ML18038B203
| ML18038B203 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/16/1995 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18038B204 | List: |
| References | |
| DPR-33-A-220, DPR-68-A-194 NUDOCS 9503200312 | |
| Download: ML18038B203 (52) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISS(ON WASHINGTON, D.C. 2055&4001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FER Y NUCLEAR PLANT UNIT 1 AMENDMENT TO FACILITY OPE ING LICENSE Amendment No. 220
'License No. DPR-33 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated March 31, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission'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 Commission; C.
There is reasonable assurance (i) 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 Commission'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 Commission's regulations and all applicable requirements have been satisfied.
95032003i2 9503ib PDR
- DOCK 05000259 P
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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.
DPR-33 is hereby amended to read as follows:
3.
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 220, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
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 Q ~c.l~
Frederick J.
Heb n, Director Project Directorate II-4 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
Ihrch 16, 1995
II J
ATTACHMENT TO LICENSE AMENDMENT NO.
220 FACILITY'PERATING LICENSE:NO'.
DPR-33 DOCKET NO. 50-259 Revise the Appendix A Technical Speci'fications by removing the pages identified below and inserting, the enclosed pages.
The. revised pages
.are identified by the captioned amendment
.number and contain marginal 1'ines indicating, the. area 'of change.
Overl'eaf pages are provided to maintain, document completeness.,
REMOVE 3.2/4'.2-18 3'. 2/4,. 2-19 3.2/4.'2-22'.2/4.2'-22a 3.2/4.2-23 3.2/4.2-24 3.2/4,2-46 3..2/4.2-47 3.2/4.2-67 3.2/4.2-68 INSERT 3.2/4.2-18 3.2/4.2-19 3,.2/4.2-22*
3.2/4.2-22a 3.2/4.2-23 3'.2/4.2-24*
3.'2/4.2-46 3.'2/4.2-47 3.2/4.2-67 3.2/4.2-68*
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TABLE 3.2.B (Continued)
Hini'mum No.
Oper'able Per
~Tri ~l 1
1(2)
Fun i
n HPCI Trip System bus power monitor RCIC Trip System bus power monitor Instrument Channel-Condensate Header Low Level (LS-73-56A 4 8)
Tri Lv1 tin N/A N/A
> Elev. 551'Ai n
R m rk 1.
Honi tors availabil ity of power to logic systems.
1.
Honi tors avail ability of power to 1 ogi c sys tems.
1.
Below trip settirig will open HPCI suction valves to the suppression chamber.
4J I
I
<0 1(2) 2(2) 3(2) 3(2)
Instrument Channel-Suppression Chamber High Level Instrument Channel-Reactor High Water 'Level Instrument Channel-RCIC Turbine Steam Line High Flow Instrument Channel-RCIC Steam Supply Pressure - Low (PS 71-.1A-D)
Instrument Channel-RCIC Turbine Exhaust Diaphragm Pressure-High (PS 71-11A-0)
< 7" above instrument zero A
< 583" above vessel zero A
< 450" H20 (7)
>50 psig
<20 psig 1.
Abbve trip setting will open HPCI suction valves to the suppression chamber.
l.
Above trip setting trips RCIC turbine.
1; Above trip setting isolates RCIC system and trips RCIC
- turbine.
l.
Below trip setting isolates RCIC system and trips RCIC turbine.
l.
Above trip setting isolates RCIC system and trips RCIC turbine.
F3 8
H AHH
~
A8
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Hinimum No.
i' Operable Per
~Tri
~sl I
2(2) 3(2) 3(2) 1 (16)
Func ion Instrument Channel-Reactor High Mater Level Instrument Channel-HPCI Turbine Steam Line High Flow Instrument Channel-HPCI Steam Supply Pressure Low (PS 73-lA-D)
Instrument Channel-HPCI Turbine Exhaust Diaphragm (PS 73-20A-D)
Core Spray System Logic RGIC System (Initiating)
Logic RCIC System
( Isol ation)
Logic ADS Logic TABLE 3.2.8 (Continued)
Tri Lv1 5
in
<583" above vessel zero.
<90 psi (7)
)100 psig
<20 psig N/A N/A N/A N/A
- ActCin, B
R mark l.
Above trip setting trips HPCI turbine.
Above trip setting isolates HPCI system and trips HPCI turbine.
l.
BeloQ trip setting isolates HPCI system and trips HPCI turbine.
l.
