosemount 710DU (9-17) +
Permissive B2T-A12SS (25-110) (9-84)
(25-110) .
PS-1-91B FT-1-91B PIS 1-91B IA RPS/RPT Barksdale Rosemount 1153 Rosemount 710DU (9-15)
B2T-A12SS (25-110) (9-83)
(25-110)
Reactor High PS-3-22A FT-3-22AA PIS-3-22AA IA RPS Pressure Barksdale Tobar 32PA2212 Rosemount 710DU (9-15)
B2T-A12SS (25-5A) (25-5A) (983)
PS-3-22B PT-3-22BB PIS-3-22BB IB RPS Barksdale Tobar 32PA2212 Rosemount 710DU (9-17)
B2T-A12SS (25-5A) (25-5A) (9-84)
PS-3-22C FT-3-22C PIS-3-22C IIA RPS Barksdale Tobar 32PA2212 Rosemount 710DU (9-15)
B2T-A12SS (25-6A) (25-6A) (9-85)
PS-3-22D Irr-3-22D PIS-3-22D IIB RPS Barksdale Tobar 32PA2212 Rosemount 710DU (9-17)
B2T-A12SS (25-6A) (25-6A) (9-86)
Reactor Low LITS-3-52 LT-3-52 LIS-3-52 I cont.
Water Level Yarway 4418CE Rosemount 1153 Rosemount 710DU Spray (Level 0) (25-51B) (25-51B) (9-81) (9-32) s Reactor Low LIS-3-56A LT-3-56A LIS-3-56A IA PCIS Low Water Level Yarway 4418C Rosemount 1153 Rosemount 710DU (9-15)
(Level 1) (25-5A) (25-5D) (9-83)
LIS-3-56B LT-3-56B LIS-3-56B IB PCIS Yarway 4418C Rosemount 1153 Rosemount 710DU (9-17) -
(25-5A) (25-5D) (9-84)
LIS-3-56C LT-3-56C LIS-3-56C IIA PCIS Yarway 4418C Rosemount 1153 Rosemount 710DU (9-15)
(25-6A) (25-6D) (9-85)
LIS-3-56D LT-3-56D LIS-3 56D llB PCIS Yarway 4418C Rosemount 1153 Rosemount 710DU (9-17)
(25 6A) (25-6D) (9-86)
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l DELETED NEW TRANSMITTER NEW ATU j INSTRUMENT TVA LOOP NO. TVA 1,00P NO. SYSTEM i VA.RIABLE VENDOR / NO. VENDOR / NO. VENDOR /NO. INPUTS NAME (14) CATION) (LOCATION) (LOCATION) DIVISION (LOCATION)
Reactor Low LIS-3-58A LT 3-58A LIS-3 58A I IIPCI,RCIC Water Level Yarway 4418C Rosemount 1153 GE 184C5988G (9-32) l (Level 2) (25-5A) (25-5D) (9-81)
LS-3 58A I CSS,LPCI GE 184C5988G (9-32) l (9-81) ADS (9-30)
LS-3-58A1 1 ATWS GE 184C5988G (ARI/RIT) i (9--81) (25-416)
LITS-3-58B LT-3 58B LIS-3-58B I HPCI, RCIC Yarway 4418C Rosemount 1153 GE 184C5988G (9-32)
(25-5A) (25-5 D) (9-81)
LS-3-58D 1 CSS,LPCI GE 184C5988G (9 32)
(9-81) ADS (9-30)
LS-3-58B1 1 ATWS GE 184C5988G (ARI/RIT)
(9-81) (25-416)
LIS-3-58C LT-3-58C LIS-3-58C II HPCI (9-39)
Yarway 4418C Rosemount 1153 GE 184C5988G RCIC (9-33)
(25-6A) (25-6D) (9-82)
LS-3-58C 11 CSS, LPCI GE 184C5988G (9-33)
(9-82) ADS (9-33)
LS-3-58C1 II ATWS GE 184C5988G (ARI/RPT)
(9-82) (25-613)
LITS-3-58D LT-3-58D LIS-3-58D 11 llPCI (9-39)
Yarway 4418C Rosemount 1153 GE 184C5988G RCIC (9-33)
(254A) (25-6D) (9-82)
LS-3-58D 11 CSS,LPCI GE 184C5988G (9-33)
(9-82) ADS (9-33)
LS 3-58D1 11 ATWS GE 184C5988G (ARl/RIT)
(9-82) (25-613)
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1 DELETED NEW TRANSMITTER NEW ATU INSTRUMENT TVA LOOP NO. TVA LOOP NO. SYSTEM VARIABLE VENDOR / NO. VENDOR / NO. VENDOR /NO. INPLTIS NAME (LOCATION) (ll) CATION) (LOCATION) DIVISION (LOCATION)
Reactor Low LITS-342 LT-3-62A LIS 342A !! Cont. Spray Water Level Yarway 4418CE Rosemount 1153 Rosemount 710DU (9-33)
(Level 0) (25-52B) (25-528) (9-82)
Reactor Pressure PS-3-74A PT-3-74A PIS-3-74A I CSS,LPCI Barksdale Tobar 32PA1212 Rosemount 710DU (9-32)
B2T-M12SS (25-5A) (9-81)
(25-5A)
PS-3-74A I LPCI (9-32)
Rosemount 710DU (9-81)
PS-3-74B FT-3-74 B PIS-3-74B 11 CSS, L.PCI Barton 288 Tobar 32 pal 212 Rosemount 710DU (9-33)
(25-6A) (25-6A) (9-82)
PS-3-74B 11 LPCI (9-33)
Rosemount 710DU (9-82)
Reactor Low LIS-3184 LT-3-184 LIS-3-184 1 ADS (9-30)
Water Level Yarway 4418C Rosemount 1153 Rosemount 710DU (Level 3) (25-58) (25-5 D) (9-81)
LIS-3-185 LT-3-185 LIS-3-185 II ADS (9-33)
Yarway 4418C Rosemount !!53 Rosemount 710DU (254B) (25-6D) (9-82)
Reactor Low LIS-3-203A LT-3-203A LIS-3 203A lA RPS, PCIS Water Level Barton 288A Rosemount 1153 Rosemour.-. 710DU (9-15)
(Level 3) (25-5-1) (25-5 C) (9-83)
LIS-3-203B LT-3-203 B LIS-3-203B IB RPS, PCIS Barton 288A Rosemount 1153 Rosemount 710DU (9-17)
(25-5-1) (25-5C) (9-84)
LIS-3 203C LT-3-203C LIS-3-203 C llA RPS, PCIS Barton 288A Rosemount 1153 Rosemount 710DU (9-15)
(25-6-1) (25-6C) (9-85)
LIS-3-203 D LT-3-203D LIS-3-203D !!B RPS, PCIS Barton 288A Rosemount 1153 Rosemount 710DU (9-17) )
(2541) (254C) (9-86) i l
'l i
I i
El-66 l l
DELETED NEW TRANSMITTER NEW ATU INSTRUMENT TVA IJX)P NO. TVA LOOP NO. SYSTEM VARIABLE VENDOR / NO. VENDOR / NO. VENDOR /NO. INPUTS NAME (LOCATION) (LOCATION) (LOCATION) DIVISION (LOCATION)
Reactor liigh PS-3 204A FT-3-204A PIS-3-204A I ATWS Pressure Static-O-Ring Rosemount 1153 GE 184C5988G (ARI/RirT) 9N-AA45 (25-5-1) (25-5C) (9-81) (25-416)
PS-3-204B PT-3-204B PIS-3-204B I ATWS Static-O-Ring Rosemount 1153 GE 184C5988G (ARl/RPT) 9N-AA45 (25-5-1) (25-5C) (9-81) (25-416)
