ML20070A033
| ML20070A033 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/10/1991 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070A034 | List: |
| References | |
| NUDOCS 9101170069 | |
| Download: ML20070A033 (21) | |
Text
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'o UNITED STATES
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8"3 NUCLE AR REGULATORY COMMISSION e
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- j WA$HINGTON, D. C. 20555
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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.187 License No. DPR-52 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated July 13, 1990 as supplemented September 17, 1990, complies with the standards :nd requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurante (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirement: have been satisfied, i
1 9101170069 910110 ADOCK OD0 g G DR
9 2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the. attachment to this license amendment and paragraph 2.C.(2) of facility Operating License No. OPR-52 is hereby nended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.187, are hereby incorporated in the license. The licensee shall operate the. facility in accordance with
-the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the.-date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
%AAG.mL Frederick J. Hebj' on, Director Project Directorate !!-4, NRR Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of. Issuance: January 10, 1991 l
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ATTACHMENT T0 LICENSE AMEN 0 MENT NO.
. FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines _ indicating-the area of change. Overleaf pages* are provided to maintain document completeness.
REMOVE-INSERT 3.2/4.2-18 3.2/4.2-18 3.2/4.2-19 3.2/4.2-19 3.2/4.2-22 3.2/4.2-22*
3.2/4.2-22a 3.2/4.2-23 3.2/4.2-23 3.2/4.2-24_
3.2/4.2-24*
3.2/4.2-44 3.2/4.2-44*
3.2/4.2-45 3.2/4.2-45 3.2/4.2-46 3.2/4.2-46 3.2/4.2-47 3.2/4.2-47 3.2/4.2-67 3.2/4.2-67 f
3.2/4.2-68 3.2/4.2-68*
3.2/4.2-69 3.2/4.2-69*
3.2/4.2-70 3.2/4.2-70*
3.2/4.2-71 3.2/4.2-71*
3.2/4.2-72 3.2/4.2-72*
3.2/4.2-73 3.2/4.2-73*
3.2/4.2-73a*
TABLE 3.2.8 (Continued)
Hinimum No.
ew
- s Operable Per
- y
. Trio Sysf11 Function Trio Level Settino _
Action Remarks 1
. HPCI Trip System bus power N/A C
1.
Monitors availability of monitor power to logic systems.
1 RCIC Trip Systen but power N/A C
1.
Monitors availability of monitor power to logic. systems.
1(2)
Instrument Channel -
1 Elev. 551' A
1.
Below trip setting will Condensate Header Low open HPCI suction valves Level (LS-73-55A & 8) to the suppression chamber.
i 1(2)
Instrument Channel -
i 7" above instrument zero A
1.
Above trip setting will open Suppression Chamber High HPCI suction valves to the.
Level suppression chand>er.
2(2)
Instrument Channel -
1 583" above vessel zero A
1.
Above trip setting trips RCIC Reactor High Water Level turbine.
(LIS-3-208A and y
LIS-3-208C)
U 1
Instrument Channel -
1 450" H O (7)
A 1.
Above trip setting isolates 2
RCIC Turbine Steam Line RCIC system and trips RCIC w
High Flow turbine.
1 (PDIS-71-1A and 18) e 1
en 3(2)
Instrumer.t Channel -
150 psig A
1.
Below trip setting isolates RCIC Steam Supply RCIC system and trips RCIC Pressure.- Low :
turbine.
(PS 71-1A-D)
$a 3(2)
Instrument Channel -
120 psig A
1.
Above trip setting isolates RCIC Turbine Exhaust RCIC system and trips.RCIC a
Diaphraya Pressure - High turbine.
(PS 71-11A-D)
T>
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-. +..
TABLE 3.2.B (Continued)
Minimum No.
c: ex, Operable Per Dy Trio Sys(11 Function Trio Level Settino Action Remarics 2(2)
Instrument Channel -
1583" above vessel zero.
A 1.
Above trip setting trips HPCI g
Reactor High Water Level turbine.
(LIS-3-2088 and LIS-3-20BD)
~
1 Instrument Channel -
190 psi (7)
A 1.