Above trip setting isolates HPCI sy'tem and trips HPCI turbine.
l.
Includes testing auto initiation inhibit to Core Spray Systems in other units.
1.
Includes Group 7 valves.
2.
Group 7:
A Group 7 i sol ati on is automatically actuated by only the following condition:
l.
The respective turbine steam supply valve n'ot fully closed.
l.
Includes Group 5 valves.
2.
Group 5:
A Group 5 isolation is actuated by any of the following conditions:
a.
RCIC Steamline Space High Temperature b.
RCIC Steamline High Flow c.
RCIC Steamline Low Pressure d.
RCIC Turbine Exhaust Diaphragm High Pressure O
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TABLE 3.2.8 (Continued)
M ID Ql Minimum No.
Operable Per
~Tri c
1 1(10) 1(10) 1(10) 1(12) 1(12) 1(13) 1(16) 1(16) func ion Instrument Channel
'hermostat (Core Spray Area Cooler Fan)
RHR Area Cooler Fan Logic Core Spray Area Cooler Fan Logic Instrument Channel Core Spray Hotors A or C
Start Instrument Channel Core Spray Hotors 8 or 0 Start Instrument Channel-Core Spray Loop 1 Accident Signal (15)
Instrument Channel Core Spray Loop 2 Accident Signal (15)
RHRSW Initiate Logic RPT i.ogic AOS Timer AOS High Drywell Pressure Bypass Timer Tri Lev 1
in 100oF N/A N/A N/A N/A N/A N/A N/A N/A t
< 115 sec.
t < 322 sec.
~cc icc (14)
R mark 1.
Above trip setting starts Core Spray area cool e r fans.
1.
Starts RHRSW pumps Al, 83, Cl, and D3 1.
Starts RHRSW pumps Al, 83, Cl, and 03 1.
Starts RHRSW pumps Al, 83, Cl, and 03 1.
Starts RHRSW pumps Al, 83, Cl, and D3 1.
Trips recirculation pumps on turbine control valve fast closure or stop valve closure
> 30$ power.
l.
Above trip setting in conjunction with 1 ow reactor water level permissive, low r'eactor water level; high drywell pressure or ADS high drywell pressure bypass timer timed out, and RHR or CSS pumps running, initiates ADS.
l.
Above trip setting, in conjunction with low reactor water level permissive, low reactor water level, AOS timer timed out and RHR or CSS pumps running, initiates AOS.
0 IE
TABLE 3.2.B (Continued)
Hinimum No.
Operable Per
~Tri ~l Fn in RCIC Steam Line Space Torus Area Hi gh Temperature RCIC Steam Line Space RCIC Pump Room Area High Temperature HPCI Steam Line Space Torus Area High Tem'perature HPCI Steam Line,Space HPCI Pump Room Area High Temperature
-Tri L v 1
in
<1550 F
<180'F
<180'F
<200'F
~Ai n R
m rk l.
Above trip setting isolates RCIC system and trips RCIC turbine.
l.
Above trip setting isolates RCIC system and trips RCIC turbine.
l.
Above trip setting isolates HPCI system and trips HPCI turbine.
l.
Above trip setting isolates HPCI system and trips HPCI turbine.
O
NOTES OR L
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1.,
Whenever any CSCS System is required by Section 3.5 to be OPERABLE, there shall be two OPERABLE trip systems except as noted. If a requirement 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 RHRS pump is scheduled to start immediately and each CSS pump is sequenced to start about 7 sec. later.
6.
With normal power, one CSS and one RHRS pump is scheduled to start instantaneously, one CSS and one RHRS 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.
7.
The RCIC and HPCI steam line high flow trip level settings are given in terms of differential pressure.
The RCICS setting of'50" of water 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.
Similarly, the HPCIS setting of 90 psi corresponds to at least 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 item.
9.
The head tank is designed to assure that the discharge piping from the CS and RHR 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.
BFN Unit 1 3.2/4.2-23 AIfENDIfEHT'NO. 220
NOTES FOR TABLE 3.2 B
(Cont'd) 10.
Only one trip system for each cooler fan.
ll.
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.
RHRSW 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 reactor pressure or the vessel low water level (g 378" 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 ei'ght 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 exceeds 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 30 percent within four hours.
18.
Not required to be OPERABLE in the COLD SHUTDOWN CONDITION.