PS-3-204C FT-3-204C PIS-3-204C 11 ATWS Static-O-Ring Rosemount 1153 GE 184C5988G (ARl/ RIFT) 9N-AA45 (25-6-1) (25-6C) (9-82) (25-613)
PS-3-204D PT-3 204D PIS-3-204D 11 ATWS Static-O-Ring Rosemount 1153 GE 184C5988G (ARI/RFT) 9N-AA45 (25-6-1) (25-6C) (9-82) (25-613)
Reactor liigh LIS-3-208A LT-3-208A LIS-3-208A I RCIC water Level Barton 288A Rosemount 1153 Rosemount 710DU (25-31)
(Level 8) (25-5-1) (25-5C) (9-81)
LIS-3-208 B LT-3-208B Lis-3 208B 11 liPCI Barton 288A Rosemount i153 Rosemount 710DU (9-39)
(25-5-1) (25-5C) (9-82)
LIS 3-208C LT-3-208C LIS-3-208C I RCIC Barton 288A Rosemount 1153 Rosemount 710DU (25-31)
(25-6-1) (25-6C) (9-81)
LIS-3-208D LT-3-208D LIS-3-208D 11 IIPCI Barton 288A Rosemount 1153 Rosemount 710DU (9-39)
(25-6-1) (25-6C) (9-82)
Primary PS-64-56A irr-64-56A PIS-64-56A IA RPS/PCIS l Containment Static-O-Ring Rosemount i153 Rosemount 710DU (9-15) 1 liigh Pressuce 12N-AA4 (25-58) (25-5A) (9-83) l l
PS-64-56B PT-64-56B PIS-64-56B IB RPS/PCIS l;
Static-O-Ring Rosemount !!53 Rosemount 710DU (9-17) 12N-AA4 (25-5B) (25-5A) (9-84)
PS-64-56C frT-64-56C PIS-64-56C llA RPS/PCIS Static-O-Ring Rosemount 1153 Rosemount 710DU (9-15) 12N-AA4 (25-68) (25-6B) (9-85)
PS-64-56D frT-64-56D PIS-64-56D llB RPS/PCIS Static 4 Ring Rosemount i153 Rosemount 710DU (9-17) 12N-AA4 (25-68) (25-68) (9-86)
El-67
DELETED NEW TRANSMITTER NEW ATU INSTRUMENT TVA LOOP NO. TVA LOOP NO. SYSTEM VARIABLE VENDOR / NO. VENDOR / NO. VENDOR /NO. INPUTS NAME (LOCATION) (LOCATION) (LOCATION) DIVISION (LOCATION)
Primary PS44-57A Irf-64-57A PIS&57A II ADS Containment Static-O-Ring Rosemount 1153 Rosemount 710DU (9-33) liigh Pressure 12N-AA4 (25-6B) (254B) (9-82)
PS & 57B PT-64-57B PIS-64-57B i ADS Static O-Ring Rosemount 1153 Rosemount 710DU (9-30) 12N-AA4 (25-5B) (25-5A) (9-81)
PS44-57C frT44-57C PIS&57C 11 ADS Static-O-Ring Rosemount 1153 Rosemount 710DU (9-33) 12N-AA4 (254B) (25-6A) (9-82)
PS&57D FT & 57D PIS&57D I ADS Static-O Ring Rosemount i153 Rosemount 710DU (9-30) 12N.AA4 (25-58) (25-5A) (9-81)
Primary PS44-58A IYT-64-58A PIS-64 58A Il CSS, llPCI, Containment Static-O-Ring Rosemount 1153 Rosemount 710DU LPCI Low Pressure 12N-AA4 (25-68) (25-6B) (9-82) (9-33)
PS-64-58B FT & 58B PIS44-58B - 1 CSS, llPCI, Static-O-Ring Rosemount 1153 Rosemount 710DU LPCI 12N-AA4 (25-58) (25-5B) (9-81) (9-32)
PS&58C FT-64-SSC PIS&58C 11 CSS, HPCI, Static-O-Ring Ru:emount 1153 Rosemount 710DU LPCI 12N-AA4 (254B) (25-6B) (9-82) (9-33)
PS-64-58D IrT&58D PIS&58D I CSS, liPCI, Static-O-Ring Rosemount 1153 Rosemount 710DU LPCI 12N-AA4 (25-5B) (25-5B) (9-81) (9-32)
Primary PS44-58E PT & 58E PIS&58E I Cont.
Containment Static-O-Ring Rosemount 1153 Rosemount 710DU Spray Low Pressure 12N-AA4 (25 58) (25-5 B) (9-81) (9-32)
PS44-58F IrT & 58F PIS&58F !! Cont.
Static-O-Ring Rosemount 1153 Rosemount 710DU Spray 12N-AA4 (25-68) (25-6A) (9-82) (G 13)
PS&$8G PT & 580 PIS-64-58G I Cont.
Static-O-Ring Rosemount 1153 Rosemount 710DU Spray 12N-AA4 (25-5B) (25-5B) (9-81) (9-32)
PS & $8H PT&58H PIS&58H 11 Cont.
Static-O-Ring Rosemount 1153 Rosemount 710DU Spray 12N-AA4 (25-6B) (25-6A) (9-82) (9-33)
El-68
DELETED NEW TRANSMITTER NEW ATU INSTRUMENT TVA L(X)P NO. TVA LOOP NO. SYSTEM VARIABLE VENDOR / NO. VENDOR / NO. VENIX)R/NO. INPUTS NAME (LOCATION) (LOCATION) (LOCATION) DIVISION (LOCATION)
Reactor Low PS48-95 Irr-68-95 PIS-68-95 I CSS, Pressure Barksdale Tobar32PA1212 Rosemount 710DU LPCI B2T-M12AA (25-51B) (9-81) (9-32)
(25-51B)
PS48-95 i LPCI Rosemount 710DU (9-32)
(9-81)
PS-68-96 IFT-68-96 PIS-68-96 11 CSS, Barksdale Tobar 32 pal 212 Rosemount 710DU LPCI B2T-M 12AA (25-52B) (9-82) (9-33)
(25-528)
PS-68-96 11 LPCI Rosemount 710DU (9-33)
(9-82)
RCIC Steam Line PDIS-71-I A PDT-71-1 A PDIS-71-1 A I RCIC liigh Flow Barton 288 Rosemount 1153 Rosemount 710DU (25-31)
(25-7A) (25-7A) (9-81)
PDIS-71-1B PDT-71-1 B PDIS-71-1B 11 RCIC Barton 288 r osemount i153 Rosemount 710DU (9-33)
(25-7A) (25-7A) (9-82) llPCI Steam Line PDIS-73-I A PDT-73-1 A PDIS-73-I A I HPCI liigh Flow Barton 288A Roacmount i153 Rosemount 710DU (9-33)
(25-7B) (25-7B) (9-81)
PDIS-73-1 B PDT 73-1B PDIS 731B D llPCI Barton 288A Rosemount 1153 Rosemount 710DU (9 39)
(25-7B) (25-7B) (9-82)
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.--. , - . . ~ - . - .
l Plant Specific'Information Reauired Section 5.4.2 - Trip Unit Cabinet :!
l I
Supply information for each trip unit' cabinet as identified below:
- 1. Cabinet layout showing location areas of the power supplies, trip relays, and trip units.