Above trip setting isolates HPCI Turbine Steam Line HPCI system and trips HPCI High Flow turbine.
(PDIS-73-1A and IB) 3(2)
Instrument Channel -
1100 psig A
1.
Below trip setting isolates HPCI Steam Supply HPCI system and trips HPCI Pressure - Low (PS 73-1A-0) turbine.
3(2)
Instrument Channel.-
120 psig A
1..
Above trip setting isolates
.HPCI Turbine Cxhaust HPCI system and trips HPCI Diaphrague (PS 73-20A-D) turbine.
F 1
Core Spray System Logic N/A B
1.
Includes testing aisto w
initiation inhibit to p
Core Spray Systems in*
other units.
1 RCIC System (Initiating)
N/A B
1 Includes Group 7 valves.
Logic Refer to Table 3.7.A for list of valves.
1 g
RCIC System (Isolation)
N/A B
1.
Includes Group S valves.
i Logic Refer to Table 3.7.A for eg list of valves.
M 1 (16)
ADS Logic N/A A
f (Initiation)
N D
TABLE 3.2,5 (Continued)-
Minim m No.
c 08 Operable Per E@
Trio Sys(1)
Function Trio Level Settine Action Remarts a
w 1(10)
Instrument Channel -
1 100*F A
1.
Above trip setting starts Core Thermostat (Core Spray Area Spray area cooler fans.
Cooler Fan) 1(10) met Area Cooler Fan Logic N/A A
1(10)
Core Spray Area Cooler Fan N/A A
Logic -
1(11)
Instrtment Char:nel -
N/A A
1.
Starts IIHRSW pumps A1, 83, Core Spray Motors A or D C1, and D3 Start 1(11)
Instrument Channel N/A A
1.
Starts llHRSW pumps A1, 83, Core Spray Motor B or C C1, and D3 l
Start 1(12)
Instrument Channel -
N/A A
- 1.
Starts RHRSW pumps A1, 83, u
Core Spray Loop 1 Accident C1, and D3 Signal (15)
~
1(12)
Instrument Channel -
N/A A
1.
Starts llHRSW pumps A1,*B3, w
Cere Spray Loop 2 Accident C1, and D3
,8, Signal (15) w 1(13)
IHtSW Initiate Logic N/A (14)
~1 RPT Logic N/A (17) 1.
Trips recirculation pumps r
on turbine control valve fact closure or stop valve closure > 301 power.
e e
i.
E" TABLE 3.2.8 (Continued) rE f
Minimum No.
Oper2ble Per N
I Trio Svs(1)
Function Trio Level Settino Action Remarks 1(16)
ADS Timer ti115 sec.
A 1.
Above trip settir.g in conjunction with low reactor i
water Tevel permissive, low reactor water level; high drywell pressure or ADS high drywell pressure bypass timer timed out ?.1d IHt or CSS i
pumps running, initiates ADS.
1(16)
ADS High Drywell ti322 sec.
A 1.
Above trip setting, in i
Pressure Bypass Timer conjuntion with low reactor water level permissive, low reactor water level, ADS timer timed out and RHR or CSS pumps running, initiates ADS.
4*
N 2
RCIC Steam Line Space il55* F E
1.
Above trip settleg isolates RCIC ysteen and trips RCIC Torus Area o
turbine.
I High Teeperature g
l t
N 2
- tCIC Steam Line Space 1180* F E
1.
Above trip setting isolates l
RCIC Pump Room Area RCIC system and trips RCIC c
turbine.
High Temperature 2
HPCI Steam Line Space 1180* F E
1.
Above trip setting isolates HPCI system and trips HPCI Torus Area turbine.
g High Temperature I
2 HPCI Steam Line Space 1200* F E
1.
Above trip setting isolates r+
HPCI Pump Room Area HPCI system and trips HPCI turbine.
High Temperature l
O l
I I.
~
I t
_m NOTES FOR TABLE 3.2.B 1.
Whenever any CSCS System is required by Section 3.5 to be OPERABLE, there shall be two OPERABLE trip systems arcept as noted. If a require' ment of the first column is reduced by one,.the indicated action shall be taken.
If the same function is inoperable in more than one trip system or the first column reduced by more than one, action B shall be taken.