BFN Unit 1 3.2/4.2-24 AMENDMM'0.I 8 0 l
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TABLE 4.2.8 (Continued)
SURVEILLANCE RE()UIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fun i n Instrument Channel-RHR Pump Discharge Pr'essure Instrument Channel-Core Spray Pump Discharge Pressure Func ion 1
T libr i n once/3 months once/3 months In trumn h
k none none Core Spray Sparger to RPV d/p Trip System Bus Power Honitor Instrument Channel Condensate Header Low Level (LS-73-56A, 8)
Instrument Channel-Suppression Chamber High Level Instrument Channel Reactor High Mater Level Instrument Channel-RCIC Turbine Steam Line High Flow Instrument Channel-RCIC Steam Supply Low Pressure Instrument Channel-RCIC Turbine Exhaust Diaphragm High Pressure RCIC Steam Line Space Torus Area High Temperature RCIC Steam Line Space RCIC Pump Room Area High Temperature once/operating Cycle once/31 days once/31 days once/3 months N/A once/3 months once/3 months once/3 months once/3 months once/18 months once/18 months once/3 months once/3 months once/day none none none once/day none once/day once/day none none O
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TABLE 4.2.8 (Continued)
SURVEILLANCE RE()UIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fun i n HPCI Steam Line Space Torus Area High Temperature HPCI Steam Line Space HPCI Pump Room Area High Temperature Instrument Channel-HPCI Turbine Steam Line High Flow Instrument Channel-HPCI Steam Supply Low Pressure Instrument Channel
HPCI Turbine Exhaust Diaphragm High Pressure Core Spray System Logic RCIC System (Initiating) Logic RCIC System (Isolation) Logic HPCI System (Initiating) Logic HPCI System (Isolation) Logic ADS Logic LPCI (Initiating) Logic LPCI (Containment Spray)
Logic Core Spray System Auto Initiation Inhibit (Core Spray Auto Initiati on)
LPCI Auto Initiation Inhibit (LPCI Auto Initiation)
Fn i nl T
once/31 days once/31 days once/18 months once/18 months once/18 months once/18 months once/18 months once/18 months once/18 months once/18 months once/18 months (7) once/18 months (7) libr i n once/3 months once/3 months once/3 months once/18 months once/18 months (6)
N/A (6)
(6)
(6)
(6)
(6)
(6)
N/A N/A In rmn h
k none none none once/day once/day N/A N/A N/A N/A N/A K/A N/A N/A N/A N/A O
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3.2 BASES (Cont'd)
The setting of 200'F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks.
For large breaks, the high steam flow instrumentation is. a, backup to the temperature instrumentation.
In the event of a loss of the reactor building ventilation system, radiant heating 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 scram.
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.
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 bimetallic 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 channels.
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 psid and the RCIC high steam flow trip setting of 450" H20 have been selected such that the trip 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 isol'ate 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 Cleanup System floor drain could indicate a break in the cleanup system.
When high temperature
- occurs, the cleanup system is isolated.
The instrumentation which initiates CSCS action is arranged in a,dual bus system.
As for other vital instrumentation arranged in this fashion, the 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.
BFN Unit 1 3.2/4.2-67 AtfHNDHENT NO.
220
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3.2 BASES (Cont'd)
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 availabie,. and the RBM is a backup system to the written sequence for'ithdrawal of control rods.
The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced'low.
The APRM provides gross core protection; i.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 is 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 adequate to assure the above criteria are met.
The specification preserves the effectiveness of the system during.periods of maintenance,
- testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.
Two 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 exhaust ducts and in the refueling zone.
BFN Unit 1 3.2/4.2-68 NENOMgffPg. p p5
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gP,S RE00 Cg
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++*++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 2055&0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
194 License No.
DPR-68 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The appl'ication for amendment by Tennessee Val.ley Authority (the licensee) dated March 31,
- 1994, compl,ies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act)',
and the Commission'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 Commission; C.
There is reasonable assurance (i) 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 Commission'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 Commission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly, the license is amended'y changes to the Technical Specifications as indicated in the attachment to this l.icense amendment and paragraph.2.C.(2) of 'Facil'ity Operating License No.
DPR-68 is hereby amended'o read as follows:
3.
(2)'echnical S ecifications The Techn'ical Specifications contained in Appendices A. and B,, as. revised
,through Amendment 'No. 194, are hereby incorporated in the license.