- 2. Division of which the cabinet is. assigned.
- 3. Layout of each card file in the-trip unit' cabinet ;
showing the trip variable for each card file.
slot.
TVA Response
- 1. The cabinet layout and the location of' trip relays and units is shown in Figure 1 and isL listed in ths Section 5.4.1 table. Panels that begin with the prefix 25-?? are local (Physically y located near the instrument).1 Panels that begin i with the prefix 9-?? are located in:the auxiliary 1 ,
instrument room (No' pipe 1from the nuclearLsystem: l or the primary' containment penetrates the- q Auxiliary Instrument ~ Room). j i
2.- The division assigned to-each cabinet is shown in Figure 2.
- 3. The layout of each' card file in the trip' unit cabinet and the trip variable for each card file' j slot is shown in Figure 2. .!
El-70
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FIGURE 1-TRIP CABINET ASSEMBLY <
i 2'.6' 1.34' ID EYESOLTS (ItD40VARE) to PLACES) .
(ch UPPER RELAY PANEL o
(( (22* WIDE X I* 3 3/4" LG.) ,
b \
a g 4__ q _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ q _ _3,,
E m 7et ; _ _ _ _ _ _ _. _ _ _ _ _ _ _ q _ _ l h g-I ..*
7 A* 11 I l Il O 11 g l
~ ~
gi g.
Il TERMINAL BOARD PANEL -
~'
,I _;g ___________ ____a__ 83 (18* WIDE X 6'*4-l/2* LG)-
. ,._ q _ _ _ __ _ _ _ _ _ _ _ _ _ _ q _' _ 1
~ ll
._______________c ll ,
+_____________________y
- * * ~~~ '
j l 28 9
- I CARD FILE I Il [ SEE COvCR DETAIL TAMPER PROOF -
1 tj . (TYP) i T i il r- lg [ i. Il L H ~_ ~_ . " 1_~_- T :
k lll f(CAROFILE2 8'*TP enwu sump D g- . Il MDT.ATOR LNW4TA 17" CUTOUT .6so m.
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GROUND BUS
-1 El-71 r
FIGURE 2 TRIP CABINET ASSEMBLY ECCS DIVI ECCS DIV II RPS Al RPS 88 RPS A2 RPS $2 PANEL 9 88 PANEL 9-82 PANEL 9 83 PANEL 9-84 PANEL 9-85 PAMEL 9 86 g gp DC g px.yg.60 1 6 IA PX-78-60-2 6 2A PX-99-Al 6 Als PX-99-54 6 SIA PX-99 A2 1. A2A PX-99-52 6 82A pg gyppty IL-]I-60 1 IL-78-60-2 IL*99 Al IL-99-Of IL-99*A2 IL-99-52 180 LIClilS IL-71 60=lA IL-78-60 2A IL-99 AIA IL-99-8IA IL-99-A2A IL-99-82A AreiuMCIA10R XA-71 60 ISEE NOTE 12) XA-99-l $$EE NOTE 121 4 LI6-3-52 LIS 3-62A PDIS-4-63A PDI6*l-430 PDIS-8-83C PDIS 4-83D 2 LIS-3 585 LIS 3-58D GuTuRL F uT u ti. _ lul uI'(~ ~~- F uTum,R]
3 LS-3 586 LS-3 58D POIS-t-25A POIS-1-258 PDIS-I-25C PDIS*l-25D LS-3 58BI LS-3-58DI
~
4 Oututt i FuTUAt_ l_ __. fuwnt. . FututL~)
5 CivTu @ Z fvwaf.) FDI6Al -36A PDIS-l=365 PDISAl -36C_ PDItal-360 6 LIS-3-58A LIS-3*58C (IJSat I
- Fututto. _ ruTutt - FVTuM 7 LS-3-58A LS-3-58C PDlS3 *'50A PDIS-l!506 ~ PDIS-8-000 ~ ~ PDIS-t-500 8 LS-3-58Al LS-3-50Cl Qpfutt ' _ _ JuTuti 'FiffUt1
' ~
FUTuM.)
9 @gruaf ~ ~ fufunD PIS-I-72 PIS-I-7f. PI s- 478 '2' -PIS-F-86 IO LIS-3-184 LIS-3-185 Ounset _ .MuTune.' _ _ rut _ute.; Firt Jith 11 PIS-64-578 PIS-64 57A PI S' 9 3 8 PIS-l'-9 0 A PIS-l 8IB ~ ~ PISMA
~
42 PIS64-570 PIS64-57C OUTu r.L ' FuTutt
. HTuRL ~ "MrfuRL7 XIS 71-60 4 XIS-78-60-2 XIS-99-1 XIS-99-IB XIS-99-2 XIS 99-28 la i pIS-64-588 PIS-64-58A PIS-3-22AA PIS-3-2288 PIS-3-22C PIS-3-2&D 2 (JOnut t. _-. FUTURE") LIS 3-56A LIS-3-568 LIS-3-56C LIS-3 56D 3 PIS-64-580 PIS-64-680 . LIS-3-203A ' LIS-3 203B LIS-3-203C LIS-3 2030 4 Q uTute__ __. JuTuiiC) (ftifUlE P FUTuRL.~ ~PFWTuP P. ' -- M 5 PIS-64 58E PIS-64-58F b 9
)
6 ( Futhnt 2_ fuT M k wa % _am.' a 7 d 7 PIS64-58G. PIS-tJ-SSH PIS-64*S6A PIS*64*565 PIS-64-56C PIS*64-56D k 8 (EutofLil FurifLD ( ruhiKt. t vi Dr.t- - Eu1Daf- "' TUTU M h 9 PIS-68-95 PIS-68-96 h }
10 PS-68 95 PS-68-96 / /
11 PDIS-78-IA PDIS-78-48 ) /
12 PDIS-73 lA PDIS-73-10 \' m%% l A.. m u . i' /
13 XIS-71-60-IA XIS-78-60-2A - XIS-99-lA XIS-99.l85 XIS-99-2A XIS-99-255 i LIS-3-208A LIS-3 2088 (FuC' su1utt ' -- -
' rurusti"
- --NRS.9 LIS-3-208C LIS-3 208D - --
2 3 / W uct 679tt T 4 f ruiuu. FuTuRL /
5 (rpugt._ __ __ruto.RF) 6 PIS-3-204A PIS-3 204C 7 PIS-3-2048 PIS-3 204D 8 [rutuat3 [Futuetth 9 QSurunt.) QuiuntJ
$0 PIS-3-74A PIS-3-740 II PS-3-74A PS-3-748 32 frvTuRtl [FuTutQ 4 g3 w- v< --
XIS-73-91 x!S-73-92 14 El-72
- Plant Specific Informatipn Reauired ,
Section 5.4.3 - Environmental Interface-The environment at each location where the retrofit hardware will be located must be compared to the maximum environment as stated in the topical report for the following factors:
- 1. Normal operation and-post-accident temperature and humidity.