Action:
A.
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 />, take action B.
B.
Declare the system or component inoperable.
C.
Immediately take action B until power is verified on the trip system.
D.
No action required; indicators are considered redundant.
E.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore the inoperable channel (s) to OPERABLE status or place the inoperable channel (s) in the tripped condition.
2.-
In only one trip system.
3.
Not considered in a trip system.
4.
Deleted.-
5.-
With diesel power, each RERS pump is scheduled'to start immediately and each CSS pump is sequenced to start about 7 seconds-later.
6.
With normal power, one CSS and one RERf pump is scheduled to start instantaneously, one CSS and one RERS pump is sequenced to start after about 7 sec with similar pumps starting after about 14 sec. and 21 sec.,
at which time the full complement of CSS and RHRS pumps would be operating.
i 7.
The RCIC and HPCI steam line high flow trip level settings are given in l
terms of differential pressure. The RCICS setting of 450" of water 4
corresponds to at least 150 percent above maximum steady state steam flow to assure that spurious isolation does not occur while ensuring the initiation of isolation following a postulated steam line break.
l:
Similarly, the HPCIS setting of 90 pai corresponds to at least L
150 percent above maximum steady state flow while also ensuring the initiation of isolation following a postulated break.-
8.
Note 1 does not apply to this ites.
9.
The head tank is designed to assure that the discharge piping from the CS and RER pumps are full. The pressure shall be maintained at or above the values listed in 3.5.H, which ensures water in the discharge piping and up to the head tank.
l BFN.
3.2/4.2-23
-Amendment 187 Unit 2
.l
NOTES FOR TABLE 3.2.B (Crnt'd) 10.
Only one trip system for each cooler fan.
11.
In only two of the four 4160-V shutdown boards.
See note 13, 12.
In only one of the four 4160-V shutdown boards. See note 13.
13.
An emergency 4160-V shutdown board is considered a trip system.
14.
RKRSW pump would be inoperable. Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.
15.
The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low res.ctor pressure or the vessel low water level (1 398" above vessel zero) originating in the core spray system trip system.
16.
The ADS circuitry is capable of accomplishing its protective action with one OPERABLE trip system. Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours.
17.
Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly.
If the test period for one RPT system arceeds two consecutive hours, the system will be declared inoperable.
If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly power reduction shall be initiated and reactor power shall be less than 85 percent within four hours.
l l
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l BFN 3.2/4.2-24 AMIN 0 MENT N0.183 Unit 2 i
(
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~
TABLE 4.2.8 SURVEILLANCE REQUIREMENTS FDR INSTRtMENTATION THAT INITIATE OR CONTROL THE CSCS e ce 5, E r
Function Functional Test Calibrati on Instrument Check Instrument Channel (1) (27)
Once/18 Months (28)
Once/ day Reactor low Water Level (LI5-3-58A-D)
Instrument Channel (1) (27)
Once/18 Honths (28)
Once/ day Reactor low Water Level (LIS-3-184 & 185)
Instrument Channel (1) (27)
Once/18 Months (28)
Once/ day Reactor Low Water Level (LIS-3-52 & 62A)
Instrument Channel (1) (27)
Once/18 Months (28) none Drywell High Pressure u
(PIS-64-58E41)
Instrument Channel (1) (27)
Once/18 Months (28) none Drywell High Pressure o
w (PIS-64-58A-D) i Instrument Channel (1) (27)
Once/18 Months (28) none Drywell High Pressure (PIS-64-57A-0)
Instrument Channel (1) (27)
Once/6 Months (28) none l
g Reactor Low Pressure (PIS-3-74A&B, PS-3-74A&B)
(PIS-68-95, PS-48-95)
(PIS-68-96, PS-68-%)
ME o55
-s 2":
9 N
C3 4
TABLE 4.2.8 (Continued)
SURVEILLANCE REQUIREMENTS FOR INSTRtpfMTATION THAT INITIATE OR CONTROL THE CSCS c en function Functional Test Calibration Instrument Checli
- E Core Spray Auto Sequencing Timers (4)
Once/ operating cycle none (Normal Power) g Cere Spray Auto Sequencing Timers (4)
Once/ operating cycle none
~
(Diesel Power)
LPCI Auto Sequencing Timers (4)
Once/ operating cycle none (Normal Power)
LPCI Auto Sequencing Timers (4)
Once/ operating cycle none (Diesel Power)
RHRSW A1, 83, C1, 03 Timers (4)
Once/ operating cycle none (Normal Power)
RHRSW A1, 83, C1, D3 Timers (4)
Once/ operating cycle none (Diesel Power)
ADS Timer (4)
Once/ operating cycle none u.