The
,licensee shall operate the facil.ity in accordance with the Technical Specifications.
This 1:icense amendment:is effective as of its date of issuance and shal,l be implemented within 30 days 'from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Frederick J.
Heb n, Director Project Directorate II;4 Divis'ion of 'Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to, the Technical Specifications Date of Issuance:
March 16, 1995
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ATTACHMENT TO LICENSE AMENDMENT NO.
194 FACILITY OPERATING LICENSE NO.
DPR-68 DOCKET NO. 50-296 Revise the Appendix A Technical Specifications by removing the pages identified beTow and inserting the enclosed pages.
The revised pages are identi.fied by the captioned amendment number and contain marginal lines indicating the are'a of change.
Overleaf and spillover pages are provided to maintain document completeness.
REMOVE 3.2/4.2-18 3.2/4.2-19 3.2/4.2-21a 3.2/4.2-2lb 3.2/4.2-22 3.2/4.2-23 3.2/4.2-45 3.2/4.2-46 3.2/4.2-66 3.2/4.2-67 3.2/4.2-68 3.2/4.2-69 INSERT 3.2/4.2-18 3.2/4.2-19 3.2/4.2-21a 3.2/4.2-21b*
3.2/4.2-22 3.2/4.2-23*
3.2/4.2-45 3.2/4.2-46 3.2/4.2-66 3.2/4.2-67**
3.'2/4.2-68**
3.2/4'.2-69**
- 1 0
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TABLE 3.2.B (Continued)
Hinimum No.
Operable Per
~Tri ~l 1(2)
Fun i
n HPCI Trip System bus power moni toi RCIC Trip System bus power monitor Instrument Channel Condensate Header Low Level (LS-73-56A & 8)
Tri L v l.
in N/A N/A
> Elev.551'Ai n
R rk 1.
Honi tors avail ability of power to logic systems.
1.
Honitors availability of poqer to logic systems.
1.
Below trip setting will open HPCI suc'tion valves to the suppression chamber.
2(2) 2(2)
Instrument Channel-Suppression Chamber High Level Instrument Channel Reactor High Water Level
< 7" above instr'ument zero A
< 583" above vessel zero l.
Above trip setting will open HPCI suction valves to the suppression chamber.
l.
Above trip setting trips RCIC turbine.
3(2)
Instrument Channel-RCIC Turbine Steam Line High Flow
< 450" H20 (7)
Instrument Channel-
>50 psig RCIC Steam Supply Pressure Low (PS 71-lA-D) 1.
Above trip setting isolates RCIC system and trips RCIC turbine.
1.
Below trip setting isolates RCIC system and trips RCIC
- turbine, 3(2)
Instrument Channel-RCIC Turbine Exhaust Diaphragn Pressure High (PS 71-11A-0)
< 20 psig l.
Above tr'ip setting isolates RCIC system and trips RCIC turbine.
O
II 0
~i~
U
Hinimum No.
g ~
Operable Per
~Tri S
1 2(2)
, 3(2) 3(2) 4J 4
I i
1 (16)
Func ion instrument Channel-Reactor High Mater Level Instrument Channel-HPCI Turbine Steam Line High Flow Instrument Channel-HPCI Steam Supply Pressure - Low (PS 73-lA-0)
Instrument Channel-HPCI Turbine Exhaust Diaphragm (PS 73-20A-D)
Cor'e Spray System Logic RCIC Sy'tem (Initiating)
Logic RCIC System (Isolation)
L091c ADS Logic TABLE 3.2.B (Continued)
Tri Level Se tin Action
<90 psi (7)
>100 psig
<20 psig N/A N/A N/A N/A
<583" above vessel zero.
A R marks l.
Above trip setting trips HPCI turbine.
l.
Above trip setting isolates HPCI system and trips HPCI turbine.
l.
Below trip setting isolates HPCI system and trips HPCI turbine.
l.
Above trip setting isolates HPCI system and trips HPCI turbine.
1.
Includes testing auto initiation inhibit to Core Spray Systems in other units.
l.
Includes Group 7 valves.
2.
Group 7:
A Group 7 isolation is automatically actuated by only the following condition:
1.
The respective turbine steam supply valve not fully closed.
l.
Includes Group 5 valves.
2.
Group 5:
A Group 5 isolation is actuated by any of the following condi tions:
a.
RCIC Steamline Space High Temperature b.