TVA Resoonse
- 1. The response to Item 1, regarding the normal operation and post-accident temperature and humidity, is provided in the attached table.
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.n SECTION 5.4.3, ITEM 1 - ENVIRONMENTAL INTERFACE TEMPERATURE AND HUMIDITY Maximum Maximum Maximum Maximum Maximum Maximum Transmitter Normal Normal Post-Accident Post-Accident Qualified Qualified Number"' Temperature Humidity Temperature Humidity Temperature Humidity PDT-1-13 95'F 80% 110*F 100% 415'F 100%
A, B, C, D PDT-1-25 95'F 80% 110*F 100% 415'F 100%
A, B, C, D PDT-1-36 95*F 80% 110*F 100% 415'F 100%
A, B, C, D PDT-1-50 95*F 80% 110*F 100% 415'F 100%
A, B, C, D PT-1-72, 105'F 80% 105'F 80% 318'F 100%
76, 82, 86*
PT-1-81ASB 105*F 80% 105*F 80% 318*F 100%
PT-1-91A&B*
PT-3-22" 90*F 80% 100*F 90% 420*F 100%
A, B, C, D LT-3-52 90*F 80% 195'F 100% 415'F 100%
LT-3-62A El-74
Maximum Maximum Maximum Maximum Maximum Maximum Transmitter Normal Normal Post-Accident Post-Accident Qualified Qualified Number
- Temperature Humidity Temperature Humidity Temperature Humidity-i LT-3-56 90*F -80% 185'F 100% 318*F 100%
A, B, C, D LT-3-58 90*F 80% 185'F 100% 318'F 100%
A, B, C, D PT-3-74A&B 90*F 80% 185*F 100% 420*F 100%
LT-3-184 90*F 80% 125'F 80% 318'F 100%
LT-3-185 j LT-3-203 90*F 80% 185*F 100% 318'F 100%
A, B, C, D :,
PT-3-204" 90*F 80% 100*F 90% 318'F 100%
A, B, C, D LT-3-208 90*F 80% 195*F 100% 318'F 100%
A, B, C, D PT-64-20 90*F 80% 126*F 90% 318'F 100%
PT-64-21 i
PT-64-56 90*F 80% 125'F 100% 318'F 100% I A, B, C, D PT-64-57 90*F 80% 125'F 90% 318'F 100%
A, B, C, D El-75 .
Maximum Maximum Maximum Maximum Maximum Maximum Transmitter Normal Normal Post-Accident Post-Accident Qualified Qualified Number
- Temperature Humidity Temperature Humidity Temperature Humidity PT-64-58 90*F 80% 125*F 90% 318'F 100%
A, B, C, D PT-64-58 90*F 80% 125'F 90% 318*F 100%
E, F, G, H PT-68-95 90*F 80% 185'F 100% 420*F 100%
PT-68-96 PdT-71-1A 95'F 80% 136*F 90% 415*F 100%
PdT-71-1B PdT-73-1A 95'F 80% 165*F 100% 415'F 100%
PdT-73-lb Footnote: * - Worst case radiation exposure (40 year plus 100' day post-accident) is 1. 37 x 10' rade.
Minimum qualification for the transmitter models to be installed is 2.62 x 10' rade.
"' - This equipment is not within the scope of the 10 CFR 50.49 program. The specified qualified temperature / humidity values are from the test report of this manufacturer /model.
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Plant Specific Information Reauired Section 5.4.3 - Environmental Interface The environment at each location where the retrofit hardware will be located must be compared to the maximum environment as stated in the topical report for the following factors:
- 2. Comparison of the floor seismic response spectra-of the cabinet mounting location for the specific plant to seismic test envelope that the cabinet was tested to.
TVA Response
- 2. Figure 3 provides the comparison of the floor seismic response spectra of the cabinet mounting location to the seismic test envelope that the cabinet was tested to.
Plant Specific Information Recuired Section 5.4.3 - Environmental Interface The environment at each location where the retrofit hardware will be located must be compared to the maximum environment as stated in the topical report for the following factors:
- 3. If the trip unit cabinets are not located in the preferred location as per paragraph 5.1.4, provide justification for the alternate selected location.
TVA Response
- 3. The preferred location recommended by NEDO-21617-A, for the trip unit cabinets is the auxiliary room or control room. The Unit-3 equipment is located in the preferred location.
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J 1
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FIGURE 3 ENVIRONMENTAL INTERFACE - SEISMIC RESPONSE Cabinet Seismic Test Envelope i '. . . ........
2t Damping SSE ..
~
Reference:
Structural Dynamics Research Corporation .
j i *
- Report No. 11417-1 i"-------*
1 -
t -
prepared for Nutherm International ..i..
Contract No. 827232 - . .
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Floor Seismic Response Spectrum of the Cabinet Mounting Location for the Specific Plant Browns Ferry Nuclear Plant
. . . . _i " _ . " . " .'.
. ~.
. . .._ _ _ . Reactor Building, Elevation S93'
_ _.I.. .. a __..' ._ ... '..'. ._.! ~.~ ~ 2% Damping SSE
./ ,
./ / Jo Frequency (Hz)
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Plant Specific Information Reauired v
Section 5.4.4 - Specific Plant Interconnections An interconnection diagram which shows the interconnections between the existing logic cabinets and instrument cabinets and the new trip unit cabinets is to be provided the NRC. The content of the information is to be similar to the information shown of Figures 5-3, 5-4 and 5-5 as applicable. The detail of interconnection shown of the retrofit elementary and interconnection block diagram should be sufficient.
TVA ResDonse The BFN plant specific interconnection diagram is represented by NEDO-21617-A, Figure 5-5 .
Plant Specific Information Reauired Section 5.4.5 - Field Calibration Rack The design and operational information on the " Field Calibration Rack" is to be supplied to the NRC if such a device is purchased and used for transmitter calibration.
TVA ResDonse BFN did not purchase and does not use a " Field Calibration Rack."