ADS High Drywell Pressure (4)
Once/ operating cycle none Bypass Timer n
[
RCIC Steam Line Space (1)
Once/3 months none w
Torus Area High Temperature RCIC Steam Line Space (1)
Once/3 months none RCIC Pump Room Area t
g High Temperature M
E e
e 3
e 9
~
e l
l TABLE 4.2.8 (Continued)
^
1 SURVElltANCE REQUIR90ffS FOR INSTRUMENTATION THAT INITIATE OR CONTROL THE CSCS e en EE l
to Function Functfonal Test Calibration Instrument Ches_k Instrument Channel -
(1)
Once/3 months none RHR Pump Discharge Pressure Instrument Channel -
(1)
Once/3 senths none l
Core Spray Pump Discharge Pressure Core Spray Sparger to RPV d/p (1)
O xe/3 senths Oxe/ day Trip System Bus Power Monitor Once/ operating Cycle N/A none Instrument Channel -
(1)
Once/3 wenths none Condensate Header Low Level (LS-73-56A. 8) i ta Instrument Channel -
(1)
Once/3 months none Suppression Chamber High Level Instrument Channel -
(1)
Once/3 months Once/ day y
Reactor High Water Level Instrument Channel -
(1)
Once/3 months none RCIC Turbine Steam Line High Flow Instrument Channel -
Once/31 days Once/18 months none f
RCIC Steam Supply Low Pressure 5
Instrument Channel -
Once/31 days Once/18 months none RCIC Turbine Ethaust Olaphreipu 3
High Pressure HPCI Steam Line Space (1)
Once/3 months nene Torus Area High Temperature HPCI Steam Line Space (1)
Once/3 months none HPCI Pump Room Area High Temperature A
6
TA8tE"4.2.8 (Continued)
SUPNEILLANCE REQUIRDIENTS FOR INSTRUPENTATION THAT INITIATE OR CONTROL THE CSCS c: en E, 2 '
r Function Functienal Test Calibration Instrumant Check Instrument Channel -
-(1)
Once/3 months none
- HPCI Turbine Steam Line High Flow Instrunent Channel -
Once/31 days Once/18 months e --
HPCI Steam Supply Low Pressure Instrument Channel -
Once/31 d'ays Once/I3 months none HPCI Turbine Enhaust Diaphrstpi High Pressure Core Spray System Logic.
Once/18 months (6)
N/A RCIC System (Initiating) Logic Once/18 months-N/A N/A RCIC System (Isolation) Logic Once/18 months (6)
N/A HPCI System (Initiating) Logic Once/18 months (5)
N/A
[
HPCI System (Isolation) Logic Once/18 months (6)
N/A ADS Logic
~
Once/18 months (6)
N/A LPCI (Initiating) Logic Once/18 months (6)
N/A LPCI (Containment Spray) Logic Once/18 months (6)
N/A R
Core Spray System Auto Initiation once/18 months (7)
N/A N/A a
Inhibit (Core Spray Auto Initiation) c+
LPCI Auto Initiation Inhibit once/18 months (7)
VA N/A m
1 g
(LPCI Auto Initiation) 4 I
g n
3.2 R&&E1 (Crnt'd) flow instrumentation is a backup to the temperature instrumentation. In the event of' a loss of the reactor building ventilation system, r'adiant heatins in the vicinity of-the main. steam lines raises the ambient.
temperature above 200'F. The temperature increases can cause an unnecessary main steam line isolation and reactor scraa. Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation..