RCIC Steamline High Flow c.
RCIC Steamline Low Pressure d.
RCIC Turbine Exhaust Diaphragm High Pressure O
II II
TABLE 3.2.B (Continued)
Hinimum No.
Operable Per Tri~i~l 1(16) 1(16)
Function ADS Timer AOS Hi gh Orywel 1 Pressure Bypass Timer RCIC Steam Line Space Torus Area High Temperature RCIC Steam Line Space RCIC Pump Room Area High Temperature HPCI Steam Line Space Torus Area High Temperature HPCI Steam Line Space HPCI Pump Room Area High Temperature Tri Lev 1
in t
< 115 sec.
t
< 322 sec.
<155OF
<180 F
<180'F
<200 F
~Ai n R
m rk 1.
Above trip setting in conjunction with low reactor water level permissive, low reactor water level; high drywell pressure or AOS high'drywell pressure bypass timer timed out, and RHR or CSS pumps runriing, initiates AOS.
l.
Above trip setting, in conjunction with low reactor water level permissive, low reactor water level, ADS timer timed out and RHR or CSS pumps running, initiates ADS.
l.
Above trip setting isolates RCIC system and trips RCIC turbine.
1.
Above trip setting i sol ates RCIC syste'm an'd trips RCIC turbine.
l.
Above trip setting isolates HPCI system and trips HPCI turbine.
1.
Above trip setting i sol ates HPCI system and trips HPCI turbine.
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THIS PAGE.IHTEHTIOHALLYLEFT BLANK BFH Unit 3
- 3. 2/4. 2-21b ANENDMElfFNO. X 7 8
0
NOTES FOR TABLE 2
B 1.
Whenever any CSCS System is required by Section 3.5 to be OPERABLE, there shall 'be two OPERABLE trip systems except as noted. If a requirement 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'y 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 'hours 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 RHRS pump is scheduled to start immediately and each CSS pump i's sequenced to start about 7 seconds later.
6.
With normal power, one CSS and one RHRS pump is, scheduled to start instantaneously, one CSS and one RHRS pump is sequenced-to start after about 7 seconds with similar pumps starting after about 14 seconds and 21 seconds, at which time the full complement of CSS and'HRS pumps would be operating.
7.
The RCIC and HPCI steam line high flow trip level.settings are given in terms of differential pressure.
The RCICS setting of 450" of water 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 1'ine break.
Similarly, the HPCIS setting of 90 psi corresponds to at least 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 item.
9.
The head tank is designed to assure that the discharge piping from the CS and RHR 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.
BFN Unit 3 3.2/4.2-22 AlfHNDIIENT NO.
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NOTES FOR TABL'E 2
B (Continued) 10.
Only one trip system for each cooler fan.
ll.
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.
RHRSW 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 reactor pressure or the vessel low water level (g 378" 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 exceeds 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 30 percent within four hours.
,)
18.
BFN Unit 3 3.2/4.2-23 AMENOMENTN. I52
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TABLE 4.2.B (Cont'd)
SURVEILLANCE REQUIREMENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fun i
n Instrument Channel-RHR Pump Discharge Pressure Instrument Channel-Core Spray Pump Discharge Pressure Core Spray Sparger to RPV d/p Trip System Bus Power Monitor Instrument Channel Condensate Header Level (LS-73-56A, 8)
Instrument Channel-Suppression Chamber High Level Instrument Channel-Reactor High Hater Level Instrument Channel-RCIC Turbine Steam Line High Flow Instrument Channel-RCIC Steam Supply Low Pressure Instrument Channel-RCIC Turbine Exhaust Diaphragm High Pressure RCIC Steam Line Space Torus Area High Temperature RCIC Steam Line Space RCIC Pump Room Area High Temperature HPCI Steam Line Space Torus Area High Temperature HPCI Steam Line Space HPCI Pump Room Area High Temperature Fn i nl T once/operating Cycle once/31 days once/31 days libr i n
once/3 months once/3 months once/3 months N/A once/3 months once/3 months once/3 months once/3 months once/18 months once/18 months once/3 months once/3 months once/3 months once/3 months In rmn h
k none none once/day none none none once/day none once/day once/day none none none none
II 41
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TABLE 4.2.8 (Cont'd)
SURVEILLANCE RE()UIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fn in Instrument Channel-HPCI Turbine Steam Line High Flow Instrument Channel-HPCI Steam Supply Low Pressure Instrument Channel-HPCI Turbine Exhaust Diaphragm High Pressure Core Spray System Logic RCIC System (Initiating) Logic RCIC System (Isolation) Logic HPCI System
( Initiating) Logi c HPCI System (Isol ati on) Logic ADS Logic LPCI (Initiating) Logic LPCI (Containment Spray) Logic Fn i nl T
once/31 days once/31 days once/18 months once/18 months once/18 months once/18 months once/18 months once/18 months once/18 months once/18 months libr i n once/3 months once/18 months once/18 months (6)
N/A (6)
(6)
(6)
(6)
(6)
(6)
In rmn h
k none once/day once/day N/A N/A N/A N/A N/A N/A N/A N/A 8
3.2 BASES (Cont'd)
The setting of 200 F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks.