In general, the instrumentation that will be installed as part of the ATTS on Unit 3 is the same instrumentation previously reviewed by NRC and installed on Unit 2, with the following exception: ,
The transmitters for the Reactor High Water Level instrument channel (Equipment identifiers LT-3-208 A-D, which include the level indicating switches LIS-3-208 A-D that appear in the Technical Specifications) are being replaced with environmentally qualified Rosemount 1153 transmitters.
Gould transmitters were used in this application on Unit 2 due to the unavailability of the Rosemount transmitters at the time of the Unit 2' modifications.
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As previously committed in TVA's March 15, 1993, response to NRC Bulletin 90-01, Supplement 1, TVA will replace or refurbish the Rosemount Model 1153 Series B and D and Model 1154 transmitters in safety related or ATWS applications prior to the restart of Unit 3.
The specific Technical Specification changes associated with the installation of the ATTS on Unit 3 (unless otherwise annotated) are discussed below. Also, unless otherwise noted, these changes make the Unit 3 Technical Specifications consistent with changes previously approved to the Unit 2 Technical Specifications.
- 1. Table 3.1.A, Reactor Protection System (SCRAM)
Instrumentation Requirements, is being revised to add instrument identifiers for the equipment installed as part of the ATTS modifications. These instrument identifiers enhance the useability of the Technical Specifications.
- 2. Table 4.1.A, Reactor Protection System (Scram)
Instrumentation Functional Tests Minimum Functional Test Frequencies for Safety Instr. and Control Circuits, is being revised to add or revise instrument identifiers, change the functional testing group designator for the instruments, and update the footnotes associated with the functional test type and minimum test frequency.
The addition of, or revision to, the instrument identifiers reflects the equipment installed'as part of the ATTS modifications. These instrument identifiers enhance the useability of the Technical Specifications.
The functional testing group designator for the instrumentation being installed as part of the ATTS modifications is being changed to "B". As described in the Bases for this section, the Group B designator is for analog devices, coupled with bistable trips, that provide a scram function.
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Note 7 is being added to'the Functional Test column to' E define the new functional test requirements for the instrumentation being installedgas part of.the ATTS modifications. Note 7 states that the functional test consists of the injection of a simulated signal into the electronic trip circuitry,.in place of the sensor :
signal, to verify the operability of the trip and alarm functions. The inclusion of Note 7 is consistent with the definition for the Channel Functional Test of Analog / Digital Channel, which is contained in Technical Specification Definition 1.0.V.12a.
Note 1 is'bei'ng deleted from the Minimum Frequency column in order'to clarify the Technical Specifications. Note-1 states that the minimum- .
l frequency for the indicated tests shall initially be once per month. Since the functional test frequency ,
for the instrumentation being installed as part of;the ATTS modifications is once per month, the inclusior. of Note 1 is redundant.
- 3. Table 4.1.B, Reactor Protection System (Scram)
Instrumentation Calibration Minimum Calibration '
Frequencies for Reactor Protection < Instrument . .
[
Channels, is being revised to add or. revise instrument identifiers, change the functional testing group .
designator for the instruments,.and change the minimum calibration frequencies.
The addition of, or revision to, the instrument ,
identifiers reflects the equipment installed as part of the ATTS modifications. These instrument identifiers enhance the useability of the Technical '
Specifications.
The functional testing group designator for the '
instrumentation being installed as'part'of the ATTS modifications is being changed to "B". As. described-in the Bases for this section, the Group _B designator l 1s for analog devices', coupled with' bistable-trips, that provide a scram function.
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In general, surveillance frequencies are based on industry accepted practices and engineering judgement.
Consideration is given to the conditions' required to perform a given test, the ease of performing the test, and the likelihood of a change in the system / component status during the performance of the test.
. Instrumentation calibration frequencies consist of an j
optimum selection of. time versus drift. Setpoint scaling calculations are performed:to provide assurance that there is adequate margin between the required-trip setpoint and the limiting safety system settings to account for inaccuracies in the instrument loop.
The minimum calibration frequency for the High Drywell Pressure, Reactor Low Water Level, and Turbine First Stage Pressure Permissive are being changed to once i per 18 months. The extension of the minimum
~
calibration frequency to once per 18 months reflects the high reliability of the analog instrumentation .
systems. .
The minimum calibration frequency for the High Reactor Pressure instrument channel is being revised to once per 6 months. This reflects the installation of instrument loops that contain transmitters manufactured by Tobar, Incorporated. These ,
instruments only permit the extension'to a six month calibration frequency. ,
These calibration' frequencies'are in accordance with j the Unit 3 specific setpoint and scaling calculations-for the ATTS instrumentation. Note 9 is being added' ;
to the Minimum Frequency column to specify the ;
calibration methods required for this new- i instrumentation.
- 4. A paragraph is being added to Unit.3 Bases'3.1.in !
order to describe the Reactor Protection. System (RPS) power supply.' This change describes the ability of _
the'RPS to' tolerate a single failure of a.non-class 1E power supply and makes the Unit 3 Technical Specifications consistent with changes previously E performed on the Units 1 and 2 Technical Specifications.
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- 5. Table 3.2.A, Primary Containment and Reactor Building Isolation Instrumentation, is being revised to add or revise instrument identifiers and to correct notes in the Remarks column.
The addition of, or revision to, the instrument identifiers reflects the equipment. installed as part of the ATTS modifications. These instrument identifiers enhance the useability of the Technical Specifications.
The listing of the Primary Containment Isolation System valve groups, which are initiated by the reactor low water level, is being deleted. This change makes this entry consistent with the other entries in'this table, which do not list the specific valve groups initiated by a trip function.
- 6. Table 3.2.B, Instrumentation that Initiates or Controls'the Core and Containment Cooling Systems, is being revised to add or revise instrument identifiers.
The addition _of, or revision to, the instrument identifiers reflects the equipment installed as part of the ATTS modifications. These instrument identifiers enhance the useability of the Technical Specifications.
- 7. Table 3.2.F, Surveillance Instrumentation, is being revised to correct instrument identifiers and to change the indicated range of the reactor pressure instruments.
The revision to the instrument identifiers reflects the equipment installed as part of the ATTS modifications. These instrument identifiers enhance the useability of the Technical Specifications.
The revision to the reactor pressure indication range reflects the newly installed equipment. The newly installed range, 0 - 1200 psig, includes the full range of pressures for which operator actions would be initiated during accident conditions. This range was approved for BFN in the NRC's May 10, 1991 Safety Evaluation of Emergency Response Capability and Conformance to Regulatory Guide 1.97, Revision 3.
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1
Surveillance Instrumentation, is being revised to add instrument identifiers.
The addition of the instrument identifiers reflects the equipment installed as part of the ATTS modifications. These instrument identifiers enhance the useability of the Technical Specifications (NOTE:
Similar changes are also being proposed for the Unit 2 Technical Specifications).
- 9. Table 4.2.A, Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation, is being revised to correct instrument identifiers, change the functional testing description for the instruments, and change the minimum calibration frequencies.
The correction of the instrument identifiers reflects the equipment installed as part of the ATTS modifications. These instrument identifiers enhance the useability of the Technical Specifications.