High radiation monitors in the main steam line tunnel have been provided to detect gross' fuel failure as in the control rod drop accident. With the establinhed nominal setting of three times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference
~
Section 14.6.2 FSAR. An alars with a nominal setpoint of 1.5 x normal full-power background is provided also.
Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig.
The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in' actuation of HPCI isolation-valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE.
High temperature in the vicinity of the HPCI equipment is sensed by four sets of four binetallic temperature switches. The 16 temperature switches are arranged.in two trip systems with eight' temperature switches in-each trip system. Each trip-system consists of two elements..Each channel contains one temperature switch located in the pump room and three temperature switches located-in the torus area. The.RCIC high flow and high area temperature sensing instrument channels are arranged in the same manner as the HPCI system.
The HPCI high steam flow trip setting of 90 paid and the RCIC high steam flow trip setting of 450" H O have been selected such that the trip 2
setting is high enough to prevent spurious tripping during pump startup-but low enough to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.
The HPCI and RCIC steam line space. temperature switch trip settings are high enough to prevent spurious isolation due to normal temperature excursions.in the vicinity of the steam supply piping., Additionally, these trip settings ensure that'the primary. containment isolation steam supply valves isolate a break within an acceptable time period to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.
~
High. temperature at the Reactor Water Cleanup (RWCU) System floor drain in the space near the RWCU system or in the space near the pipe trench containing RWCU piping could indicate a break in the cleanup system.
high temperature occurs, the cleanup system is isolated.
When BFN 3.2/4.2-67 Amendment 187 Unit 2
d 3.2 BASIS (Cont'd)
,The instrumentation which initiates CSCS action is arranged in,a dual bus As for.other vital instrumentation arranged in this fashion, the system.
specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed.
The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.07.
The trip logic for this function is 1-out-of-n e.g., any trip on one of six APRMs, eight IRMs, or four SRMs will result in a rod block.
The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.
The APRM rod block function is flow biased and prevents a s'lgnificant reduction in MCPR, especially during operation at reduced flow.
The APRM provides gross core protection; 1.e.,
limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than 1.07.
The RBM rod block function provides local protection of the core; i.e.,
the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.
If the IRM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IRM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.
A downscale indication is an indication the instrument has failed or the instrument la not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.
The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.
For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time. The automatic-pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are BFH 3.2/4.2-68 Unit 2
i 3.2 RASES (Cent'd) adequate to assure the above criteria are met. The specification preserves t'he effectiveness of the system during periods of maint'enance, testing, or calibration, and also minimizes the risk of inadvertent.
operation; i.e.,
only-one instrument channel out of service.
Two post treatment off-gas radiation monitors are provided and, when their trip point is reached, cause an isolation of the off-gas line.
Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip or both have a downscale trip.
Both instruments are required for trip but the instruments are set so that the instantaneous stack release rate limit given in Specification 3.8 is not exceeded.
Four radiation monitors are provided for each unit which initiate Primary Containment Isolation (Group 6 isolation valves) Reactor Building Isolation and operation of the Standby Gas Treatment System. These instrument channels monitor-the radiation in the reactor zone ventilation axhaust ducts and in the refueling zone.
Trip setting of 100 ar/hr for the monitors in the refueling zone are based upon initiating normal ventilation isolation and SGTS operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the SGTS.
Flow integrators and sump fill rate and pump out rate timers are used to determine leakage in the drywell. A system whereby the time interval to fill a known volume will be utilized to provide a backup.. An air sampling system is also provided to detect leakage inside the primary containment (See Table 3.2.E).
For each parameter monitored, as listed in Table 3.2.F, there are two 6hannels of instrumentation except as noted. By comparing readings between the two channels, a near continuous surveillance of instrument performance is available. Any deviation in readings vill-initiate an-early recalibration, thereby maintaining the quality of the instrument readings.
A Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control room. An accident signal that isolates primary containment will also automatically isolate the control room and initiate the emergency pressurization system. In addition, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressurization rystem. Activity. required to cause automatic actuation is about one:mRam/hr.
Because of the constant surveillance and control exercised by TVA cvar the Tennessee Valley, flood levels of large magnitudes can be predicted in BFN.