For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.
In the event of a loss of the reactor building ventilation system, radiant heating 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. scram.
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.
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 bimetallic 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 channels.
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 psid and the RCIC high steam flow trip setting of 450" H20 have been selected such that the trip 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'o prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.
High temperature at the Reactor Water Cleanup (RWCU) System in the main steam valve vault, RWCU pump room 3A, RWCU pump room 3B, RWCU heat exchanger room or in-the space near the pipe trench containing RWCU piping could indicate a break in the cleanup 'system.
When high temperature
- occurs, the cleanup system is isolated.
BFN Unit 3 3.2/4.2-66 AliEi%MfENT HO.
194
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3.2 BASES (Cont'd)
The instrumentation which initiates CSCS action is arranged in a dual bus system.
As for other vital instrumentation arranged in this fashion, the 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 wil'1 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 significant reduction in MCPR, especially during operation at reduced flow.
The APRM provides gross core protection; i.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.
1'f the IRM channels are in the worst condition of allowed bypass, the seal'ing 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 is 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 adequate to assure the above criteria are met.
The specification preserves the effectiveness of the system during periods of maintenance,
- testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.
BFN Unit 3 3.2/4.2-67
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AItENDIIE¹ NO. 194
3.2 BASES (Cont'd)
Two 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 exhaust ducts and in.the refueling zone.
Trip setting of 100 mr/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.
The allowed inoperable time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for functional testing or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for calibration and maintenance (with the downscale trip of the inoperable channel in the tripped condition) of the.Reactor Building Ventilation system 'is based upon a Probabilistic Risk Assessment (PRA).
The assessment considered the failures, relay failures and the probability of an accident occurring for which the RBVRMs are required to operate.
Flow integrators and sump fillrate and pump out rate timers are used to determ'ine leakage in the drywell.,
A system whereby the time interval to filla 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 channels 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 will initiate an early recalibration, thereby maintaining the quality of the instrument
,readings.
Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control'oom.
An accident signal that isolates primary containment will also automatical'ly isolate the control room and initiate the emergency pressuri'zation system.
In addition, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressurization system.
Activity required to cause automatic actuation is about one mRem/hr.
Because of the constant surveillance and control exercised by TVA over the Tennessee Valley, flood levels of large magnitudes can be predicted in 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.
Therefore, during flood conditions, the plant will be p'ermitted to operate until water begins to run across the top of the pumping station at elevation 565.,
Seismically BFN Unit 3 3.2/6.2-68 AtfENDEIENT NO.
194
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3.2 BASES (Cont'd) 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 orderly shutdown of the plant will be initiated, although surges even to a depth of several feet
'ver the pumping station deck will not cause the loss 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 Browns Ferry, Nuclear Plant and to determine whether the plant can continue to be operated'afely.
The instrumentation provided is consistent with specific portions of the recommendations of Regulatory Guide 1.12 "Instrumentation for Earthquakes."
The instrumentation in Tables 3.2.K/4.2.K monitors the concentration of potentially explosive gas mixtures in the offgas holdup system.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 63 of Appendix A to 10 CFR Part 50.
ATWS/RPT, Anticipated Transients without Scram/Recirculation Pump Trip system provides a means of limiting the consequences of the unlikely 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 Staff Safety Evaluation Report.
ATWS/RPT utilizes the engineered safety feature (ESF) master/slave analog trip units (ATU) which consists of four level and'our pressure channels total.
The initiating logic consists of two independent trip systems each consisting of two reactor dome 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 BFN Unit 3 3.2/4.2-69 AIIZNDIIENT NO.
194
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