The addition of Note 28 in the Functional Test column describes the type of function test performed on the ATTS instrumentation.
The minimum calibration frequencies for the Reactor Low Water Level, High Drywell Pressure, Low Pressure Main Steam Line, and High Flow Main Steam Line instrument channels are being revised in accordance with the Unit 3 specific setpoint and scaling calculations for the ATTS instrumentation. The addition of Note 29 in the Calibration Frequency column describes the type of calibration performed on the ATTS instrumentation.
- 10. Table 4.2.B,-Surveillance Requirements for Instrumentation that Initiate or Control the CSCS, is being revised to add or correct instrument identifiers, change the functional testing description for the instruments, and change the minimum calibration frequencies.
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The addition or correction of the instrument .l identifiers reflects the equipment installed as part of the ATTS moditications. These instrument '
identifiers enhance the useability of the Technical Specifications (NOTE: Instrument identifiers are also being added or corrected in the Unit 2 Technical Specifications for a Reactor Low Water Level and Reactor High Water Level instrument channel).
The addition of Note 28 in the Functional Test column describes the type of function test performed on the ATTS instrumentation (NOTE: A similar note is also being. proposed for the Unit 2 Technical Specifications for the Reactor High Water Level and RCIC and HPCI Turbine Steam Line High Flow instrument channels).
The minimum calibration frequencies for the Reactor Low Water Level, Drywell High Pressure, Reactor Low Pressure, Reactor High Water Level, and RCIC and HPCI Turbine Steam Line High Flow instrument channels are raing revised in accordance with the Unit 3 specific setpoint and scaling calculations for the ATTS instrumentation (NOTE: A similar note is also being proposed for the Unit 2 Technical Specifications for the Reactor High Water Level and RCIC and HPCI Turbine Steam Line High Flow instrument channels).
The addition of Note 29 in the Calibration Frequency column describes the type of calibration performed on the ATTS instrumentation (NOTE:. A similar note is also being proposed for he Unit 2 Technical Specifications for the Reactor High Water Level and RCIC and HPCI Turbine Steam Line High Flow instrument channels).
The minimum calibration frequency for the Reactor Low Pressure instrument channel is being revised to once per 6 months in accordance with the Unit 3 specific setpoint and scaling calculations for the ATTS instrumentation. This reflects the installation of instrument loops that contain transmitters manufactured'by Tobar, Incorporated. These instruments only permit the extension to a six month calibration frequency.
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- 11. Table 4.2.F, Minimum Test and Calibration Frequency for Surveillance Instrumentation, is being revised to add or correct instrument identifiers and change the minimum calibration frequencies.
The addition or correction of the instrument identifiers reflects the equipment installed as part of the ATTS' modifications. These instrument identifiers enhance the useability of the Technical Specifications. ,
The minimum calibration frequencies for the Reactor Water Level and Drywell Pressure instrument channels are being revised in accordance with the Unit 3 ,
specific setpoint and scaling calculations for the ATTS instrumentation (NOTE: A similar change is also being proposed for the Unit 2 Technical Specifications).
Part B: The Units 1 and 3 Reactor Vessel Water Level Safety Limit and the Level 1 Low Reactor Vessel Water Level setpoint are being revised.
- 1. The analytical reactor vessel water level safety limit determined by General Electric (the NSSS supplier) calculations has always been greater than or equal'to 372.5 inches above vessel zero. Lowering the current safety limit (378 inches) to match the analytical limit is supported by calculation and will make the Units 1 and 3 value consistent with the current Unit 2 !
setting.
- 2. The Level 1 low reactor vessel water level instruments actuate the Core Spray and Low Pressure Coolant Injections systems in order to mitigate the consequences of a loss of coolant accident. rney also ,
isolate the main steam lines to reduce inventory loss. !
l j
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\
During the process of generating setpoint and accuracy-calculations for plant parameters in support of Unit 2, a determination was made that the Level 1 Reactor Vessel Level 1-Low Water Level setpoint was not conservative based on the current calculation methodology, which is based on Regulatory Guide 1.105, Instrument Setpoints for Safety Related Systems.
Regulatory Guide 1.105 endorses Instrument Society of America.(ISA) Standard ISA-S67.04 - 1982, Setpoints for Nuclear Safety Related Instrumentation Used in Nuclear Power Plants, as an acceptable method for ensuring that setpoints stay within technical specification limits.
As discussed above, the analytical safety limit for the Reactor Vessel Water Level Safety Limit is 372.5 inches above vessel zero. This limit was used as a design input to a scaling and setpoint calculation that determined the nominal trip setpoint and trip level setting based on inaccuracies associated with the instrument loops. The allowance for instrument inaccuracies in the determination of' the actual trip setpoint provides conservative assurance that the trip function will be performed at or before reaching the analytical limit. This scaling and setpoint calculation is in accordance with the guidance contained in Regulatory Guide 1.105.
The proposed change to the Level 1 Low Reactor Vessel Water Level setpoint guarantees that core cooling is maintained and fission product loss minimized _during a design basis event by ensuring that the associated trips occur within the process parameter value (analytical limit) utilized to confirm the design bases of the plant.
The specitic Technical Specification changes associated with the change to the Level 1 Low Reactor Vessel Water Level setpoint makes the Units 1 and 3 Technical Specifications consistent with changes previously approved to the Unit 2 Technical Specifications.
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Part C: For Unit 2, instrument identifiers are being added or corrected to reflect the equipment previously installed as part of the ATTS modifications. These changes are administrative in nature and do not involve a design change or other physical change to the plant. These instrument identifiers enhance the useability of the Technical Specifications.
Part D: For Unit 2, Reactor High Water Level, RCIC and HPCI turbine steam line high flow, and drywell pressure instrumentation calibration frequencies and functional test descriptions are being revised to reflect current calculations and test methods. These changes do not reflect a change in equipment, operation of the associated system, or the safety function of that system. The revised scaling and setpoint calculations are in accordance with the guidance contained in Regulatory Guide 1.105. The ,
allowance for instrument inaccuracies in the determination of the actual trip setpoint provides conservative assurance that the trip function will be performed at or before reaching the analytical limit.
In addition, the Rosemount Model 1153 transmitters, which are used for the RCIC and HPCI turbine steam line high flow, and drywell pressure instrument loops, are used in other safety related application at BFN with the new calibration frequency (18 months). These instruments have performed acceptably over the last operating cycle with this longer calibration frequency. The Gould transmitters, which are used for the Reactor High Water Level instrument loops, are not used in other safety related applications at BFN. However, vendor derived data supports the determination that these instruments will perform acceptably with the longer calibration frequency.
The instrumentation in the affected loops was upgraded as part of the installation of the ATTS, which was installed prior to Cycle 6 operation. The new calibration requirements, together with the new instrumentation, are expected to provide a more reliable instrumentation system.