3.2/4.2-69 Unit 2 4+>-
.+4
3.2 BASES (Ccnt'd) advance of their actual occurrence. In all cases, full advantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted; however, the plant flood e
protection is alitnya in place and does not depend in any way on advanced warning. Theref? ore, during flood conditions, the plant will be permitted to operate until water begins to run across the top of the pumping station at elevation 565.
Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition. At that time an ordeltly shutdown of the plant will be initiated, although surges even to a depth of several fast over the pumping station deck will not cause the losn of the main condenser circulating water pumps.
The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation dose to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.
The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the seismic response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for Brovns Ferry Nuclear Plant and to determine whether the plant can continue to be operated safely. The instrumentation provided is consistent with specific portions of the recommendations of Regulatory Guide 1.12
" Instrumentation for Earthquakes."
The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments will be calculated in accordance with guidance provided in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
This instrumentation also includes provisions for monitoring the concentration of potentially explosive gas mixtures in the off-gas holdup systam. The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CPR Part 50.
The radioactive liquid effluent instrumentation is provided to monitor and i
control, as spplicable, the releases of radioactive materials in liquid effluents during actual or potential _ releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with guidance provided in the ODCM to ensure that the alarm / trip vill occur prior to exceeding the limits of 10 CFR Part 20 Appendix B, Table II, Column 2.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design-Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
BTN 3.2/4.2-70 Unit 2 J
4 deme >9 egeen -
3.2 BA111 (C:Ent'd)
ATWS/RPT, Anticipated Transients without Scram / Recirculation Pump Trip system provides a means of limiting the consequences of the unlik'ely 1
occurrence of a failure to scram during an ATWS event. The response.of the plant to this postulated event (ATWS/RPT) follows the BWR Owners Group Report by General Electric NEDE-31096-P-A and the accompanying NRC Str ? Safety Evaluation Report.
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ATLJ/RPT utilizes the engineered safety feature (EST) master / slave analog trip units (ATU) which consists of four level and four pressure channels total. The initiating logic consists of two independent trip systems each consisting of two reactor done high pressure channels and two reactor vessel low level channels. A coincident trip of either two low levels or two high pressures in the same trip system causes initiation of ATWS/RPT. This signal from either trip system opens one of two EOC (and-of-cycle) breakers in series (the other system opens the other breaker) between the pump motor and the Motor Generator set driving each recirculation pump. Both systems are completely redundant such that only one trip system is necessary to perform the ATWS/RPT function. Power comes from the 250 VDC shutdown boards.
Setpoints for reactor dome high pressure and reactor vessel low level are such that a normal Reactor Protection System scram and accompanying recirculation pump trip would occur before or coincident with the trip by ATWS/RPT.
4.2 BASES The instrumentation listed in Tables 4.2. A through 4.2.F will be functionally tested and calibrated at regularly scheduled intervals. The same design reliability goal as the Reactor Protection System of 0.99999 generally applies for all applications of (1-out-of-2) X (2) logic.
Therefore, on-off sensors are tested once/3 months, and bistable trips associated with analog sensors and amplifiers are tested once/ week.
Those instruments which, when tripped, result in a rod block have their contacts arranged in a 1-out-of-n logic, and all are capable of being bypassed. For such a tripping arrangement with bypass capability provided, there is an optimum test interval that should be maintained in order.to maximize the reliability of a given channel (7). This takes
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account of the fact that testing degrades reliability and the optimum l i interval between tests is approximately given by:
2j i=
Where:
i= the optimum interval betwean tests.
t=
the time the trip conteets are disabled l
from performing-their function while l-the test is in progress.
r=
the expected failure rate of the relays.
l l
BFN 3.2/4.2-71 Unit 2 lll
4.2 BASES (Cent'd)
To test the trip relays requires that the channel be bypassed, the test made, and the system returned to its initial state. It is assumed'this task requires an estimated 30 minutet to complete in a thorough and.
workmanlike manner and that the relays have a failure rate of 10-6 failures p'er hour. Using this data and the above operation, the optimum test interval ist 2(0.5) 3 i =h 10-6
= 1 x 10
= 40 days For additional marmin a test interval of once eer month will be used initially.