Part E: For Units 1, 2, and 3, the differential pressure '
instrumentation, which actuates the pressure suppression chamber-reactor building vacuum breakers (Rosemount Model 1153 transmitters), calibration frequency is being revised to reflect current Unit 2 and 3 calculations. The Unit 1 change is based on the similarity of this system and equipment between the three units. A Unit 1 specific calculation will be performed to confirm the calibration frequency prior to Unit 1 restart, i
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The allowance for instrument inaccuracies in the ,
determination of the actual trip setpoint provides 1 conservative assurance that_the trip function will be !
performed at or before reaching the analytical limit. The l scaling and setpoint calculations are in accordance with the guidance contained in Regulatory Guide 1.105. In addition, the specified minimum number of instrument channels per trip system, function, trip level setting, actions required, remarks, functional test, and instrument check reflect current operational requirements are being added in order to be consistent with the treatment of other electronic trip circuitry in the Technical Specifications.
Part F: The capitalization of terms used on the affected Units 1, 2, and 3 Technical Specification pages is an administrative change. This change conforms with the current TS Definitions section. The correction of spelling and capitalization of other words on the same pages is also administrative in nature.
IV. H9 SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION DZgglMPTION OF THE PROPOSED TE9]INICAL BPECIFLCATION CHANQE This proposed change to BFN Technical Specifications consists of six parts.
Part A: The Unit 3 mechanical pressure and differential pressure indicating switches in the Reactor Protection System (RPS) and Emergency Core Cooling System (ECCS) are being replaced with an Analog Transmitter / Trip System (ATTS).
Part B: The Units 1 and 3 reactor vessel water level safety limit is being revised to reflect the analytical limit provided by General Electric and the Level 1 Low Reactor Vessel Water Level setpoint is being revised to provide a more conservative limit.
Eart_Q: For Unit 2, RPS and ECCS instrument identifiers are being added or corrected to enhance useability of the Technical Specifications. These changes do not reflect a change in equipment, operation of the associated system, or the safety function of that system.
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i Part D: For Unit 2,' Reactor High Water Level, Reactor Core )
Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) Turbine Steam Line High Flow, and Drywell Pressure instrumentation calibration frequencies and functional test descriptions are being revised to reflect current calculations and test methods. These changes do not reflect a change in equipment, operation of the associated system, or the safety function of that system.
Part E: For Units 1, 2, and 3, the differential pressure instrumentation, which actuates the pressure suppression chamber-reactor building vacuum breakers, calibration frequency is being revised. In addition, tables that specify the minimum number of instrument channels per trip system, function, trip level setting, actions required, remarks, functional test, and instrument check are being added.
Eart F: Corrects the capitalization of terms used on the affected Units 1, 2, and 3 TS pages in order to conform with the current TS Definitions section. This part also corrects spelling and capitalization of other words on the same pages.
TVA has concluded that operation of Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 in accordance with the proposed change to the technical specifications does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50. 91(a) (1) , of the three standards set forth in 10 CFR 50.92(c).
A. The proposed amendmont does not involve a sianificant increase in the probability or consecuepces of an accident previousiv evaluated.
Part A: The Unit 3 modification, which involves the installation of an Analog Transmitter / Trip System (ATTS), replaces older devices with devices of more modern design that perform the same function.
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The initiation of control rod insertion to mitigate a design basis accident is contained in Chapter 14 of the BFN Final Safety Analysis Report (FSAR). There is no change in design bases, protective function (initiation of control rod insertion), redundancy, setpoints, or logic associated with the installation of the ATTS. The consequences of a failure of this equipment are no different than that of the original equipment. Since there is no. change in any protective functions, nor the creation of any new operational conditions, the proposed amendment does not involve a-significant increase in the probability or consequences of any accident previously evaluated.
Part B: The revision to the Units 1 and 3 reactor vessel water level safety limit and the Level'1 low reactor vessel water level setpoint do not reflect any change in plant equipment. The safety limit is being changed to reflect the actual analytical safety limit calculated by General Electric.
The Level 1 low reactor vessel water level trip initiates the Core Spray and Low Pressure Coolant Injection Systems and isolates the Main Steam lines.
These actions are taken to mitigate the consequences of a Loss of Coolant Accident. The change in the setpoint affects the timing of the operation of equipment necessary to mitigate the consequences of an accident. A setpoint calculation has been generated which ensures these safety functions are initiated in accordance with the design basis accident analysis presented in Chapter 14 of the Browns Ferry FSAR.
Therefore, the proposed. amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated.
Part C: The addition or correction of Unit 2 instrument identifiers is administrative in nature:and does not reflect any modification ~to plant equipment.
These administrative changes do not reflect any change to any precursor for the design basis events or-operational transients analyzed in the Browns Ferry FSAR. There is also no change to any protective function or mitigating action for the design basis events or operational transients analyzed in the Browns Ferry FSAR.- Therefore, the probability or consequences of an accident previously evaluated is not significantly increased.
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Part D: The change in Unit 2 reactor high water level and Reactor Core Isolation Cooling (RCIC) instrumentation functional test descriptions reflects the equipment currently installed and the functional tests currently being performed.
The changes in calibration frequencies are being made to reflect current setpoint calculations. There are no modifications to plant equipment or changes in instrument setpoints associated with these changes.
The calibration frequencies specified by the current setpoint calculations ensure that the associated safety functions are initiated in accordance with the design basis accident analysis presented in Chapter 14 of the Browns Ferry Final Safety Analysis Report (FSAR). Therefore, the probability or consequences of an accident previously evaluated is not significantly increased.
Part E: The changes in Units 1, 2, and 3 calibration frequency for the differential pressure instrumentation, which actuates the pressure suppression chamber-reactor building vacuum breakers, is being made to reflect current setpoint calculations. The specified minimum number of instrument channels per trip system, function, trip level setting, actions required, remarks, functional test, and instrument check reflect current operational requirements. There are no modifications to plant equipment or changes in instrument setpoints associated with these changes. The calibration frequencies specified by the current setpoint calculations ensure that the associated safety functions are initiated in accordance with the design basis accident analysis presented in Chapter 14 of the Browns Ferry Final Safety Analysis Report (FSAR).
Therefore, the probability or consequences of an accident previously evaluated is not significantly increased.
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Part F: The proposed correction of the capitalization of terms in order to conform with the current TS Definitions section is administrative'in nature and does not: reflect any modification.to plant equipment..- '
The correction of spelling and capitalization of.other words on the same pages is also administrative in nature-and does not reflect any modification to plant equipment. These. administrative changes do not-reflect any change to any precursor for the design basis c/ents or operational transients analyzed in the
~
Browns Ferry FSAR. There is also no change to any protective function or mitigating action for the .
design basis events or operational transients analyzed- R in the Browns Ferry FSAR. Therefore, the probability or consequences of an accident previously' evaluated is not significantly'increasad. <
B. The cronoggd amendment does not create the nossibility of a new or different kind of accident from any accident nreviously evaluated.
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Part A: The installation of the ATTS replaces older devices with devices of more modern design that perform the same function. No new control functions-are added. No credible equipment failure modes or single failure are introduced which could result in the inability of redundant safety components or systems to perform their safety. functions.in accordance with the design basis accident analysis presented in Chapter 14 of the Browns Ferry FSAR.
Therefore, the proposed amendment does not create the
~
possibility of a new or different kind of accident from any accident previously evaluated.-
Part B: The revision to the Units 1 and 3 reactor vessel water level-safety limit and the Level 1 low reactor vessel water level setpoint do not reflect any change in plant equipment. The safety limit is being changed to reflect the actual analytical safety limit calculated by General Electric.
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The change in the Level 1 low reactor vessel water level setpoint affects the timing of the operation of equipment necessary to mitigate the consequences of an accident. No new failure modes or system interactions are introduced. The same protection _ functions will still occur at the Level 1 low reactor water level setpoint. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Part C: The addition or correction of Unit 2 instrument identifiers is: administrative in nature and does not reflect any modification to plant equipment. 4 The correction of instrument identifiers does not require new system alignments, modifications, or changes in operating procedures. Therefore, no new external threats, system interactions, release pathways, equipment failure modes, or types-of operator errors are created. Therefore, the proposed-amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Part D: The change in Unit 2 reactor high water level and RCIC instrumentation functional test descriptions reflects the equipment currently installed and the functional tests currently being performed. The changes in calibration frequencies are being made to reflect current setp3 int calculations. There are no modifications to plant equipment or changes in instrument setpoints associated with these changes.
No new failure modes or system interactions are introduced. The same protection functions will still occur at the same setpoints. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident'from any accident previously evaluated.
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Eart E: The changes in Units 1, 2, and 3 calibration frequency for the differential pressure instrumentation, which actuates the pressure suppression chamber-reactor building vacuum breakers, is being made to reflect current setpoint calculations. The specified minimum number of instrument channels per trip system, function, trip level setting, actions required, remarks, functional ,
test, and instrument check reflect current operational requirements. There are no modifications to plant-equipment or changes in instrument setpoints associated with these changes. No new failure modes or system interactions are introduced. The same protection functions will still occur at the same setpoints. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Part F: The proposed correction of the capitalization of terms in order to conform with the current TS Definitions section is administrative in nature and does not reflect any modification to plant equipment.
The correction of spelling and capitalization of other words on the same pages is also administrative in nature and does not reflect any modification to plant equipment. The correction of spelling and capitalization does not require new system alignments, modifications, or changes in operating procedures.
Therefore, no new external threats, system interactions, release pathways, equipment failure modes, or types of operator errors are created.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
C. The proposed amendment does not involve a sicnificant reduction in a marcin of safety.
Part A: The installation of the ATTS replaces older devices with devices of more modern design that perform the same function. The replacement equipment will improve reliability, accuracy and response times.
There are no changes in-the systems' design basis, protective function, or logic arrangement. Instrument.
setpoints and calibration frequencies are supported by Unit 3 specific calculations. Therefore, the proposed amendment does not involve a significant reduction in the margin of safety.
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Part' B: The revision to the Units 1 and 3 reactor vessel water level safety limit and the Level 1 low.
reactor vessel water level setpoint do not reflect any change in plant equipment. The safety limit is being changed to reflect the actual analytical safety limit.
L calculated by General Electric.
The change in the Level 1 low reactor vessel water level setpoint is supported by a Unit 3 specific -
setpoint calculation that-has been performed in-accordance with the methodology endorsed by Regulatory Guide 1.105, Instrument Setpoints for Safety Related Systems.
Therefore, the proposed amendment'does not involve a significant reduction in the margin of safety.
Part C: The addition or correction of Unit 2-instrument identifiers is administrative in nature and.
does not reflect any modification to plant' equipment.
Therefore, the proposed amendment does not' involve a i significant reduction in the margin of' safety.
1 Part D: The change in Unit 2-reactor high water level and RCIC instrumentation functional test descriptions reflects the equipment currently installed and the functional tests currently being performed. The-l changes in calibration frequencies are.being made to
! reflect Unit 2 specific setpoint calculations. These calculations:have been performed in3accordance with the methodology endorsed by Regulatory Guide 1.105.
There are no modifications to plant equipment or changes in instrument setpoints associated with.these-changes. Therefore, the proposed amendment does not involve a significant reduction in the margin of.
safety.
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Part E: The changes in Units 1, 2, and 3 calibration frequency for the differential pressure instrumentation, which actuates the pressure suppression chamber-reactor building vacuum breakers, is being made to reflect current setpoint calculations. The specified minimum number of instrument channels per trip system, function, trip level setting, actions required, remarks, functional test, and instrument check reflect current operational requirements. The setpoint calculations have been performed in accordance with the methodology endorsed by Regulatory Guide 1.105. There are no modifications to plant equipment or changes in instrument setpoints associated with these changes. Therefore, the proposed amendment does not involve a significant reduction in the margin of safety.
Part F: The proposed correction of the capitalization of terms in order to conform with the current TS Definitions section is administrativo in nature and~
does not reflect any modification to plant equipment.
The correction of spelling and capitalization of other words on the same pages is also administrative in nature and does not reflect any modification to plant equipment. Therefore, the proposed amendment does not' involve a significant reduction in the margin of safety.
V. ENVIRONMENTAL IMPACT CONSIDERATION -
The proposed change does not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure.
Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR
- 51. 2 2 (c) (9) . Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change.is not required.
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VI. REFERENCES BFN Unit 2 Analog Transmitter / Trip System Technical i Specification Approval -
- 1. TVA letter to NRC, dated August 23, 1984, in regards to Unit 2 Technical Specification No. 199
- 2. TVA letter to NRC, dated May 8, 1985, in regards to additional information on the analog trip system .
- 3. TVA letter to NRC, dated November 20, 1985, in regards to additional information on the analog trip system
- 4. TVA letter to NRC, dated December 30, 1985, in regards to Unit 2 Technical Specification No. 199, Supplement 2
- 5. TVA letter to NRC, dated April 29, 1986, in regards to Unit 2 Technical Specification No. 199, Supplement 3
- 6. NRC letter to TVA, dated August 19, 1986, in regards to Amendment No. 125 BFN Unit 2 Tobar Transmitters Technical Specification Approval -
- 7. TVA letter to NRC, dated February 24, 1989, Technical Specification No. 263 - Tobar Transmitters
- 8. NRC letter to TVA, dated July 7, 1989, Technical Specification for Tobar, Inc. Transmitters (TS 263)
BFN Unit 2 Reactor Pressure Vessel Low Water Level Trip Technical Specification Approval -
- 9. TVA letter to NRC, dated August 6, 1990, TVA BFN Technical Specification (TS) No. 291 - Revision to Level 1 Low Reactor Pressure Vessel (RPV) Water Level
- 10. TVA letter to NRC, dated October 9, 1990, TVA BFN Technical Specification (TS) No. 291 - Revision to .
Level 1 Low Level Reactor Pressure Vessel (RPV) Water {
Level
{
- 11. NRC letter to TVA, dated January 2, 1991, Issuance of Amendment (TS 291)
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