The sensors and electronic apparatus have not been included here as these are analog devices with readouts in the control room and the sensors and electronic apparatus can be checked by comparison with other like instruments. The checks which are made on a daily basis are adequate to assure operability of the sensors and electronic apparat.s, and the test interval given above provides for optimum testing of the relay circuits.
The above calculated test interval optimizes each individual channel, considering it to be independent of all others. As an azample, assume that there are two channels with an individual technician assigned to each. Each technician tests his channel at the optimum frequency, but the two technicians are not allowed to communicate so that one can advise the other-that his channel is under test. Under these conditions, it is possible for both channels to be under test simultaneously. Now, assume that the technicians are required to communicate and that two channels are never tested at the same time.
(7).UCRL-50451, Improving Availability and Readiness of Field Equipment Through Periodic Inspection, Benjamin Epstein, Albert Shiff, July 16, 1968, page 10, Equation (24), Lawrence Radiation Laboratory.
Forbidding simultaneous testing improves the availability of the system over that which would be achieved by testins each channel independently.
These one-out-of-n trip systama vill be tested one at a time in order to take advantage of this inherent improvement in availability.
Optimizing each channel in' dependently may not truly optimize the system considering the overall rules of system operation. However, true system optimization is a compiez prooien. The optimums are broad, not sharp, and optimizing the: individual channels is generally adequate for the system.
The formula given above minimizes the unavailability of a single channel which must be bypassed during testing. The minimization of the unavailability is illustrated by curve No.1 of Figure 4.2-1 which assumes that a channel has a failure rate of 0.1 x 10-6/ hour and 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is required to test it.
The unavailability is a minimum at a test l
interval 1, of 3.16 x 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />.
l l
BFN 3.2/4.2-72 Unit 2
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4.2 BASES (CCut'd)
If two similar channels are used in a 1-out-of-2 configuration, the test interval for minimum unavailability thanges as a function of the r'les for u
testing. The simplest case is to test each one independent of the other.
In this case, there is assumed to be a finite probability that both may be bypassed at one time. This case is shown by Curve No. 2.
Note that the unortilability is lower as arpected for a redundant system and the minimum occ',rs at the same test intervs1. Thus, if the two channels are tested independently, the equation above yields the test interval for minimum unavailability.
A more usual case is that the testing is not done independently. If both channels are bypassed and tested at the same time, the result is shown in Curve No. 3.
Note that the minimum occurs at about 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, much longer than for cases 1 and 2.
Also, the minimum is not nearly as low as Case 2 which indicates that this method of testing does not take full advantage of the redundant channel.
Bypassing both channels for simultaneous testing should be avoided.
The most likely case would be to stipulate that one channel be bypassed, tested, and restored, and then immediately following, the second channel be bypassed, tested, and restored. This is shown by Curve No. 4.
Note that there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel.
The best test procedure of all those examined is to perfectly stagger the tests. That is, if the test interval is four months, test one or the other channel every two months. This is shown in Curve No. 5.
The difference between Cases 4 and 5 is negligible. There may be other arguments, however, that more strongly support the perfectly staggered tests, including reductions in human error.
The conclusions to be drawn are these 1.
A 1-out-of-n system may be treated the same as a single channel in terms of choosing a test interval; and 2.
more than one channel should not be bypassed for testing at any one time.
The radiation monitors in the refueling area ventilation duct which initiate building isolation and standby gas treatment operation are arranged in two 1-out-of-2 logic systems. The bases given for the rod blocks apply here also and were used to arrive at the functional testing frequency. The off-gas post treatment monitors are connected in a 2-out-of-2 logic arrant,ement. Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate.
The automatic pressure relief instrumentation can be considered to be a 1-out-of-2 logic system and the discussion above applies also.
BFN 3.2/4.2-73 Unit 2
,w 4
9
%< RA1E& (Cont'd)
. The criteria for ensuring the reliability and accuracy of the radioactive gaseous effluent instrumentation is listed in Table 4.2.K.
The criteria for ensuring the reliability and accuracy of the radioactive liquid effluent instrumentaiton is listed in Table 4.2.D.
4 BFN 3.2/4.2-73a Unit 2
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